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	claims	1. A charged-particle beam apparatus comprising:a chamber having an interior evacuated by an intra-chamber evacuating means; anda lens-barrel emitting a charged-particle beam onto a sample placed in the chamber,wherein the lens-barrel comprises:a cylindrical body having a distal end at which an emission outlet is formed for communication with the chamber and from which the charged-particle beam is released;a charged-particle supplier housed at a side of a proximal end in an interior of the cylindrical body to release the charged-particle beam; andan objective lens housed at a side of a distal end in the interior of the cylindrical body and having an electrostatic lens generating an electric field by voltage application and converging the charged-particle beam released from the charged-particle supplier, anda gas supplying means in the cylindrical body of the lens-barrel for supplying an inert gas into the cylindrical body at a side of a proximal end of the objective lens and configured to provide an inert gas pressure in the chamber sufficient to substantially prevent dust present in the chamber from contact with the objective lens. 2. The charged-particle beam apparatus according to claim 1, wherein the gas supplying means includes:a gas supply pipe having a distal end provided at the proximal end side of the objective lens;a gas supplier connected to a proximal end of the gas supply pipe to supply the gas; anda gas supply valve controlling an open/close state of the gas supply pipe. 3. The charged-particle beam apparatus according to claim 2, wherein the gas supply pipe of the gas supplying means is provided with a filter removing dust mixed in the gas supplied from the gas supplier. 4. The charged-particle beam apparatus according to claim 2, wherein in the gas supply pipe of the gas supplying means, at least the distal end arranged inside the cylindrical body comprises a metal. 5. The charged-particle beam apparatus according to claim 2, wherein in the gas supply pipe of the gas supplying means, at least the distal end arranged inside the cylindrical body is subjected to vacuum baking. 6. The charged-particle beam apparatus according to claim 1, further comprising:a cylindrical body valve interposed between the gas supplying means and the charged-particle supply part to control an open/close state of each of the distal end side and the proximal end side in the cylindrical body; anda supply part evacuating means provided at the proximal end side of the cylindrical body to maintain the proximal end side of the cylindrical body, in which the charged-particle supply part is housed, at an ultra high vacuum state which is higher than a vacuum state of the chamber. 7. The charged-particle beam apparatus according to claim 1, wherein the objective lens further has a magnetic field lens generating a magnetic field so as to superimpose the magnetic field on the electric field generated by the electrostatic lens and converging the charged-particle beam. 8. The charged-particle beam apparatus according to claim 1, wherein the gas supplying means is configured to supply the inert gas at a distal end of the cylindrical body. 9. The charged-particle beam apparatus according to claim 8, wherein the gas supplying means is configured to supply the inert gas such that an atmospheric pressure is created inside the chamber and the distal end of the cylindrical body.
	description	This application is a continuation of U.S. patent application Ser. No. 12/003,145, filed Dec. 20, 2007, the contents of which are incorporated by reference in their entirety. 1. Field of the Invention Example embodiments generally relate to fuel structures and materials used in nuclear power plants. 2. Description of Related Art Generally, nuclear power plants include a reactor core having fuel arranged therein to produce power by nuclear fission. A common design in U.S. nuclear power plants is to arrange fuel in a plurality of fuel rods bound together as a fuel assembly, or fuel bundle, placed within the reactor core. These fuel rods typically include several elements joining the fuel rods to assembly components at various axial locations throughout the assembly. As shown in FIG. 1, a conventional fuel bundle 10 of a nuclear reactor, such as a BWR, may include an outer channel 12 surrounding an upper tie plate 14 and a lower tie plate 16. A plurality of full-length fuel rods 18 and/or part length fuel rods 19 may be arranged in a matrix within the fuel bundle 10 and pass through a plurality of spacers 15. Fuel rods 18 and 19 generally originate and terminate at the same vertical position, all rods continuously running the length of the fuel bundle 10, with the exception of part length rods 19, which all terminate at a lower vertical position from the full length rods 18. An upper end plug 20 and/or lower end plug 30 may join the fuel rods 18 and 19 to the upper and lower tie plates 14 and 16, with only the lower end plug 30 being used in the case of part length rods 19. As shown in FIGS. 2A and 2B, conventional upper and lower tie plates 14 and 16 may be generally solid and flat. A plurality of holes, called bosses, 25 may receive lower end plugs of all rods in an assembly in the lower tie plate 16. Similarly, a plurality of bosses 25 may receive the upper end plug of all full-length rods in the upper tie plate 14. Part length rods may not terminate at a tie plate. In this way, upper and lower tie plates 14 and 16 may axially join fuel rods to the fuel assembly and hold fuel rods at a constant and shared axial displacement in the core. Because bosses and corresponding fuel rods may begin and/or terminate at the same axial position within the bundle, fluid flow may be restricted at these axial positions. Example embodiments are directed to tiered tie plates and fuel bundles that use tiered tie plates. Example embodiment tie plates may include upper and lower tiered tie plates. Example embodiment tiered tie plates may have a plurality of bosses divided into groups, or tiers, having differing vertical (axial) displacement. In this way, bosses may receive fuel rods at varying vertical displacements depending on how the bosses are grouped and displaced. Example embodiment tiered tie plates may reduce fluid flow pressure drop by increasing the minimum cross sectional flow area available through a fuel bundle. Example embodiment fuel bundles may use tiered tie plates such that fuel rods in example bundles may originate and terminate at different vertical displacements, based upon the vertical displacement of the bosses receiving the fuel rods into the tiered tie plates. Alternatively, shanks may be used to further vary fuel rod axial displacement and diameter, allowing, for example, same-length fuel rods to originate at different axial displacements from a lower tiered tie plate and terminate at a shared axial displacement at an upper flat tie plate. In this way thermo-hydraulic characteristics of example embodiment fuel bundles may be modified based on the vertical displacement of rods and/or shanks placed therein. Detailed illustrative embodiments of example embodiments are disclosed herein. However, specific structural and functional details disclosed herein are merely representative for purposes of describing example embodiments. The example embodiments may, however, 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.). The terminology used herein is for the purpose of describing particular embodiments only and is not intended to be limiting of example embodiments. As used herein, the singular forms “a”, “an” and “the” are intended to include the plural forms as well, unless the language explicitly indicates otherwise. It will be further understood that the terms “comprises,” “comprising,” “includes,” and/or “including,” when used herein, specify the presence of stated features, integers, steps, operations, elements, and/or components, but do not preclude the presence or addition of one or more other features, integers, steps, operations, elements, components, and/or groups thereof. It should also be noted that in some alternative implementations, the functions/acts noted may occur out of the order noted in the figures. For example, two figures shown in succession may in fact be executed substantially concurrently or may sometimes be executed in the reverse order, depending upon the functionality/acts involved. FIG. 3 illustrates an example tiered tie plate 100 useable as a lower tiered tie plate. As shown in FIG. 3, example embodiment tiered tie plate 100 may include a body 115 having an inlet 110 configured to allow fluid coolant and/or moderator to pass into and through the body 115 of example tiered tie plate 100. Inlet 110 may take on a variety of configurations based on the application; for example, inlet 110 may include positioning and holding elements 112, shaped as hoops, that may retain the example embodiment tiered tie plate in a specific orientation and position relative to the core and other fuel structures. Inlet 110 may connect to a nozzle 111 that permits fluid flow to pass into a midsection 113. Midsection 113, while shown as hexahedral in FIG. 3, may be shaped in any form depending on the application and may include conventional flow-mixing and conditioning equipment. Inlet 110, nozzle 111, and midsection 113 may make up the body 115 of example embodiment tiered tie plate 110. Example embodiment tiered tie plate 100 further includes bosses 120 defining holes shaped to receive ends or end plugs of nuclear fuel useable with example embodiment tiered tie plates. In this embodiment, the bosses 120 are annular. Axial support members 130 may connect the bosses 120 to each other and to the body 115 of the example embodiment tiered tie plate 100. Some bosses 120 may be integrated into the body 115 midsection 113. Lateral support members 135 may connect the bosses 120 to each other in a transverse direction perpendicular to the axial (vertical) direction. The bosses 120 may be spaced at a desired interval. For example, bosses 120 may be spaced in a square lattice formation with axial and lateral support members 130 and 135 extending at least transversely from each boss 120 at 90-degree intervals, as shown in FIG. 3, or bosses 120 and support members 130 and 135 may be spaced radially, triangularly, etc., depending on where fuel rods may be placed in the assembly. Similarly, bosses 120 and support members 130 and 135 are not necessarily spaced at regular intervals. Bosses 120 and axial and lateral support members 130 and 135 may be omitted over larger intervals to allow for other components or flow to pass through example embodiment tiered spacer 100. For example, some or all bosses 120 may be omitted, resulting in intersecting members 130 and 135, and an alternative fuel end plug may be used. Axial and lateral support members 130 and 135 are shown with thin transverse profiles so as to form several flow areas through example embodiment tiered tie plate 100. In this way fluid may pass through the inlet 110 and out around the bosses 120. While support members 130 and 135 are shown as thin extending members, any connection between bosses 120 and body 115 that permits flow through the example embodiment tiered tie plate may be used. As shown in FIG. 3, bosses 120 are positioned at a plurality of vertical (axial) displacements, called tiers. At a first vertical displacement is tier 150, at a second vertical displacement is tier 160, and at a third vertical displacement is tier 170. Each tier 150, 160, and 170 may contain a plurality of bosses 120, so that different bosses 120 occupy different tiers 150, 160, or 170. Axial support members 130 may permit bosses 120 to occupy different vertical displacements by extending diagonally in both a transverse and axial direction. For example, axial support members 130 may have a parallelogram cross-section, as shown in FIG. 3. Although three different tiers are shown in FIG. 3, example embodiment tiered tie plates may include more distinct tiers, limited only by the number of bosses 120, such that each boss 120 may have its own tier. Further, although FIG. 3 shows tiers 150, 160, and 170 each having bosses 120 continually and laterally joined in a square pattern by lateral and axial support members 130 and 135, tiers in example embodiment plates need not have all continuous and/or joined bosses 120. By locating bosses 120 at different tiers of vertical displacement, example embodiment tiered tie plates may receive and hold nuclear fuel at several different axial displacements relative to each other. Several different three-dimensional configurations of bosses 120 and fuel rods may be possible with example embodiment tiered tie plates, due to the ability of the bosses 120 to be placed at any combination of different transverse and axial positions. FIG. 4 is an illustration of another example embodiment tiered tie plate 200 that may be used as an upper tiered tie plate. As shown in FIG. 3, example embodiment tiered tie plate 200 may share several elements with previously described example embodiment tiered tie plate 100, whose redundant description is omitted. Tiered tie plate 200 may include bosses 220 joined by axial and lateral support members 230 and 235 similar to the description of FIG. 3. Also, bosses 220 may be at a plurality of different axial displacements, or tiers, 250, 260, or 270. However, example embodiment tiered tie plate 200 may lack any body or lower inlet, and fluid may flow through and around bosses 220 and support members 230 and 235. Example embodiment tiered tie plate 200 may include a handle 240 to facilitate handling and moving example embodiment fuel bundles including example embodiment tiered tie plate 200. FIG. 4 shows example embodiment tiered tie plate 200 having three tiers 250, 260, 270 with different axial displacements and a square lattice of bosses 220 at each tier. Tiers 250, 260, and 270 may correspond in relative axial displacement and boss position to tiers 150, 160, and 170 and their associated boss 120 position. In this way, as described below, example embodiment fuel bundles may use two example embodiment upper and lower tiered tie plates 200 and 100 with several fuel elements or rods of a single length, with each rod equally seating in bosses 120 and 220 of example embodiment tiered tie plates. Example embodiment tiered tie plates may be fabricated from a material that provides sufficient material strength to support fuel rods in different axial positions with bosses 120/220 and support members 130/135/230/235 and that substantially retains its physical characteristics in an operating nuclear reactor environment. For example, zirconium-aluminum alloys, stainless steel alloys, etc., may be used to fabricate example embodiment tiered tie plates. As shown in FIG. 5, example embodiment fuel assembly 300 may use example embodiment tiered tie plates 100 and 200 as lower and upper tie plates, respectively. Several fuel rods 310 may seated at both ends in example embodiment tiered tie plates 100 and 200, such that the tie plates 100 and 200 join the fuel rods 310 together and to the example embodiment assembly 300. Other fuel rod designs and conventional features, such as spacers and end plugs, may be used in conjunction with example embodiment fuel assembly 300. Because example embodiment tie plates 100 and 200 may have complementary tiers and bosses, fuel rods 310, while rigidly attached to fuel assembly 300, may be at different axial displacements corresponding to the different tiers of example embodiment tiered tie plates 100 and 200 as described above. Further, fuel rods 310 may all have a same length due to the mirrored axial displacement configuration between example embodiment tie plates 100 and 200. In this way, example embodiment fuel assembly 300 may have various three-dimensional configurations for fuel rod 310 starting and terminating points at either end of the assembly. Although example embodiment fuel bundles are shown with the same three-tiered configuration described with respect to example embodiment tiered tie plates, different tier and different corresponding rod configurations may be used in example embodiment bundle designs, depending on the thermo-hydraulic and nuclear properties of the bundle to be affected. Example embodiment fuel bundles may use less than two example embodiment tiered tie plates. Conventional flat tie plates may be used in example bundles through the use of rod shanks or multiple length fuel rods that account for the difference between a conventional flat tie plate and an example embodiment tiered tie plate. An example shank 500 is shown in FIG. 6. Example shanks 500 may be generally cylindrical so as to present a continuous exterior when joined with fuel rods in example embodiment fuel bundles. Alternatively, example shanks may have a variety of shapes depending on the fuel shape and desired flow characteristics around example shanks. Example shank 500 may have a receptive end 502 and an insertive end 501. Receptive end 502 may join through several known connection means, including a screw and threaded hole, to a fuel rod. Insertive end 501 may be seated in an example embodiment tiered tie plate or conventional tie plate in the same manner that a fuel rod would so seat. Example shank 500 may have a length 504 equal to a displacement between tiers and a body of example embodiment tiered tie plates. In this way, example shanks may attach to some of the fuel rods seated at an opposite end in example embodiment tiered tie plates and account for the differences in axial displacement and rod termination caused by the tiers. Conventional flat tie plates may then be equally seated against each example shank and fuel rod despite the different axial displacements of the rods. Example shank 500 may have a diameter 503 that presents a continuous outer boundary with any fuel rod it may be attached to. Alternatively, diameter 503 may be decreased to less than that of any fuel rod attached to example shank 500. In this way, example embodiment fuel bundles using example shanks 500 with smaller diameters may have decreased fluid coolant pressure drop and require less pumping head. Diameter 503 may be varied and shaped in other ways depending on the desired thermo-hydraulic characteristics of example embodiment fuel bundles. FIG. 7 shows an example embodiment fuel bundle 600 using one example embodiment tiered tie plate 100 as a lower tie plate and a conventional or otherwise flat upper tie plate 610. Fuel rods 620 may be positioned between the two tie plates 100 and 610. Because example embodiment tiered tie plate 100 may have varied vertical displacement of its bosses and corresponding starting position of lower fuel rods 620, fuel rods 620 may terminate at different vertical displacements. Example shanks 500 may compensate for these differing terminal vertical displacements in the fuel rods 620 and present a planar termination at flat tie plate 610. In this way, example embodiment fuel bundles may use fuel rods of a single length with example embodiment tiered tie plate as a lower tie plate while still presenting a conventional flat upper tie plate for compatibility with conventional core structures requiring a flat upper tie plate. Also, as shown in FIG. 7, diameters of example shanks 500 may be varied throughout the example bundle 600. FIG. 8 shows another example embodiment fuel bundle 700 using one example embodiment tiered tie plate 200 as an upper tie plate and a conventional or otherwise flat lower tie plate 710. Similarly to FIG. 7, example shanks 500 may compensate for the axial differences between example embodiment upper tiered tie plate and the lower flat tie plate. Example embodiment fuel bundles may thus possess fuel rods that originate and/or terminate at differing axial (or vertical) displacements. As such, fluid flow through example embodiment fuel bundles may not be subject to a dramatic pressure drop at example embodiment upper and lower tiered tie plates, because example embodiment tiered tie plates essentially present a larger flow cross-section. Further, example shanks of smaller diameters may decrease pressure drop by similarly increasing flow cross-section and hydraulic diameter. In this way, example embodiments may improve hydrodynamic flow properties through an operating nuclear core and reduce pumping energy consumed. Example embodiments thus being described, it will be appreciated by one skilled in the art that example embodiments may be varied through routine experimentation and without further inventive activity. For example, other fuel types, shapes, and configurations may be used in conjunction with example embodiment fuel bundles and tiered tie plates. Variations are not to be regarded as departure from the spirit and scope of the exemplary embodiments, and all such modifications as would be obvious to one skilled in the art are intended to be included within the scope of the following claims.
	description	This application claims priority to Japanese Patent Application No. JP 2005-194771, filed on Jul. 4, 2005 and also to JP 2006-111315 filed Apr. 13, 2006, the disclosures of which are incorporated herein by reference. The present invention relates generally to writing apparatus and methodology and, more particularly, to exposure techniques for use in variable-shaped electron beam lithography systems. In recent years, highly integrated semiconductor devices decrease in on-chip patterns with an increase in integration density of large-scale integrated (LSI) circuits. To form a desired circuit pattern on such semiconductor devices, a microlithographic exposure apparatus is used. A prior known approach to doing this is to employ a process having the steps of placing a chosen one of masks (e.g., reticles or photomasks) with many kinds of original or “master” image patterns being formed thereon, mounting on a stage structure a workpiece such as a semiconductor wafer having its exposure surface area, performing highly accurate position alignment of the master pattern relative to the wafer exposure area, and driving a light source to emit a beam of laser light, which falls onto the wafer so that the mask circuit pattern is transferred onto the exposure area of the target wafer. An example of such apparatus is a reduced image projecting exposure tool. The master pattern is imaged and formed on a fine-finished glass plate by resist processes and others. Usually a glass substrate is prepared which has its one-side surface with a thin film of chromium (Cr) being vapor-deposited. Then, a film of resist material is uniformly deposited on the substrate. An energy particle beam, such as electron beam or laser beam, is used to perform the sensitization of the resist material at selected surface portions thereof. After completion of known development, the Cr film is selectively etched to thereby write or “form” or “draw” the pattern. Microlithography processes that take on advances in miniaturization of semiconductor devices are unique in creation of patterns and thus are very important among semiconductor device fabrication processes. While optical lithography techniques have traditionally been employed in the manufacture of semiconductor devices as stated above, it is an urgent need to develop an advanced high-resolution exposure technique in view of the fact that leading-edge devices with extra-high integration densities, such as ultralarge-scale integration (ULSI) chips, are coming closer to the limit of resolution. Electron ray (electron beam) exposure techniques offer in nature superior resolution properties and, for this reason, are used for development and mass-production of leading-edge devices, such as dynamic random access memory (DRAM) chips and some of application-specific integrated circuit (ASIC) devices. The EB exposure is also used in combination with standard optical lithography processes to form master patterns for fabrication of such leading-edge ULSI devices. See FIG. 27, which shows some major components of a prior known variable-shaped electron beam exposure apparatus for explanation of an operation thereof. As shown herein, this EB exposure tool includes a couple of spaced-apart aperture plates 410 and 420. The upper aperture 410 has a rectangular opening 411, which shapes an electron beam leaving a charge particle source 430 and then falling onto aperture 410. The lower aperture 420 has a variable shaping opening 421 for shaping the electron beam 330 that has passed through the upper aperture opening 411 into a desired rectangular cross-section. After penetration of the aperture opening 411, the electron beam 330 is deflected by a deflector to pass through part of the lower aperture opening 421, resulting in irradiation onto a target workpiece which 340 is placed on a stage that is movable in a one direction, e.g., X-axis direction. In short, only a specific pattern with a rectangular cross-section capable of passing through both aperture openings 411 and 420 is irradiated or “shot” onto the workpiece surface so that the intended pattern is drawn in the aimed exposure region of workpiece 340 on the stage moving along the X direction. This approach to forming a desired shape by letting the beam penetrate both the aperture holes 411 and 421 is called the “variable shaping,” as disclosed, for example, in JP-A-2000-58424. Another EB exposure apparatus is found in JP-A-4-171714, which involves teachings as to the shot size and current density. The above-stated EB exposure apparatus is designed so that the current density and maximal shot size are subjected to value setting along the scale of the nodes of interest in semiconductor road map at the stage of manufacture thereof in such a way as to preclude decreases in beam resolution otherwise occurring due to “space charge” effects when shooing the beam with the maximum shot size. Thereafter, in the apparatus, such initially determined current density value is used to form the electron beam, which is shaped to have its shot size that is less than or equal to the first defined maximum shot size for execution of the pattern drawing on wafers or masks or else. Unfortunately, as the to-be-drawn pattern becomes finer and more complicated, the shot size in each shot event gets smaller, resulting in an increase in total number of beam shots. This poses a problem as to unwanted increase in time as taken to draw an entirety of the pattern. This leads to degradation of throughput. To shorten the drawing time of each shot (i.e., shot time), it is effective to increase the current density. However, a mere increase in current density results in an increase in degradation of beam resolution due to space-charge effects—i.e., defocusing of electron beam pattern images. Thus, the scheme for simply increasing the current density is hardly employable in practical applications. It is an object of a currently preferred form of the present invention to avoid the problems stated above to thereby provide a scheme for achieving maximally increased throughputs while at the same time suppressing degradation of beam resolution. In accordance with one aspect of this invention, there is provided a writing apparatus which includes a selector unit responsive to receipt of input data of a pattern to be written by shots of irradiation of an electron beam, configured to select a current density of the electron beam being shot and a maximal shot size thereof based on the input data of the pattern to be written, and a writing unit configured to create an electron beam with the current density selected by said selector unit, shape the created electron beam into a shot size less than or equal to said maximal shot size in units of the shots, and shoot the shaped electron beam onto a workpiece to thereby write said pattern. In accordance with another aspect of this invention, a writing method includes, analyzing a value of a writing time pursuant to a pattern data while using as variables a current density and a maximal shot size being in a relationship that a beam current value is less than or equal to a preset value, selecting, based on a result of said analyzing, a current density and a maximal shot size so as to be in a vicinity of a point of inflexion at which the writing time value changes in concavity, and shooting an electron beam onto a workpiece with the selected current density and a shot size less than or equal to said maximal shot size to thereby write thereon a pattern pursuant to said pattern data. In accordance with a further aspect of this invention, a writing apparatus includes a selector unit responsive to receipt of input data of a pattern to be written through more than two electron beam shots, configured to select a current density of an electron beam being shot and a maximal shot area thereof based on the inputted data of the pattern to be written, and a writing unit configured to form the electron beam with the current density as selected by said selector unit, shape the formed electron beam to have a shot area less than or equal to said maximal shot area, and shoot the shaped electron beam onto a workpiece to thereby write said pattern. In accordance with another further aspect of this invention, there is provided an apparatus for writing a prespecified pattern on a workpiece through more than two shots of an electron beam, includes means for variably shaping shot size of a shot, and means for varying a current density in accordance with each shot size so that a current value of a beam being shot onto the workpiece is less than or equal to a value as preset in each shot. Referring to FIG. 1, a variable-shaped electron beam (EB) microlithographic pattern writing/exposure apparatus 100 in accordance with a first embodiment of this invention is shown in schematic block diagram form. This EB lithography apparatus is generally made up of a scanning beam pattern writing unit 150 and a system control unit 160 as operatively associated therewith. The pattern writing or “imaging” unit 150 includes a tower-like housing 102 called the electron optical colum, a stage structure 105 that is movable in X and Y directions, an electron gun assembly 201, an illumination lens 202, an upper beam-shaping aperture plate 203 as will be referred to as “first aperture” hereinafter, a projection lens 204, a deflector 205, a lower beam-shaping aperture plate (second aperture) 206, an objective lens 207, a deflector 208, and a Faraday cup 209. The system controller 160 includes a draw data processing circuit 310 that functions as a data selector, a beam deflection control circuit 320, a digital-to-analog converter (DAC) 332, and an electron optics control circuit 342. Obviously the EB lithography tool 100 includes other known components, which are eliminated from the illustration for brevity purposes only. The electron gun 201 is driven to emit an electron beam 200, which is guided by the illumination lens 202 to illuminate an entirety of the first aperture 203 having a square opening or hole. The electron beam is shaped by aperture 203 to have a square-shaped cross-section. This shaped beam 200 that passed through aperture 203—namely, first aperture image beam—is then projected by the projection lens 204 onto the second aperture 206. A position of the first aperture image on second aperture 206 is controlled by the deflector 205 so that its beam shape and size are made variable. The resultant electron beam 200 of a second aperture image which passed through second aperture 206 is focus-adjusted by the objective lens 207 and deflected by deflector 208 to fall onto a workpiece 101 being presently placed on the movable X-Y stage 105 at a desired position thereon. The electron beam 200 irradiated in such case is shaped to have a rectangular illumination area on workpiece 101 along the shape of a pattern, with a shot size being less than or equal to a prespecified maximum shot size. As shown in FIG. 2, the workpiece 101 has an exposure surface on which a pattern is to be written. This surface is virtually subdivided into a plurality of strip-like beam-deflectable portions. When writing the pattern on workpiece 101, the XY stage 105 is driven by a known actuator (not shown) to move or slide continuously in the X direction so that the incoming electron beam 200 scans one stripe area along the length thereof. During the movement of XY stage 105 in the X direction, let the shot position of electron beam on workpiece 101 follow or “trace” the stage motion in a way synchronous therewith. This makes it possible to shorten a time taken to complete the pattern writing required. After having written a pattern in one stripe of workpiece 101, the XY stage 105 is driven by the actuator (not shown) to move in step along the Y direction and then move backward in X direction so that the incoming beam scans the next stripe for execution of pattern writing thereto. In other words, XY stage 105 alternately performs continuous forward/backward movements in X direction and movements in steps along Y direction, thereby permitting the beam 200 to scan respective exposure strips of workpiece 101 on XY stage 105 in a serpentine fashion. With this serpentine stage motion control, it is possible to minimize any possible idle time during movement of XY stage 200, thereby to increase or maximize the efficiency of stage motion-control operation. As shown in FIG. 3, a total time T as required for completion of the writing of a pattern is represented by Equation (1) which follows:T=Tc·Nshot, (1)where, Tc is the shot cycle, and Nshot is the total shot number. The shot cycle Tc is given as:Tc=Tset+Tshot, (2)where, Tset is a settling time. The shot time Tshot is given by:Tshot=DOSE/J, (3)where, “DOSE” is the dose of an electron beam hitting the workpiece 101, and J is the current density. The moving speed of XY stage 105 in X direction is calculated based on the writing time T, although in practical applications the writing time is somewhat increased with an addition of the XY stage's stepping times between stripes and also of other extra time components, such as an overhead time(s). As shown in FIG. 4, the pattern writing time T is finally represented as:T=(Tset+DOSE/J)Nshot (4)Note here that the settling time Tset is “automatically” determined depending on the specifications of apparatus used. The dose amount DOSE is determined by the performance of a resist material to be exposed. Thus, the writing time T may be lessened in value by finding an optimal value while letting the current density J and the total shot number Nshot be parameters or “variables.” The variable-shaped EB exposure apparatus 100 shown in FIG. 1 employs an electrolytic deflector(s) as the deflector 205 and/or deflector 208 for position control of the electron beam 200. Use of such electrolytic deflector(s) makes it possible to control the beam deflection amount by varying a voltage to be applied to the deflector 205 or 208. When varying this voltage, a certain length of time is needed to stabilize it to a preset potential level. This time is called the settling time Tset stated previously. In the variable-shaping EB exposure tool 100, the settling time Tset becomes different in value in accordance with the magnitude of the voltage to be varied, which corresponds to the change quantity of a beam position. For instance, several tens of us is required. The optimal settling time is verified at system setup events, so it is recommendable to set it as a parameter in the writing data processor circuit 310 shown in FIG. 1. The current density has certain relation to the beam shot size as will be described with reference to FIGS. 5A to 5C below. The current density J is given as a beam current value I per unit area. Suppose that an electron beam 200 with a square cross-section is irradiated onto the workpiece 101 while letting the maximum shot size be L1. If this is the case, as shown in FIG. 5A, a maximal shot area S is indicated by a square value of the maximum shot size L1. Note here that the beam defocusing or degradation of beam resolution occurrable due to space charge effects as described in the introductory part of the description is dependent on the beam current value I. Also note that in each shot the shot area is maximized in the event that the beam is shaped into a square under the setting of the maximum shot size L1. Taking these facts into consideration, it is preferable to determine the beam current value I in such a way as to prevent unwanted increase of such beam fogging. An example is that an acceptable beam resolution was obtained by setting the maximum shot size L1 to 2.5 μm and letting the current density J1 range from 10 to 20 A/cm2. This encourages us to believe that setting the beam current value I1 to 62.5 to 125 nanoamperes (nA) or less makes it possible to obtain the allowable beam resolution. As previously stated, the shot size in each shot event decreases as a writing pattern decreases in minimum feature size and increases in complexity. While the apparatus has its performances with an ability to shape the electron beam 200 up to the maximum shot size of the value L1, the shot size in practical applications is limited to L2 that is one-half (½) of L1 in maximum as shown in FIG. 5B. In the example of FIG. 5B, L2 is half of L1, so the maximum shot area is actually defined by a square value of L2, that is, it becomes one-forth (¼) in area. Thus, the current value I2 of a beam passing through it is ¼ of I1. As the writing time T is shortened by increasing the current density J as indicated in Equation (4), it is effective to enlarge the current density J. No appreciable degradation of beam resolution is found even when increasing the current density J so that a beam current value I3 becomes equal to I1 (I3=I1), as shown in FIG. 5C. More specifically, in the example of FIG. 5C, it is possible to set J3=4J1. Exemplary plots of the beam shot density versus technology node for sample patterns A to C are graphically shown in FIG. 6. As shown herein, the beam shot density was measured for the patterns A-C while letting the maximum shot size L1 be set at L1. The measurement result reveals that the shot density rapidly increases at a certain point “nodel” as the pattern to be written becomes finer and complicated. This shot density increase indicates that the total shot number Nshot rises up in value. See FIG. 7, which shows plots of shot density versus maximum shot size for the sample patterns A to C. As apparent from viewing this graph, when lessening the maximum shot size from the value used in FIG. 6 for the patterns A-C, the shot density exhibits no appreciable value changes within a certain range. In other words, in this situation, the total shot number Nshot does not increase. Hence, by simply decreasing the maximum shot size and increasing the current density J by a degree equivalent to such decrease in maximum shot size, it is possible to shorten the writing time T without having to deteriorate the beam resolution required. However, as shown in FIG. 7, further lessening the maximum shot size results in a rapid increase in shot density, i.e., an appreciable increase in total shot number Nshot. Obviously, increasing the total shot number Nshot requires a likewise increase in settling time Tset. In this case, even when enlarging the current density J by the degree corresponding to the decrease in maximum shot size, the total shot number Nshot increases undesirably, so it is no longer possible to simply shorten the writing time T as shown in Equation (4). See next FIG. 8 which is a three-dimensional (3D) graph showing an exemplary shot-size distribution in a given pattern. In this graph, the longitudinal size of an actually employed shot size is plotted along x axis in case the maximum shot size has its predefined longitudinal and lateral size dimensions. Its y axis indicates the lateral size of the actual shot size, whereas z axis shows the shot number. Assume that for sample pattern A, the electron beam is shaped to have a rectangle whose actual shot size is set so that a long-side length is long the x-axis and a shot-side length is y as shown in FIG. 7. It can be seen that the shot size dimensions are distributed with the x and y values being less than or equal to predetermined values respectively. Another sample pattern B's shot size distribution is shown in FIG. 9. In a similar way to the graph of FIG. 8, the longitudinal and lateral sizes of an actual shot size and shot number are plotted along the x, y and z axes respectively when the longitudinal and lateral length values of maximum shot size are each set to a predefined value. As apparent from FIG. 9, the long-side length (x) and short-side length (y) of the pattern B with a rectangular shot size are distributed so that these fall within certain ranges having different values from those of FIG. 8, respectively. Exemplary plots of maximum shot size versus total shot number for sample patterns A-C are graphically shown in FIG. 10. For the patterns A-C, the maximum shot size of each is plotted along the longitudinal axis, whereas an increase ratio of the total shot number is along the lateral axis. When looking at the maximum shot size at a certain point, the pattern A is such that the total shot number is approximately 2.5 in increase rate. For the pattern B, its total shot number is about 4.5. Thus, for some patterns, even when attempts are made to minimize the maximum shot size L and increase the current density J by a degree equivalent to such shot-size minimization, this do not always result in cut-down of the writing time T. This can be said because the total shot number Nshot increases accordingly. In other words, it has been found by the inventors as named herein that a curve-change or “inflexion” point at which the writing time T turns into increase from decrease must exist in the process of lessening the maximum shot size L (alternatively, enlarging the current density J). A curve indicating variation of the maximum shot size L versus the current density J is shown in FIG. 11. As shown herein, when letting the maximum shot size L and current density J have a specific relation therebetween so that the beam current value stays at its preset value, the maximum shot size L decreases with an increase in current density J. A plot of the total shot number Nshot versus maximum shot size L is shown in FIG. 12. As shown in this graph, the total shot number Nshot decreases with a decrease in maximum shot size L, and thereafter does not increase and is almost “saturated” irrespective of any further changes in maximum shot size L. A plot of the pattern writing time T versus the current density J is shown in FIG. 13. As apparent from this graph, it is possible, by selecting appropriate values of the maximum shot size L that causes the total shot number Nshot to be kept constant as shown in FIG. 12 and the current density J of FIG. 11 at such size L, to adjust the writing time T so that it is at the inflexion point, i.e., a minimal writing time Tmin, in the process of increasing the current density J—at the point Tmin the decreasing writing time T changes to increase in value. Accordingly the current density J at this writing-time inflexion point Tmin is determined as the optimum current density Jbest, which ensures accomplishment of the maximum or “best” throughput. Then, find the maximum shot size L at this optimum current density Jbest, which becomes an optimal maximum shot size Lbest. Although in the graph of FIG. 13 a change curve of the writing time T is plotted with the current density J being as a variable therefor, the maximum shot size L may alternatively be used as such variable. This can be said because the current density J and maximum shot size L are in the relationship which permits the beam current value to stay constant in value, so similar results are obtainable by use of any one of them as the variable. Turning to FIG. 14, an exemplary internal configuration of the writing data processor circuit 310 of FIG. 1 is partly shown in block diagram form. As shown herein, the writing data processor 310 is generally made up of an analyzer unit 314, a data selector unit 316 and a data setter unit 318. The analyzer 314 includes a graphic/figure dividing unit 312, current value calculator 352, maximum shot size setter 354, writing time calculator 356, total shot number calculator 358, and current density setter 364. As shown in FIG. 15, a system procedure for variable-shaped electron beam pattern writing of this embodiment starts with step S1402, which receives an input writing data indicative of a pattern to be written or “drawn” or “imaged” onto a target workpiece, such as a wafer or else. Then, the procedure goes to step S1404 which causes the current density setter 364 to set up an initial or “default” value as the current density J. This default current density value setting step is part of an analysis process. Then, the routine proceeds to step S1406 which permits the maximum shot size setter 354 of FIG. 14 to set a default value K for the maximum shot size L. This maximum shot size setting step is part of the analysis process. Next, go to step S1408, which causes the current value calculator 352 to determine through computation a beam current value I, thereby letting I=J×L2. This step is part of the analysis process. At step S1410, the writing data processor circuit 310 compares the calculated beam current value I to a preset maximal beam current value Imax to determine whether the former is less than the latter. This current value judgment step is part of the analysis process. If the value I is less than the preset value Imax, then return to the step S1406 via step S1412 which follows. At step S1412, the maximum shot size setter 354 causes its built-in adder (not shown) to add a prespecified value—here, one (1)—to the current value K to provide an incremented value K+1, which is then used as an “updated” candidate for the maximum shot size L at step S1406. This addition step is part of the analysis process. After re-execution of the processing at step S1406, the routine again proceeds to the decision step S1410 via the current value calculation step S1408. The addition value at step S1412 should not exclusively be “1” and may be any other values as far as an ability remains to permit the value of maximum shot size L to vary as the variable while achieving the required analyzability of the drawing time T. The subroutine of from the size setup step S1406 to decision step S1410 will be repeated until the beam current value I becomes equal to the predefined beam current value Imax. At step S1406, attempts are made to redo the setting of the maximum shot size L. By using in combination the resultant current density J and maximum shot size L which are obtained after having affirmed that the beam current value I is equal to the preset beam current value Imax, it is possible to prevent degradation of the beam resolution. If NO at step S1410 then the routine goes next to step S1414, which causes the graphic divider 312 to subdivide the input writing data into graphical portions each having a shot size in a way pursuant to the maximum shot size L being presently set up. This graphic division step is also part of the analysis process. These resultant graphic portions—say, shot figures—are formed and laid out so that the length of one side edge of a rectangle is less than or equal to the maximum shot size L. Any graphic figures to be shaped in accordance with the shape of an aperture are changeable. For example, shot figures are formable into squares, rectangles or right-angled triangles. For squares and rectangles, let them have side edges each being less than or equal to the maximum shot size L, followed by disposing them in an appropriate layout. As for right triangles, form and dispose them so that the longer one of two adjacent sides crossing together at right angles is not greater than the maximum shot size L. In step S1416 which is part of the analysis process, the total shot number calculator 358 is rendered operative to count up an exact number of the shot figures that are divided at the previous step S1414, thereby determining through calculation the total shot number needed to write the pattern indicated by the input writing data. Then, at step S1418 that is part of the analysis process, the writing time calculator 356 determines through computation the length of a time period as required for writing the input writing data pattern. This writing time, T, has a value appropriate for the input draw pattern data, which is calculable by using the presently defined current density J and the total shot number Nshot as calculated at step S1416 in accordance with Equation (4). Usually the settling time Tset and the dose amount DOSE are preset at adequate values. At step S1420 which is part of the analysis process, the writing data processor circuit 310 determines whether the calculated value of the writing time T is identical to the minimum value Tmin, that is, whether the time value is at the inflexion point whereat the concavity changes as shown in FIG. 13. If NO at step S1420, then proceed to step S1422. If YES then go to step S1424. At step S1422 in the analysis process, the current density setter 364 uses its built-in adder to add a predetermined value, e.g., 1, to the current density J. The addition value at this step should not exclusively be “1” and may be other values as far as an ability remains to permit the value of current density J to vary as the variable while achieving the required analyzability of the writing time T. Then, return to the current density setup step S1404 which again performs the setting of current density J, followed by execution of the above-stated steps S1404 to S1420 until T=Tmin is verified at step S1420. Varying the values of the current density J and maximum shot size L in this way makes it possible to finally obtain an “ideal” value of the total shot number Nshot at the value of maximum shot size L as changed while being less than or equal to the preset beam current value Imax. This enables obtainment of the intended value of writing time T from such variable current density J and total shot number Nshot values. More specifically, it is possible to analyze the value of writing time T suitable in compliance with the input pattern data while using as parameters the current density J and maximum shot size L, which are in a relationship that forces the beam current value to stay less than or equal to the preset beam current value Imax. Thereafter, the routine enters a selection processing stage. More precisely, at step S1424, the selector unit 316 of writing data processor circuit 310 operates based on the analysis results to perform value selection of the current density J and maximum shot size L which cause the writing time T to stay at the inflexion point shown in FIG. 13. At this writing-time inflexion point, the writing time has its minimum value Tmin. To maximize throughputs, the selector 316 is preferably designed to select the current density J and maximum shot size L which cause the writing time T to be at the minimum value Tmin, although such point is not the only one. Similar results are obtainable by selection of other sets of values of the current density J and maximum shot size L which are available when the value of writing time T falls within a specific range including the inflexion point as its center point—i.e., the T value is a vicinity of the minimum value Tmin. In this case also, superior advantages than the prior art are achievable. For example, a range of 10% plus of the minimum value Tmin. to the minimum value Tmin. is desirable as the specific range. Especially, a range of 5% of plus of the minimum value Tmin. to the minimum value Tmin. is more desirable as the specific range. After completion of the value selection, the value setter 318 sets up the selected current density J and maximum shot size L. Then, the routine goes to step S1426 which writes a pattern on the workpiece 101 shown in FIG. 1. More specifically, the electron gun 210 in pattern writing unit 150 emits an electron beam 200 with the selected current density J, which is then shaped on a per-shot basis to have a specific shot size that is less than or equal to the maximum shot size L. The shaped electron beam 200 is then irradiated or “shot” onto the workpiece 101 to thereby depict and form thereon a desired pattern corresponding to the input writing data. Once the maximum shot size L is set up, this is notified to the deflection controller 320. In responding thereto, this controller 320 sets up a voltage appropriate for operation control of the deflector 205. This voltage is applied to deflector 205 through digital-to-analog conversion by DAC 332, thereby enabling deflection of the electron beam 200 in a way such that this beam is shaped by the second aperture 206 to have a specific shot size which causes the size of a graphic being shot onto workpiece 101 to be less than or equal to the maximum shot size L. Additionally, upon setting of the current density J, this is sent forth to the electron optics control circuit 342 shown in FIG. 1. This circuit 342 controls the electron gun 201 for adjustment of an emission current and filament temperature to thereby cause the current density J to become equal to the setup value. Alternatively or in addition thereto, the electron optics controller 342 adjusts the focusing of electron beam 200 at the illumination lens 202 to provide control so that the current density J is at the set value. Whether the current density J is at the specified value is verifiable by irradiation of electron beam onto Faraday cup 209. Although in FIG. 15 one specific case is described which varies the value of current density J with the maximum shot size L as a variable, the current density J may alternatively be varied as the variable with respect to the maximum shot size L. By varying the current density J and maximum shot size L in a way pursuant to a pattern to be written in the way stated supra, it is possible to obtain the best possible throughput. Although it has been stated that the value change of the maximum shot size L is achieved by control of the deflector 205 on a software or hardware basis so that the electron beam 200 varies in deflection position, similar results are obtainable by replacement of either one or both of the first and second apertures 203 and 206 of FIG. 1. A procedure for the aperture replacement will be discussed with reference to FIG. 16. Supposing that several first apertures are replaceably disposed in an aperture cassette (not shown). One of such first apertures, 213, is replaced by another first aperture 223, thereby to change an opening 214 to another opening 224 that is different in size therefrom. For example, the opening 214 is a rectangle, and the opening 224 replaced is a rectangle that is smaller in each side than opening 214 with the center being the same in position as that of the former, resulting in a decrease in deflection amount of the electron beam 200 falling onto a variable shaping opening 217 of second aperture 216 in the event of beam shaping by use of the maximum shot size. Such deflection reduction causes the settling time to become shorter, thereby enabling shortening of the writing time. Alternatively in FIG. 16, several second apertures are replaceably disposed in a cassette (not shown). One of such second apertures, 216, is replaced with another second aperture 226, thereby changing a variable shaping opening 217 to another variable shaping opening 227 that is different in size therefrom. For instance, the opening 217 is a rectangle, and the opening 227 replaced is a rectangle that is smaller in each side than opening 217 with the center being the same in position as that of it, resulting in a decrease in deflection amount when shooting electron beam 200 to the variable shaping opening 227 of aperture 226 in the event of beam shaping with the maximum shot size. Such deflection reduction causes the settling time to become shorter, thereby enabling shortening of the drawing time. Both the first and second apertures 213 and 216 are replaceable at a time by the first and second apertures 223 and 226, respectively. By lessening both the beam-shaping openings while letting respective aperture centers stay at the same position, it is possible to further decrease the deflection amount upon shooting of the electron beam 200. This results in a further decrease in settling time, thereby enabling further shortening of the writing time T. Although in the example of FIG. 16 the apertures themselves are replaced with others, an alternative approach is employable, which provides an aperture plate having more than two beam-shaping holes at different locations while permitting one of them to be used interchangeably. More specifically, as shown in FIG. 17, a first aperture 233 and second aperture 236 are provided, either one or both of which is/are arranged to have rectangular holes that are different in size from each other for achieving the changeability of maximum shot size L. For example, the first aperture 233 has a large rectangular hole 214 and small hole 224. Similarly second aperture 236 has a large rectangular hole 217 and small hole 227. For each aperture plate, changing between the holes makes it possible to vary the maximum shot size L. This hole change is accomplishable by use of an aperture driver (not shown) or alternatively by control of the deflection position of electron beam 200 for changing its irradiation position. The variable shaping holes of first and second apertures should not exclusively be limited to rectangles and may be other shapes as far as these enable formation of a desired shot shape. While the first embodiment stated above is arranged to perform value setup by selecting in combination the current density J and maximum shot size L which minimize the writing time T, similar results are obtainable by selection of a maximal shot area S in place of the maximum shot size L. A variable-shaped electron beam pattern writing/exposure method and an apparatus for use therewith in accordance with a second embodiment of the invention are arranged to incorporate this principle, although detailed explanations thereof are eliminated herein as these are understandable from the description of the first embodiment while reading it by changing the term “maximum shot size L” to “maximum shot area S.” The current density J is given by a beam current value I per unit area. Additionally, as previously stated, the degree of beam defocusing or increasing of beam blur due to space charge effects is variable depending on the beam current value I. Letting the maximum beam current value Imax without any increasing of beam blur be Imax, the maximum shot area S—i.e., a shot area that is maximized while preventing degradation of the space charge effect-increasing beam blur—is given as S=Imax/J in each shot under an assumption that the current density J is kept constant. In light of this, the beam defocus degradation is avoidable by setup of such specific maximum shot area S (=Imax/J) and then graphic division while letting it be less than or equal thereto, resulting in the per-shot beam current value I being less than or equal to the preset value Imax. This enables accomplishment of increased beam resolution acceptable for practical applications. A technique for writing a pattern by defining the maximum shot size L is as follows. As shown in FIG. 18, suppose that the maximum shot size L is set to L1. Assume that the pattern of interest is a line pattern segment having its width less than the value L1, e.g., half of L1. In this case, the pattern is divided into portions (shot figures) having a lateral size along the “x” direction of L1 and a longitudinal size in “y” direction of L1/2. These portions are indicated by hatching in FIG. 18. Obviously, shooting an electron beam thereto needs execution of two separate shot processes. Another example is shown in FIG. 19, which shows a similar line pattern that is subjected to pattern writing with the definition of a maximal shot area S. This area S is defined equal to the area S1 of a square segment with its each side length being equal to the maximum shot size L1 of FIG. 18. Suppose that the line pattern has a width which is less than the value L1, e.g., one-half of it. When looking at its portion indicated by hatching in FIG. 19, this is divided into shot figures of different shapes each having the same area S1. Hence, shooting a beam onto the hatched portion can be done by a mere single shot. Accordingly, defining the maximum shot area S in place of the maximum shot size L makes it possible to reduce the total shot number. A curve of the maximum shot area S versus current density J is shown in FIG. 20. As shown in this graph, when setting the maximum shot area S and the current density J so that these are in a certain relationship which causes the beam current value I to stay at a preset value, the larger the current density J, the smaller the maximum shot area S. A plot of total shot number Nshot versus maximum shot area S is shown in FIG. 21. As shown, when the maximum shot area S is lessened in value, the total shot number Nshot generally is kept almost unchanged until the area value S reaches a specific point. Curves indicating writing time versus current density characteristics are plotted in FIG. 22. A specific value of the maximum shot area S is selected which falls within the “constant value” range of the total shot number Nshot shown in FIG. 21. Also select the current density J shown in FIG. 20 at the selected value of maximum shot area S. These value settings make it possible to adjust the writing time T to the inflexion point (i.e., minimum writing time Tmin) at which the decreasing writing time T turns to increase in the process of increasing the current density J. Thus, the current density value J at such inflexion point becomes an optimal current density Jbest2 which ensures achievement of the best possible throughput. The value of maximum shot area S at this optimum current density Jbest2 is an optimal shot area Sbest. As the maximum shot area S is defined as an alternative to the maximum shot size L, the resulting total shot number Nshot decrease, so the writing-time inflexion point is shiftable downward as shown in FIG. 22. Thus it is possible by defining the maximum shot area S to provide a minimal writing time Tmin2 that is made shorter than the minimum writing time Tmin1 in the case of defining the maximum shot size L. To maximize throughputs, the selector 316 is preferably designed to select the current density J and the maximum shot area S which cause the writing time T to be at the minimum value Tmin2, although such point is not the only one. Similar results are obtainable by selection of other sets of values of the current density J and the maximum shot area S which are available when the value of writing time T falls within a specific range including the inflexion point as its center point—i.e., the T value is a vicinity of the minimum value Tmin2. In this case also, superior advantages than the prior art are achievable. For example, a range of 10% plus of the minimum value Tmin2. to the minimum value T min2. is desirable as the specific range. Especially, a range of 5% of plus of the minimum value Tmin2. to the minimum value Tmin2. is more desirable as the specific range. Although in FIG. 22 a change of the writing time T is shown with the current density J being as a variable, such change in the writing time T may alternatively be shown with the maximum shot area S as the variable. Since the current density J and maximum shot area S are set in the specific relation which forces the beam current value to stay at a preset value, similar results are obtained by use of any one of them. In the second embodiment, the value of the current density J of an electron beam to be shot in accordance with input pattern data and the value of maximum shot area S are specifically selected so that the writing time T is at its minimum value or at approximate values thereof. After this value selection, the pattern writing unit 150 creates an electron beam with the current density thus selected, which beam is shaped to have a shot area less than or equal to the maximum shot area S and is then shot onto a target workpiece so that a pattern of the input data is written or “drawn” thereon. With such an arrangement, it is possible to increase the throughput while at the same time suppressing degradation of the beam resolution. While in the above embodiments the current density J is set so that the beam current I flowing in a region of either the squared value of maximum shot size L or the maximum shot area S becomes equal to a specific value which avoids degradation of the beam resolution, every shot area does not always become such expected area when consideration is given for respective shots. In view of this, it is also preferable to vary the current density J on a per-shot basis. One desirable approach is to vary the current density J pursuant to each shot size or each shot area in a way such that the value I of a beam current being shot onto a workpiece 101 is in maximal proximity to a preset beam current value Imax without the risk of beam resolution degradation while letting the former be less than the latter. Varying the current density J in deference to each shot size or area makes it possible to increase the current density of a shot having its area less than either the squared value of the maximum shot size or the maximum shot area. This results in appearance of a shot capable of shortening the shot time, which is contributed to the shortening of the writing time. An electron beam photolithography apparatus incorporating this principle in accordance with a third embodiment of the invention is similar in hardware configuration to that of the embodiments stated previously, so its detailed explanation will be eliminated herein. Each of the embodiments stated above is such that the beam current value Imax is set in advance to a unique value irrespective of the kinds of patterns to be written. However, the invention should not exclusively be limited thereto. Several patterns being written on a mask or wafer or else include a pattern without a need for high accuracy. Respective arrangements in a fourth embodiment may be similar to those in the previous embodiments, so explanations thereof will be omitted here. See FIG. 23. This diagram shows a plan view of an exemplary workpiece 101 to be subjected to the pattern writing. As shown, a pattern to be written on the workpiece 101 has a central region 10 which is required to have a high accuracy level and is under the requirement for precision compensation and a peripheral region 20 that is relatively low in precision. For example, in case the workpiece 101 is a mask used to fabricate highly integrated semiconductor circuitry on wafers, higher precision is required for a region in which is formed a semiconductor circuit pattern(s). On the contrary, an identification code pattern that permits users to identify this mask is free from strict precision requirements. Examples of the precision-free pattern are a bar code 22 shown in FIG. 23, numerals, ID number, date and serial number (S/N). Barcode 22 is to be optically read by a barcode reader and thus may be roughly sized as far as users can visually recognize it. The others, such as the numerals, ID number, date and serial number (S/N), are merely visually recognized by users, so these may be sized to permit users to do so. For these patterns with such rough sizes, high accuracy is not required, so it is possible to enlarge the beam current value Imax. For example, let it be two times greater than the beam current value Imax used for the high accuracy-required region 10. Consequently, in the fourth embodiment, the beam current value Imax is made variable in compliance with a to-be-written region or pattern. Varying it depending on the precision required makes it possible to shorten the writing time. As shown in a flowchart of FIG. 24, a system procedure of the fourth embodiment starts with step S1402, which causes the writing data processor circuit 310 of FIG. 1 to receive an input pattern writing data in a similar way to that shown in FIG. 15. The procedure goes next to step S1403, which permits the writing data processor 310 to set up a default beam current value. At this step, the maximum beam current value Imax is set up in accordance with a writing region, pattern kind and pattern accuracy level or else, unlike the previous embodiments which are arranged so that the maximum beam current value Imax is preset to a unique value irrespective of the kind of a pattern to be written. As shown in FIG. 25, a predetermined one-to-one correspondence relation between a pattern identifier indicative of pattern names or the like and the beam current value Imax to be used therefor is prepared in advance in the form of a “look-up” table 30. In a column of this table, several patterns to be written in the high-accuracy-required area 10 and low-precision area 20 shown in FIG. 23 are distinguished in name from each other. Then, the writing data processor circuit 310 extracts such pattern identifiers from the input pattern data for setup of its corresponding value of maximum beam current value Imax. Then, the routine proceeds to step S1404, at which the current density setter 364 sets up a default value for the current density J. The following steps S1406 through S1426 are similar to those shown in FIG. 15. By setup of the beam current value Imax in accordance with the pattern to be written, the selector 316 selects the value of each shot size or shot area and the current density J for co-use therewith in such away that the electron beam is nor greater in its current value than the beam current value Imax thus determined. In other words, while varying the maximum beam current value Imax to be determined depending on pattern data, the selector 316 sets up the current density J and the maximum shot size L or area S so that the electron beam current is less than or equal to the value Imax. In this way the selector 316 performs the value setting in units of patterns to be written on the target workpiece 101. With such an arrangement, it is possible to further shorten the drawing time T. The pattern name versus Imax correspondence table 30 shown in FIG. 25 may be replaced by a table 30 shown in FIG. 26. This table indicates in a one-to-one correspondence way the relation between writing areas and values of the beam current value Imax. With use of this table, it is possible for the writing data processor 310 to determine the maximum beam current value Imax directly from an area or region to be subjected to the pattern writing. While letting the exposure surface of workpiece 101 be virtually divided into a plurality of pattern-writing regions, the beam current value Imax to be set per region is made variable in the way stated above. Then, the setter 316 selects the current density J and the maximum shot size L or area S so that an electron beam being shot per region has its current value less than or equal to the value thus selected. With this arrangement also, it is possible to lessen the writing time. Although in FIG. 23 one specific example was discussed for division into the high-accuracy pattern area 10 and low-accuracy area 20, the invention is not limited thereto. For example, it is also preferable to set the beam current value Imax per strip region while making variable the beam current value Imax to be determined per strip region. This is effective because some strip regions can be different in pattern accuracy level from each other. Using the “Imax variable” scheme for selecting the current density J and the maximum shot size L or area S so that the beam has its current value less than or equal to the per-region determined beam current value, it is possible to shorten the writing time. As apparent from the foregoing discussions, the pattern writing apparatus in accordance with the first embodiment is characterized by including a selector unit operable to receive input data of a pattern to be written by shots of an electron beam and then select based on the input pattern data a current density of the electron beam being shot and a maximal shot size thereof, and a pattern writing unit for creating the electron beam with the current density as selected by the selector unit and for shaping the formed electron beam to have its shot size that is less than or equal to the maximal shot size in units of shots, which beam is then irradiated or “shot” onto a workpiece to thereby write the pattern required. It has been stated that the invention provides a technique for appropriately determining in a way pursuant to pattern data the beam current density and the maximum shot size which optimize the throughput while suppressing degradation of the beam resolution. The best possible throughput is accomplishable by selecting the “best” combination of such current density and maximum shot size values in compliance with the writing pattern data and then using these values to write a pattern(s). Another important feature lies in that the selector unit is arranged to select the current density and maximum shot size in a way such that even upon inputting of different pattern data, an electron beam being shot onto a workpiece which is shaped to less than or equal to the maximum shot size has its current value less than or equal to the preset value. By selecting the specific current density and maximum shot size values so that the beam current value is not greater than the preset value, it is possible to avoid unwanted occurrence of degradation of the beam resolution (i.e. beam defocusing or blur) otherwise occurring due to space charge effects. In addition, the above-stated pattern writing method of the first embodiment is featured by including the steps of analyzing the value of a writing time adequate for pattern data while using as variables a current density and maximal shot size which are in the relation that a beam current is less than or equal to a preset value, selecting based on an analysis result the values of the current density and maximal shot size which cause the writing time to fall within a range in which its value changes in rate of change, and shooting an electron beam onto a workpiece while letting its shot size be less than or equal to the selected current density and maximal shot size to thereby write a pattern as indicated by the pattern data. Performing the analysis process makes it possible to exactly specify how the writing time varies in value. By the selection of the current density and maximal shot size which permit the value of writing time is within a limited range including the inflexion point at which the concavity changes, i.e., the decreasing draw-time value changes to increase, it is possible to improve the throughput in the pattern writing process. The pattern writing apparatus of the second embodiment stated supra is featured by including a selector unit responsive to receipt of input data of a pattern to be formed by shots of an electron beam for selecting a current density of the electron beam being shot and a maximal shot area thereof, and a pattern writing unit for creating or “forming” an electron beam with the current density selected by the selector unit and for shaping the created electron beam into a shot area less than or equal to the maximal shot area in units of shots and then irradiating the shaped electron beam onto a workpiece to thereby write the pattern. Using the maximum shot area in place of the maximum shot size makes it possible to reduce a total number of beam shots. The pattern writing apparatus of the third embodiment stated supra is arranged to use a plurality of shot-size variably shaped electron beams to write a specified pattern or patterns on a workpiece, which apparatus is featured in that the beam being shot onto the workpiece has a current value which is made variable in current density in a way pursuant to each shot size so that the beam current value in each shot is less than or equal to a preset value. Varying the current density on a per-shot basis makes it possible to maximize the value of the current density while at the same time avoiding occurrence of any appreciable space charge effects. This enables further shortening or cut-down of a shot time required. It has been stated that the embodiments above are capable of using the best possible combination of current density and maximal shot size or area which is chosen to maximally increase the throughput while suppressing degradation of the beam resolution. This in turn makes it possible to achieve better throughputs then ever before. The parts or components as expressed by “units” in the above description may also be configured by computer-executable software programs or any combinations with hardware or firmware or both. Such software programs are storable or recordable on recording media, such as magnetic disk devices, magnetic tape recorders, floppy diskettes (FDDs) or read-only memories (ROMs) or equivalents thereto. Additionally the input data and/or output data stated previously may be stored in storage devices, such as a register or memory or else. Each arithmetic processing may be done by use of an adder or multiplier or the like. It must be noted that in case the maximum shot size L in the first embodiment is replaced by the maximum shot area S, the current value calculator 352 of FIG. 1 operates to compute the beam current value I=J×S, rather than I=J×L2. While this invention has been particularly shown and described with reference to specific embodiments, the invention should not exclusively be limited thereto. Although explanations are omitted of apparatus components and control schemes which are deemed unnecessary for discussion of the principles of the invention, these are realizable by using known appropriate ones on a case-by-case basis. For example, regarding the configuration of the system controller for control of the variable-shaped electron beam (EB) writing apparatus 100, its detailed explanation is eliminated as such is achievable by adequate use of controller arrangements ad libitum. Any other pattern writing/imaging systems and methods which incorporate the principles of the invention and which are modifiable through design changes by skilled persons are included in the coverage of the invention. 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 described and illustrated herein. Various modification and alterations may be made without departing from the spirit and scope of the general inventive concept as defined by the appended claims and their equivalents.
058754067	summary	BACKGROUND OF THE INVENTION The present invention relates to the field of treatment of radioactive waste, particularly oils and solvents, in particular from nuclear power plants and in military research centers, and has for its object a process for reduction of waste suitable for this purpose. The invention also has for its object a device for practicing this process. DESCRIPTION OF THE RELATED ART At present, the elimination of contaminated oils or solvents poses a great problem in nuclear power plants and in military research centers. Active oils used in these power plants are of the mineral or synthetic type. There is meant by contaminated oil an oil containing radioactive substances. Until now, the known processes for the decontamination of contaminated oils and solvents are centrifugation, decantation, filtration over rare earth and bacterial destruction. To this end, there is known from DE-A 3 522 126, a process for the centrifugation of radioactive wastes and drying the solid residues. Also, JP-A-63 204 198 discloses a process of centrifugation and filtration of radioactive effluents containing an oily portion. There is also known, from WO-A-89 08316, a process for treatment of effluents of low radioactivity by preheating to eliminate volatile solvents, then by special centrifugation followed by filtration. Finally, the publications INSTITUTE OF ELECTRICAL ENGINEERS, STEVENAGE, GB Inspec. No. 3106590 SIMIELE G A et al. "Radioactive decontamination of waste oil by filtration, centrifugation, and chelation" and NUCLEAR AND CHEMICAL WASTE MANAGEMENT, 1987, UK, Vol. 7, No. 3-4, page(s) 257-263, ISSN 0191-815X disclose the decontamination of oily wastes from pressurized water reactors. This process suggests prefiltration which removes large particles, this prefiltration being followed by centrifugation adapted to eliminate water. The oily residue is then treated by chelating agents such as E.T.D.A., and the emulsion is again subjected to centrifugation. However, the use of chelating agents gives rise to serious problem of elimination of wastes from nuclear power plants. Moreover, these processes are often very long or incomplete, in particular in the case of centrifugation and filtration. Thus, these processes, which are suggested for the treatment of aqueous effluents, permit lowering the radioactivity of the oils but not eliminating or reaching levels equal to natural radioactivity. As a result, these oils, as well as the contaminated solvents, are generally simply stored at the site, awaiting a technical solution, whereby the active wastes encumber the sites and constitute a permanent pollution danger. SUMMARY OF THE INVENTION The invention has for its object to overcome these drawbacks. It thus has for its object a process for the reduction of radioactive wastes, particularly oils and solvents, particularly from nuclear power plants and military research centers, characterized in that it consists essentially in dumping said wastes into a reservoir, in which they are subjected to continuous agitation, preheating them, carrying out a chemical precipitation treatment, passing the mixture into a centrifuge, carrying out electrostatic or conventional filtration and then testing the level of radioactivity, the treatment being performed continuously until the desired level of decontamination is reached. The invention also has for its object a device for practicing this process, essentially constituted by a contaminated waste reservoir connected by means of forward and return conduits and by a circulation pump, to a filtration element, characterized in that it is provided moreover, between the contaminated waste reservoir and the filtration element, with a centrifuge adapted to carry out separation of water that may be contained in the wastes and to carry out the separation of the large contaminated particles, so as to limit the consumption of the filtering elements, in that the reservoir is provided with a mixture and a heating means and in that the filtration element is constituted by at least one electrostatic collector or a conventional filter, the assembly of the elements being controlled and monitored by means of a control panel.
	summary
	abstract	In some embodiments, a method includes receiving, in a processor, information indicative of (i) a treatment plan defining planned treatment beams, (ii) a patient volume relative to a reference, (iii) ideal intersections of the planned treatment beams with the patient volume at the time the patient is to be treated, (iv) any constraints that prevent achievement of the recommended repositioning using only the patient support, (v) an allowable change to a gantry position from a planned value and an allowable change to a collimator position from a planned value; defining, in the processor, a plurality of alternatives based at least in part on the information indicative of any constraints of the patient support and the information indicative of allowable movement of the gantry and collimator, each alternative defining a modified patient support position and modified beams, each modified beam being based at least in part on a respective one of the planned treatment beams, the change to the position of the gantry for the respective planned treatment beam and the change to the position of the collimator for the respective planned treatment beam; determining, in the processor, for each modified beam of each alternative, an intersection of the patient volume and the modified beam, with the patient volume positioned on the patient support and the patient support having the modified patient support position defined by the alternative; and defining, in the processor, for each alternative, a measure of difference between the ideal intersections and the intersections for the modified beams of the alternative.
	claims	1. A system to control characteristics of a proton beam emitted from a nozzle of a proton treatment system, comprising:a plurality of beam modifying members to define a characteristic of an emitted proton beam; anda clamping member configured to be mounted to the nozzle, the clamping member having a plurality of receiving portions disposed on the clamping member to respectively receive one of the plurality of beam modifying members therein, the plurality of receiving portions being configured to selectively and interchangeably receive multiple ones of the plurality of beam modifying members within the clamping member when the clamping member is mounted to the nozzle. 2. The system of claim 1, wherein the multiple ones of the beam modifying members are configured as a first set of plates having a first size, and the receiving portions include a plurality of slots spaced apart from one another on opposing surfaces of the clamping member to respectively receive opposing ends of each first plate. 3. The system of claim 2, wherein the clamping member includes one or more detector units to detect the presence of one or more of the beam modifying members within the clamping member. 4. The system of claim 1, wherein at least one of the beam modifying members includes a second clamping member having at least one receiving portion smaller than the receiving portions of the first clamping member to respectively receive one or more other beam modifying members therein. 5. The system of claim 4, wherein the one or more other beam modifying members are configured as a second set of plates smaller than the first set of plates, and the receiving portions of the second clamping member include a plurality of slots spaced apart from one another on opposing surfaces of the second clamping member to receive opposing ends of each second plate. 6. The system of claim 4, wherein the second clamping member includes one or more detector units to detect the presence of at least one beam modifying member within the second clamping member. 7. The system of claim 1, wherein the clamping member is configured to be mounted adjacent to a proton delivery nozzle aperture of the proton treatment system and downstream from the nozzle aperture. 8. The system of claim 1, wherein the plurality of beam modifying members include one or more of a place-holder plate, aperture plate, collimator plate, compensator plate, degrader plate, or combinations thereof, and wherein the place-holder plate is configured to not significantly modify the proton beam, the aperture plate is configured to define a cross sectional area of the proton beam, the collimator plate is configured to align the proton beam, the compensator plate is configured to affect a Bragg peak distance of the proton beam, and the degrader plate is configured to reduce an intensity of the proton beam. 9. The system of claim 1, wherein the multiple ones of the beam modifying members are stacked together side by side within the clamping member. 10. The system of claim 1 wherein the plurality of beam modifying members include annular shielding. 11. The system of claim 1, further comprising:a compensator integrated with at least one beam modifying member. 12. The system of claim 3, further comprising:an output unit in communication with the one or more detector units to output presence information of at least one of the beam modifying members within the clamping member. 13. The system of claim 6, further comprising:an output unit in communication with the one or more detector units to output presence information of at least one of the beam modifying members within the clamping member. 14. The system of claim 1, wherein each of the plurality of beam modifying members include a handle portion to facilitate gripping of the beam modifying members by an operator of the proton treatment system. 15. The system of claim 5, wherein the proton treatment system includes a snout, and the plurality of beam modifying units can be interchanged in the clamping members without removing the snout from the proton treatment system. 16. The system of claim 1, further comprising a locking member to locate and secure the multiple ones of the beam modifying members within the clamping member relative to the nozzle, the locking member configured to cooperate with each of the plurality of receiving portions of the clamping member. 17. A method of controlling characteristics of a proton beam emitted from a nozzle of a proton treatment system, comprising:mounting a clamping member to a nozzle of a proton treatment system, the clamping member configured to be mounted to the nozzle, the clamping member having a plurality of slotted receiving portions disposed on the clamping member to respectively receive one of the plurality of beam modifying members therein, the plurality of receiving portions being configured to selectively and interchangeably receive multiple ones of the plurality of beam modifying members within the clamping member when the clamping member is mounted to the nozzle; andinterchangeably sliding one or more of the plurality of beam modifying members into a respective receiving portion, the one or more beam modifying members being selected to define a characteristic of a proton beam emitted from the nozzle,wherein the one or more beam modifying members are configured as a first set of plates having a first size, and the receiving portions include a plurality of slots spaced apart from one another on opposing surfaces of the clamping member to receive opposing ends of each first plate, andwherein at least one of the beam modifying members includes a second clamping member having at least one receiving portion smaller than the receiving portions of the first clamping member to respectively receive one or more other beam modifying members therein. 18. The method of claim 17, further comprising:detecting the presence of at least one beam modifying member within the clamping members; andoutputting presence information of the least one beam modifying unit to an output unit of the proton treatment system.
	abstract	The invention relates to a radiation protection screen having two main sides. According to the invention, the radiation protection screen includes a framework which stiffens the screens and on either sides holds a pane; one of the panels being radioprotective. The invention further relates to a system including at least two screens which are or can be joined together.
048225592	description	The fuel element as shown in FIG. 1 conforms in all details--except for the ribs on the inside surface of the cladding tube--with a standard type LWR fuel element. The sintered pellets 3 of the encapsuled column are thus centreless ground to very precise diameter dimensions, generally within a tolerance range of .+-.0.010 mm, the pellet diameter being of the order of 8-12 mm depending on the actual design. The pellet heights can vary from 7-15 mm and the end surfaces of the pellets are often made concave, "dished", to minimize the axial thermal expansion of the pellet column during approach to power. The full pellet column height is quite considerable, usually of the order of about 4 meters. The cladding tube 1 is made of a zirconium alloy, generally Zircaloy, as are also the two end plugs 4 and 8. The end plugs are provided with axially protruding pins 6 and 10, respectively. Said pins 6, 10 have for a purpose to maintain the fuel element in a fixed position in the reactor proper in a conventional manner. The tube 1 fits around the pellet column with a certain cold assembly annular (radial) gap clearance, e.g. within the range of 0.05 mm to 0.15 mm, and is preferably about 0.10 mm. In this disclosure, the dimensions of this annular gap between cladding and pellets, reference is always made to the nominal cold assembly clearance between the circular nominal inner periphery of the cladding tube, disregarding the ribs, and the outer cylinder surface of the fuel pellets. These precise small gap dimensions are required for minimizing the temperature drop across the gap during operation and simultaneously avoiding a more severe mechanical interaction, which represents a potential source of performance problems in the conventional LWR fuel element. During operation the cold assembly gap closes up more or less, depending on actual heat rating and extent of burn up. Often circumferential ridges form on the cladding surface at the positions of the pellet-pellet interfaces, due to mechanical interactions and distortion--"hour-glassing"--of the individual pellets. In the severe case cladding fractures develop at these ridges in conventional prior art fuel elements. To improve the annular gap heat transfer an inert gas is added to the void volume of the fuel element. This inert gas is normally prepressurized for the purpose of preventing an early creep down of the cladding tube onto the pellet column during operation, and also for minimizing the effect of contamination by fission gases on the gas thermal conductivity. Again referring to FIG. 1, a plenum space (14) is arranged for accomodating released fission gases and excess inert gas. Also a spring coil 12 acting on the pellet column is located in this plenum space. This spring coil 12 has mainly for its purpose to keep the fuel pellets in place during transportation and handling of the fuel element. Such plenum spaces (14) can also be arranged at both ends of the fuel element. As an inert gas there may be used a noble gas such as helium or argon or a mixture thereof. FIG. 2 shows a detail of the adjoining area between two fuel pellets 3 and the adjacent section of the cladding tube 1. FIG. 2 shows the chamfer provided at the adjoining corners of fuel pellets 3. The lateral depth of chamfer 5 indicated by distance (a) is at least about 10% of the pellet radius, whereas the axial depth (b) of chamfer 5 preferentially is about a quarter of the lateral depth (a). The chamfering of pellets 3 form together with the surrounding cladding tube 1 a toroidal space 7a. FIGS. 3A and 3B also show a detail cross section of the fuel element of the invention illustrating also dish cavities 7b, and these figures also indicate the crack pattern of the ceramic fuel arising after a period of operation of the nuclear reactor. The pellet-pellet interface volumes 7a and 7b occupy together a certain fraction of the void volume of the fuel element. The invention calls for the use of a special shape of the fuel pellets. The fuel pellets should preferably have a length/diameter (L/D) ratio near 1 (one), such as from about 0.8 to about 1.2, in order to avoid the "hour-glass" distortion illustrated in FIG. 7 of the previously mentioned British patent, where the distorted pellets will become locked and actually nearly close the axial channels, as actually intended, by using a quite large L/D ratio. Moreover, according to experimental findings and operational experience the pellet end faces need to be properly chamfered because of the impact of the compressive axial forces on the form and integrity of the pellets under severe PCI conditions. Positioning of the axial loading to the circumferential part of the endface must be avoided, and by chamfering the loading position inwards a distance of at least 10% of the pellet radius as illustrated on FIG. 2 this will be attained. In doing so also a non-permissible "chipping" of the pellet corners into tiny fragments will be avoided and the "hour-glass" distortion of the pellets minimized. The interfacial gas communication system of the rib cladding comes into useful operation already at low burnup levels, at which state the fuel pellet/clad gap has been occupied to a substantial degree by pellet fragments which have relocated radially outward in "soft mechanical contact" with the cladding ribs. By "soft mechanical contact" is meant that the pellet fragments are spaced apart loosely and do not exert any pressure on the cladding. Under this condition the helium filler gas distributes itself between the voids within the pellet stack and the exterior channel volume in the pellet/clad interface. The helium gas in the voluminous top plenum 14 will now continuously communicate by axial diffusion and convection with the gas contained in the whole fuel section below and thus dilute any released FG. However, the rate of gas migration will be slower in the cracked pellet bodies than in the pellet/clad interface location, because of the more tortuous communication paths in the interior of the pellets. Any FG in the toroidal spaces or torus chambers 7a in the chamfer positions of the pellets will, of course, continuously undergo a faster rate of dilution with helium from the plenum than the FG in the interior voids of the pellet column. In fact each torus chamber will act as a local helium rich plenum to the nearby couple of cracked fuel pellets. In a conventional type fuel rod, however, where the relocated pellet fragments exhibit an almost perfect geometric fitting to the clad bore the rate of axial gas communication with the top plenum along the narrow pellet/clad interface remains very restricted. In this case the torus chambers and, in particular the voids within the cracked pellets will dilute their FG contents only slowly. This means that the residence time of any released FG will be long and results in a degradation of the rod thermal conduction and a consequent rise in fuel temperature. The beneficial effect of the aforementioned channel system as regards axial gas communication with the helium in the top plenum is retained at any burnup level where a release of fission gases as a consequence of high power operation or occasional power transients occurs. A restoration of the gas thermal conductivity thus takes place fairly fast and continuously along the whole fuel rod length by the axial gas transport mechanism. However, at high or extended burnup levels where the accumulated amount of fission gases in the pellet structure has become considerably higher, the consequences of an uncontrolled transient fission gas release may approach or exceed certain life performance limitations in standard type LWR fuel, as for example the permissible internal gas pressure at end-of-life or in case of a LOCA event. Under such conditions the original fuel element design using ribbed cladding needs to be fundamentally modified as described below in order to perform adequately. The fuel design according to the invention provides for effective control, specifically under mechanically closed pellet/clad conditions, the thermal feed back effect and its consequences in the form of induced high fuel temperature and high internal FG pressure. These effects are known to limit the performance of standard type LWR fuel rods following power ramps that cause the fuel temperature to exceed the critical temperature of significant FGR. According to the invention the integral volumes of the two sets of channel systems involved, i.e., the axial clad bore channels 9 and the pellet torus channels 7a and any dish volumes 7b are balanced in the relative terms so that the former integral volume represents only a fraction of the later integral volume. By this feature of the invention a "burst" (large and sudden) release of fission gases at one section of the fuel rod following a power ramp will cause the internal gas pressure at the position of FGR to increase for a moment. The local gas overpressure will then decline rapidly (within a few seconds) until equilibration because of the fast axial gas flow to the plenum through the narrow but open and straight axial clad bore channels. No thermal feed back effects can be initiated during this short time, but a certain modest temperature rise will follow at the position of FGR, which causes only a local (non-propagating) feed back effect thereat. In a conventional fuel element design the same amount of released FG are forced to penetrate, under a slow pressure drop for some minutes, the tightly compressed cracked pellet column through long and tortuous flow paths. During that long flow time the fuel temperature will readily rise to cause a propagating thermal feed back and more gas will be released. (When the pressure equilibrates for a longer period of time than that of the fuel rod thermal time constant, i.e., 4-8 seconds, a temperature rise takes place). In the fuel element according to the invention the speed of the axial pressure driven gas flow will be magnified by the expanding helium gas contained in the pellet-pellet interface volume due to the raised temperature induced by the power ramp. The flow speed will become greater the smaller the fractional axial channel volume is relative the torus volume including any dish spaces. The fraction of the volume contained in the axial channels around one pellet can according to experience readily be kept to a lower value than 1. The axial channels as illustrated in FIGS. 4A-C and further described below will have still lower fractional values. This means that the ratio of axial to pellet-pellet interface volumes should be lower than about 1 (one) and preferably lower than about 0.5 to be as effective as feasible. In the conventional fuel design the contained helium gas of the torus chambers and any dish spaces have only a limited effect in terms of shortening the flow time until pressure equilibration, because of the inherent very constrained gas flow in that case. After complete termination of the gas flow period a diffusional mixing of the fission gases with the helium gas follows. In the conventional fuel designs local mixing at each pellet elevation takes place only slowly between the torus plena and the internal and external voids due to the tortuous diffusion paths. In addition, essentially no axial mixing takes place within a reasonable time. Hence the thermal feedback mechanism may easily initiate, or continue to operate, if initiated already during the gas flow period. In the fuel element design according to the invention essentially no pressure driven penetration of fission gases into the pellet column takes place outside the axial position of FGR. This is due to the fast pressure equilibration. During the flow phase a considerable dilution of the FG with He of the torus plena actually takes place. The diffusional He/FG mixing starts effectively after the short gas flow period. During the diffusional axial mixing, which may take some hours for completion, the thermal conductivity over the pellet/clad interface will be kept fairly high because of the high interfacial contact pressure. The good heat transfer will according to experience keep the temperature rise in the fuel pellets below the critical temperature of FGR or, at least minimize a possible temperature overshoot. In the irradition experiments underlying this invention it has been noted that the integral thermal conduction of the fuel rods remains fairly unaffected by the power rise and hence the temperature increase kept quite modest, perhaps about 50.degree. C. Also, it has been noted that after about 12 hrs holding at the ramp terminal level the fuel temperature had decreased with about the same magnitude due to the efficient axial diffusional mixing. As mentioned earlier no such discernable mixing takes place within several days in fuel rods of conventional design under mechanically closed gap conditions. Thus, it is evident that the fuel element design according to the invention performs superior to conventional designs as regards pellet/clad heat transfer, axial gas mixing and thermal feedback effects following transient fission gas release. The prior art design will not at all perform effectively and reliably in these respects. The potential excessive FGR under load follow/power cycling operation is a very important issue from a fuel performance point-of-view in particular, at high and extended fuel burnup conditions. The risk of entering into a thermal feed back situation with consequent high FG pressure and high fuel temperature is obvious. The fuel element according to the invention will perform excellently under these operating conditions. The ease by which gas flows along the axial channels during the power changes will generate continuous delivery of gas to and removal of gas from the torus chambers so that an effective He/FG mixing is maintained and FG:s are continuously removed from the pellet interior voids. The load follow/power cycling seems actually to be the most beneficial mode of operation when using the fuel element according to the invention. For conventional fuel the opposite seems true. In this context it should be pointed out that the channel communication system according to the invention with the same advantage can be combined with various PCI failure remedies, like the Zr-liner concept, the graphite interlayer concept, "soft" pellet designs, etc. Another related objective of the invention is the design means available to adjust and control the fuel thermal response time (or "time constant"). As mentioned earlier smaller diameter rods, e.g., 9.times.9 instead of 8.times.8 fuel rod arrays in the assemblies for BWRs suffers of a too short response time in certain types of applications. The invention offers a unique feasibility to adjust and control the fuel response time for any type of water reactor fuel elements by simply combining a certain cold assembly gap size with a certain filler gas composition of a given pressure. Thus, in order to reduce the thermal conductivity of the filler, the helium can be admixed with a heavier inert gas, such as argon or neon, preferably argon. As for example in the irradiation tests performed the clad rib design was measured to increase the fuel thermal time constant with 0.8 seconds as compared to the conventional design which typically measured 5.0 seconds. In both cases 3 bar pure He and a fabrication annular gap size of 0.075 mm were used. Using a 33% Ar 67% He fill gas mixture under otherwise similar conditions the time constant was measured to be only 0.2 seconds longer, i.e., 6.0 seconds. Should a wider radial gap, say 0.10 mm have been used, the same Ar/He mixture would have produced a still larger time constant. These results indicate that by increasing the fuel burnup, under which some contamination of the He with FG (Kr/Xe) and some closing of the pellet/cladding gap occurs, the fuel thermal time constant will not change significantly. In conventional fuel designs an increase of the time constant is generally expected, however, starting from a lower value (occasionally too low in 9.times.9 type rods). At higher burnups, however, a very irregular behaviour is noted. One important new feature is the redesigned forms and proportions of the two channel systems provided for internal gas communication, as described earlier, resulting in lower local FG concentration in various operating situations. The lower FG concentration achieved will have as an effect to minimize the availability and chemical activity of the aggressive corrodents, like iodine and cadmium, thus depressing the PCI/SCC initiation correspondingly. Also the higher thermal conduction achieved by minimizing the FGR causes the fuel temperature to stay low and hence restricts the rate of FGR and maintains the PCI induced clad stresses low. Only by using the modification in design as represented by the invention to control the FGR a substantial and reliable improvement in PCI/SCC failure resistance can be obtained. In the fuel element according to the invention the use of a graphite interlayer between the pellets and the clad bore is additionally very efficient as a PCI/SCC remedy. On applying a 5 micron coating on the fuel pellets no failure at exceedingly large over-power ramping hs been obtained. This behaviour is probably due to an inhibiting chemical effect of the graphite via the gas atmosphere inside the fuel rod. A protective surface layer may have been formed at the clad bore surface inbetween the contacting points where the cladding is being stressed far beyond critical PCI/SCC levels, but remains unaffected by the nearby fuel pellets. In the conventional fuel design using graphite interlayers the PCI/SCC cracks are seen to initiate at clad surface areas, where the pellets have been "scrubbing" the cladding. Additional new features of significance to the PCI/SCC phenomenon relate to the detailed configuration of the ribbed (or undulated) cladding bore surface. In the conventional art, such as the techniques described in the above-identified British patent, the design is associated with certain disadvantages among which the following may be mentioned. The flat rib face configuration as of the prior art referred to above may cause splitting of the ceramic pellets at the position of contact under severe mechanical pressure and thus calls for an impractical multitude of such flat ribs around the circumference in order to be avoided. The splitting takes the form of cracks originating at the interaction point. This effect is highly undesirable because it impairs the pellet/cladding heat transfer process and may also produce tiny pellet fragments that enter into the pellet/cladding gap, where they can act as crack initiation points for PCI/SCC. A wider geometrically matching contact area is hence requested. The heat transfer through the pellet/rib contact sites has proven to be relatively efficient already when using flat ribs. According to the invention significant improvement will be obtained by introducing ribs of an inward concave form. The protruding faces will then enter into contact with the pellet column under both "soft" and "hard" interfacial pressure. Characteristically the ribs will be fewer around the bore than by using flat ribs for the same total pellet/cladding contact area. Hence also the average gap size (and contained channel volume) becomes smaller. Accordingly, the invention provides for a nuclear fuel element wherein each of the ribs provided on the inside of the cladding tube in a cross-section has an interior contour of which at least a part thereof lies inside a circle segment formed between a circle circumscribing the deepest points of channels 9 and a cord connecting adjacent points, see FIG. 4A, deepest points 11. Thus, at least part of each rib lies radially seen outside a corresponding cord to the said circle through said points. Said part of the rib as seen in cross-section has preferably a circular shape approximately matching that of the juxtaposed fuel pellet. Said part of the rib suitably has circumferential extension which is about one fifth to one half of the distance between the centre lines of two adjacent ribs. According to a preferred embodiment of the invention said ribs are evenly distributed over the inside surface of the cladding tube so as to form an inner contour of the cladding tube conforming to a regular polygon. The corners of such polygon may be rounded (FIG. 4C) and the intermediate part deviates from a straight line. For further illustration examples of embodiments of rib configurations are shown in FIGS. 4A, 4B and 4C. These figures show enlarged fractional sections of cladding tubes 1, the insides of which are provided with longitudinally extending ribs of different shapes. In the embodiment of FIG. 4A the longitudinally extending rib 9a has a concave circular shape, the radius of curvature, .sup.r Rib, being greater than the pellet radius. In the embodiment according to FIG. 4B only a fraction of the rib 9b has a concave circular shape, the radius of curvature (.sup.r Rib) of which is equal to the pellet radius, .sup.r Pellet. This circular part of rib 9b is about one third of the distance between the centres of adjacent ribs. In the embodiment of FIG. 4C rib 9c also has a central part of a concave cylindrical shape, but in this embodiment the corners corresponding to the polygon formed inside the cladding tube 1 by the ribs are rounded instead of fairly sharp. The optional configurations of FIGS. 4A, B and C are specific to the fabrication process being applied. The ribs according to FIGS. 4A and B can be produced by a conventional tube reducing technique and those of FIG. 4C e.g. a plug drawing method. The rib contour of FIG. 4C produces a more favorable channel form with respect to axial gas flow under closed gap conditions as it gives less gas flow resistance. The flow resistance is roughly inversly proportional to the third power of the gap width.
	description	FIG. 1 is a sectional view, with parts cut away, of a boiling water nuclear reactor pressure vessel (RPV) 10. RPV 10 has a generally cylindrical shape and is closed at one end by a bottom head 12 and at its other end by a removable top head 14. A side wall 16 extends from bottom head 12 to top head 14. A cylindrically shaped core shroud 20 surrounds a reactor core 22. 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 shroud 20 and side wall 16. A pump deck 30, which has a ring shape, extends between shroud support 24 and RPV side wall 16. Pump deck 30 includes a plurality of circular openings 32, with each opening housing a jet pump assembly 34. Jet pump assemblies 34 are circumferentially distributed around core shroud 20. Heat is generated within core 22, which includes fuel bundles 36 of fissionable material. Water circulated up through core 22 is at least partially converted to steam. Steam separators 38 separates steam from water, which is recirculated. Residual water is removed from the steam by steam dryers 40. The steam exits RPV 10 through a steam outlet 42 near vessel top head 14. The amount of heat generated in core 22 is regulated by inserting and withdrawing control rods 44 of neutron absorbing material, such as for example, hafnium. To the extent that control rod 44 is inserted into fuel bundle 36, it absorbs neutrons that would otherwise be available to promote the chain reaction which generates heat in core 22. Control rod guide tubes 46 maintain the vertical motion of control rods 44 during insertion and withdrawal. Control rod drives 48 effect the insertion and withdrawal of control rods 44. Control rod drives 48 extend through bottom head 12. Fuel bundles 36 are aligned by a core plate 50 located at the base of core 22. A top guide 52 aligns fuel bundles 36 as they are lowered into core 22. Core plate 50 and top guide 52 are supported by core shroud 20. FIG. 2 shows a front view of a shroud repair apparatus 60, in accordance with an exemplary embodiment of the present invention, mounted on shroud 20 of RPV 10. FIG. 3 shows a side view of shroud repair apparatus 60. Referring to FIGS. 2 and 3, shroud repair apparatus 60 includes an upper stabilizer assembly 62, a lower stabilizer assembly 64, and a tie rod 66 extending between upper and lower stabilizer assemblies 62 and 64. Shroud repair apparatus 60 is attached to shroud 20 to provide positive positioning of all segments of shroud 20 and fuel bundles 36. In one exemplary embodiment, four shroud repair apparatus 60 are attached to shroud 20 and are circumferentially distributed around shroud 20. In other embodiments, more than four or less than four shroud repair apparatus can be used to repair shroud 20 and overcome the problems associated with shroud weld stress corrosion cracking. Shroud 20 includes a shroud head flange 68, an upper shroud section 70, a top guide support 72, mid shroud sections 74, 76, and 78, a core plate support 80, and a lower shroud section 82. Circumferential welds 84, 86, 88, 90, 92, 94, and 96 couple the shroud elements together. A circumferential weld 98 attaches lower shroud section to shroud support 24. Welds 84, 86, 88, 90, 92, 94, 96, and 98 are sometimes referred to as welds H1, H2, H3, H4, H5, H6A, H6B, and H7 respectively. It has been generally observed that the shroud welds nearest to reactor core 22, for example welds 86, 88, 90, 92, and 94, are more likely to experience stress corrosion cracking. Shroud repair apparatus 60 is effective in providing positive positioning of all segments of shroud 20 affected by failures in welds 86, 88, 90, 92, and/or 94. Referring also to FIGS. 4, 5, and 6, upper stabilizer assembly 62 includes an upper stabilizer block 100 and an upper stabilizer wedge 102 slidably coupled to upper stabilizer block 100. Upper stabilizer block 100 is configured to couple to shroud lugs 104 located circumferentially around shroud head flange 68. Particularly, upper stabilizer block 100 includes a first portion 106 and a second portion 108 extending from a first side 110 of first portion 106. A second side 112 of upper stabilizer block first portion 106 is configured to engage shroud 20. Specifically, second side 112 is configured to engage shroud head flange 68, upper shroud section 70, and top guide support 72. Upper stabilizer second portion 108 is configured to engage wedge 102. Particularly, second portion is tapered to provide a wedge engagement surface 114. Upper stabilizer first portion 106 includes a slotted opening 116 sized to receive a pair of shroud lugs 104. Additionally, slotted opening 116 accommodates the subsequent installation of an existing shroud T-bolt 117, which is used to secure shroud head 26 to shroud 20. First portion 106 also includes a bore 118 extending from a first end 120 to slotted opening 116. Bore 118 is sized to receive tie rod 66. Upper stabilizer wedge 102 includes a tapered first side 122 and an opposing second side 124. First side 122 includes a channel 126 sized to receive wedge engagement surface 114 of upper stabilizer block second portion 108. Channel 126 extends from a first end 128 of wedge 102 at least partially towards a second end 130 of wedge 102. Upper stabilizer assembly 62 further includes a jack bolt 132 extending through a jack bolt opening 134 in upper stabilizer wedge 102 and threadedly engaging a jack bolt opening 136 in upper stabilizer block second portion 108. Upper stabilizer wedge jack bolt opening 134 extends from second end 130 of wedge 102 into channel 126. Upper stabilizer wedge 102 also includes a ratchet lock spring 138 configured to engage jack bolt 132 to maintain the tightness of jack bolt 132. Channel 126 maintains alignment of upper stabilizer wedge 102 and permits vertical position adjustment of wedge 102 by jack bolt 132. This adjustment along wedge tapered side 122 is used to install wedge 102 with a specified tight fit between upper stabilizer block 100 and pressure vessel side wall 16 while accommodating for any variations in the width of annulus 28. Upper stabilizer wedge 102 further includes an integral leaf spring portion 140 formed by a slot 142 in wedge 102. Leaf spring portion 140 is configured to engage side wall 16 of reactor pressure vessel 10. Leaf spring portion 140 provides flexibility for tightening jack bolt 132 at assembly and absorbing operating variations in the width of annulus 28, while also limiting radial and friction interaction loads for various reactor operating conditions. Referring to FIGS. 7, 8, and 9, lower stabilizer assembly 64 includes a stabilizer block 144 and a lower stabilizer wedge 146 slidably coupled to lower stabilizer block 144. Lower stabilizer block 144 is configured to engage shroud 20. Particularly, lower stabilizer block 144 includes a first side 148, configured to engage shroud 20, and a tapered second side 150 configured to engage lower stabilizer wedge 146. A lip portion 152 extends from first side 148 of lower stabilizer block 144. Lip portion 152 is configured to engage a ledge 154 formed by core plate support 80. First side 148 is configured to engage mid shroud sections 76 and 78. Lower stabilizer block 144 also includes a threaded opening 156 sized to receive tie rod 66. Lower stabilizer wedge 146 includes a tapered first side 158 and a second side 160. First side 158 includes a channel 162 sized to receive lower stabilizer block 144. Channel 162 extends from a first end 164 of wedge 146 at least partially towards a second end 166 of wedge 146. Lower stabilizer assembly 64 further includes a jack bolt 168 extending through a jack bolt opening 170 in lower stabilizer wedge 146 and threadedly engaging a jack bolt opening 172 in lower stabilizer block 144. Lower stabilizer wedge jack bolt opening 170 extends from second end 166 of wedge 146 into channel 162. Lower stabilizer wedge 146 also includes a ratchet lock spring 176 configured to engage jack bolt 168 to maintain the tightness of jack bolt 168. A horizontal stabilizing spring 178 is attached to second side 160 of wedge 146. Horizontal stabilizing spring 178 and second side 160 of wedge 146 is configured to engage side wall 16 of reactor pressure vessel 10. Horizontal stabilizing spring 178 maintains the orientation of lower stabilizer assembly 64 square to RPV side wall 16. Stabilizing spring 178 is sized to provide the same flexibility, preload, and clearance as integral leaf spring 140, described above, provides in upper stabilizer wedge 102. Referring again to FIGS. 2 and 3, tie rod 66 is threaded at each end. A first end 180 threadedly engages threaded tie rod opening 156 in lower stabilizer block 144. A second end 182 is received by upper stabilizer block bore 118 and is secured by a tie rod nut 184. Tie rod nut 184 reacts the tie rod load against upper stabilizer block 100. In one embodiment, tie rod 66 is fabricated from Nixe2x80x94Crxe2x80x94Fe alloy X-750 steel. Tie rod 66 preload increases at operating temperatures due to the differential expansion between X-750 steel tie rod 66 and stainless steel shroud 20. With an X-750 tie rod 66, more thermal differential contraction of tie rod 66 is produced than needed for the desired operating preload. To compensate, a belleville spring washer 186 is positioned between tie rod nut 184 and the upper stabilizer block 100. Spring washer 186 deflects only slightly with the low mechanical installation preload, for example, 5000 pounds (2268 kg), but compresses additionally to seat flat against upper stabilizer block 100 under full thermal preload, for example 40,000 pounds (18,144 kg). A limit stop 188 is attached near second end 182 of tie rod 66. Limit stop 188 includes shear pins 190 and 192 which fit mating holes 194 and 196 in upper stabilizer block first portion 106. Shear pins 190 and 192 provide a torque restraint for tie rod 66 when tightening tie rod nut 184 as well as a pinned anti-vibration connection to support tie rod 66 during operation. Limit stop 188 limits possible horizontal displacement of shroud section 74 if welds 88 and 90 fail. In known shroud repair apparatus using tie rods, it is usually necessary to provide a mid support to increase the tie rod natural frequency to be sufficiently higher than the vortex shedding frequency due to annulus cross flow to avoid flow induced vibration (FIV) of the tie rod. Tie rod 66 includes a plurality of longitudinal grooves 198 spaced around the periphery of tie rod 66 to reduce the magnitude and frequency of the alternating flow forces, and thus eliminate the need for a mid support. Grooves 198 have the affect on the coolant cross flow stream of dividing cylindrical tie rod 66 into a number of smaller objects which reduces the vortex shedding frequency to a value which is a safe margin below the natural vibration frequency of tie rod 66. As a result, resonant excitation of tie rod 66 does not occur. Grooves 198 can be regularly or irregularly spaced. FIG. 10 shows grooves 198 irregularly spaced to interfere with cumulative resonant interaction of the flow vortices. Grooves 198 extend longitudinally along a central portion 200 of tie rod 66. Because the orientation of grooves 198 is parallel to tie rod 66 axial loading, grooves 198 produce no structural stress concentration in tie rod 66. In another embodiment, as shown in FIGS. 11 and 12, grooves 198 are replaced by a sleeve 202 attached to tie rod 66. Longitudinal fins 204 project radially from sleeve 202 Sleeve 202 is positioned on tie rod 66 adjacent a suction inlet 206 of jet pump assembly 34, where cross flow velocity is high enough to cause concern for FIV of tie rod 66. The length of sleeve 202 is less than tie rod 66 because the anti FIV attribute that sleeve 202 imparts to tie rod 66 is needed only over a relatively short length of tie rod 66 which is in annulus 28 flow region near jet pump suction inlet 206. Sleeve 202 inhibits FIV by two means. First, the Strouhal number for the finned shape of sleeve 202 is lower than for a cylinder. Second, projecting fins 204 increase the characteristic diameter of tie rod 66 without addition of significant mass or machining over the remaining length of tie rod 66. Both of these effects directly reduce the vortex shedding frequency which provides a margin separating the excitation frequency from the higher natural frequency of tie rod 66. The above described shroud repair apparatus 60 is quickly and easily installed in reactor pressure vessel 10 because it does not require any installation machining of existing reactor components. Lower stabilizer assembly 64 and tie rod 66 are pre-assembled with tie rod 66 threaded into lower stabilizer block 144. Tie rod 66 and lower stabilizer block 144 are then lowered into position in annulus 28 with lower stabilizer block 144 engaging protruding core plate support ledge 154. Lower stabilizer wedge 146 is then lowered into place on lower stabilizer block 144 and adjusted by tightening jack bolt 168. Ratchet lock spring 184 prevents jack bolt 168 from loosening. Upper stabilizer assembly 62 is lowered into position in annulus 28 between shroud 20 and reactor pressure vessel outer wall 16, engaging tie rod 66 through bore 118 in upper stabilizer block 100. Slotted lug opening 116 at the top of upper stabilizer block 100 is then engaged onto a shroud lug pair 104. Belleville spring washer 186 and tie rod nut 184 is then lowered in place and tightened to tie rod 66 which causes lower stabilizer block 144 to seat against the bottom surface of shroud core plate support ledge 154. A ratchet lock spring (not shown) prevents tie rod nut 184 from loosening during reactor operation. Upper stabilizer wedge 102 is then lowered into position and adjusted by tightening jack bolt 132. Ratchet lock spring 138 prevents jack bolt 132 from loosening during reactor operation. Typically four repair apparatus 60, equally spaced around shroud 20, are installed in reactor pressure vessel 10 to repair cracked shroud welds 86, 88, 90, 92, and 94. The above described shroud repair apparatus 60 does not require any installation machining of existing reactor components prior to installation, and therefore is quickly and easily installed in reactor pressure vessel 10. Repair apparatus provides lateral support for shroud 20 and imparts a clamping force to shroud 20 to maintain shroud joint integrity and overcome the effects of any stress corrosion cracking in circumferential shroud welds 86, 88, 90, 92, and 94. Additionally, slotted lug opening 116 permits the use of the existing shroud lug T-bolt 117 attachment without significant additional loading. The normal upward load applied by existing shroud T-bolt 117 to lug 104 is about one half the new downward tie rod 66 preload, so the net operating load on lug 104 is not increased. Further, shroud T-bolt 117 contributes to the total available strength in the load path connecting to tie rod 66, which offers additional margin for higher LOCA accident condition pressure uplift loading. While the invention has been described and illustrated in terms of various specific embodiments, those skilled in the art will recognize that the invention can be practiced with modification within the spirit and scope of the claims.
047145854	claims	1. A spacer grid assembly for supporting a plurality of nuclear fuel pins in a closely-spaced array, comprising: (a) a first grid strip comprising a plurality of segments, disposed at an angle to each other with alternate segments being substantially parallel to each other, each of said segments carrying at least one dimple for contacting said fuel pins; (b) a second grid strip comprising a plurality of segments disposed at an angle to each other with alternate segments being substantially parallel to each other, each alternate segment of said second strip carrying a resilient contact for contacting said fuel pins; (c) a third grid strip comprising a plurality of segments disposed at an angle to each other with alternate segments being substantially parallel to each other, each of said segments of said third strip carrying at least one dimple for contacting said fuel pins; and (d) interlocking means formed on said grid strips for interlocking said first and third grid strips with said second grid strip to form an array of grid cells having a hexagonal cross section and said segments of said first and third grid strips are substantially coplanar, and wherein said interlocking means includes first tabs positioned on alternate segments of said first and third grid strips for interlocking with first slots formed on other alternate segments of said second grid strips that do not carry said resilient contacts. (a) a first grid strip comprising a plurality of segments, disposed at an angle to each other with alternate segments being substantially parallel to each other, each of said segments carrying at least one dimple for contacting said fuel pins; (b) a second grid strip comprising a plurality of segments disposed at an angle to each other with alternate segments being substantially parallel to each other, each alternate segment of said second strip carrying a resilient contact for contacting said fuel pins; (c) a third grid strip comprising a plurality of segments disposed at an angle to each other with alternate segments being substantially parallel to each other, each of said segments of said third strip carrying at least one dimple for contacting said fuel pins; and (d) interlocking means formed on said grid strips for interlocking said first and third grid strips with said second grid strip to form an array of grid cells having a hexagonal cross section and said segments of said first and third grid strips are substantially coplanar, and wherein said interlocking means further includes second slots formed at a vertex of said angles of said first and third strips and wherein each of said alternate segments of said second strip carries second tabs for interlocking with said second slots. (a) a first grid strip comprising a plurality of segments, disposed at an angle to each other with alternate segments being substantially parallel to each other, each of said segments carrying at least one dimple for contacting said fuel pins; (b) a second grid strip comprising a plurality of segments disposed at an angle to each other with alternate segments being substantially parallel to each other, each alternate segment of said second strip carrying a resilient contact for contacting said fuel pins; (c) a third grid strip comprising a plurality of segments disposed at an angle to each other with alternate segments being substantially parallel to each other, each of said segments of said third strip carrying at least one dimple for contacting said fuel pins; and (d) interlocking means formed on said grid strips for interlocking said first and third grid strips with said second grid strip to form an array of hexagonal grid cells, said interlocking means includes first tabs positioned on alternate segments of said first and third grid strips for interlocking with first slots formed on other alternate segments of said second grid strips that do not carry said resilient contacts, said interlocking means further includes second slots formed at a vertex of said angles of said first and third grid strips and wherein each of said alternate segments of said second strip carries second tabs for interlocking with said second slots, and wherein the alternate segments of the first and third grid strips, when assembled, are coplanar with each other and form at least a portion of a wall of a grid cell. an elongated first grid strip comprising a plurality of segments in which the adjacent segments are arranged at obtuse angles with each other, said first grid strip having a first longitudinal edge and a second longitudinal edge; an elongated second grid strip comprising a plurality of segments in which the adjacent segments are arranged at obtuse angles with each other, said second grid strip having a first longitudinal edge and a second longitudinal edge, a plurality or spaced first flanges extending from said first longitudinal edge and a plurality of spaced second flanges extending from said second longitudinal edge; an elongated third grid strip comprising a plurality of segments in which the adjacent segments are arranged at obtuse angles with each other, said third grid strip having a first longitudinal edge and a second longitudinal edge; and interlocking means formed on said grid strips for interlocking said first and third grid strips with said second grid strip to form an array of grid cells having hexagonal cross sections so that said first and third grid strips are coplanar and so that portions of said second longitudinal edge of said first grid strip abut portions of said first longitudinal edge of said second grid strip, and portions of said first longitudinal edge of said third grid strip abut portions of said second longitudinal edge of said second grid strip. 2. The spacer grid assembly of claim 1, wherein said first slots comprise S-shaped bends formed along edges of said other alternate segments. 3. The spacer grid assembly of claim 1, wherein said interlocking means further includes second slots formed at a vertex of said angles of said first and third strips and wherein each of said alternate segments of said second strip carries second tabs for interlocking with said second slots. 4. The spacer grid assembly of claim 3, wherein said peripheral boundary strip, together with said grid strip, forms a series of peripheral grid cells, at least some of which are pentagonal. 5. The spacer grid assembly of claim 3, in which said peripheral boundary strip comprises third slots and said grid strips terminate in third tabs, wherein said third tabs are positioned in said third slots. 6. The spacer grid assembly of claim 5, wherein segments of said second strip means carrying said resilient contacts substantially form a wall of a grid cell. 7. The spacer grid assembly of claim 5, wherein each cell wall comprises only a single thickness of grid strip material. 8. A spacer grid assembly for supporting a plurality of nuclear fuel pins in a closely-spaced array, comprising: 9. The spacer grid assembly of claim 8 further including a peripheral boundary strip at which said grid strips terminate. 10. A spacer grid assembly for supporting a plurality of nuclear fuel pins in a closely-spaced array, comprising: 11. A spacer grid assembly for supporting a plurality of nuclear fuel pins in a closely-spaced array, comprising: 12. The spacer grid assembly of claim 11, wherein said first and second flanges extend in opposite directions and said interlocking means further includes tabs formed on said first and second flanges of said second grid strip and slots formed in said first and third strips, said tabs engage said slots when said first, second, and third grid strips are interlocked.
047740493	claims	1. A sensor monitoring system for displaying a profile of enthalpy rise over a defined area comprising at least a part of a core of a nucelear reactor, which system comprises: a plurality of core exit temperature sensors positioned to monitor at least a portion of the defined area; an inlet temperature sensor outside the core and monitoring the temperature of core coolant at an inlet to the reactor; means, responsive to the outputs from both said temperature sensors and said inlet temperature sensor, for generating corresponding representative fractinal deviation values of the enthalpy rise from reference values; means for interpolating the generated values of the enthalpy rise to provide an output representaitive of interpolated values of the fractional deviations in enthalpy rise over the entire defined area; means for classifying the generated values and the interpolated values into a number of different classes; and means for multidimensionally displaying the classes of the generated values and the classes of the interpolated values to present the enthalpy rise profile over the defined area. a plurality of core exit temperature sensors positioned to monitor at least a portion of the defined area; an inlet temperature sensor outside the core and monitoring the temperature of core coolant at an inlet of the reactor; means, responsive to the outputs from both said temperature sensors and said inlet temperature sensor, for generating corresponding representative values of the enthalpy rise; means for interpolating the values of the enthalpy rise to provide an output representative of the interpolated values of the enthalpy rise over the entire defined area; means, responsive to the values, the interpolated values, and reference values of the enthalpy rise, for generating fractional deviations of the values and the interpolated values from the reference values; means for classifying the deviations into a number of different classes; and means for multidimensionally displaying the classes of the deviations. the display means displays the class of the values and the class of the interpolated values in a visual map designed to correspond to the defined area, said map having subdivisions corresponding to portions of the defined area and said subdivisions being color modulated according to the class of the value or the class of the interpolated value corresponding thereto, whereby a relative color variable profile of the enthalpy is displayed over the defined area. the display means displays the class of the deviations in a visual map designed to correspond to the defined area, said map having subdivisions corresponding to portions of the defined area and said subdivisions being color modulated according to the class of the deviation corresponding thereto, whereby a relative color variable profile of the deviations is displayed over the defined area. monitoring flux emanating from the core at positions exterior of the core; monitoring the temperature of reactor coolant flowing into the core and the temperatures of the reactor coolant exiting the core assemblies; generating fractional deviations of enthalpy rise from the temperatures; determining the three dimensional core power distribution in response to the flux measurements; adjusting the three dimensional power distribution in light of the deviations; and multidimensionally presenting the adjusted three dimensional core power distribution. classifying the deviations into a number of different classes; and displaying the class of deviations in color modulated form according to the class of the deviation. a core inlet temperature sensor; a plurality of core-exit thermocouples positioned to monitor at least a portion of the defined area; means, responsive to the outputs from said thermocouples and said core inlet temperature sensor, for generating corresponding representative fractional deviation values of relative enthalpy rise from reference values; means for interpolating the generated values of relative enthalpy rise to provide an output representative of interpolated values of fractional deviations in enthalpy rise in the defined area; a plurality of neutron sensors for monitoring core flux; means for modifying the generated values and the interpolated values in dependence on the core flux to produce power values; means for classifying the power values into a number of different classes; and means for multidimensionally displaying the classes of the power values in a visual map designed to correspond to the defined area, said map having subdivisions corresponding to portions of the defined area and said subdivisions being color modulated according to the class of the power value corresponding thereto, the color classifications being chosen so that lower values of power are more blue and higher values of power contain more red, whereby a relative color variable profile of power is displayed over the defined area. a plurality of neutron sensors for monitoring flux in the core; and means for combining the generated values and the interpolated values of the fractional deviation in enthalpy rise with the flux outputs of said neutron sensors and producing power values, said classifying means classifying the power values and said display means displaying the classes of the power values. a plurality of neutron sensors for monitoring flux in the core; and means for combining the deviations with the flux outputs of said neutron sensors and producing power deviations, said classifying means classifying the power deviations and said display means displaying the classes of the power deviations. (a) monitoring core coolant inlet temperature; (b) monitoring core coolant exit temperatures of the coolant exiting core assemblies; (c) determining fractional deviations in enthalpy rise from the core coolant inlet temperature, the exit temperatures and reference values; and (d) multidimensionally displaying the deviations. a plurality of core exit coolant temperature sensors positioned to monitor at least a portion of the defined area; an inlet temperature sensor outside the core which monitors the temperature of core coolant at an inlet to the reactor; means, responsive to the outputs from both said core exit temperature sensors and said inlet temperature sensor, for generating corresponding representative values of acutal coolant enthalpy rise and corresponding values of relative enthalpy rise at each location in the defined area at which a core exit coolant temperature sensor is available; means, responsive to the generated values of relative enthalpy rise and to reference values of relative enthalpy rise at corresponding locations in the defined area, for generating values fo the fractional deviation of the measured values of relative enthalpy rise from the corresponding reference values; means for interpolating the generated values of fractional deviation in relative enthalpy rise to provide interpolated values of fractional deviation in relative enthalpy rise at locations in the defined area of the core other than those at which core exit coolant temperature sensors are available; and means for multidimensionally displaying the generated and interpolated values. means for classifying the generated values and the interpolated values of fractional deviation in relative enthalpy rise into a number of different classes; and means for displaying the classes of generated and interpolated values to present the fractional deviation relative enthalpy rise in profile over the defined area. a plurality of neutron sensors for monitoring neutron flux emanating from the reactor core; means for determining a first estimate of core three-dimensional power distribution by adjusting a reference three-dimensional core power distribution such that an axial component of the power distribution conforms both in shape and in amplitude to the axial power distribution constructed from the flux; an core inlet temperature sensor; a plurality of core exit temperature sensors positioned to monitor at least a portion of the defined areea; means, responsive to the outputs of said core exit temperature sensors and said core inlet temperature sensor, for generating corresponding representative values of relative enthalpy rise; means for genrating corresponding values of fractional deviations of the generated relative enthalpy rise values from coresponding reference relative enthalpy rise values; means for interpolating among the generated values of fractional deviation in relative enthalpy rise to provide interpolated values of fractional deviation in relative enthalpy rise at locations in the defined area that are not directly monitored by core exit temperature sensors; means for adjusting the first estimate of core three-dimensional core power distribution such that the radial component of power distribution incorporates the generated and interpolated values of fractional deviation in relative enthalpy rise; and means for multidimensionally displaying the adjusted first estimate. means for classifying the power values into a number of different classes; and means for multidimensionally displaying the classes of power vlaues in a visual map corresponding to the defined area, said map having subdivisions being color modulated according to the class of the power value corresponding thereto, the color classifications being chosen so that lower values of power are more blue and higher values of power contain more red, and a relative color vairable profile of power is displayed over the defined area. 2. A sensor monitoring system for displaying a profile of enthalpy rise over a defined area comprising at least a part of a core of a nucelear reactor, which system comprises: 3. The sensor monitoring system of claim 1 wherein: 4. The sensor monitoring system of claim 2 wherein: 5. The sensor monitoring system of claim 3 further comprising means for indicating a value or an interpolated value which is outside the limits of a preselected range. 6. The sensor monitoring system of claim 4 further comprising means for indicating a deviation which is outside the limits of a preselected range. 7. The sensor monitoring system of claim 5 wherein said indicating means flashes a light within at least a portion of the map corresponding to the respective subdivision associated with the out-of-limits value or interpolated value. 8. The sensor monitoring system of claim 6 wherein said indicating means flashes a light within at least a portion of the defined area corresponding to the respective subdivision associated with the out-of-limits deviation. 9. The sensor monitoring system of claim 3 wherein said visual map is formed on a two-dimensional color graphics monitor. 10. The sensor monitoring system of claim 4 wherein said visual map is formed on a two-dimensional color graphics monitor. 11. The sensor monitoring system system of claim 1 wherein said interpolation means comprises means for performing a surface splines interpolation. 12. The sensor monitoring system system of claim 2 wherein said interpolation means comprises means for performing a surface splines interpolation. 13. The sensor monitoring system of claim 9 wherein the color classifications are chosen so that lower values of the parameter contain more blue and higher values of the parameter contain more red. 14. The sensor monitoring system of claim 10 wherein the color classifications are chosen so that lower values of the deviation contain more blue and higher values of the deviation contain more red. 15. The sensor monitoring system of claim 1 wherein the reference values are periodically updated. 16. The sensor monitoring system of claim 1 wherein the reference values are periodically updated. 17. A method of monitoring the three dimensional power distribution in a core of a nuclear reactor in a manner to representatively reconstruct the three dimensional power distribution which comprises the steps of: 18. The method of claim 17 wherein the step of presenting comprises displaying the adjusted three dimensional core power distribution. 19. The method of claim 17 wherein the step of presenting comprises storing the adjusted three dimensional core power distribution. 20. The method of claim 17 wherein the step of presenting comprises presenting the adjusted three dimensional core power distribution to means for determining and presenting the burnup distribution in the core. 21. The method of claim 20 wherein presenting the burnup distribution comprises displaying the burnup distribution. 22. The method of claim 18 wherein the step of displaying comprises displaying two dimensional representations of deviations of the adjusted three dimensional core power distribution from predetermined reference values, the two dimensional representations correspond to planes in the reactor core which are perpendicular to a given axis of the reactor core. 23. The method of claim 22 wherein the step of displaying further comprises: 24. The method of claim 23 wherein the color classifications are chosen so that lower values of the deviation contain more red. 25. A sensor monitoring system for displaying a profile of power over a defined area comprising at least a part of a core of a nuclear reactor, which system comprises: 26. The sensor monitoring system of claim 1, further comprising: 27. The sensor monitoring system of claim 2, further comprising: 28. The sensors monitoring system of claim 27, wherein said combining means adjusts a reference three-dimensional core power distribution and produces three-dimensional power values. 29. The sensor monitoring system of claim 28, wherein said dispaly means displays the three-dimensional power values. 30. A method of monitoring enthalpy rise in a core of a nucelar reactor, comprising the steps of: 31. A sensor monitoring system for displaying a profile of fractional deviations in relative coolant enthalpy rise over a defined area comprising at least a part of a core of a nucelar reactor, which system comprises: 32. The sensor monitoring system of claim 31, further comprising: 33. A sensor monitoring system for displaying a profile of power over a defined area comprising at least a part of a core of a nuclear reactor, which system comprises: 34. The sensor monitoring system of claim 33, further comprising:
054250641	abstract	Coolant flow (900) in the core (103) of a natural circulation boiling water reactor (100) is monitored by making use of a detector (208) for nuclear radiation and a turbine device (204) which comprises a rotor (304) which can modulate the flux (206) of the nuclear radiation on the detector. The turbine rotor can be installed in a fuel bundle (104), while the detector can be placed in an adjoining instrumentation tube (320). Included in the turbine rotor is material (308) which modulates the radiation field (neutron field and/or gamma field). The radiation detector detects (602) variations in the radiation field when the turbine rotor device is set into rotation by the coolant flow (900). The coolant flow rate is then calculated (603) from the speed of these variations.
	summary
047568682	abstract	Mechanism for displacing and securing the rod of a nuclear reactor control bar, comprising an enclosure coaxial with the rod and forming a hydraulic cylinder in which the rod slides while defining a decompression chamber, a device for mechanically securing the rod in its high position, and a positive displacement reciprocating pump inserted between the decompression chamber and the inside of the tank, whose piston is actuated electromagnetically means and causes at each backward and forward movement lifting of given amplitude of the rod and the device for securing the rod in the high position disengagable also electromagnetically.
	description	The present application is a CIP of international applications number PCT/FR2015/052475, filed Sep. 16, 2015, and PCT/FR2015/052476, filed Sep. 16, 2015, both of which claim benefit to U.S. provisional application, 62/051,913, filed Sep. 17, 2014, the entire contents of all three of which are hereby incorporated herein by reference. The present invention belongs to the field of the materials employed in the nuclear field, in particular materials intended to exhibit the best possible resistance to the physicochemical conditions encountered under nominal conditions and during an accident scenario of a nuclear reactor, such as, for example a Pressurized Water Reactor (PWR) or a Boiling Water Reactor (BWR), or a reactor of “Canadian Deuterium Uranium” (CANDU) type. The invention relates more particularly to nuclear fuel claddings, to their processes of manufacture and uses against oxidation and/or hydriding. The constituent zirconium alloy of nuclear fuel claddings oxidizes on contact with the water constituting the coolant of PWR or BWR or of CANDU type nuclear reactors. As the oxide formed is brittle and the take-up of hydrogen associated with the oxidation results in the precipitation of zirconium hydrides which cause embrittlement, the lifetime of the claddings is in large part limited by the maximum thickness of oxide acceptable and the content of associated absorbed hydrogen. In order to guarantee satisfactory residual mechanical properties of the cladding targeted at ensuring an optimum confinement of the nuclear fuel, the residual thickness of healthy and ductile zirconium alloy has to be sufficient and the fraction of hydrides sufficiently limited. The possibility of limiting or of delaying such an oxidation and/or the hydriding may thus prove to be crucial in accident conditions. These conditions are, for example, reached in the case of hypothetical accident scenarios of the RIA (“Reactivity Insertion Accident”) or LOCA (Loss Of Coolant Accident) type, indeed even in conditions of dewatering of the spent fuel storage pool. They are characterized, among others, by high temperatures which are generally greater than 700° C., in particular between 800° C. and 1200° C., and which may be reached with a high rate of rise in temperature. At such temperatures, the coolant is in the form of steam. The oxidation in accident conditions is much more critical than in conditions of normal operation of the nuclear reactor, as the deterioration in the cladding, the first barrier for confinement of the fuel, is faster and the associated risks greater. These risks are, among others, as follows: emission of hydrogen; embrittlement of the cladding at high temperature, by the oxidation, indeed even, under certain conditions, the hydriding of the cladding; the embrittlement of the cladding on quenching, brought about by the sudden decrease in temperature during the massive supplying of water for making the core of the nuclear reactor safe; low mechanical strength of the cladding after the quenching or the cooling, in the case, among others, of post accident handling operations, of earthquake aftershocks. In view of these risks, it is essential to limit as far as possible the high temperature oxidation and/or hydriding of the cladding in order to improve the safety of nuclear reactors using, among others, water as coolant. The solution proposed by the patent application “WO 2013/160587” consists in producing a nuclear fuel cladding in which the zirconium-based substrate is covered with a multilayer coating comprising metal layers composed of chromium, of a chromium alloy and/or of a ternary alloy of the Nb—Cr—Ti system. Nevertheless, additional experiments have shown that the resistance to high-temperature oxidation, although improved with respect to the prior claddings, proves to be insufficient at very high temperature, typically for temperatures of greater than or equal to 1200° C. when the deposition of the multilayer coating on the zirconium-based substrate was carried out by a process of physical vapor deposition (PVD) by magnetron cathode sputtering of conventional type. These very high temperatures lie at the extreme of, indeed even beyond, those of the high temperatures between 700° C. and 1200° C. which are set by accident regulatory conditions. In point of fact, the regulatory criteria governing the dimensioning accidents according to the scenario of “LOCA” type defined from the 1970s require that the maximum temperature of the cladding does not exceed 1204° C. (2200° F.) and a maximum degree of “ECR” oxidation of 17%. The degree of “ECR” (“Equivalent Cladding Reacted”) oxidation is the percentage of thickness of metal cladding converted into zirconia (ZrO2) resulting from the oxidation of the zirconium contained in the nuclear fuel cladding, it being assumed that all the oxygen which has reacted forms stoichiometric zirconia. In order to take into account the additional embrittling effect related to the in service hydriding of the cladding, this acceptable residual degree of “ECR” oxidation may henceforth be much lower than 17% under certain conditions, such as, for example, a cladding hydrided in service up to several hundred ppm by weight, which corresponds in practice to a duration of oxidation of the cladding which should not exceed a few minutes at 1200° C. An improvement in the resistance to oxidation and/or hydriding at very high temperature would advantageously make it possible to obtain additional margins of safety, among others by preventing or delaying all the more the deterioration in the cladding in the event of aggravation or of persistence of the accident situation. One of the aims of the invention is thus to prevent or to alleviate one or more of the disadvantages described above by providing a nuclear fuel cladding and a process for the manufacture thereof which makes it possible to improve the resistance to oxidation and/or hydriding, among others in the presence of steam. Another aim of the invention is to improve this resistance to oxidation and/or hydriding at very high temperature, namely above 1200° C., particularly between 1200° C. and 1400° C., more particularly between 1200° C. and 1300° C.; among others when these temperatures are reached with a rate of rise in temperature which is between 0.1° C./second and 300° C./second. Another aim of the invention is to improve the duration of the resistance to oxidation and/or hydriding, beyond which duration the confinement of the nuclear fuel is no longer certain. The present invention thus relates to a process for the manufacture of a nuclear fuel cladding comprising i) a substrate containing a zirconium-based internal layer coated or not coated with at least one interposed layer placed over the internal layer and ii) at least one external layer placed over the substrate and composed of a protective material chosen from chromium or a chromium alloy; the process comprising the following successive steps: a) ion etching of the surface of the substrate; b) deposition of the at least one external layer over the substrate with a high power impulse magnetron sputtering (HiPIMS) process in which the magnetron cathode is composed of the protective material. The nuclear fuel cladding thus manufactured can be composite (presence of an interposed layer) or not. In comparison with the processes of the state of the art, the manufacturing process of the invention has among others the distinctive feature that it uses a high power impulse magnetron sputtering (HiPIMS) process in order to deposit, according to step b), the at least one chromium-based external layer over the zirconium-based internal layer. Such a process is known to a person skilled in the art and described, for example, in the document “Techniques de l'ingénieur, La pulvérisation cathodique magnétron en régime d'impulsions de haute puissance (HiPIMS) pulvérisation cathodique magnétron, Réference IN207” [Techniques of the Engineer, Magnetron cathode sputtering under high power impulse conditions (HiPIMS) magnetron sputtering, Reference IN207]. The HiPIMS sputtering process differs in a number of aspects from conventional magnetron cathode sputtering processes. In accordance with the conventional magnetron cathode sputtering process as used in “WO 2013/160587” (mentioned hereinbelow as conventional magnetron PVD process), a difference in potential is applied between a negatively polarized chromium target (magnetron cathode) and the walls of the cathode sputtering reactor which are connected to earth. For this type of process, the continuous polarization voltage applied to the target is typically between −600 V and −200 V. The discharge current is a few amperes. Under these conditions, the rarefied atmosphere, generally composed of argon, is then partially ionized and forms a cold plasma. It then essentially comprises argon atoms Ar and a low proportion of argon ions Ar+, but no metal ions or metal ions in an infinitesimal amount far below 10−6. The Ar+ ions are subsequently accelerated by the electric field of the target, which they impact, which results in the ejection of chromium atoms which are deposited on the substrate to be coated which generally faces the target. The HiPIMS sputtering process differs in particular from the conventional magnetron PVD process used in “WO 2013/160587” in several characteristics, among others: high frequency polarization impulses are applied to the chromium target constituting the magnetron cathode. The impulses last, for example, from 1/1000 to 1/100 of the total polarization duration; the instantaneous power delivered by each impulse is from several tens of kilowatts to a few megawatts. This results in the ejection of a large amount of Cr+ metal ions, although the power averaged over the entire duration of polarization is at most a few kilowatts, for example less than 1.2 kW; the production of an atmosphere essentially composed of Cr− metal ions. Unexpectedly, the inventors have discovered that a nuclear fuel cladding obtained by the manufacturing process of the invention made it possible to confer on it an improved resistance to oxidation and/or hydriding, in particular at very high temperature, among others in the presence of steam. Such properties cannot be anticipated from the viewpoint of the specific chemical and metallurgical qualities of zirconium and the zirconium alloys used for nuclear applications, among others their chemical composition, surface condition, crystal texture, final metallurgical state (work-hardened or more or less recrystallized), which properties are liable to influence the quality and the behavior of the coatings. In particular, the α phase of a zirconium alloy (denoted “Zr-α”, of compact hexagonal crystallographic structure) at low temperature is converted into the β phase (denoted “β-Zr”, of cubic centered crystallographic structure) in a temperature range typically ranging from 700° C. to 1000° C. On changing from the Zr-α structure to the β-Zr cubic structure, the alloy undergoes local dimensional variations. These variations are a priori unfavorable to the mechanical strength of an external layer which would cover a zirconium-based internal layer, due among others to the incompatibility in their coefficients of expansion. These difficulties of adhesion are accentuated by the mechanisms of diffusion of chemical entities, which are faster in the β-Zr phase than in the Zr-α phase, and which may modify the interface between the substrate and its coating. The invention also relates to a nuclear fuel cladding (composite or not) obtained or obtainable by the manufacturing process of the invention. The invention also relates to a composite nuclear fuel cladding comprising i) a substrate containing a zirconium-based internal layer and at least one interposed layer placed over the internal layer and composed of at least one interposed material chosen from tantalum, molybdenum, tungsten, niobium, vanadium, hafnium or their alloys and ii) at least one external layer placed over the substrate and composed of a protective material chosen from chromium or a chromium alloy. In this case, the interposed material or the protective material may be respectively deposited on the internal layer or on the substrate by any type of process, for example a process of physical vapor deposition by magnetron cathode sputtering different from the HiPIMS sputtering process, or by a high power impulse magnetron sputtering (HiPIMS) such as described, among others in its alternative forms, in the present description wherein the magnetron cathode is composed of the interposed material or of the protective material. The internal layer of the composite nuclear fuel cladding can be composed of a zirconium alloy comprising, by weight, from 100 ppm to 3000 ppm of iron, the cladding comprising an interface layer positioned between the internal layer and the external layer and composed of an interfacial material comprising at least one intermetallic compound chosen from ZrCr2 of cubic crystal structure, Zr(Fe,Cr)2 of hexagonal crystal structure or ZrFe2 of cubic crystal structure. Naturally, it is in particular positioned between the internal layer and the first external layer. The interface layer is formed during the HiPIMS deposition according to step b) of the external layer on the substrate in the absence of at least one interposed layer placed over the internal layer and when the constituent zirconium alloy of the internal layer comprises, by weight, from 100 ppm to 3000 ppm of iron. Unexpectedly, it does not damage the adhesion of the external layer with regard to the substrate, whereas intermetallic compounds are known for their mechanical properties of brittle type. Furthermore, under oxidation in the presence of steam, at high temperature, indeed even very high temperature (for example at 1200° C.), the inventors have found that the interfacial layer predominantly or completely comprising at least one intermetallic compound thickens. Here again, unexpectedly, no generalized exfoliation is found, despite the assumed intrinsic brittleness of an intermetallic compound and the interfacial stresses which theoretically may develop during the manufacturing steps, indeed even subsequently in service and/or in nominal conditions or certain accident conditions. See, for example, example 2.4 Preferably, the interface layer has a mean thickness of 10 nm to 1 μm. These types of nuclear fuel claddings according to the invention, namely obtained or liable to be obtained by the manufacturing process of the invention, composite or noncomposite or with an interface layer, may be provided according to one or more of the alternative forms described in the present description for the abovementioned manufacturing process of the invention, among others the alternative forms relating to the structure and/or to the composition of a nuclear fuel cladding. These alternative forms relate among others and not exclusively to: the internal layer, the internal coating, the composition of the zirconium alloy or of the chromium alloy, the structure of the external layer, such as are described in detail in the present description, among others the description of the manufacturing process of the invention. The geometry of these claddings is such that they may be provided in the form of a tube, or of a plate resulting more particularly from the assembling of two subunits. The invention also relates to a process for the manufacture of a composite nuclear fuel cladding, comprising the following successive steps: A) production of a substrate by deposition, on a zirconium-based internal layer, of at least one interposed layer composed of at least one interposed material chosen from tantalum, molybdenum, tungsten, niobium, vanadium, hafnium or their alloys; B) deposition, on the substrate, of at least one external layer composed of a protective material chosen from chromium or a chromium alloy. The deposition according to step A) and/or B) can be carried out by a physical vapor deposition or a pulsed electrolysis. The physical vapor deposition may be a cathode sputtering, more particularly of magnetron type, still more particularly an HiPIMS sputtering process preferably according to one or more of the characteristics shown in the present description. When the interposed layer is composed of hafnium, its thickness is from 1 nm to 1 μm. The invention also relates to the use of these types of claddings, for combating oxidation and/or hydriding in a humid atmosphere comprising water, in particular in the form of steam. The invention also relates to the use of these types of claddings (in particular the composite nuclear fuel cladding), for combating hydriding in a hydrogenated atmosphere comprising hydrogen, in particular an atmosphere comprising more than 50 mole % of hydrogen and/or water, in particular in the form of steam. The humid atmosphere or the hydrogenated atmosphere may further comprise an additional gas chosen from air, nitrogen, carbon dioxide or their mixtures. Preferably, the purpose of these uses is to combat oxidation and/or hydriding: in which the humid atmosphere or the hydrogenated atmosphere is at a temperature of between 25° C. and 1400° C., indeed even between 25° C. and 1600° C., more particularly a temperature of between 200° C. and 1300° C., still more particularly between 1200° C. and 1300° C., indeed even between 1300° C. and 1600° C.; and/or up to at least 5000 seconds, more particularly between 1000 seconds and 5000 seconds, in particular when the temperature is between 1200° C. and 1300° C.; and/or in the presence of a rate of rise in temperature which is between 0.1° C./second and 300° C./second; and/or subsequent to a quenching with water of the nuclear fuel cladding, in particular in which the quenching takes place at a temperature of between 25° C. and 400° C. In the present description of the invention, a verb such as “to comprise”, “to incorporate”, “to include”, “to contain”, “composed of” and its conjugated forms are open terms and thus do not exclude the presence of additional element(s) and/or step(s) which are added to the initial element(s) and/or step(s) stated after these terms. However, these open terms are targeted in addition at a specific embodiment in which only the initial element(s) and/or step(s), with the exclusion of any other, are targeted; in which case, the open term is targeted in addition at the closed term “to consist of”, “to constitute ” and its conjugated forms. The expression “and/or” is targeted at connecting elements for the purpose of simultaneously denoting just one of these elements, both elements, indeed even their mixture or their combination. The use of the indefinite article “one” or “a” for an element or a step does not exclude, unless otherwise mentioned, the presence of a plurality of elements or steps. Any reference sign in brackets in the claims should not be interpreted as limiting the scope of the invention. Furthermore, unless otherwise indicated, the values of the limits are included in the ranges of parameters shown and the temperatures shown are considered for an implementation at atmospheric pressure. The manufacturing processes of the invention are targeted at producing a nuclear fuel cladding (composite or not) comprising: i) a substrate containing a zirconium-based internal layer intended to be in contact with or facing the nuclear fuel, the internal layer being coated or not with at least one interposed layer; and ii) at least one chromium-based external layer placed over the substrate and intended to protect the cladding with regard to the exterior environment, in particular the coolant. As previously mentioned, the alternative embodiments described hereinafter, among others the alternative forms relating to the structure and/or to the composition of a nuclear fuel cladding are also alternative forms for all the types of nuclear fuel claddings of the present description. Preferably, at least one interposed layer is placed between the internal layer and the external layer with regard to which layers it acts as diffusion barrier. In this embodiment, the substrate is formed by the combination of the internal layer and of the at least one interposed layer. The cladding may also comprise an internal coating placed under the internal layer, the thickness of which is, for example, between 50 μm and 150 μm. The internal coating may comprise one or more layers. It constitutes an internal “liner” which improves the strength of the cladding with regard to the physicochemical and mechanical interactions with the fuel. It is generally obtained by hot coextrusion during the manufacture of the internal layer. The internal layer is zirconium-based, namely that it is composed of zirconium to more than 50% by weight, particularly more than 90%, indeed even more than 95%. More specifically, the internal layer and/or the internal coating is composed of zirconium or of a zirconium alloy. The zirconium alloy may comprise, by weight: from 0% to 3% of niobium; preferably 0% to 1.2%; from 0% to 2% of tin; preferably 0% to 1.3%; from 0% to 0.5% of iron; preferably from 100 ppm to 2000 pm; from 0% to 0.2% of chromium; from 0% to 0.2% of nickel; from 0% to 0.2% of copper; from 0% to 1% of vanadium; from 0% to 1% of molybdenum; and from 0.05% to 0.2% of oxygen. The zirconium alloy is, for example, Zircaloy-2 or Zircaloy-4. The zirconium alloy may in particular be chosen from an alloy meeting the constraints of the nuclear field, for example Zircaloy-2, Zircaloy-4, Zirlo™, Optimized-Zirlo™ or M5™. The compositions of these alloys are such that they comprise, by weight, for example: Zircaloy-2 alloy: 1.20% to 1.70% of Sn; 0.07% to 0.20% of Fe; 0.05% to 1.15% of Cr; 0.03% to 0.08% of Ni; 900 ppm to 1500 ppm of O; the remainder is zirconium. Zircaloy-4 alloy: 1.20% to 1.70% of Sn; 0.18% to 0.24% of Fe; 0.07% to 1.13% of Cr; 900 ppm to 1500 ppm of O; less than 0.007% of Ni; the remainder is zirconium. Zirlo™ alloy: 0.5% to 2.0% of Nb; 0.7% to 1.5% of Sn; 0.07% to 0.28% of at least one element chosen from Fe, Ni, Cr; up to 200 ppm of C; the remainder is zirconium. Optimized-Zirlo™ alloy: 0.8% to 1.2% of Nb; 0.6% to 0.9% of Sn; 0.090% to 0.13% of Fe; 0.105% to 0.145% of O; the remainder is zirconium. M5™ alloy: 0.8% to 1.2% of Nb; 0.090% to 0.149% of O; 200 ppm to 1000 ppm of Fe; the remainder is zirconium. The at least one external layer placed over the substrate is composed of a protective material chosen from chromium or a chromium alloy, in particular any chromium alloy capable of being used in the nuclear field and/or under irradiation. More particularly, the chromium alloy composing the protective material may comprise at least one alloying element chosen from silicon, yttrium or aluminum, for example at a content of 0.1% to 20% by atoms. The at least one external layer optionally has a columnar structure. Preferably, the constituent columnar grains of the columnar structure have a mean diameter of 100 nm to 10 μm. The at least one chromium-based external layer is deposited over the substrate with the manufacturing process of the invention according to the following successive steps: a) ion etching of the surface of the substrate; b) deposition over the substrate of the at least one external layer with a high power impulse magnetron sputtering (HiPIMS) process. The magnetron cathode constituting the target is then composed of the protective material. Steps a) and b) are carried out over the final layer of the substrate, namely over the zirconium-based internal layer or over the final interposed layer, according to whether the substrate respectively contains an internal layer coated or not coated with at least one interposed layer. In order to place at least one interposed layer over the internal layer, it is possible to carry out the following successive steps carried out before the etching step a): a′) ion etching of the surface of the internal layer; b′) production of a substrate by deposition of the at least one interposed layer over the internal layer with a high power impulse magnetron sputtering (HiPIMS) process in which the magnetron cathode is composed of the at least one interposed material. This embodiment constitutes a specific case of the process for the manufacture of the composite nuclear fuel cladding according to the invention in that the at least one interposed layer is deposited with an HiPIMS sputtering process. The distance separating the substrate and the magnetron cathode used according to the etching step a) or a′) and/or the deposition step b) or b′) may be between 40 mm and 150 mm. The ion etching according to step a) and/or a′) may be carried out with an HiPIMS etching process or a cathodic arc etching process. The use of the HiPIMS sputtering process according to step b) or b′) requires the establishment of a polarization voltage using polarization impulses which are applied to the target present in the cathode sputtering reactor. The magnetron cathode may be a flat cathode, or a hollow cathode for example a cylindrical cathode. The ranges of values which follow relating to the polarization voltages and the polarization impulses are given by way of indication for a magnetron cathode with a surface area of 300 cm2. A person skilled in the art may in particular adjust the values shown for the polarization impulses applied to the magnetron cathode in order to observe the recommended power density range, it being known that the polarization voltage to be applied varies inversely in proportion to the surface area of the target. The HiPIMS etching process used in step a) and/or a′) may comprise the polarization of the magnetron cathode with a voltage of between −1000 V and −500 V. The cathodic arc etching process according to step a) and/or a′) may comprise the polarization of an arc cathode with a voltage of between −20 V and −50 V or according to an intensity of 50 A to 250 A. The HiPIMS or cathodic arc etching process according to step a) and/or a′) may comprise the polarization of the substrate with a voltage of between −800 V and −600 V. The Cr+ ions produced during step a) and/or a′) etch the surface of the substrate for the purpose of improving the adhesion of the external layer to be deposited. The HiPIMS sputtering process according to step b) and/or b′) generally comprises the maintenance of the polarization of the magnetron cathode or so that the voltage remains between −1000 V and −500 V. The polarization of the substrate is for its part decreased with respect to the step of etching according to step a) and/or a′), for example so that the HiPIMS sputtering process according to step b) and/or b′) comprises the polarization of the substrate with a voltage of between −200 V and 0 V. The HiPIMS sputtering process according to step b) and/or b′) may comprise the application to the magnetron cathode of polarization impulses, each of which may exhibit at least one of the following characteristics: duration of 10 ps to 200 ps; instantaneous mean peak intensity of 50 A to 1000 A, for example of 50 A to 200 A; instantaneous power of 50 kW to 2 MW, more particularly of 100 kW to 2 MW; power density of 0.2 kW/cm2 to 5 kW/cm2, more particularly 1 kW/cm2 to 5 kW/cm2. The polarization impulses may be applied to the magnetron cathode according to a frequency of 50 Hz to 600 Hz, more particularly of 100 Hz to 600 Hz. The HiPIMS etching process according to step a) and/or a′) or the HiPIMS sputtering process according to step b) and/or b′) is carried out with a carrier gas comprising at least one rare gas. The rare gas may be chosen from argon, xenon or krypton. The carrier gas is, for example, at a pressure of between 0.2 Pa and 2 Pa. According to a specific embodiment of the manufacturing process of the invention, after the deposition over the substrate of the first external layer with the HiPIMS sputtering process according to step b) and/or b′), at least a part of the additional external layer(s) are deposited during step b) and/or b′) with a magnetron cathode sputtering process of a different type from the HiPIMS which is carried out simultaneously with the HiPIMS sputtering process according to step b) and/or b′). The magnetron cathode sputtering process of a different type from the HiPIMS is, for example, such that the polarization of the target is continuous (“DC” for “Direct Current”) or pulsed at a medium frequency (“pulsed DC”), resulting in a polarization voltage delivering an instantaneous power of a few kilowatts. The deposition of the additional external layers with a conventional magnetron PVD process combined with the HiPIMS sputtering process according to step b) and/or b′) makes it possible to improve the industrial operation of the manufacturing process of the invention by increasing the rate of deposition of the additional external layers. On the conclusion of the manufacturing process of the invention, at least one external layer is obtained which has a thickness of 1 μm to 50 μm, preferably of 3 μm to 25 μm, still more preferably of 3 μm to 10 μm. The cumulative thickness of the external layers is generally from 1 μm to 50 μm, indeed even 2 μm to 50 μm. Several external layers may be deposited on the substrate. For example, the nuclear fuel cladding comprises from 1 to 50 external layers, in order to constitute a multilayer external coating. If appropriate, the external layers may merge to give a single external layer in order to constitute a monolayer external coating, for example after application of a heat treatment to the external layers or by varying the etching and deposition conditions. According to a preferred embodiment of the manufacturing process of the invention, the at least one interposed layer is composed of at least one interposed material chosen from tantalum, molybdenum, tungsten, niobium, vanadium, hafnium or their alloys. Such an interposed layer constitutes a diffusion barrier which limits, indeed even prevents: the diffusion of the chromium from the external layer toward the zirconium-based internal layer, which results in an accelerated consumption of the external layer(s) in addition to its oxidation to give chromic oxide; the formation of a eutectic above 1330° C. approximately, which may potentially harm the mechanical strength of the fuel claddings and their ability to be cooled. Preferably, the interposed material is tantalum. Tantalum or its alloy can be replaced or combined with at least one refractory metal element or its alloy, the physicochemical properties of which are compatible with the zirconium-based internal layer up to 1300° C. In particular, up to 1300° C., the refractory metal element or its alloy does not form a eutectic and exhibits a limited diffusion into the zirconium and/or the chromium. Apart from tantalum, such a refractory metal element is, for example, molybdenum, tungsten or niobium, vanadium, hafnium. Other subject matters, characteristics and advantages of the invention will now be specified in the description which follows of specific embodiments of the process of the invention, given by way of nonlimiting illustration, with reference to the appended figures. 1. Manufacture of a Plate using the Process of the Invention This example of implementation of the manufacturing process of the invention is carried out in a cathode sputtering reactor sold by Balzers (BAK 640 model) and equipped with a Hüttinger generator. The experimental conditions applied may nevertheless vary as a function of the reactor used or of its magnetic configuration, of the shape and of the size of the target, . . . . In accordance with his general knowledge, a person skilled in the art may, however, easily adapt himself to these variations by modifying at least one of the parameters, such as, for example, the polarization voltage of the substrate which is applied during the step a) of ion etching or the step b) of deposition of the internal layer, the duration, the frequency, the intensity or the polarization voltage of the polarization impulses, the distance between the chromium target and the substrate, or the pressure of the carrier gas. More particularly, these parameters influence the mean energy of the Cr+ Pions which are produced during steps a) or b). This energy can condition the density, the homogeneity, the texture, the microstructure or the state of stress of the external layer. 1.1. Step of Ion Etching A Zircaloy-4 plate with dimensions of 45 mm×14 mm×1.2 mm is degreased in an alkaline solution, rinsed with water and cleaned ultrasonically in ethanol. It is subsequently placed in an HiPIMS cathode sputtering reactor containing a chromium magnetron cathode placed at a distance generally of between 6 cm and 8 cm and in this instance of 8 cm. The chamber of the reactor is placed under a vacuum of less than 2.10−5 mbar and then filled with a carrier gas composed of argon at a pressure of 0.5 Pa. The plate constituting the internal layer and thus the substrate to be coated is polarized negatively with a polarization voltage of −800 V. The chromium target is subsequently supplied using the HiPIMS generator according to a polarization voltage of −800 V in order to generate a strongly ionized discharge. The chromium is then sputtered in the form of ions which are accelerated by the electric field of the substrate. The adsorbed carbon-based entities and the nanometric layer of native zirconium oxide or hydroxide are then removed from the surface of the substrate in order to improve the adhesion of the external layer. This ion etching of the substrate lasts 3 minutes in order to limit the heating up of the plate. 1.2. Step of Deposition of the External Layer by HiPIMS Sputtering The polarization voltage applied to the etched plate is, for example, decreased between −50 V and 0 V, in the present case to −50 V for 8 hours. As the rate of deposition is generally between 0.5 μm/h and 1 μm/h, these conditions lead to the deposition of an external chromium layer with a thickness of 6 μm. The polarization voltage of the chromium target is maintained at −800 V. Several polarization impulses are applied to the magnetron cathode according to the following characteristics: duration of an impulse=40 μs; frequency of the impulses=500 Hz; overall mean intensity=approximately 2 A; instantaneous mean intensity=approximately 100 A; overall mean power=approximately 1 kW; and instantaneous mean power for an impulse=60 kW. The surface condition of the plate coated with the external chromium layer is illustrated by FIG. 1B. By way of comparison, FIG. 1A illustrates a very different surface morphology obtained for a control Zircaloy-4 plate on which a chromium coating of the same thickness was deposited using a conventional magnetron PVD process similar to that described in example 1 of “WO2013/160587”. FIG. 1C illustrates the structure in columnar grains of the chromium-based external layer. FIG. 1D again illustrates the columnar structure of the external layer, and also the presence of an interface layer. FIG. 1E shows the zones 1, 2 and 3 of the interface region which are respectively positioned in the middle of the interface region, between the zirconium-based substrate and the interface region, and between the chromium-based external layer and the interface region. It also illustrates the zone 1 and the structural parameters which identify the intermetallic compound Zr(Fe,Cr)2 of hexagonal crystal structure of which this zone is composed. FIG. 1F illustrates the zone 2 and the structural parameters which identify the intermetallic compound ZrFe2 of cubic crystal structure of which this zone is composed. FIG. 1G illustrates the zone 3 and the structural parameters which identify the intermetallic compound ZrCr2 of cubic crystal structure of which this zone is composed. The combined FIGS. 1E, 1F and 1G thus show that the composition of the interface layer gradually changes from the interface with the Zircaloy-4 substrate toward the interface with the external chromium layer according to the following order of the intermetallic compounds: ZrFe2, Zr(Fe,Cr)2 and ZrCr2. Contrary to all expectations, the iron-containing nanometric phases ZrFe2 and Zr(Fe,Cr)2 were formed by interaction of the zirconium alloy of the substrate with the chromium. 2. Properties with Regard to the Oxidation/Hydriding 2.1. Evaluation of the Resistance to Oxidation in Accident Conditions at 1200° C. In order to evaluate its resistance to oxidation, a plate based on Zircaloy-4 provided with a single external chromium layer of 6 μm in accordance with example 1 stays for 300 seconds in a furnace in which steam brought to 1200° C. circulates. By way of comparison, the same experiment is carried out with a control Zircaloy-4 plate which has been covered with a chromium coating of the same thickness using a conventional cathode sputtering process in accordance with example 1 of “WO2013/160587”. The condition of the plates obtained on completion of this oxidation is illustrated by FIGS. 2A and 2B. FIG. 2A shows that a layer of chromium oxide Cr2O3 of limited thickness is formed. The chromium coating deposited according to the process of the state of the art thus has a partially protective nature with regard to the oxidation at 1200° C. Nevertheless, porosities and cracks are present in the residual metallic chromium layer lying under the layer of chromium oxide Cr2O3. They result among others from an exfoliation at the interface between the Zircaloy-4 substrate and the chromium coating, which reflects an embrittling of the residual metallic chromium layer of the control plate and a deterioration in its properties of resistance to oxidation, among others by loss of leaktightness with regard to the diffusion of oxygen. On the other hand, even if FIG. 2B illustrates the formation of a peripheral layer of chromium oxide Cr2O3 of greater thickness, the underlying layer of residual metallic chromium initially deposited with the HiPIMS sputtering process is for its part not damaged. This is confirmed by the absence of exfoliation at the interface with the internal Zircaloy-4 layer, and also by a very low amount of porosities. This preserved microstructure confirms the protective nature of the chromium coating with regard to the oxidation/hydriding of the zirconium alloy up to at least 1200° C., in particular its ability to limit the diffusion of the oxygen into the internal Zircaloy-4 layer. Such results are confirmed by FIGS. 3A and 3B, which illustrate the profiles of diffusion of the zirconium, chromium and oxygen measured with the electron microprobe in the thickness of the coating and in the vicinity of the Zircaloy-4/chromium interface. The absence of measurable diffusion of oxygen within the residual metal layer of the HiPIMS coating and a fortiori in the zirconium-based internal layer may be observed. For the control plate, the measurements of FIG. 3A show a significant diffusion of oxygen through the chromium coating, which exhibits a mean oxygen content of approximately 1% by weight. This diffusion continues actually within the Zircaloy-4, in which the oxygen content is of the order of 0.3% to 0.4% by weight in the first 100 μm from the chromium coating/Zircaloy-4 substrate interface. For the plate produced in accordance with the manufacturing process of the invention, the measurements of FIG. 3B show that the mean oxygen content in the Zircaloy-4 is virtually identical to the initial value of 0.14% by weight. The absence of diffusion of oxygen within the residual chromium layer and a fortiori the Zircaloy-4 substrate makes it possible to preserve the mechanical properties of the substrate, among others the residual ductility and the residual toughness. This thus provides a much better margin of protection with regard to the harmful consequences of the oxidation at 1200° C. Furthermore, similar experiments of an oxidation at 1200° C. for 300 seconds, followed by a quenching with water at ambient temperature, have confirmed such a behavior when the plate geometries are replaced by tubular cladding geometries nevertheless involving a different crystal texture: the gain in weight representative of the take up of oxygen is from 10 to 30 times less for the tube produced by the process of the invention in comparison with that measured for the tube covered with a chromium coating with a conventional cathode sputtering process. 2.2. Evaluation of the Resistance to Oxidation in Accident Conditions at 1300° C. Another sample of the plate produced in accordance with the manufacturing process of the invention stays for 5600 seconds in an equimolar oxygen/helium atmosphere brought to 1300° C. In this specific oxidation temperature domain, such an atmosphere composition is reasonably representative of the oxidation conditions under steam as, except for in a particular case (confined steam, alloy of mediocre quality, degraded surface condition, . . . ), no significant hydriding of the substrate occurs during the oxidation at 1300° C. Although these temperature conditions lie more than 100° C. above the “LOCA” regulatory limits, the photograph of FIG. 4 shows that the plate with a thickness of approximately 1 mm is not destroyed and that only a portion of the external chromium layer with an initial thickness of 15 μm to 20 μm has been oxidized to give chromic oxide Cr2O3. The Zircaloy-4 substrate predominantly exhibits a structure of Zr-ex-β type which provides most of the residual ductility of the plate. In comparison, the metallic residual internal layer of Zircaloy-4 of a control plate not coated with an external chromium layer and subjected to the same oxidation conditions for its part exhibits a wholly α-Zr(O) structure which is brittle at low temperature and which is responsible for a loss of integrity by transverse splitting. Even in oxidizing conditions at 1300° C., far above regulatory safety limits, a nuclear fuel cladding obtained by the manufacturing process of the invention may retain its mechanical integrity and exhibit a comfortable residual margin of resistance to oxidation/hydriding. 2.3. Evaluation of the Resistance to Oxidation with an Interposed Tantalum Layer A plate is produced under conditions similar to those of example 1, apart from the fact that an interposed layer with a thickness of 2 μm to 3 μm approximately composed of tantalum is deposited on the internal layer using an HiPIMS sputtering process. The deposition of the interposed tantalum layer is carried out according to conditions similar to those of the deposition of the external chromium layer for example 1, apart from the fact that the tantalum target is polarized at −800 V for an impulse duration of 25 μs. After carrying out the ion etching (according to step a) of the manufacturing process of the invention) of the interposed tantalum layer, an external chromium layer with a thickness of 4 μm is subsequently deposited on this interposed layer in accordance with step b) of the manufacturing process of the invention. By way of comparison, several corresponding control plates, apart from the fact that they are devoid of interposed tantalum layer, are produced. After a residence of 300 seconds in a furnace in which steam at 1200° C. circulates, the profiles of diffusion of the chromium of the external layer toward the internal Zircaloy-4 layer are measured from the interface between these layers. These measurements illustrated in FIG. 5 show: a very good reproducibility of the results obtained with regard to the control plates devoid of interposed tantalum layer; a diffusion of the chromium into the internal Zircaloy-4 layer which is greater for the control plates. This is because, at 1200° C., the external chromium layer is consumed at relatively similar proportion via an internal phenomenon of diffusion of the chromium toward the zirconium alloy and via the external oxidation of the chromium to give chromium oxide; the beneficial effect in oxidizing conditions at 1200° C. of the interposed tantalum layer, which acts as a diffusion barrier: in comparison with the control plate, the total amount of chromium which diffuses from the external layer toward the internal Zircaloy-4 layer is thus divided by approximately 4 and the lifetime of the external layer may optionally be multiplied by 2. Generally, the interposed layer reduces, indeed even eliminates, the phenomenon of diffusion, which increases the lifetime of the external layer and thus of the corresponding nuclear fuel cladding, amongst others in accident conditions, such as, for example, the dewatering of a spent fuel storage pool or those defined by the criteria of dimensioning accident of LOCA type. Furthermore, the impact of the interposed layer on the diffusion of the chromium toward the zirconium alloy also has the advantage of delaying the formation of a eutectic between the zirconium and the chromium above 1330° C. and thus the production of a surface liquid phase, which makes it possible to avoid or limit the potentially negative consequences which might result therefrom in the event of an incursion above ˜1320° C. 2.4. Evaluation of the Resistance to Hydriding at 1000° C. Hydriding is a phenomenon which occurs within a nuclear fuel cladding in nominal conditions or in certain accident conditions. The hydriding results from the sequence of the following reactions (1) and (2): the zirconium present in the nuclear fuel cladding is oxidized by the pressurized water or the steam according to the reaction,Zr+2H2O→ZrO2+2H2 (1) then a portion of the hydrogen thus released diffuses into the zirconium alloy of the cladding and may form a hydride with the zirconium of the cladding which has not yet oxidized, according to the reactionZr+xH→ZrHx. (2) The index “x” indicates that hydrides of variable stoichiometry may be formed, this index being in particular equal to or less than 2. According to the overall hydrogen content and/or the temperature, all or a portion of the hydrogen will precipitate, the remainder remaining in solid solution (in insertion in the α-zirconium crystal lattice). For example, at 20° C., virtually all of the hydrogen is precipitated in the form of hydrides, whereas their dissolution may be total at high temperature (typically greater than 600° C.) Hydrogen in solid solution, but especially in the form of zirconium hydride precipitate, has the disadvantage of decreasing the ductility of zirconium alloys and thus of causing embrittlement of the cladding, among others at low temperature. This embrittlement is all the more to be feared when it is desired to reach high burn-up rates as, at these rates, an increase in the proportion of zirconium oxidized according to the reaction (1) and thus in the amount of hydrides formed according to reaction (2) is found. It then generally results in the corrosion of the usual industrial alloys at levels which are harmful with regard to the criteria of safety and integrity of the cladding, and may present problems for post-service transportation and storage. Observed in normal conditions with regard to the zirconium alloys M5™ or Zirlo™ of a nuclear fuel cladding, hydriding is generally observed in accident conditions only in the vicinity of 1000° C., or toward 800° C. for longer oxidation times. This phenomenon, known as “breakaway”, is associated with an increase in the kinetics of oxidation beyond a certain critical time. It results from the appearance of cracks and/or porosities in the ZrO2 phase related to the presence of stresses generated at the Zr/ZrO2 interface probably related to the reversible transformation of tetragonal ZrO2 into monoclinic ZrO2. The consequences of this uptake of hydrogen are, in the same way as in normal conditions, an embrittlement of the material in the vicinity of 1000° C. which can result in the fracturing thereof during a quenching or after returning to low temperature. The “breakaway” phenomenon generally occurs after 5000 seconds at 1000° C. for a zirconium alloy, such as the Zircaloy-4 or M5™. In order to evaluate the resistance to the hydriding of a nuclear fuel cladding according to the invention, another sample of the plate produced in accordance with the manufacturing process of the invention stays for 15 000 seconds in an atmosphere of steam brought to 1000° C. By way of comparison, the same experiment is carried out with a control plate of Zircaloy-4 which has been covered with a chromium coating of the same thickness using a conventional cathode sputtering process in accordance with example 1 of “WO 2013/160587”. The results obtained are illustrated by FIGS. 6A and 6B. FIG. 6A shows that the control plate exhibits a local exfoliation of the external chromium layer which has partly oxidized to give Cr2O3. The content of hydrogen dissolved in the plate is measured by analysis of the gases after reduction melting in an analyzer provided for this purpose, instead of the indirect and imprecise evaluation according to example 1 of the document “WO 2013/160587”: this content is approximately 1000 ppm by weight. The Zircaloy-4 plate also comprises brittle phases of α-Zr(O) structure at low temperature due to the diffusion of oxygen into the Zircaloy-4. In point of fact, it is known that, after a quenching with water from the β domain (>900-1000° C.), Zircaloy-4 loses its residual ductility at low temperature (20-150° C.), when the hydrogen content increases in weight above 600 ppm approximately. On the other hand, FIG. 6B shows that the plate produced in accordance with the manufacturing process of the invention exhibits a layer of oxide Cr2O3 with a thickness 5 times smaller than for the control plate. Furthermore, the content of dissolved hydrogen is at most from 60 ppm to 80 ppm by weight and no phase of α-Zr(O) structure appears in the zirconium-based substrate. The plate has a significant residual ductility since its mechanical strength is for its part approximately 900 MPa and the mode of fracture is transgranular ductile dimple with an elongation at break of several %. These results confirm the very good resistance to hydriding of a nuclear fuel cladding according to the invention, for example under “post-breakaway” conditions. 3. Geometry of the Nuclear Fuel Cladding According to the Invention The nuclear fuel cladding obtained by the manufacturing process of the invention is described with reference to FIGS. 7A and 7B, in the nonlimiting specific case of a tubular geometry. According to a first embodiment of the invention, the cladding illustrated by FIG. 7A is composed of a zirconium-based internal layer (1), the internal surface of which delimits a closed volume capable of receiving the nuclear fuel. The internal layer (1) forms a substrate on which is placed an external layer (2) composed of a chromium-based protective material which makes it possible to improve the resistance to oxidation of the cladding at very high temperature. According to a second embodiment illustrated by FIG. 7B, the cladding may be provided with an interposed layer (3) placed between the internal layer (1) and the external layer (2). In this case, the combination of the internal layer (1) and of the interposed layer (3) forms the substrate. The interposed layer (3) is composed of at least one interposed material, such as, for example, tantalum, capable of preventing or limiting the diffusion of the chromium from the external layer (2) toward the internal layer (1). According to a third nonillustrated embodiment, an internal coating is placed under the internal layer (1), and thus directly facing the volume capable of receiving the nuclear fuel. It emerges from the preceding description that the process of the invention makes it possible to manufacture a nuclear fuel cladding exhibiting an improvement in the resistance to oxidation at very high temperature. The additional safety margins thus obtained make it possible among others to prevent or delay the deterioration in the cladding in the event of worsening or persistence of the accident situation.
039829944	abstract	A typical embodiment of the invention provides a means for selectively inserting and withdrawing one or more fuel rods from a fuel element cellular grid structure. The transverse stubs on one side of a long, thin bar are turned through 90.degree. to extend across the gap between mutually perpendicular grid structure plates. The extreme ends of these stubs engage the adjacent portions of the associated plates that form part of the grid cells. Pressing the stubs against the plate portions through the application of appropriate force in a longitudinal direction relative to the bar deflects the engaged plates through a sufficient distance to enable fuel rods to be inserted into, or withdrawn from, respective cells. After rod insertion, the force applied to the bar is released to enable the plates to relax and engage the fuel rods. The bars are rotated once more through 90.degree. and withdrawn from the grid structure. A similar procedure is employed to withdraw fuel rods from the grid structure.
	description	Embodiment 1 FIG. 1 schematically depicts a lithographic projection apparatus according to the invention. The apparatus comprises: a radiation system LA, Ex, IN, CO for supplying a projection beam PB of extreme ultraviolet radiation (e.g. with a wavelength of about 10 nm); a mask table MT provided with a mask holder for holding a mask MA (e.g. a reticle); a substrate table WT provided with a substrate holder for holding a substrate W (e.g. a resist-coated silicon wafer); and a projection system PL for imaging an irradiated portion of the mask MA onto a target portion C (die) of the substrate W. The radiation system comprises a radiation source LA which produces a beam of radiation. This beam is passed along various optical components,xe2x80x94e.g. beam shaping optics Ex, an integrator IN and a condenser COxe2x80x94so that the resultant beam PB is substantially collimated and uniformly intense throughout its cross-section. The beam PB subsequently intercepts the mask MA which is held in a mask holder on a mask table MT. Having passed through the mask MA, the beam PB passes through the projection system PL, which focuses the beam PB onto a target area C of the substrate W. With the aid of the interferometric displacement and measuring means IF, the substrate table WT can be moved accurately, e.g. so as to position different target areas C in the path of the beam PB. The condenser CO and the projection system PL are schematically shown as refractive components in FIG. 1. In practice, however, they will generally comprise reflective components for a beam of extreme ultraviolet radiation. The components shown in FIG. 1 should only be considered as a schematic representation. The depicted apparatus can be used in two different modes: 1. In step mode, the mask table MT is fixed, and an entire mask image is projected in one go (i.e. a single xe2x80x9cflashxe2x80x9d) onto a target area C. The substrate table WT is then shifted in the x and/or y directions so that a different target area C can be irradiated by the (stationary) beam PB; and 2. In scan mode, essentially the same scenario applies, except that a given target area C is not exposed in a single xe2x80x9cflashxe2x80x9d. Instead, the mask table MT is movable in a given direction (the so-called xe2x80x9cscan directionxe2x80x9d, e.g. the x direction) with a speed v, so that the projection beam PB is caused to scan over a mask image; concurrently, the substrate table WT is simultaneously moved in the same or opposite direction at a speed V=Mv, in which M is the magnification of the projection system PL (typically, M=xc2xc or ⅕). In this manner, a relatively large target area C can be exposed, without having to compromise on resolution. FIG. 2 schematically shows a radiation source LA of the radiation system, comprising primary and secondary jet nozzles 10 and 20 and a supply of primary and secondary gases 11 and 21 to the primary and secondary jet nozzles, respectively. Both jet nozzles are in the embodiment shown pulsed jet nozzles, in which both supply lines comprise valves which are opened at certain instants in time to supply a pulse of primary and secondary gases to the respective jet nozzles. The outflow of the primary and secondary gasses is designated by the reference numerals 15 and 25, respectively, in FIG. 2 FIG. 2 further shows a laser 30 for supplying a laser beam 31, which in general will be a pulsed laser beam. The laser beam 31 is directed to and focussed in the outflow of the primary gas from its jet nozzle 10. The frequency of the light in the laser beam and the intensity of the laser beam are chosen such that a plasma will be created in the region in which the laser beam 31 crosses the outflow of the primary gas 15. In the plasma electrons are detached from the atoms of the primary gas. After some time the electrons and nuclei will recombine under the emission of electromagnetic radiation which may have a large contribution in the extreme ultraviolet range of the radiation spectrum. The electromagnetic radiation is collected and redirected by optical elements, such as a condenser system, and not shown in FIG. 2. FIG. 4 shows a longitudinal section through the jet nozzle source for the primary and secondary gases. FIG. 6 shows a front view of the nozzle source. The primary and secondary jet nozzles are arranged co-axial, the secondary jet nozzle 20 enclosing the primary jet nozzle 10. The primary jet nozzle 10 has a circular outlet 13 and the secondary nozzle 20 has a annular outlet 23. Plungers 12 and 22 are arranged in the supply of the primary and secondary gases 11 and 21, respectively, and may be independently operated to close of their respective supply by abutting against a tapered end of the supply. In this way valves are obtained for opening and closing the respective supplies to yield a pulsed outflow of the primary and secondary gasses. However, pulsed nozzles may be obtained in various other configurations. The plungers 12, 22 are operated by means which are not shown in the drawings. When a pulse of primary gas and no pulse of secondary gas is ejected from the nozzle source, the outflow of the primary gas 15 from the jet nozzle outlet 13 will be strongly divergent. Also ejecting a pulse of secondary gas 25 results in a less divergent or even parallel or convergent outflow of the primary gas 15. An optimum outflow of the primary gas for the radiation source can be reached by varying one or more of several parameters. One of these parameters is the supply rate of secondary gas to the secondary jet nozzle 20 with respect to the supply rate of primary gas to the primary jet nozzle 10. Another parameter is the timing of the pulse of secondary gas with respect to the timing of the pulse of primary gas. It appears that an appropriately delayed pulse of primary gas with respect to the pulse of secundary gas results in a less divergent beam in case the secondary gas is a lighter gas than the primary gas as compared to a non-delayed pulse at the same flow rates of primary and secondary gasses. Other relevant parameters are the backing pressures of the gases in the nozzle source and the jet geometry. The optimum parameters will depend on the gases or liquids used and on the specific geometry of the primary and secondary jet nozzles. The primary gas of the first embodiment of the radiation source comprises krypton or xenon, which may be supplied pure or in a mixture with other (inert) gases. A xenon plasma, for instance, has been shown to emit a large contribution of extreme ultraviolet radiation. In an alternative embodiment water droplets or cryogenic liquids, such as liquid xenon, in a carrier gas may be used as a primary liquid. The secondary gas may be selected from the group comprising helium, neon, argon, krypton, methane, silane and hydrogen. In the preferred embodiment the secondary gas is hydrogen, because hydrogen hardly absorbs extreme ultraviolet radiation. Since hydrogen has favorable absorption characteristics with respect to extreme ultraviolet radiation, a very large outflow of hydrogen from the secondary nozzle can be employed, resulting in a high local density in the outflow. A lighter secondary gas is expected to yield a worse confinement of xenon as a primary gas with respect to a heavier secondary gas due to a smaller transfer of momentum in a collision. The much larger outflow and higher pressure of hydrogen which can be employed in the radiation source according to the invention overcompensates for the smaller mass of hydrogen with respect other secondary gasses, due to the considerably larger local pressures which can be tolerated. With the above jet nozzles an approximate parallel outflow of the primary gas from the primary jet nozzle 10 may be obtained. The laser beam 31 of the laser 30 for creating a plasma in the primary gas is crossed with and focussed in the outflow of primary gas at a distance from the nozzle outlet which is sufficient not to produce debris from the jet nozzle by interaction of the plasma with the nozzle. Embodiment 2 FIG. 5 schematically shows another configuration for the nozzle source of the radiation source according to a second embodiment of the invention. The second embodiment differs from the first in that it comprises a continuous nozzle source for a continuous outflow of primary and secondary gasses from the respective outlets 13 and 23 of the primary and secondary jet nozzles 10 and 20. An embodiment comprising a continuous jet nozzle for the primary jet nozzle and a pulsed nozzle for the secondary jet nozzle, or vice versa, may also be envisaged. Embodiment 3 FIG. 7 schematically shows a front view of a nozzle source of a radiation source according to a third embodiment of the invention. The third embodiment differs from the first and second embodiment in that the secondary nozzle is positioned at one side of the primary nozzle. The figure shows the outlets 13 and 23 of the primary and secondary jet nozzles, respectively. The divergence of the outflow from the primary nozzle may for this embodiment only controlled at this one side. Shielding of a plasma created in the primary gas will also only be present at this one side. An embodiment in which the secondary jet nozzle partly encloses the primary jet nozzle, or having, for instance, outlets of the secondary jet nozzle on two opposite sides of the outlet of the primary jet nozzle may also be envisaged. Embodiment 4 FIG. 3 schematically shows another embodiment of the radiation source according to the invention. The fourth embodiment differs from the first, second and third embodiment in that the plasma in the outflow of the primary gas from the nozzle is created by an electrical discharge in the primary gas ejected from the primary nozzle 10. The discharge is generated in between electrodes 40 which are connected to a high voltage source 41. However, other means for creating a plasma in the outflow of the primary gas may also be employed. Whilst specific embodiments of the invention are disclosed above it will be appreciated that the invention may be practiced other than described. The description is not intended to limit the invention.
058928090	summary	This application is based on provisional application Ser. No. 60/058,409 which was filed Sep. 10, 1997 and bears the same title. BACKGROUND--FIELD OF THE INVENTION This invention relates to instruments for microanalysis in which a small region at the surface of a specimen is bombarded with monochromatic x-rays. Measurement of the scattered or emitted x-rays or charged particles is used to characterize the material making up the specimen. BACKGROUND--PRIOR ART It is well known that microanalysis can be performed by bombarding a small region with a focused beam of charged particles or electromagnetic radiation. At the present time, most commercially available instruments for accomplishing this in the laboratory use electron or ion beams for excitation because of the ease with which charged particles can be focused. The use of X-rays has been limited up to now because of the difficulties of focusing X-rays. However, some success has been achieved in localizing the analysis in instruments for X-ray fluorescence analysis by the use of apertures for the X-ray beam or by using total reflection inside a capillary or a monolithic polycapillary optic. Unfortunately these approaches do not provide monochromatic radiation. This results in the detection limits in microanalysis by X-ray fluorescence being degraded by the presence of background due to scattering by the specimen of the X-ray continuum from the source. The use of doubly curved X-ray diffractors has been investigated for a long time as a way to obtain focussing of a monochromatic X-ray beam for microanalysis. A suitable geometry for point-to-point focussing ideally involves rotating either the Johann or the Johansson geometries about a line joining the source and image. In the Johann geometry, a curved crystal has lattice planes curved to a radius 2R and the source and its image lie on a focal circle of radius R that is tangent to the crystal's surface. In the Johansson geometry, the crystal planes are similarly shaped but the crystal surface is curved to a radius of R. Some of the other approaches that have been proposed have been referenced in Wittry's U.S. Pat. No. 4,599,741 issued in 1986, for example patents by Berreman (1958), Hammond (1973), Furnas (1975), and Carrol (1980). A paper describing the focusing of monochromatic radiation was presented by P. S. Ong at the 29th Annual conference of the Microbeam Analysis Society in 1974. This paper described an experimental set-up for X-ray fluorescence analysis that had been constructed using a singly curved diffractor and another one under construction that would use a doubly curved diffractor. Both diffractors employed the Johansson geometry in the plane of the focal circle but satisfactory operation of the doubly curved diffractor was never reported subsequently. Practical development of point-focusing diffractors for obtaining a small spot of monochromatic X-radiation was accomplished by Larson and Palmberg as described in U. S. Pat. Nos. 5,315,113 and 5,444,242. Their diffractor was used for aluminum K.sub..alpha. radiation in a commercial instrument for ESCA. Because of the relatively long wavelength of radiation needed for ESCA, it was possible to use a Bragg angle close to 90 degrees. The large Bragg angle facilitates implementation of the point focussing diffractor which was made of quartz. For X-ray fluorescence analysis, the use of Bragg angles close to 90 degrees is not usually possible. In this application, the desired radiation has sufficiently small wavelengths that no commonly available crystal materials yield high diffraction efficiency at large Bragg angles. Also, for this application, it appeared that the Johansson geometry would be needed in the plane of the focal circle in order to obtain a sufficiently high collection solid angle. As a result, much of the work on point-focusing diffractors during the past 10 years was concentrated on trying to construct diffractors that had this geometry and also subtended a large solid angle at the source. The difficulties of achieving this configuration greatly inhibited the development of practical diffractors for X-ray microprobe fluorescence analysis. A different approach which was used recently by Chen and Wittry led to a successful demonstration of microprobe X-ray fluorescence analysis as described in the Journal of Applied Physics vol. 84, pp 1064-1073 (1998). In this approach, the emphasis was on utilizing a diffractor that was more accurately made and aligned, rather than one that had the largest possible collection solid angle. It was found that a small toroidally curved diffractor based on the Johann geometry when used with a 3 watt microfocus X-ray source provided enough intensity in a focused beam of Cu K.sub..alpha. radiation for X-ray fluorescence analysis. The successful results were due to a number of factors. First, work by Wittry and his coworkers provided a theoretical basis for understanding the precision required in the fabrication and alignment of curved diffractors (Journal of Applied Physics: vol. 67, pp 1633-38, 1990; vol. 71, pp 564-8, 1992; vol. 73, pp 601-07, 1993; vol. 74, pp 2999-3008, 1993). Second, detection of the fluorescence-excited radiation by an energy dispersive spectrometer with its high collection solid angle reduced the X-ray microprobe intensity required compared with the intensity required if a wavelength dispersive spectrometer were used. Finally, comparison data for X-ray fluorescence analysis obtained with an X-ray microprobe based on the use of single and polycapillary optics were available to indicate that the diffractor was capable of providing superior performance than these other methods. The present invention is an improvement and simplification over the Wittry U.S. Pat. No. 4,599,741. As in U.S. Pat. No. 4,599,741, several diffractors are provided. But selection of one of the multiple diffractors can be accomplished without the need for the diffractors to move. This results in simpler and more reliable operation. OBJECTIVES AND ADVANTAGES OF THE PRESENT INVENTION The objectives of the present invention are to provide a simplified system whereby microanalysis with high sensitivity and low detection limits for impurities can be performed in the laboratory. These characteristics are obtained by the used of monochromatic x-rays from characteristic x-ray lines which can be selected to optimize the photon energy for excitation of particular ranges of energy levels of elements in the specimen. In contrast to similar systems that have previously been described, the present system is more compact, less expensive to manufacture, and easier to operate.
	description	The present application claims benefit of priority to U.S. Provisional Patent Application No. 62/438,323, entitled “Passive Reactivity Control in a Nuclear Fission Reactor” and filed on Dec. 22, 2016, which is specifically incorporated by reference herein for all that it discloses or teaches. A fast spectrum nuclear fission reactor (“a fast neutron reactor”), such as a sodium fast reactor, generally includes a reactor vessel containing a nuclear reactor core. The nuclear reactor core includes an array of device locations for placement of fuel assembly devices and other reactor support and control devices. Fissile nuclear fuel within the nuclear reactor core is subjected to neutron collisions that result in fission reactions. In a breed-and-burn fast neutron reactor, a fission chain reaction yields “fast spectrum neutrons” that, in turn, collide with fertile nuclear fuel, thereby transmuting (“breeding”) the fertile nuclear fuel into fissile nuclear fuel. Liquid coolant flows through the nuclear reactor core, absorbing thermal energy from the nuclear fission reactions that occur in the nuclear reactor core. The heated coolant then passes to a heat exchanger and a steam generator, transferring the absorbed thermal energy to steam in order to drive a turbine that generates electricity. Design of such nuclear reactors involves combinations of materials, structures, and control systems to achieve desirable operational parameters, including nuclear reactor core stability, efficient thermal generation, long-term structural integrity, etc. The described technology provides a fast-acting passive reactivity control nuclear fuel device that functions by thermal expansion of a liquid/molten nuclear fuel under high neutron flux and introduces a negative power feedback for a nuclear fission fast reactor. A nuclear reactor includes a passive reactivity control nuclear fuel device located in a nuclear reactor core. The passive reactivity control nuclear fuel device includes a multiple-walled fuel chamber including an outer wall chamber and an inner wall chamber contained within the outer wall chamber. The inner wall chamber is positioned within the outer wall chamber to hold nuclear fuel in a molten fuel state within a high neutron importance region of the nuclear reactor core. The inner wall chamber is further configured to allow at least a portion of the nuclear fuel to move in a molten fuel state to a lower neutron importance region of the nuclear reactor core while the molten nuclear fuel remains within the inner wall chamber as the internal temperature of the inner wall chamber satisfies a negative reactivity feedback expansion temperature condition. A duct contains the multiple-walled fuel chamber and flows a heat conducting fluid through the duct and in thermal communication with the outer wall chamber. The heat conducting fluid operates as a coolant, and the flow temperature of the heat conducting fluid is typically less than the temperatures inside the inner wall chamber during a nuclear reaction. This Summary is provided to introduce a selection of concepts in a simplified form that are further described below in the Detailed Description. This Summary is not intended to identify key features or essential features of the claimed subject matter, nor is it intended to be used to limit the scope of the claimed subject matter. Other implementations are also described and recited herein. Fast nuclear reactors are typically designed to increase the utilization efficiency of nuclear fuel (e.g., uranium, plutonium, thorium) and to limit moderations of neutrons in fission reactions. In many implementations, fast nuclear reactors can capture significantly more of the energy potentially available in natural uranium, for example, than typical light-water reactors. Nevertheless, the described technology can be employed in different types of nuclear reactors, including light water reactors. A particular classification of fast nuclear reactor, referred to as a “breed-and-burn” fast reactor, includes a nuclear reactor capable of generating (“breeding”) more fissile nuclear fuel than it consumes. For example, the neutron economy is high enough to breed more fissile nuclear fuel from fertile nuclear reactor fuel, such as uranium-238 nuclear or thorium-232 fuel, than it burns. The “burning” is referred to as “burnup” or “fuel utilization” and represents a measure of how much energy is extracted from the nuclear fuel. Higher burnup typically reduces the amount of nuclear waste remaining after the nuclear fission reaction terminates. Another particular classification of a fast nuclear reactor is based on the type of nuclear fuel used in the nuclear fission reaction. A metal fuel fast nuclear reactor employs a metal nuclear fuel, which has an advantage of high heat conductivity and a faster neutron spectrum than in ceramic-fueled fast reactors. Metal fuels can exhibit a high fissile atom density and are normally alloyed, although pure uranium metal has been used in some implementations. In a fast nuclear reactor, minor actinides produced by neutron capture of uranium and plutonium can be used as a metal fuel. A metal actinide fuel is typically an alloy of zirconium, uranium, plutonium, and minor actinides. FIG. 1 illustrates a partial-cutaway perspective view of an example nuclear fission reactor 100 with a nuclear reactor core 102 containing one or more passive reactivity control nuclear fuel devices, such as a passive reactivity control nuclear fuel device 104. Other elements within the nuclear reactor core 102 include nuclear fuel assembly devices (such as a nuclear fuel assembly device 106) and movable reactivity control assembly devices (such as a movable reactivity control assembly device 108). Certain structures of the example nuclear fission reactor 100 have been omitted, such as coolant circulation loops, coolant pumps, heat exchangers, reactor coolant system, etc., in order to simplify the drawing. Accordingly, it should be understood that the example nuclear fission reactor 100 may include different and/or additional structures not shown in FIG. 1. Implementations of the example nuclear fission reactor 100 may be sized for any application, as desired. For example, various implementations of the example nuclear fission reactor 100 may be used in low power (˜5 Mega Watt thermal) to around 1000 Mega Watt thermal) applications and large power (around 1000 Mega Watt thermal and above) applications, as desired. In one implementation, the example nuclear fission reactor 100 is a fast spectrum nuclear fission reactor having an average neutron energy of greater than or equal to 0.1 MeV, although other configurations are contemplated. It should be understood, however, that the described technology can be employed in different types of nuclear reactors, including light water reactors. Some of the structural components of the nuclear reactor core 102 may be made of refractory metals, such as tantalum (Ta), tungsten (W), rhenium (Re), or carbon composites, ceramics, or the like. These materials may be selected to address the high temperatures at which the nuclear reactor core 102 typically operates. Structural characteristics of these materials, including creep resistances, mechanical workability, corrosion resistance, etc., may also be relevant to selection. Such structural components define an array of device locations within the nuclear reactor core 102. The nuclear reactor core 102 is disposed in a reactor vessel 110 containing a pool of heat conducting fluid, such as a coolant. For example, in various implementations, a reactor coolant system (not shown) includes a pool of liquid sodium coolant (not shown) disposed in the reactor vessel 110. In such cases, the nuclear reactor core 102 is submerged in the pool of liquid sodium coolant within the reactor vessel 110. The reactor vessel 110 is surrounded by a containment vessel 116 that helps prevent loss of the liquid sodium coolant in the unlikely case of a leak from the reactor vessel 110. In alternative implementations, liquid coolant can flow through coolant loops throughout the nuclear fission reactor 100. The nuclear reactor core 102 contains the array of device locations for receiving various reactor core devices, such as nuclear fuel assembly devices, movable reactivity control assembly devices, and passive reactivity control fuel assembly devices within the central core region 112. An in-vessel handling system 114 is positioned near the top of the reactor vessel 110 and is configured, under control of a reactivity control system (not shown), to shuffle individual reactor core devices in and/or out of the device locations within the nuclear reactor core 102. Some reactor core devices may be removable from the nuclear reactor core 102, while other reactor core devices may not be removable from the nuclear reactor core 102. The nuclear reactor core 102 can include a neutron source and a larger nuclear fission reaction region. The neutron source provides thermal neutrons to initiate a fission reaction in the fissile nuclear fuel. The larger nuclear fission reaction region may contain thorium (Th) or uranium (U) fuel and functions on the general principles of fast neutron spectrum fission breeding. In one implementation, the nuclear fuel within a nuclear fuel assembly device may be contained within fissile nuclear fuel assembly devices or fertile nuclear fuel assembly devices. The difference between fissile nuclear fuel assembly devices or fertile nuclear fuel assembly devices is effectively the enrichment level of the nuclear fuel, which can change over time within the nuclear reactor core 102. Structurally, fissile nuclear fuel assembly devices or fertile nuclear fuel assembly devices can be identical in some implementations. The nuclear fuel assembly device 106 in the nuclear reactor core 102 can include a solid hexagonal duct or tube surrounding a plurality of fuel elements, such as fuel pins, which are organized into the nuclear fuel assembly device 106. Non-hexagonal ducts may also be used in some implementations. The ducts in a nuclear fuel assembly device 106 allow coolant to flow past the fuel pins through interstitial gaps between adjacent duct walls. Each duct also allows individual assembly orificing, provides structural support for the fuel bundle, and transmits handling loads from a handling socket to an inlet nozzle. Fuel pins typically consist of multiple nuclear fuel rods (such as uranium, plutonium or thorium) surrounded by a liner and cladding (and sometimes an additional barrier), which prevents radioactive material from entering the coolant stream. Individual pins of a nuclear fuel assembly device 106 in the nuclear reactor core 102 can contain fissile nuclear fuel or fertile nuclear fuel depending on the original nuclear fuel rod material inserted into the pin and the state of breeding within the pin. The movable reactivity control assembly device 108 can be inserted into and/or removed from the central core region 112 by the in-vessel handling system 114 to provide real-time control of the fission process, balancing the needs of keeping the fission chain reaction active while preventing the fission chain reaction from accelerating beyond control. The state of a fission chain reaction is represented by an effective multiplication factor, k, which indicates the total number of fission events during successive cycles of the chain reaction. When a reactor is in a steady state (i.e., each individual fission event triggers exactly one subsequent fission event), k equals 1. If k>1, the reactor is supercritical and the reaction rate will accelerate. If k<1, the reactor is subcritical and the fission rate will decrease. Conditions within the central core region 112 change over time. Hence, movable reactivity control assemblies may be used to adjust the multiplication factor of the fission chain reaction as conditions change. Such assemblies can be moved to and from different locations in the nuclear reactor core to influence the multiplication factor of the fission chain reaction. In addition, the axial position (e.g., up/down) position of such assemblies can also be adjusted to influence the multiplication factor of the fission chain reaction. The movable reactivity control assembly device 108 is a highly effective neutron absorbing mechanical structure that can be actively inserted into or removed from the central core region 112 while the fission process is occurring. A movable reactivity control assembly device includes chemical elements of a sufficiently high neutron capture cross-section to absorb neutrons in the energy range of the nuclear fission reaction, as measured by its absorption cross-section. As such, the movable reactivity control assembly device 108 influences the number of neutrons available to cause a fission reaction within the nuclear reactor core 102, thereby controlling the fission rate of the fissile nuclear fuel within the nuclear reactor core 102. Example materials used in movable reactivity control assembly devices of the nuclear fission reactor 100 include without limitation boron carbide or an alloy of silver, indium, and cadmium, europium, or a hafnium-hydride. By controlling the portion of the movable reactivity control assembly device 108 (as well as the number of movable reactivity control assemblies) that interacts with the fission reaction within the central core region 112, the multiplication factor can be tuned to maintain reactor criticality. Accordingly, a movable reactivity control assembly device 108 represents an adjustable parameter for controlling the nuclear fission reaction. The passive reactivity control nuclear fuel device 104 contains nuclear fuel material in a solid fuel state that can achieve a molten fuel state when the nuclear fuel material temperature exceeds the melting temperature of the nuclear fuel material (e.g., satisfying a nuclear fuel melting temperature condition). Example nuclear fuel material may include fuel salts, eutectics, pure metals, etc. In some implementations, the nuclear fuel material will be alloyed with a bonding/carrier material (e.g., Mg) such that the alloyed nuclear fuel and bonding/carrier material melts at the same temperature and time, although other implementations may involve the nuclear fuel material melting before the bonding/carrier material as temperature rises. The molten fuel within the passive reactivity control nuclear fuel device 104 can then move within the passive reactivity control nuclear fuel device 104. By moving in and out of a high neutron importance region of the nuclear reactor core 102, the molten nuclear fuel material can increase or decrease, respectively, the reactivity within the nuclear reactor core 102. Neutron importance represents a magnitude of the contribution of a neutron to power generated by a nuclear reactor. When the molten fuel satisfies a negative reactivity feedback fuel expansion temperature condition (e.g., a temperature high enough to cause the expansion of the nuclear fuel to move a portion of the molten fuel into a lower neutron importance region of the nuclear reactor core 102), the reduction in volume of fissile material in the high neutron importance region provides negative reactivity feedback in the nuclear reactor core 102. For example, when fissile nuclear fuel within the passive reactivity control nuclear fuel device 104 moves into a lower neutron importance region of the nuclear reactor core 102 (such as through thermal expansion of molten fuel as temperature increases), reactivity within the nuclear reactor core 102 decreases. Movement of the molten fuel within the passive reactivity control nuclear fuel device influences the reactivity of the nuclear reactor core 102. Passive reactivity control nuclear fuel devices can also be moved to and from different locations in the nuclear reactor core 102 to influence the reactivity of the nuclear reactor core 102. In addition, the axial position (e.g., up/down) position of such assemblies can also be adjusted to influence the reactivity of the nuclear reactor core 102. It should be understood that the molten fuel can also act as a neutron poison within the nuclear reactor. Accordingly, moving molten fuel out of a high neutron importance region and into a lower neutron importance region of a nuclear reactor can also provide some level of positive reactivity feedback. Nevertheless, the passive reactivity control nuclear fuel device 104 can be designed such that the negative reactivity feedback of moving the molten fuel to a lower neutron importance region exceeds the positive reactivity feedback of removing the molten fuel (as a poison) from the high neutron importance region. In contrast, when the molten fuel no longer satisfies a negative reactivity feedback fuel expansion temperature condition (e.g., a temperature that is no longer high enough to cause the expansion of the nuclear fuel to move a portion of the molten fuel into a lower neutron importance region of the nuclear reactor core 102), the increase in volume of fissile material in the high neutron importance region provides positive reactivity feedback in the nuclear reactor core 102. For example, when fissile nuclear fuel within the passive reactivity control nuclear fuel device 104 moves back into the high neutron importance region of the nuclear reactor core 102 (such as through densification of molten fuel as temperature decreases), reactivity within the nuclear reactor core 102 increases. Accordingly, in one implementation, a passive reactivity control nuclear fuel device, such as the passive reactivity control nuclear fuel device 104, can provide negative feedback to the fission reaction as the temperature of the nuclear fuel material increases and positive feedback to the fission reaction as the temperature of the nuclear fuel material decreases. FIG. 2 illustrates a cross-sectional view of an example nuclear reactor core 202 having an array of locations (such as device location 204) of nuclear reactor core devices, including passive reactivity control assembly devices. It should be understood that a fast nuclear reactor core typically has more device locations and devices than shown in the example core of FIG. 2, but a reduced number of device locations and devices is shown to facilitate description and illustration. Each device is inserted into a structurally-defined device location within the array. Reflector devices, such as a replaceable radiation reflector device at the device location 204, and permanent radiation reflector material 214 are positioned at the boundary of the central reactor core region to reflect neutrons back into the central reactor core region. Nuclear fuel assembly devices, such as a nuclear fuel assembly device 206 and a nuclear fuel assembly device 208, occupy the majority of the device locations in the nuclear reactor core 202. As the nuclear reaction progresses, an atom of fertile nuclear fuel can be converted or transmuted to fissile nuclear fuel by the capture of a neutron within a certain energy range. For example, a fertile nucleus, such as a U-238 nucleus, can capture fast neutrons and be transmuted to a fissile nucleus, such as Pu-239, by beta-decay. Meanwhile, in this case, the Pu-239 nucleus can capture a neutron, resulting in a fission reaction that yields multiple fast neutrons. This neutron multiplication with each fission reaction provides enough neutrons for the transmutation of new fissile nuclear fuel from the fertile nuclear fuel. As such, the fission reaction drives the breeding of new fissile nuclear fuel from fertile nuclear fuel that it consumes. The reactivity of the fission reaction can be controlled to some extent by one or more movable reactivity control assembly devices, such as a movable reactivity control assembly device 210. By introducing passive reactivity control nuclear fuel devices, such as passive reactivity control nuclear fuel device 212 into the nuclear reactor core 202, the reactivity of the fission reaction can be decreased after the temperature of the nuclear fuel satisfies a negative reactivity feedback expansion temperature condition (e.g., the temperature of the fuel has exceeded a particular temperature threshold) by thermally expanding molten fuel to move into a lower neutron importance region of the nuclear reactor core 202. In contrast, the reactivity of the fission reaction can be increased after the temperature of the nuclear fuel fails to satisfy a negative reactivity feedback expansion temperature condition (e.g., the temperature of the fuel has decreased below a particular temperature threshold) by densification of molten fuel to move back into the high neutron importance region of the nuclear reactor core 202. In one implementation, eighteen passive reactivity control nuclear fuel devices may be employed throughout the nuclear reactor core 202, although different reactor designs may benefit from a greater or lesser number of passive reactivity control nuclear fuel devices. FIG. 3 illustrates a cross-sectional view of an example passive reactivity control nuclear fuel device 300 containing nuclear fuel 318 in a solid fuel state within a high neutron importance region of a nuclear reactor core. The passive reactivity control nuclear fuel device 300 includes a duct 304 through which a heat conducting fluid (e.g., coolant 306, such as molten sodium) can flow. The duct 304 may be manufactured from HT9 stainless steel although other materials may be employed. The duct 304 also contains a multiple-walled fuel chamber 308, which is an example of a multiple-walled fuel chamber. In one implementation, an outer wall chamber 310 of the multiple-walled fuel chamber 308 primarily includes HT9 stainless steel, although other materials may be employed, at least in part. Further, in one implementation, an inner wall chamber 312 of the multiple-walled fuel chamber 308 primarily includes molybdenum, although other materials may be employed, at least in part. In one implementation, the inner wall chamber 312 is fitted with one or more thermally conductive contacts 314, which may also be formed from HT9 stainless steel or other materials. The contacts 314 can improve thermal communication between the inner wall chamber 312 and the outer wall chamber 310 (and therefore, the coolant 306) as one or more of the contacts 314 approach and/or physically contact the outer wall chamber 310. Corresponding contacts 315 on the interior of the outer wall chamber 310 may also be provided, as shown. The gap region 316 between the inner wall chamber 312 and the outer wall chamber 310 may contain a vacuum or a gas, such as a tag gas that can be detected if the outer wall chamber 310 mechanically fails and potentially compromises the molten fuel storage. The gap region 316 between the outer wall chamber 310 and the inner wall chamber 312 thermally isolates the inner wall chamber 312, such as from thermal communication with the coolant. The flow temperature of the heat conducting fluid or coolant is typically less than the temperatures inside the inner wall chamber during a nuclear reaction. The multiple-walled fuel chamber 308 contains nuclear fuel 318 within the inner wall chamber 312. In the state shown in FIG. 3, the nuclear fuel 318 is in an initial state (e.g., at reactor start-up or initial insertion into the nuclear reactor core). In one implementation, the nuclear fuel 318 is in a solid porous form including a combination of fertile nuclear fuel and a bonding material (which may or may not be neutronically translucent). For example, the bonding material may perform as a nuclear translucent carrier medium (e.g., when melted). In FIG. 3, the nuclear fuel 318 remains in a solid state because the temperature of the nuclear fuel 318 has not exceeded its melting temperature (e.g., not satisfying a nuclear fuel melting temperature condition). Other example forms of the nuclear fuel materials may include without limitation a solid non-porous slug, a powder, a slurry, and a suspension. For example, the fertile nuclear fuel may include 238U (uranium), such as a uranium foam, and the bonding material may include magnesium (Mg). Other material types and structures may be employed. (In some implementations, the nuclear fuel 318 may also include a quantity of fissile nuclear fuel, such as 239Pu (plutonium), particularly if the passive reactivity control nuclear fuel device 300 is intended to undergo fission early in its life cycle (e.g., at startup).) In a breed-and-burn fast reactor, however, the fertile nuclear fuel can eventually be transmuted into fissile nuclear fuel that can undergo fission. A plenum region 320 is also located within the inner wall chamber 312 to receive gaseous fission products as well as molten fuel as the fuel temperature rises and the fuel material expands into the plenum or plenum region 320. As shown, at least a portion of the plenum region 320 is located outside the high neutron importance region. Thus, as molten fuel expands into the plenum region 320 (and out of the high neutron importance region), reactivity in the nuclear reactor core decreases. One method of increasing the internal temperature of the nuclear fuel 318 involves initiating and sustaining a nuclear fission reaction within the example passive reactivity control nuclear fuel device 300. Within the nuclear reactor core, neutrons within the nuclear reactor core can collide with fissile nuclear fuel residing within the example passive reactivity control nuclear fuel device 300 to produce a nuclear fission reaction sufficient to increase the internal temperature of the example passive reactivity control nuclear fuel device 300 to exceed the melting temperature of the bonding material. Once the bonding material is melted into a carrier material, the fissile nuclear fuel can go into solution with the carrier material to provide a fissile nuclear fuel solution (the molten fuel). Other methods of increasing the temperature of the nuclear fuel 318 may be employed. FIG. 4A illustrates a perspective view of a passive reactivity control nuclear fuel device 400 along a long axis 402, and FIG. 4B illustrates a cross-sectional view of the passive reactivity control nuclear fuel device 400 along the long axis 402. In one implementation, an outer structural wall 404 of the passive reactivity control nuclear fuel device 400 forms a duct for containing a flow of coolant 406, such as liquid sodium. The outer structure wall 404 of the duct may be manufactured from HT9 stainless steel, although other materials may be employed. Within the outer structural wall 404 is located a multiple-walled fuel chamber 408 having an outer wall chamber 410 and an inner wall chamber 412. In one implementation, an outer wall chamber 410 of the multiple-walled fuel chamber 408 primarily includes HT9 stainless steel, although other materials may be employed, at least in part. Further, in one implementation, an inner wall chamber 412 of the multiple-walled fuel chamber 408 primarily includes molybdenum, although other materials may be employed, at least in part. In one implementation, the inner wall chamber 412 is fitted with one or more thermally conductive contacts, such as contact 414, which may also be formed from HT9 stainless steel or other materials. The contacts 414 can improve thermal communication between the inner wall chamber 412 and the outer wall chamber 410 (and therefore, the coolant 406) as one or more of the contacts 414 approach and/or physically contact the outer wall chamber 410. Corresponding contacts, such as contact 415, on the interior of the outer wall chamber 410 may also be provided, as shown. The gap region 416 between the inner wall chamber 412 and the outer wall chamber 410 may contain a vacuum or a gas, such as a tag gas that can be detected if the outer wall chamber 410 mechanically fails and potentially compromises the molten fuel storage. The gap region 416 between the outer wall chamber 410 and the inner wall chamber 412 thermally isolates the inner wall chamber 412, such as from thermal communication with the coolant. The flow temperature of the heat conducting fluid or coolant is typically less than the temperatures inside the inner wall chamber during a nuclear reaction. The inner wall chamber 412 contains nuclear fuel 418. In one implementation, the nuclear fuel 418 is in a solid porous form including a combination of fertile nuclear fuel and a bonding material. For example, the fertile nuclear fuel may include 238U (uranium), such as a uranium foam, and the bonding material may include magnesium (Mg). Other material types and structures may be employed. (In one implementation of this initial state, the nuclear fuel 418 may also include a quantity of fissile nuclear fuel, such as 239Pu (plutonium), particularly if the passive reactivity control nuclear fuel device 400 is intended to undergo fission early in its life cycle (e.g., at startup).) In a breed-and-burn fast reactor, however, the fertile nuclear fuel can eventually be transmuted into fissile nuclear fuel that can undergo fission. FIG. 5 illustrates a cross-sectional view of an example passive reactivity control nuclear fuel device 500 containing molten fuel 501 within a high neutron importance region of a nuclear reactor core. A solid porous fuel slug as nuclear fuel 518 resides in the example passive reactivity control nuclear fuel device 500, such that the molten fuel 501 can pass through the pores of the nuclear fuel 518. The passive reactivity control nuclear fuel device 500 includes a duct 504 through which coolant 506, such as molten sodium, can flow. The duct 504 may be manufactured from HT9 stainless steel although other materials may be employed. The duct 504 also contains a multiple-walled fuel chamber 508. In one implementation, an outer wall chamber 510 of the multiple-walled fuel chamber 508 primarily includes HT9 stainless steel, although other materials may be employed, at least in part. Further, in one implementation, an inner wall chamber 512 of the multiple-walled fuel chamber 508 primarily includes molybdenum, although other materials may be employed, at least in part. In one implementation, the inner wall chamber 512 is fitted with one or more thermally conductive contacts 514, which may also be formed from HT9 stainless steel or other materials. The contacts 514 can improve thermal communication between the inner wall chamber 512 and the outer wall chamber 510 (and therefore, the coolant 506) as one or more of the contacts 514 approach and/or physically contact the outer wall chamber 510. Corresponding contacts 515 on the interior of the outer wall chamber 510 may also be provided, as shown. The gap region 516 between the inner wall chamber 512 and the outer wall chamber 510 may contain a vacuum or a gas, such as a tag gas that can be detected if the outer wall chamber 510 mechanically fails and potentially compromises the molten fuel storage. The gap region 516 between the outer wall chamber 510 and the inner wall chamber 512 thermally isolates the inner wall chamber 512, such as from thermal communication with the coolant. The flow temperature of the heat conducting fluid or coolant is typically less than the temperatures inside the inner wall chamber during a nuclear reaction. The multiple-walled fuel chamber 508 contains nuclear fuel 518 within the inner wall chamber 512. In the state shown in FIG. 5, the nuclear fuel 518 is in an intermediate state in which the temperature of the nuclear fuel 518 has exceeded the melting temperature of the nuclear fuel 518 that has caused all or some portion of the nuclear fuel 518 to turn into molten fuel 501 (e.g., a solid portion of the fertile nuclear fuel can remain in a solid state). In one implementation, the nuclear fuel 518 can include a combination of a fissile nuclear fuel and a carrier material or a combination of fertile nuclear fuel, fissile nuclear fuel, and a carrier material. For example, the fertile nuclear fuel may include 238U (uranium), such as a uranium foam, and the bonding material may include magnesium (Mg), and the fissile nuclear fuel may include 239Pu (plutonium) in a Mg—Pu solution. Other material types and structures may be employed. In a breed-and-burn fast reactor, however, the fertile nuclear fuel can eventually be transmuted into fissile nuclear fuel that can undergo fission. It is noted that the nuclear fuel 518 includes solid fertile nuclear fuel and molten fissile fuel (e.g., a solution of fissile nuclear fuel and the carrier material). The molten state of the fissile fuel demonstrates that the temperature of the nuclear fuel 518 has exceeded the melting temperature of the nuclear fuel 518 and the position of the molten fuel 501 within and not outside the high neutron importance region of the nuclear reactor core demonstrates that the temperature of the nuclear fuel 518 has not satisfied the negative reactivity feedback expansion temperature condition. A plenum region 520 is also located within the inner wall chamber 512 to receive gaseous fission products as well as molten fuel as the fuel temperature rises and the fuel material expands into the plenum or plenum region 520. As shown, at least a portion of the plenum region 520 is located outside the high neutron importance region. Thus, as molten fuel expands into the plenum region 520 (and out of the high neutron importance region), reactivity in the nuclear reactor core decreases. In one implementation, Mg may be used as the bonding material. Mg has a melting point of about 650° C. In contrast, one configuration of a nuclear reactor includes an inlet coolant temperature of about 360° C., which is generally insufficient to melt the Mg. The internal temperature of the example passive reactivity control nuclear fuel device 500, and therefore the bonding material, can be increased to exceed the melting temperature of Mg, with some buffer to ensure that the Mg is molten. The molten Mg forms a carrier material, which can act as a solvent for fissile nuclear fuel within the example passive reactivity control nuclear fuel device 500, such as 239Pu, providing a molten fissile nuclear fuel solution (Mg—Pu). One method of increasing the internal temperature of the nuclear fuel 518 involves sustaining a nuclear fission reaction within the example passive reactivity control nuclear fuel device 500. The increased internal temperature of the nuclear fuel 518 can transition the nuclear fuel 518 from a solid fuel state to a molten fuel state. The fissile nuclear fuel may be initially stored within the inner wall chamber 512 of the example passive reactivity control nuclear fuel device 500. Alternatively, or additionally, the fissile nuclear fuel may breed up from the solid porous fertile nuclear fuel by fast neutrons resulting from fission reactions elsewhere within the operating nuclear reactor core. FIG. 6 illustrates a cross-sectional view of an example passive reactivity control nuclear fuel device 600 containing molten fuel 601 expanding outside a high neutron importance region of a nuclear reactor core. A solid porous fuel slug as nuclear fuel 618 also resides in the example passive reactivity control nuclear fuel device 600, such that the molten fuel 601 can pass through the pores of the nuclear fuel 618. The passive reactivity control nuclear fuel device 600 includes a duct 604 through which coolant 606, such as molten sodium, can flow. The duct 604 may be manufactured from HT9 stainless steel although other materials may be employed. The duct 604 also contains a multiple-walled fuel chamber 608. In one implementation, an outer wall chamber 610 of the multiple-walled fuel chamber 608 primarily includes HT9 stainless steel, although other materials may be employed, at least in part. Further, in one implementation, an inner wall chamber 612 of the multiple-walled fuel chamber 608 primarily includes molybdenum, although other materials may be employed, at least in part. In one implementation, the inner wall chamber 612 is fitted with one or more thermally conductive contacts 614, which may also be formed from HT9 stainless steel or other materials. The contacts 614 can improve thermal communication between the inner wall chamber 612 and the outer wall chamber 610 (and therefore, the coolant 606) as one or more of the contacts 614 approach and/or physically contact the outer wall chamber 610. Corresponding contacts 615 on the interior of the outer wall chamber 610 may also be provided, as shown. The gap region 616 between the inner wall chamber 612 and the outer wall chamber 610 may contain a vacuum or a gas, such as a tag gas that can be detected if the outer wall chamber 610 mechanically fails and potentially compromises the molten fuel storage. The gap region 616 between the outer wall chamber 610 and the inner wall chamber 612 thermally isolates the inner wall chamber 612, such as from thermal communication with the coolant. The flow temperature of the heat conducting fluid or coolant is typically less than the temperatures inside the inner wall chamber during a nuclear reaction. The multiple-walled fuel chamber 608 contains nuclear fuel 618 within the inner wall chamber 612. In the state shown in FIG. 6, the nuclear fuel 618 is in an intermediate state in which the molten fuel 601 is heated to a high enough temperature that the molten fuel 601 expands within the inner wall chamber 612, into a plenum or plenum region 620, but not enough to cause a large enough increase in temperature to cause the inner wall chamber 612 to thermally expand significantly. In one implementation, the nuclear fuel 618 is in a solid porous form including a combination of fertile nuclear fuel and a bonding material. For example, the fertile nuclear fuel may include 238U (uranium), such as a uranium foam, and the bonding material may include magnesium (Mg). Other material types and structures may be employed. (In one implementation of this intermediate state, the molten fuel 601 would also include a quantity of fissile nuclear fuel, such as 239Pu (plutonium) in a Mg—Pu solution.) In a breed-and-burn fast reactor, however, the fertile nuclear fuel can eventually be transmuted into fissile nuclear fuel that can undergo fission. It is noted that the nuclear fuel 618 includes solid fertile nuclear fuel and molten fissile fuel (e.g., a solution of fissile nuclear fuel and the carrier material). The molten state of the fissile fuel demonstrates that the temperature of the nuclear fuel 618 has exceeded the melting temperature of the nuclear fuel 618 and the position of the molten fuel 601 both within and outside the high neutron importance region of the nuclear reactor core demonstrates that the temperature of the nuclear fuel 601 has satisfied the negative reactivity feedback expansion temperature condition. The plenum region 620 is also located within the inner wall chamber 612 to receive gaseous fission products as well as molten fuel 601 as the fuel temperature rises and the fuel material expands into the plenum or plenum region 620. As shown, at least a portion of the plenum region 620 is located outside the high neutron importance region. Thus, as molten fuel expands into the plenum region 620 (and out of the high neutron importance region), reactivity in the nuclear reactor core decreases. FIG. 7 illustrates a cross-sectional view of an example passive reactivity control nuclear fuel device 700 containing molten fuel 701 expanding outside a high neutron importance region of a nuclear reactor core and expanding an inner wall chamber 712 within which it is contained. A solid porous fuel slug as nuclear fuel 718 also resides in the example passive reactivity control nuclear fuel device 700, such that the molten fuel 701 can pass through the pores of the nuclear fuel 718. The passive reactivity control nuclear fuel device 700 includes a duct 704 through which coolant 706, such as molten sodium, can flow. The duct 704 may be manufactured from HT9 stainless steel although other materials may be employed. The duct 704 also contains a multiple-walled fuel chamber 708. In one implementation, an outer wall chamber 710 of the multiple-walled fuel chamber 708 primarily includes HT9 stainless steel, although other materials may be employed, at least in part. Further, in one implementation, the inner wall chamber 712 of the multiple-walled fuel chamber 708 primarily includes molybdenum, although other materials may be employed, at least in part. In one implementation, the inner wall chamber 712 is fitted with one or more thermally conductive contacts 714, which may also be formed from HT9 stainless steel or other materials. The contacts 714 can improve thermal communication between the inner wall chamber 712 and the outer wall chamber 710 (and therefore, the coolant 706) as one or more of the contacts 714 approach and/or physically contact the outer wall chamber 710. Corresponding contacts 715 on the interior of the outer wall chamber 710 may also be provided, as shown. The gap region 716 between the inner wall chamber 712 and the outer wall chamber 710 may contain a vacuum or a gas, such as a tag gas that can be detected if the outer wall chamber 710 mechanically fails and potentially compromises the molten fuel storage. The gap region 716 between the outer wall chamber 710 and the inner wall chamber 712 thermally isolates the inner wall chamber 712, such as from thermal communication with the coolant. Nevertheless, as the temperature of the inner wall chamber 712 increases, the inner wall chamber 712 can thermally expand to overcome the thermal isolation. The flow temperature of the heat conducting fluid or coolant is typically less than the temperatures inside the inner wall chamber during a nuclear reaction. The multiple-walled fuel chamber 708 contains nuclear fuel 718 within the inner wall chamber 712. In the state shown in FIG. 7, the nuclear fuel 718 is in a very high-temperature state, caused by a potential combination of neutron heating, gamma heating, and direction fission of fuel material. At a sufficiently high temperature, the inner wall chamber 712 can thermally expand, such that the walls of the inner wall chamber 712 expand toward the wall of the outer wall chamber 710. In one implementation, the nuclear fuel 718 is in a solid porous form including a combination of fertile nuclear fuel and a bonding material. For example, the fertile nuclear fuel may include 238U (uranium), such as a uranium foam, and the bonding material may include magnesium (Mg). Other material types and structures may be employed. (In one implementation of this intermediate state, the nuclear fuel 718 would also include a quantity of fissile nuclear fuel, such as 239Pu (plutonium) in a Mg—Pu solution.) In a breed-and-burn fast reactor, however, the fertile nuclear fuel can eventually be transmuted into fissile nuclear fuel that can undergo fission (e.g., a 238U fertile nuclear fuel material bred to a 239Pu fissile nuclear fuel material, which can go into solution with Mg, designated as a Mg—Pu fissile fuel solution). It is noted that the nuclear fuel 718 includes solid fertile nuclear fuel and molten fissile fuel (e.g., a solution of fissile nuclear fuel and the carrier material). The molten state of the fissile fuel demonstrates that the temperature of the nuclear fuel 718 has exceeded the melting temperature of the nuclear fuel 718, and the position of the molten fuel 701 within and outside the high neutron importance region of the nuclear reactor core demonstrates that the temperature of the molten fuel 701 has satisfied the negative reactivity feedback expansion temperature condition. In addition, the expansion of the inner wall chamber 712 demonstrates that the temperature of the fuel and/or the inner wall chamber 712 satisfy an inner wall chamber expansion condition in a very high-temperature condition. As the temperature of the nuclear fuel 718 decreases, the temperature may no longer satisfy the inner wall chamber expansion condition, such that the thermal expansion rate of the inner wall chamber 712 can also decrease and/or reverse. A plenum region 720 is also located within the inner wall chamber 712 to receive gaseous fission products as well as molten fuel as the fuel temperature rises and the fuel material expands into the plenum or plenum region 720. As shown, at least a portion of the plenum region 720 is located outside the high neutron importance region. Thus, as molten fuel expands into the plenum region 720 (and out of the high neutron importance region), reactivity in the nuclear reactor core decreases. In contrast to the state shown in FIG. 6, the state shown in FIG. 7 depicts the molten fuel 701 at a sufficiently high temperature to force the inner wall chamber 712 to expand. In order to avoid mechanical failure of the inner wall chamber 712 when the temperature satisfies an inner wall chamber expansion condition (e.g., causing the inner wall chamber 712 to expand toward the outer wall chamber 710), the expansion moves the contacts 714 of the inner wall chamber 712 toward the contacts 715 of the outer wall chamber 710. As the distance between the contact 714 and 715 decreases (particularly to the point of physical contact), the heat of the molten fuel 701 can radiate or conduct to the outer wall chamber 710, rapidly reducing the temperature of the molten fuel 701, even to the extent that the molten fuel 701 densifies rapidly, reducing the temperature of the molten fuel 701, even potentially to the point of freezing the molten fuel 701 into a solid nuclear fuel slug of nuclear fuel 718. In this manner, an upper limit of the inner wall chamber temperature can be maintained—as the temperature increases, the inner wall chamber 712 thermally expands into radiative or conductive thermal communication with the outer wall chamber 710 and the coolant 706, resulting in a reduction in the temperature of the inner wall chamber 712. FIG. 8 illustrates a cross-sectional view of an example passive reactivity control nuclear fuel device 800 containing nuclear fuel 818 within a high neutron importance region of a nuclear reactor core after molten fuel has densified. The passive reactivity control nuclear fuel device 800 includes a duct 804 through which coolant 806, such as molten sodium, can flow. The duct 804 may be manufactured from HT9 stainless steel although other materials may be employed. The duct 804 also contains a multiple-walled fuel chamber 808. In one implementation, an outer wall chamber 810 of the multiple-walled fuel chamber 808 primarily includes HT9 stainless steel, although other materials may be employed, at least in part. Further, in one implementation, an inner wall chamber 812 of the multiple-walled fuel chamber 808 primarily includes molybdenum, although other materials may be employed, at least in part. In one implementation, the inner wall chamber 812 is fitted with one or more thermally conductive contacts 814, which may also be formed from HT9 stainless steel or other materials. The contacts 814 can improve thermal communication between the inner wall chamber 812 and the outer wall chamber 810 (and therefore, the coolant 806) as one or more of the contacts 814 approach and/or physically contact the outer wall chamber 810. Corresponding contacts 815 on the interior of the outer wall chamber 810 may also be provided, as shown. The gap region 816 between the inner wall chamber 812 and the outer wall chamber 810 may contain a vacuum or a gas, such as a tag gas that can be detected if the outer wall chamber 810 mechanically fails and potentially compromises the molten fuel storage. The gap region 816 between the outer wall chamber 810 and the inner wall chamber 812 thermally isolates the inner wall chamber 812, such as from thermal communication with the coolant. The flow temperature of the heat conducting fluid or coolant is typically less than the temperatures inside the inner wall chamber during a nuclear reaction. The multiple-walled fuel chamber 808 contains nuclear fuel 818 within the inner wall chamber 812. In the state shown in FIG. 8, the nuclear fuel 818 is in a reset state (e.g., after sufficient heat has been extracted from the nuclear fuel 818 to transition the molten fuel into a solid state, such as nuclear fuel 818). The temperature of the nuclear fuel 818 no longer exceeds the melting temperature of the nuclear fuel 818 and no longer satisfies the negative reactivity feedback expansion temperature condition. In one implementation, the nuclear fuel 818 returns to a solid form including a combination of fertile nuclear fuel, fissile nuclear fuel, and a bonding material. For example, the fertile nuclear fuel may include 238U (uranium), such as a uranium foam, and the bonding material may include magnesium (Mg). Other material types and structures may be employed. In a breed-and-burn fast reactor, however, the fertile nuclear fuel can eventually be transmuted into fissile nuclear fuel that can undergo fission. A plenum region 820 is also located within the inner wall chamber 812 to receive gaseous fission products as well as molten fuel as the fuel temperature rises and the fuel material expands into the plenum or plenum region 820. As shown, at least a portion of the plenum region 820 is located outside the high neutron importance region. Thus, as molten fuel expands into the plenum region 820 (and out of the high neutron importance region), reactivity in the nuclear reactor core decreases. FIG. 9 illustrates a cross-sectional view of an alternative example passive reactivity control nuclear fuel device. The passive reactivity control nuclear fuel device may include a duct (not shown) through which coolant, such as molten sodium, can flow. The duct may be manufactured from HT9 stainless steel although other materials may be employed. In one implementation, a multiple-walled fuel chamber 900 includes an outer wall chamber 910 of the multiple-walled fuel chamber 900 that primarily includes HT9 stainless steel, although other materials may be employed, at least in part. Further, in one implementation, an inner wall chamber 912 of the multiple-walled fuel chamber 900 primarily includes molybdenum, although other materials may be employed, at least in part. A gap region 916 between the inner wall chamber 912 and the outer wall chamber 910 may contain a vacuum or a gas, such as a tag gas that can be detected if the outer wall chamber 910 mechanically fails and potentially compromises the molten fuel storage. The gap region 916 between the outer wall chamber 910 and the inner wall chamber 912 thermally isolates the inner wall chamber 912, such as from thermal communication with the coolant. The flow temperature of the heat conducting fluid or coolant is typically less than the temperatures inside the inner wall chamber during a nuclear reaction. The multiple-walled fuel chamber 900 contains fertile nuclear fuel 918 within the inner wall chamber 912, such as 238U (uranium) in a porous, powdered, or suspension form. The fertile nuclear fuel 918 is separated from the fissile nuclear fuel region containing a liquid metal fuel 901, such as Mg—Pu, by a permeable barrier 919, although liquid metal fuel 901 may reside on either side of the permeable barrier 919. As the fertile nuclear fuel 918 is transmuted into fissile nuclear fuel by fast spectrum neutrons resulting from fission reactions with the nuclear reactor (and potentially from within the multiple-walled fuel chamber 900 itself), the transmuted fissile nuclear fuel diffuses into solution with the liquid metal fuel 901. In the state shown in FIG. 9, the liquid metal fuel 901 is in an intermediate state (e.g., after sufficient heat has been provided to the liquid metal fuel 901 to maintain the molten state of the liquid metal fuel. A plenum region 920 is also located within the inner wall chamber 912 to receive gaseous fission products as well as molten fuel as the fuel temperature rises and the fuel material expands into the plenum or plenum region 920. As shown, at least a portion of the plenum region 920 is located outside the high neutron importance region of a nuclear reactor core. Thus, as molten fuel expands into the plenum region 920 (and out of the high neutron importance region), reactivity in the nuclear reactor core decreases. FIG. 10 illustrates a cross-sectional view of an example passive reactivity control nuclear fuel device in which liquid metal fuel 1001 has expanded outside a high neutron importance region of a nuclear reactor core. The passive reactivity control nuclear fuel device may include a duct (not shown) through which coolant, such as molten sodium, can flow. The duct may be manufactured from HT9 stainless steel although other materials may be employed. In one implementation, a multiple-walled fuel chamber 1000 includes an outer wall chamber 1010 of the multiple-walled fuel chamber 1000 that primarily includes HT9 stainless steel, although other materials may be employed, at least in part. Further, in one implementation, an inner wall chamber 1012 of the multiple-walled fuel chamber 1000 primarily includes molybdenum, although other materials may be employed, at least in part. A gap region 1016 between the inner wall chamber 1012 and the outer wall chamber 1010 may contain a vacuum or a gas, such as a tag gas that can be detected if the outer wall chamber 1010 mechanically fails and potentially compromises the molten fuel storage. The gap region 1016 between the outer wall chamber 1010 and the inner wall chamber 1012 thermally isolates the inner wall chamber 1012, such as from thermal communication with the coolant. The flow temperature of the heat conducting fluid or coolant is typically less than the temperatures inside the inner wall chamber during a nuclear reaction. The multiple-walled fuel chamber 1000 contains fertile nuclear fuel 1018 within the inner wall chamber 1012, such as 238U (uranium) in a porous, powdered, or suspension form. The fertile nuclear fuel 1018 is separated from the fissile nuclear fuel region containing a liquid metal fuel 1001, such as Mg—Pu, by a permeable barrier 1019, although liquid metal fuel 1001 may reside on either side of the permeable barrier 1019. As the fertile nuclear fuel 1018 is transmuted into fissile nuclear fuel by fast spectrum neutrons resulting from fission reactions with the nuclear reactor (and potentially from within the multiple-walled fuel chamber 1000 itself), the transmuted fissile nuclear fuel diffuses into solution with the liquid metal fuel 1001. In the state shown in FIG. 10, the liquid metal fuel 1001 is in an intermediate state (e.g., after sufficient heat has been provided to the liquid metal fuel 1001 to maintain the molten state of the liquid metal fuel 1001) to further expand the liquid metal fuel 1001 such that a considerable volume of the liquid metal fuel 1001 is located outside the high neutron importance region of the nuclear reactor core. By removing a considerable volume of liquid metal fuel 1001 outside the high neutron importance region, the passive reactivity control nuclear fuel device reduces reactivity within the nuclear reactor core. A plenum region 1020 is also located within the inner wall chamber 1012 to receive gaseous fission products as well as molten fuel as the fuel temperature rises and the fuel material expands into the plenum or plenum region 1020. As shown, at least a portion of the plenum region 1020 is located outside the high neutron importance region of the nuclear reactor core. Thus, as molten fuel expands into the plenum region 1020 (and out of the high neutron importance region), reactivity in the nuclear reactor core decreases. It should be understood that an alternative implementation similar to those illustrated in FIGS. 9 and 10 may not employ a permeable membrane. For example, a solid slug or a solid, porous slug of fertile material may be employed such that the fertile material generally maintains its shape and location without the aid of a permeable membrane. FIG. 11 illustrates a cross-sectional view of another alternative example passive reactivity control nuclear fuel device. The passive reactivity control nuclear fuel device may include a duct (not shown) through which coolant, such as molten sodium, can flow. The duct may be manufactured from HT9 stainless steel although other materials may be employed. In one implementation, a multiple-walled fuel chamber 1100 includes an outer wall chamber 1110 of the multiple-walled fuel chamber 1100 that primarily includes HT9 stainless steel, although other materials may be employed, at least in part. Further, in one implementation, an inner wall chamber 1112 of the multiple-walled fuel chamber 1100 primarily includes molybdenum, although other materials may be employed, at least in part. A gap region 1116 between the inner wall chamber 1112 and the outer wall chamber 1110 may contain a vacuum or a gas, such as a tag gas that can be detected if the outer wall chamber 1110 mechanically fails and potentially compromises the molten fuel storage. The gap region 1116 between the outer wall chamber 1110 and the inner wall chamber 1112 thermally isolates the inner wall chamber 1112, such as from thermal communication with the coolant. The flow temperature of the heat conducting fluid or coolant is typically less than the temperatures inside the inner wall chamber during a nuclear reaction. The multiple-walled fuel chamber 1100 contains liquid metal fuel 1101 within the inner wall chamber 1112, such as Mg—Pu. In the state shown in FIG. 11, the liquid metal fuel 1101 is in an intermediate state (e.g., after sufficient heat has been provided to the liquid metal fuel 1101 to maintain the molten state of the liquid metal fuel), but the liquid metal fuel 1101 has not expanded significantly outside the high neutron importance region of the nuclear reactor core. A plenum region 1120 is also located within the inner wall chamber 1112 to receive gaseous fission products as well as molten fuel as the fuel temperature rises and the fuel material expands into the plenum or plenum region 1120. As shown, at least a portion of the plenum region 1120 is located outside the high neutron importance region of the nuclear reactor core. Thus, as molten fuel expands into the plenum region 1120 (and out of the high neutron importance region), reactivity in the nuclear reactor core decreases. FIG. 12 illustrates a cross-sectional view of another alternative example passive reactivity control nuclear fuel device in which liquid metal fuel 1201 has expanded outside a high neutron importance region of a nuclear reactor core. The passive reactivity control nuclear fuel device may include a duct (not shown) through which coolant, such as molten sodium, can flow. The duct may be manufactured from HT9 stainless steel although other materials may be employed. In one implementation, a multiple-walled fuel chamber 1200 includes an outer wall chamber 1210 of the multiple-walled fuel chamber 1200 that primarily includes HT9 stainless steel, although other materials may be employed, at least in part. Further, in one implementation, an inner wall chamber 1212 of the multiple-walled fuel chamber 1200 primarily includes molybdenum, although other materials may be employed, at least in part. A gap region 1216 between the inner wall chamber 1212 and the outer wall chamber 1210 may contain a vacuum or a gas, such as a tag gas that can be detected if the outer wall chamber 1210 mechanically fails and potentially compromises the molten fuel storage. The gap region 1216 between the outer wall chamber 1210 and the inner wall chamber 1212 thermally isolates the inner wall chamber 1212, such as from thermal communication with the coolant. The flow temperature of the heat conducting fluid or coolant is typically less than the temperatures inside the inner wall chamber during a nuclear reaction. The multiple-walled fuel chamber 1200 contains liquid metal fuel 1201 within the inner wall chamber 1212, such as Mg—Pu. In the state shown in FIG. 12, the liquid metal fuel 1201 is in an intermediate state (e.g., after sufficient heat has been provided to the liquid metal fuel 1201 to maintain the molten state of the liquid metal fuel), and the liquid metal fuel 1201 has expanded significantly outside the high neutron importance region of the nuclear reactor core. By removing a considerable volume of liquid metal fuel 1201 outside the high neutron importance region, the passive reactivity control nuclear fuel device reduces reactivity within the nuclear reactor core. A plenum region 1220 is also located within the inner wall chamber 1212 to receive gaseous fission products as well as molten fuel as the fuel temperature rises and the fuel material expands into the plenum or plenum region. As shown, at least a portion of the plenum region 1220 is located outside the high neutron importance region of the nuclear reactor core. Thus, as molten fuel expands into the plenum region 1220 (and out of the high neutron importance region), reactivity in the nuclear reactor core decreases. Note that FIGS. 11 and 12 depict implementations that do not include fertile nuclear fuel within the inner wall chamber of the multiple-walled fuel chamber. While not explicitly described, other implementations of the described technology, such as that illustrated and described with regard to FIG. 3 may also be implemented without fertile nuclear fuel within the inner wall chamber of the multiple-walled fuel chamber. FIG. 13 illustrates a cross-sectional view of an example passive reactivity control nuclear fuel device 1300 including an alternatively shaped plenum region 1320 in an inner wall chamber 1312. Molten fuel 1301 has expanded outside a high neutron importance region of a nuclear reactor core and into the plenum region 1320. A solid porous fuel slug as nuclear fuel 1318 resides in the example passive reactivity control nuclear fuel device 1300, such that the molten fuel 1301 can pass through the pores of the nuclear fuel 1318. The passive reactivity control nuclear fuel device 1300 includes a duct 1304 through which coolant 1306, such as molten sodium, can flow. The duct 1304 may be manufactured from HT9 stainless steel although other materials may be employed. The duct 1304 also contains a multiple-walled fuel chamber 1308. In one implementation, an outer wall chamber 1310 of the multiple-walled fuel chamber 1308 primarily includes HT9 stainless steel, although other materials may be employed, at least in part. Further, in one implementation, the inner wall chamber 1312 of the multiple-walled fuel chamber 1308 primarily includes molybdenum, although other materials may be employed, at least in part. In one implementation, the inner wall chamber 1312 is fitted with one or more contacts 1314, which may also be formed from HT9 stainless steel or other materials. The contacts 1314 can improve thermal communication between the inner wall chamber 1312 and the outer wall chamber 1310 (and therefore, the coolant 1306) as one or more of the contacts 1314 approach and/or physically contact the outer wall chamber 1310. Corresponding contacts 1315 on the interior of the outer wall chamber 1310 may also be provided, as shown. The gap region 1316 between the inner wall chamber 1312 and the outer wall chamber 1310 may contain a vacuum or a gas, such as a tag gas that can be detected if the outer wall chamber 1310 mechanically fails and potentially compromises the molten fuel storage. The gap region 1316 between the outer wall chamber 1310 and the inner wall chamber 1312 thermally isolates the inner wall chamber 1312, such as from thermal communication with the coolant. The flow temperature of the heat conducting fluid or coolant is typically less than the temperatures inside the inner wall chamber during a nuclear reaction. The multiple-walled fuel chamber 1308 contains nuclear fuel 1318 within the inner wall chamber 1312. In the state shown in FIG. 13, the nuclear fuel 1318 is in an intermediate state in which the molten fuel 1301 is heated to a high enough temperature that the molten fuel 1301 expands within the inner wall chamber 1312, into a plenum region 1320, but not enough to cause a large enough increase in temperature to cause the inner wall chamber 1312 to thermally expand significantly. It is noted that the nuclear fuel 1318 includes solid fertile nuclear fuel and molten fissile fuel (e.g., a solution of fissile nuclear fuel and the carrier material). The molten state of the fissile fuel demonstrates that the temperature of the nuclear fuel 1318 has exceeded the melting temperature of the nuclear fuel 1318 and the position of the molten fuel 1301 both within and outside the high neutron importance region of the nuclear reactor core demonstrates that the temperature of the nuclear fuel 1301 has satisfied the negative reactivity feedback expansion temperature condition. The plenum region 1320 is also located within the inner wall chamber 1312 to receive gaseous fission products as well as molten fuel 1301 as the fuel temperature rises and the fuel material expands into the plenum or plenum region 1320. As shown, at least a portion of the plenum region 1320 is located outside the high neutron importance region. Thus, as molten fuel expands into the plenum region 1320 (and out of the high neutron importance region), reactivity in the nuclear reactor core decreases. The plenum region 1320 is shown as somewhat tapered, such that the diameter of the plenum region 1320 decreases as the distance from the high neutron importance region increases. Because the plenum region 1320 has a smaller diameter than the inner wall chamber diameter within the high neutron importance region, the plenum region 1320 allows the molten fuel 1301 to expand farther away from the high neutron importance region than a plenum region having a larger diameter, under the same temperature conditions. Other plenum configurations are contemplated. An example passive reactivity control nuclear fuel device as described, for example, with regard to FIG. 2 represents device locatable within a nuclear reactor core. In alternative implementations, however, a passive reactivity control nuclear fuel device can be manufactured on a larger scale to form a substantial volume of a nuclear reactor vessel. In such an implementation, for example, a passive reactivity control nuclear fuel device can be manufactured such that the primary nuclear reactor core is located within the inner wall chamber and molten fuel can be moved out of a high neutron importance region within the inner wall chamber to a region of lower neutron importance region through thermal expansion of the thermal fuel to provide negative reactivity feedback. Likewise, cooling the molten fuel such that the removed molten fuel moves back into the high neutron importance region can provide positive reactivity feedback in such a large-scale passive reactivity control nuclear fuel device. The above specification, examples, and data provide a complete description of the structure and use of exemplary implementations of the invention. Since many implementations of the invention can be made without departing from the spirit and scope of the invention, the invention resides in the claims hereinafter appended. Furthermore, structural features of the different implementations may be combined in yet another implementation without departing from the recited claims.
	description	The present invention relates to methods for water purification processing and the economic utilization of waste waters produced from water purification processing. The disposal of saline water has become an expensive problem for society. For example, approximately 1.61 billion gallons of water containing approximately 800,000 tons of mixed sodium, calcium and magnesium chlorides and sulfates is produced from water treatment operations and oil fields in the state of California alone. This saline water must be disposed of, costing the state millions of dollars each year. Meanwhile, the United States Geological survey recently determined that New Mexico has an astounding 15 billion acre feet of brackish ground water, and a single basin in West Texas alone was found to have 760 million acre feet of brackish ground water. Many coal beds are located where traditional mining is not feasible. Instead, the coal beds are stripped of their associated methane by pumping water from the coal bed strata. Methane migrates to gas wells where it is pumped out and transported for public use. The removed water is usually of moderate salinity, typically 900 to 1500 parts per million (ppm) of total dissolved salts (TDS). Unfortunately, the water is typically high in sodium and carbonate and/or bicarbonate. Meanwhile, the disposal of waste water has become even more problematic in other parts of the world. As a result, billions of dollars are spent each year toward efforts to dispose of waste waters. Accordingly, it would be highly advantageous to provide improved methods of disposing of salty waters. It would even be more advantageous to provide methods of utilizing salty waters which provide a benefit to society, instead of simply disposing of the unwanted waters. Water purification typically produces a first effluent of relatively “clean water” and a second effluent of “waste water” which includes unwanted contaminants. For purposes herein, clean water is defined to mean water including less than 0.05% by weight of the chloride, sulfate or carbonate salts of sodium, potassium, calcium or iron or combinations thereof. In addition to waste water, there is a substantial amount of “moderately saline water” around the world that has less salinity than waste water but which is not generally acceptable for irrigation or animal consumption. Thus, this moderately saline water has severely limited application and usefulness. As defined herein, “moderately saline water” means water that has 0.05% or more by weight and less than 1.00% by weight of the chloride, sulfate or carbonate salts of sodium, potassium, calcium or magnesium, or combinations thereof. Known water purification processes proceed by numerous methods including ion-exchange, membrane softening, electrolysis, evaporation and precipitation. The softening of hard water take place by removing calcium and magnesium which is required for both industrial and household use. Known water softening processes proceed either by way of ion-exchange, membrane softening or precipitation. In the ion-exchange processes, the calcium (Ca2+) and magnesium (Mg2+) ions are exchanged for sodium (Na+) and the regeneration of the ion-exchange resin is achieved with a large excess of NaCl. This process creates a regeneration effluent being a relatively concentrated aqueous solution of sodium, calcium and magnesium chlorides which has to be discarded. Consequently, by this method, considerable amounts of sodium, calcium and magnesium salts in solution must be disposed of. Alternatively, it is possible to soften water by using weak acid resins which exchange hydrogen (H+) for calcium (Ca2+) and magnesium (Mg2+), and to regenerate the spent resins with a mineral acid. While this method creates less waste water, it is more expensive and yields relatively acidic soft water which is corrosive. Meanwhile, membrane softening concentrates the calcium, magnesium salts and salts of other divalent ions to produce waste waters which require costly disposal. The precipitation process has traditionally been carried out by the “lime soda” process in which lime is added to hard water to convert water soluble calcium bicarbonate into water insoluble calcium carbonate. This process also results in waste water which is difficult to filter and requires cumbersome treatment. My previously issued patent, U.S. Pat. No. 5,300,123 relates to the purification of impure solid salts. Even this process produces salty waste water which must be disposed of. My later-issued patents U.S. Pat. Nos. 6,071,411 6,374,539 and 6,651,383 relate to the processing and utilization of processed waste waters. These processes preferably employ ion-exchange, preferably using sodium sulfate or calcium sulfate, to alter the salt content of treated water. Moreover, the resulting salts, clean effluents and waste water effluents are useful for various applications including the treatment of soils for improving dust control, soil stabilization, adjusting the soil's sodium adsorption ratio (SAR), and treating root rot. Unfortunately, even with all of the various water treatment processes of the prior art, there are billions of gallons of waste water and moderately saline water that are discarded or not utilized because it is far too expensive to purify such waters using known water treatment processes. This overabundance of water is troubling because there is an overwhelming world-wide need for water, particularly for human and livestock consumption. A recent report from the United Nations states that more than 50 percent of the nations in the world will face water stress or water shortages by the year 2025. By 2050, as much as 75 percent of the worlds's population could face water scarcity. Even more troubling, in impoverished countries humans and animals often suffer from calcium and magnesium deficiencies even though there may be millions of gallons of nearby saline waters. These saline waters typically contain some calcium and magnesium but are too high in sodium to be drinkable. Unfortunately, due to the expense and unavailability of equipment, this water cannot be processed for human or animal consumption. Instead, milk is recommended to provide an adequate diet of calcium and magnesium but milk is typically not affordable or available in sufficient quantities to meet the needs of children in developing countries or even the needs of children in poor areas of developed countries. It would be an incredible development if the saline water could be treated to lower the sodium but increase or maintain the calcium and magnesium to levels suitable for human and livestock consumption. Water is also in great demand for soil treatment, particularly for irrigation. Unfortunately, waste waters typically have saline content which is not suitable for nearby irrigation. Thus, it would be extraordinarily advantageous if an inexpensive process were developed for processing waste waters to produce an effluent suitable for irrigation. Wind erosion of soil is also a significant problem throughout the world. Due to small particle size and poor cohesion, finely divided soil is sensitive to the influence of wind. Such finely divided soil is found in agricultural lands, dunes, lake beds, construction sites and roads under construction. Erosion by wind causes the drifting of masses of soil in the form of dust. The erosion by wind causes the inconvenience of dust formation and the loss of valuable matter such as seed, fertilizer and plantlets. Dust storms are a danger to traffic and a health risk to persons located in the vicinity. Finally, it would be desirable if all of the aforementioned objectives could be accomplished while overcoming the expensive and problematic concerns facing this country and the rest of the world, specifically, the disposing of saline waters. It would further be desirable if this objective could be obtained while simultaneously meeting with above described needs. Briefly, in accordance with the invention, I provide methods for economically and efficiently processing moderately saline waters, particularly those produced from oil and gas wells, and irrigation drainage, to produce an effluent containing lower sodium content but the increased salts of multivalent cations, particularly calcium and magnesium. I also provide methods for utilizing the effluent produced by water purification. The process of the present invention provides for treating moderately saline water having 0.05% or more by weight of the salts of Na, K, Ca, Mg, Fe, Cl, SO4, or CO3 or combinations thereof and less than 1.00% by weight of the salts of Na, K, Ca, Mg, Fe, Cl, SO4, or CO3 or combinations thereof. The present invention is particularly suitable for treating water having high sodium content. The moderately saline water is then passed through an ion exchange resin saturated with multivalent cations to produce “useful water”. As defined herein, the term “saturated” is interpreted in a loose sense to mean that the ion exchange resin has sufficient multivalent cations to effect an ion exchange for sodium to reduce the amount of sodium in moderately saline water. Though the present process is not a water softening process, I have discovered that commercially available ion exchange resins sold for water softening may be utilized for the practice of the present invention. Moreover, though I have not discovered a preferred ion exchange resin, I have determined that a resin called Lewatit C-249 from Sybron Chemicals, a division of Bayer Chemicals, is acceptable. For practicing the water treatment process of the present invention, the ion exchange resin is saturated with multivalent cations. Various multivalent cations may be utilized. However, it is preferred that the multivalent cations are calcium (Ca2+) or magnesium (Mg2+) ions, or combinations thereof. Water softening resins are often sold saturated with sodium which is unacceptable for practicing the present invention. To saturate the resin with multivalent cations, calcium chloride or magnesium chloride solution may be utilized to flush the resin until the resin is sufficiently saturated with calcium and/or magnesium cations to effect an ion exchange for sodium. The moderately saline water is passed through the ion exchange resin to produce a useful effluent having decreased sodium cations compared to the moderately saline water. The useful effluent will also have higher calcium and magnesium. However, I have determined that the useful effluent can be utilized for both human and animal consumption. Moreover, I have determined that the useful effluent is also beneficial for treating soil for irrigation. As the moderately saline waters passes through the ion exchange resin, the sodium content of the resin rises and the multivalent cation content lowers until the resin is unacceptable for further water treatment in accordance with the present invention. To regenerate the ion exchange resin, the resin is flushed with a brine solution having more than 1.00% by weight of the salts of Na, K, Ca, Mg, Fe, Cl, SO4, or CO3. Preferably, the brine is particularly high in calcium and/or magnesium content and low in sodium. The brine solution is flushed through the ion exchange resin until the amount of multivalent cations affixed to the ion exchange resin is increased and the sodium affixed to the resin is decreased until the ion exchange resin is sufficiently saturated with multivalent cations to again process moderately saline water having high sodium content. The recharge process increases the multivalent cations in the ion exchange resin. However, the effluent produced from the recharge process has a significantly high sodium content. Advantageously, I have discovered that the recharge effluent waters which have a high sodium content are particularly suitable for soil stabilization, pond sealing and treating root rot. These high sodium waste waters are also suitable for use in cooling towers and laundry applications. I have learned that waters produced from water purification, particularly those high in calcium and magnesium can be used to control dust and to irrigate farm land, or as additive to irrigation waters where the soil has a high sodium content. Moreover, I have discovered that useful effluent and recharge effluent can be processed to create both solid and aqueous mixtures which can be applied to roads and highways for deicing. Accordingly, it is an object of the invention to provide cost effective means of processing moderately saline waters. It is a also principal object of the invention to provide new methods for utilizing the useful water produced from water purification. These and other, further and more specific objects and advantages of the invention will be apparent to those skilled in the art from the following detailed description taken in conjunction with the drawings. The presence of calcium and/or magnesium in water results in the water being considered “hard”. These mineral ions in the water react with heat, plumbing and other chemical agents to reduce the cleaning effectiveness of laundry, dish washing and bathing applications. These calcium and magnesium ions also combine with carbonates, sulfates, oils and fat to create bathtub scum, spotted dishes, gray sheets, etc. In addition, hard water has been found to cause scaling of industrial water heaters and commercial boilers causing early substantial energy losses through impaired heat transfer and early shutdown for the removal of scale. Accordingly, there has been substantial effort to remove the “hardness” of the water. With reference to FIG. 1, water softening is the removal of the “hardness” from the water which means predominantly removing or altering the calcium and magnesium ions from the water. Several methods for effecting water softening are known. The best known process for softening water is “ion-exchange”. The hard water passes through a tank containing an ion exchange resin, often containing beads which are microporous. The beads are saturated with sodium to cover both their exterior and interior surfaces. As the water passes through the resin, an ion exchange process occurs. Ion-exchange entails the exchange of sodium, which is introduced into the water from the resin, for calcium, magnesium, iron and other divalent mineral ions which are transferred out of the water and into a resin. Calcium and magnesium ions attach to the resin while the sodium on the resin is released into the water. When the resin approaches saturation with these hard ions, the resin is regenerated, most often with solutions of sodium chloride leaving an effluent containing 3 to 25% sodium, calcium and magnesium salts which must be disposed of. The exact concentration of the effluent depends on the shop practice and, in particular, on the amount of rinse water included in the effluent, if any. Less often mineral acids like sulfuric acid or hydrochloric acid are used for water softening. Meanwhile, I have discovered that the reversal of the water softening process can be practiced to treat moderately salty waters, particularly waters high in sodium content, to produce “useful waters” having higher calcium and magnesium content but lower sodium content. Because the process results in increased calcium and magnesium in the water, the process can be referred to as a “water hardening” process. With reference to FIG. 2, the process of the present invention passes the moderately saline waters through an ion exchange resin. The resin is pre-treated to be saturated with multivalent cations. Preferred multivalent cations include calcium (Ca2+) or magnesium (Mg2+) ions, or combinations thereof. The pretreatment can be achieved by various methods which can be selected by those skilled in the art. However, in a preferred embodiment, the resin may be pre-treated using a calcium chloride or magnesium chloride solution to flush the resin until it is properly saturated with calcium and/or magnesium cations. I have found that a resin of Lewatit C-249 from Sybron Chemicals can be generated using a 13% solution of calcium chloride flushed at a rate of 15 lbs of calcium chloride per cubic foot of resin. The moderately saline water is passed through the ion exchange resin to produce a useful effluent having decreased sodium cations compared to the pre-treated moderately saline water. The useful effluent will also have higher calcium and magnesium. As the moderately saline waters passes through the ion exchange resin, the sodium content of the resin rises until the resin is unacceptable for further water treatment in accordance with the present invention. To regenerate the ion exchange resin, the resin is flushed with a brine solution having more than 1.00% by weight of the salts of Na, K, Ca, Mg, Fe, Cl, SO4, or CO3. Preferably, the brine is particularly high in multivalent cation content, such as calcium and/or magnesium, but low in sodium content. The brine solution is flushed through the ion exchange resin until the amount of multivalent cations affixed to ion exchange resin is increased and the sodium is decreased until the ion exchange resin is sufficiently saturated with multivalent cations to again process moderately saline water in accordance with the present invention. In an additional preferred embodiment, the moderately saline water may be pre-treated using known water treatment techniques prior to undergoing the water “hardening” process of the present invention. For example, where the water has significant levels of calcium and magnesium, as well as sodium, the water is preferably processed through a known water softening process to remove as much calcium and magnesium as possible. Advantageously, the removal of the calcium, magnesium and other multivalent cations before the hardening treatment helps prevent the creation of precipitates which bind to membranes if membrane filtration is also utilized. The method of treating water of the present invention will now be further explained in and by the following examples. Moderately saline well water is pumped from the Wonder Valley area, east of Twenty Nine Palms, Calif. Water from the well is measured at 1960 ppm TDS and water analysis reveals the following results. Pre-Treated Moderately Saline WaterCationsResults (ppm)Calcium 51MagnesiumNot detectedSodium700 Using the chart of FIG. 3 reveals that the water has an SAR value of 29. The well water is hardened by passing it through an ion exchange resin of Lewatit C-249 from Sybron Chemicals which has been saturated with calcium cation using a 13% solution of calcium chloride flushed at a rate of 15 lbs of calcium chloride per cubic foot of resin. After treatment, water analysis reveals the following results. Post-Treated “Useful” WaterCationsResults (ppm)Calcium410Magnesium1.4Sodium380 Using the chart of FIG. 3 reveals that the water now has an SAR value of 5.0. Moderately saline well water is pumped from Red Rock Ranch, Calif. from the Department of Water Reclamation (“DWR”). The water is initially measured to have 5600 ppm TDS and water analysis reveals the following results. Pre-Treated Moderately Saline WaterCationsResults (ppm)Calcium530Magnesium110Sodium1400 Because the chart of FIG. 3 cannot be used to reveal SAR value of the water, numerical calculations are performed to determine that the SAR is approximately 15.2. As a result of its high initial calcium and magnesium levels, it is decided that the DWR water will undergo both a water softening process and a water hardening process. The water is treated through a water softening process to lower the calcium and magnesium levels to below 5 ppm as reflected in the following results. Moderately Saline Water After Water SofteningCationsResults (ppm)CalciumNot detectedMagnesiumNot detectedSodium2200 The extremely high sodium level and low calcium and magnesium levels results in an “off-the-chart” SAR level. The DWR water is then hardened by passing it through an ion exchange resin of Lewatit C-249 from Sybron Chemicals which has been saturated with calcium cation using a 13% solution of calcium chloride flushed at a rate of 15 lbs of calcium chloride per cubic foot of resin. After treatment, water analysis reveals the following results. Post-Treated “Useful” WaterCationsResults (ppm)Calcium940Magnesium2.3Sodium1100 Using the chart of FIG. 3 reveals that the water now has an SAR value of 4.9. Irrigation With “Hardened” Waters All irrigated areas suffer from a buildup of sodium. Plant evaprotranspiration and plant growth use about 70 to 90% of the irrigation water and the sodium is concentrated in the remaining 10 to 30% of the water. This water must be washed from the roots or plant growth suffers. Calcium and Magnesium have very little affinity for water while sodium has a very strong affinity for water. Sodium's affinity for water is strong enough to spread clay particles, and clays are said to “swell”. The more sodium ions in soil water, as compared to the concentration of calcium and magnesium ions, the higher the percentage of the ion exchange sites that will be occupied by sodium. This causes a greater attraction of water and the soils swell more. However, experimentation has found that there is a stopping point to the soil swelling. When about 14-16% of the change sites are occupied by sodium ions, the clay particles disperse into small units and the swelling is lost and the soil packs tightly. The clay particles plug most of the pores that remain in the soil and this further restricts the movement of air, water and nutrients and the soil's productivity is lost. U.S. Salinities Laboratory calculated the amount of ion exchange sites that would be occupied by sodium based on the amount of calcium and magnesium present. These calculations were named the Sodium Adsorption Ration (SAR). As shown in FIG. 3, the sodium buildup is predicted by the sodium absorption ratio (SAR) vs. the total salinity of the irrigation water. To use the chart in FIG. 3, the sodium concentration is marked on the left side of the nomogram. The calcium plus magnesium concentration is then marked on the right side of the nomogram. Drawing a straight line between the two marks identifies the SAR value where the line intersects the sodium adsorption scale. Due to the inverse relationship between the addition of sodium to calcium and magnesium, an increase in calcium and/or magnesium will actually lower the SAR value of the irrigation water. Though some plants are much more tolerant of high sodium content in the soil, generally a SAR value of 14 or more will cause a dispersion of the clay content within the soil and a corresponding loss in productivity. With reference to Examples 1 and 2, the moderately saline waters are tested and determined to have SAR values of 29 and 15.2, respectively. Clearly, use of these waters for irrigation would have a harmful impact on soil productivity. However, after the hardening process of the present invention, the waters are found to have SAR values of 5.0 and 4.9, respectively, which are much more conducive for irrigation. By using the waters having an increased calcium and magnesium content as irrigation water reduces the buildup of exchangeable sodium in the soil thereby maintaining the soil in proper sodium equilibrium. Moreover, the process produces water which will optimize the SAR of soil moisture in the root zone of plants while decreasing the soil's salinity. This decrease in salinity is particularly advantageous because the prior art practices of adding calcium and magnesium salts or sulfuric acids causes an undesirable increase in soil salinity. Animal Consumption of “Hardened” Waters While the U.S. Environmental Protection Agency (“EPA”) recommends that the salt content of drinking water for humans be limited to 500 ppm TDS, for most animals such as livestock, 1000 to 1500 ppm is tolerable. Moreover, the less sodium in the water, causes a corresponding increase in the total salt content that is tolerable to both humans and livestock. Meanwhile, whole milk has much higher salt content than approved drinking water. The average concentrations of milk salt constituents is listed as follows. Whole MilkConstituentResults (ppm)Calcium1230Magnesium120Sodium580 As a result of diets in many parts of the world, children are very deficient in calcium and/or magnesium. For many children, milk is not available. In addition to children, animals such as livestock are also in need of water, calcium and/or magnesium. Thus, it would be desirable if the drinking water supply in impoverished areas could provide the calcium and/or magnesium where milk is unavailable or unaffordable. With reference to Examples 1 and 2, water processing of the present invention produces water having salt contents comparable to whole milk. Moreover, the saline waters of Examples 1 and 2 contain TDS significantly higher than billions of gallons of moderately saline waters across world. Processing of the moderately saline waters in accordance with the practice of the present invention would produce waters having even lower salt contents. In many areas of the world, the only water supplies available are saline. However, the water can be improved upon by lowering the sodium while increasing the calcium and/or magnesium content by processing the available saline water in accordance with the process of the present invention. Other Uses For “Hardened” Waters In addition to use the water for human and livestock consumption, the useful effluent produced by practicing the hardening process of my invention can be utilized for various purposes. For example, I have learned that the waters produced from my water purification process, particularly those waters high in calcium and magnesium can be used to control dust. Moreover, I have discovered that useful effluent waters can be processed to create both solid and aqueous mixtures which can be applied to roads and highways for deicing. The recharge process increases the multivalent cations in the ion exchange resin. However, the effluent produced from the recharge process has a very high sodium content. Advantageously, I have discovered that the recharge effluent waters which have a high sodium content are particularly suitable for soil stabilization, pond sealing and treating root rot. These high sodium waste waters are also suitable for use in cooling towers and laundry applications. Having described the invention in such terms as to enable one skilled in the art to make and use it and having identified the presently best mode of practicing it,
	abstract	A method for measuring a demagnification of a charged particle beam exposure apparatus includes measuring a first stage position of a mask stage in accordance with a mask stage coordinate system, irradiating a first charged particle beam to a first irradiation position on a specimen through the opening portion of the mask, measuring the first irradiation position in accordance with a specimen stage coordinate system, moving the mask stage to a second stage position, measuring the second stage position of the mask stage, irradiating a second charged particle beam to a second irradiation position on the specimen through the opening portion of the mask measuring the second irradiation position in accordance with the specimen stage coordinate system, and calculating a demagnification of the charged particle beam exposure apparatus from the first and second stage positions and the first and second irradiation positions.
	description	The following discussion of the preferred embodiments directed to an EUV lithography source where an input laser beam is delivered off-axis relative to the first collection optics is merely exemplary in nature, and is in no way intended to limit the invention or its applications or uses. FIG. 3 is a plan view of an EUV source 66 shown at an angle similar to that of the source 50 shown in FIG. 2, according to an embodiment of the present invention. In this embodiment, target production hardware 68 is positioned at its usual location relative to a plasma spot 70. However, the collection optics 60 had been replaced with first collection optics 72 that is only partially dish-shaped and is positioned at a different location relative to the hardware 68 than the collection optics 60. In this embodiment, the collection optics 72 is about half the size of the collection optics 60, and is positioned above the target hardware 68. An opening 74 is provided in the collection optics 72 through which a target laser beam 76 propagates to the plasma spot 70. The input laser beam 76 is positioned off-axis or asymmetrical relative to the collection optics 72. Because the collection optics 72 is at this position, the angular distribution 78 of the generated EUV radiation directed towards the collection optics 72 is such that the strong EUV radiation is reflected from the upper edges of the optics 72 and is not obscured by the target hardware 68, as shown. In the embodiment shown in FIG. 3, the collection optics 72 is about one-half the size of the collection optics 60 shown in FIG. 2. Therefore, some of the generated EUV radiation does not get reflected from the collection optics 72 that normally would in the conventional system. FIG. 4 is a schematic plan view of an EUV source 82 including target hardware 84 and a plasma spot 86, according to another embodiment of the present invention. In this embodiment, first collection optics 88 has the same shape as the collection optics 60, but includes two openings 90 and 92 for two separate input laser beams 94 and 96, respectively. Basically, it is the embodiment shown in FIG. 3, only doubled so that strong EUV radiation is provided both above and below the target hardware 84. Particularly, the angular distribution 98 of the EUV radiation from the beam 94 is directed along the line of the input laser beam 94 and is reflected from the optics 88 below the target hardware 84, and the angular distribution 100 of the EUV radiation from the beam 96 is directed along the line of the input laser beam 96 and is reflected above the target hardware 84. Therefore, the embodiment shown in FIG. 4 provides more EUV radiation than the EUV source 50. The foregoing discussion describes merely exemplary embodiments of the present invention. One skilled in the art would readily recognize that various changes, modifications and variations can be made therein without departing from the spirit and scope of the invention as defined in the following claims.
	description	1. Field of the Invention This invention pertains generally to a nuclear reactor fuel assembly and more particularly to an improved hold down spring on the top nozzle of the fuel assembly. 2. Related Art The primary side of nuclear reactor power generating systems which are cooled with water under pressure comprises a closed circuit which is isolated and in heat exchange relationship with the secondary side for the production of useful energy. The primary side comprises the reactor vessel enclosing a core internal structure that supports a plurality of fuel assemblies containing fissile material, the primary circuit within heat exchange steam generators, the inner volume of a pressurizer, pumps and pipes for circulating pressurized water; the pipes connecting each of the steam generators and pumps to the reactor vessel independently. Each of the parts of the primary side comprising a steam generator, a pump and a system of pipes which are connected to the vessel form a loop of the primary side. For the purpose of illustration, FIG. 1 shows a simplified nuclear reactor primary system, including a generally cylindrical reactor pressure vessel 10 having a closure head 12 (also shown in FIG. 2), enclosing a nuclear core 14. A liquid reactor coolant, such as water is pumped into the vessel 10 by pump 16 through the core 14 where heat energy is absorbed and is discharged to a heat exchanger 18, typically referred to as a steam generator in which heat is transferred to a utilization circuit (not shown), such as a steam driven turbine generator. The reactor coolant is then returned to the pump 16, completing the primary loop. Typically, a plurality of the above described loops are connected to a single reactor vessel 10 by reactor coolant piping 20. An exemplary reactor design is shown in more detail in FIG. 2. In addition to the core 14 comprised of a plurality of parallel, vertical, co-extending fuel assemblies 22, for purposes of this description, the other vessel internal structure can be divided into the lower internals 24 and the upper internals 26. In conventional designs, the lower internals' function is to support, align and guide core components and instrumentation as well as direct flow within the vessel. The upper internals restrain or provide a secondary restraint for the fuel assemblies 22 (only two of which are shown for simplicity in this figure), and support and guide instrumentation and components, such as control rods 28. In the exemplary reactor shown in FIG. 2, coolant enters the reactor vessel 10 through one or more inlet nozzles 30, flows down through an annulus between the vessel and the core barrel 32, is turned 180° in a lower plenum 34, passes upwardly through a lower support plate 37 and a lower core plate 36 upon which the fuel assemblies 22 are seated and through and about the assemblies. In some designs, the lower support plate 37 and the lower core plate 36 are replaced by a single structure, the lower core support plate, at the same elevation as 37. The coolant flow through the core and surrounding area 38 is typically large on the order of 400,000 gallons per minute at a velocity of approximately 20 feet per second. The resulting pressure drop and frictional forces tend to cause the fuel assemblies to rise, which movement is restrained by the upper internals, including a circular upper core plate 40. Coolant exiting core 14 flows along the underside of the upper core plate 40 and upwardly through a plurality of perforations 42. The coolant then flows upwardly and radially to one or more outlet nozzles 44. The upper internals 26 can be supported from the vessel or the vessel head and include an upper support assembly 46. Loads are transmitted between the upper support assembly 46 and the upper core plate 40, primarily by a plurality of support columns 48. A support column is aligned above a selected fuel assembly 22 and perforations 42 in the upper core plates 40. The rectilinearly movable control rods 28 typically include a drive shaft 50 and a spider assembly 52 of neutron poison rods 28 that are guided through the upper internals 26 and into aligned fuel assemblies 22 by control rod guide tubes 54. The guide tubes are fixedly joined to the upper support assembly 46 and connected to the top of the upper core plate 40. FIG. 3 is an elevational view, represented in vertically shortened form, of a typical fuel assembly being generally designated by reference character 22. The fuel assembly 22 is of the type used in a pressured water reactor and has a structural skeleton which, at its lower end, includes a bottom nozzle 58 sometimes referred to as the lower end fitting. The bottom nozzle 58 supports the fuel assembly 22 on a lower core support plate 60 in the core region of the nuclear reactor (the lower core support plate 60 is represented by reference character 36 in FIG. 2). In addition to the bottom nozzle 58, the structural skeleton of the fuel assembly 22 also includes a top nozzle 62 (sometimes referred to as the upper end fitting or top end fitting) at its upper end and a number of guide tubes or thimbles 54 (also referred to as guide tubes), which extend longitudinally between the bottom and top nozzles 58 and 62 and at opposite ends are rigidly attached thereto. The fuel assembly 22 further includes a plurality of transverse grids 64 axially spaced along and mounted to the guide thimbles 54 and an organized array of elongated fuel rods 66 transversely spaced and supported by the grids 64. Although it cannot be seen in FIG. 3, the grids 64 are conventionally formed from orthogonal straps that are interleaved in an egg-crate pattern with the adjacent interface of four straps defining approximately square support cells through which the fuel rods 66 are supported in a transversely spaced relationship with each other. In many conventional designs, springs and dimples are stamped into the opposing walls of the straps that form the support cells. The springs and dimples extend radially into the support cells and capture the fuel rods therebetween; exerting pressure on the fuel rods cladding to hold the rods in position. Also, the assembly 22 has an instrumentation tube 68 located in the center thereof that extends between and is mounted to the bottom and top nozzles 58 and 62. With such an arrangement of parts, fuel assembly 22 forms an integral unit capable of being conveniently handled without damaging the assembly of parts. As mentioned above, the fuel rods 66 in the array thereof in the assembly 22 are held in spaced relationship with one another by the grids 64 spaced along the fuel assembly length. Each fuel rod 66 includes a plurality of nuclear fuel pellets 70 and is closed at its opposite ends by upper and lower end plugs 72 and 74. The pellets 70 are maintained in a stack by a plenum spring 76 disposed between the upper end plug 72 and the top of the pellet stack. The fuel pellets 70, composed of fissile material, are responsible for creating the reactive power of the reactor. The cladding which surrounds the pellets functions as a barrier to prevent the fission by-products from entering the coolant and further contaminating the reactor system. To control the fission process, a number of control rods 78 are reciprocably movable in the guide thimbles 54 located at predetermined positions in the fuel assembly 22. Specifically, a rod cluster control mechanism 80 positioned above the top nozzle 62 supports the control rods 78. The control mechanism has an internally threaded cylindrical hub member 82 with a plurality of radially extending flukes or arms 52. Each arm 52 is interconnected to the control rods 78 such that the control rod mechanism 80 is operable to move the control rods vertically in the guide thimbles 54 to thereby control the fission process in the fuel assembly 22, under the motive power of control rod drive shafts 50 which are coupled to the control rod hubs 80, all in a well known manner. As previously mentioned, the fuel assemblies are subject to hydraulic forces that exceed the weight of the fuel rods and thereby exert significant forces on the fuel rods and the fuel assemblies. These forces are countered by a combination of the weight of the fuel assemblies 22 and a plurality of hold down spring assemblies 56 on the top nozzles 62 which push against the upper core plate 40 (FIG. 2) of the reactor. The hold down spring assemblies 56 thereby prevent the force of the upward coolant flow from lifting the fuel assemblies into damaging contact with the upper core plate, while allowing for changes in fuel assembly length due to core-induced thermal expansion and radiation growth. Operating experience has shown that these hold down springs can be subject to stress corrosion cracking which can reduce their effectiveness. Accordingly, a new hold down arrangement is desired that will maintain its resiliency over extended fuel cycles. Furthermore, a new hold down assembly is desired that will be more resistant to stress corrosion cracking. These and other objects are achieved by an improved fuel assembly having a top end fitting and a bottom end fitting connected together by a structural assembly having an axial dimension that extends from the bottom end fitting to the top end fitting, with the top end fitting having a hold down spring assembly projecting above an upper surface of the top end fitting. The hold down spring has a primary spring member extending above the top end fitting, that includes a first straight leg portion having one end attached to a frame of the top end fitting at an acute angle to a plane orthogonal to the axial dimension of the fuel assembly, with the acute angle being greater than 0°. An arcuate transition portion extends at the other end of the first straight leg with a straight second leg portion extending from the transition portion toward the frame at an acute included angle with the first leg. The primary spring member is oriented on the top end fitting so that the transition portion is at the vertical highest elevation, whereby movement of the end fitting and an upper plate of the reactor that the nuclear fuel assembly is designed to operate in, relatively towards each other, primarily loads the transition portion and deflects the first leg portion about the attachment to the end fitting frame. The hold down spring assembly also includes at least one secondary spring that has a first and second end. The first end is attached to the top end fitting adjacent the first end of the primary spring first leg. The second end terminates adjacent the transition portion and includes means for interacting with the transition portion to resist downward movement of the transition portion as the primary spring member deflects in a cantilever fashion. Preferably, the end of the primary spring that is attached to the frame of the top end fitting is supported in a slot in the frame that extends substantially at the acute angle. Desirably, the spring is clamped on a first portion of the surface on the frame that extends substantially at the acute angle where a periphery of the first portion of the surface of the frame under the primary spring is radiused to transition to a second portion of the surface of the frame under the primary spring that extends substantially parallel to the plane orthogonal to the axial dimension. In still another embodiment, the at least one secondary spring above has a substantially flat leg that extends from the first end to an intermediate portion near the second end where the flat leg is radiused in the direction of the frame. Preferably, the radiused intermediate portion is curved at between 10° and 70°. As previously indicated, the hold down spring assemblies 56 shown in FIG. 3 are important structural members for a nuclear fuel assembly. Several leafs are assembled together to form a spring set in order to provide the needed hold down force to the fuel assembly to counteract the upward lift forces due to the hydraulic flow and to permit fuel assembly growth due to differential thermal expansion and irradiation dosage during normal plant operation. Conventional hold down springs 56 are mounted on the top fuel nozzles 62 and are retained by a pin 60 located at diametrically opposite corners of the top plate 20 as shown in FIG. 3. Typically, the top nozzle 62 supports four spring sets 56 as illustrated in FIG. 4. Each spring set has a primary spring 84 and at least one secondary spring 86 with two secondary springs being shown in FIG. 3 and three being shown in FIG. 4. In accordance with the prior art, spring leafs 84 and 86 have a flat, horizontal base 88 that is secured against the top plate 20 of the top nozzle 62 by a pin 60. The leafs then curve upward away from the top nozzle 20 with the primary spring member 84 having a first flat leg 90 that extends at an acute angle, greater than 0°, with the top plate 20 to an arcuate transition portion 92 at the other end of the first leg 90. A straight second leg portion 94 extends from the transition portion 92 toward the frame of the top nozzle 62 at an acute included angle with the first leg 90. The secondary spring leafs 86 of this prior art embodiment have a short flat section that corresponds to the flat spring base 88 of the primary spring and then curve upward under the primary spring, extending over a straight portion under the primary spring and terminating adjacent the transition portion 92 at a second end, with the second ends of the secondary leafs 86 interacting with the transition portion 92 of the primary spring 84 to resist downward movement of the transition portion 92 as the primary spring 84 deflects downward in a cantilevered fashion. The second leg 94 of the primary spring extends through an opening in the secondary spring leafs 86 to the top nozzle frame 62 where it interacts with a stop that is not shown. Fuel assemblies 22 are installed vertically in the reactor core 14 and stood upright on the lower core plate 60 (36). As can be appreciated from FIG. 2, after the fuel assemblies are set in place, the upper support structure 26 is installed. The upper core plate 40 then bears down against the hold down springs 56 on the top nozzle 62 of each fuel assembly 22 to hold the fuel assemblies in place. The springs are generally made of nickel-chromium-iron alloy 718. The retaining pin 60, which holds the spring set in place, can be either threaded into the top nozzle or welded to prevent loosening while in service. The improvement of this invention is illustrated in FIG. 5. Like reference characters are used for corresponding components of the spring 56 and top nozzle 62, though it should be appreciated that the design of the individual components will deviate from the corresponding components of the prior art illustrated in FIGS. 3 and 4, as hereafter described. In accordance with this invention, the spring base 88 of each leaf, i.e., the primary spring members and secondary spring members, are formed from a short section of straight, flat beam followed by a long straight, flat beam 90 whose thickness is tapered, extending in a direction along the leafs away from the base 88. The beams 88 and 90 thus form one continuous flat leg. Other than the top primary spring 84, there is a slight bend 96 at the end portion of the straight secondary beams 86. Since the spring set 56 is a cantilevered structural system, the maximum bending moment and stretch occur at the support end 98. From the flexure loading of a straight beam, the absolute magnitude of strain or stress on the inner and outer fibers are equal. However, as for the curved base of a conventional spring design, the absolute magnitude of strain or stress on the inner and outer fibers are not equal due to a curvature effect which can be appreciated from the graphical representation of the strain distribution for the prior art leaf spring design shown in FIG. 7 and the strain distribution of the leaf spring design of this invention illustrated in FIG. 8. This analysis assumes an elastic-plastic deflection up to the operation condition. Based on the same loading analysis, the maximum absolute strain for the straight (flat) end spring design is equally distributed on the inner and outer fibers. The maximum strains are 0.014247 and 0.010104, respectively for the curved base of the prior art and the straight base design of this invention. This means the maximum strain is reduced by approximately 29% for the straight base design. FIG. 6 provides another view of the spring design illustrated in FIG. 5 taken from another angle. The sections 88 and 90 shown in FIGS. 5 and 6 provide a straight, flat beam leaf spring set that extends from a slanted slot 100 in the top nozzle. The straight beams extend until the transition portion 92 in the primary spring leaf and the slightly curved sections 96 in the second end of the secondary spring leafs. The slanted slot 100 extends at an acute angle, greater than 0°, to a plane orthogonal to the longitudinal axis of the fuel assembly. The bends 96 are radiused at between 10° and 70°. Similarly, the lower lip 102 of the slot 100 is similarly radiused between 10° and 70°. In other respects, the top nozzle 62 is similar to that shown in FIG. 4. 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. A method for preparing a captured trace, the captured trace being stored for later use, of I/O workload activity experienced on one or more data storage volumes included with a data storage system for being analyzed by a computerized trace analysis process, the method comprising the steps of:preparing the captured trace for being analyzed by categorizing information from the captured trace into categories related to (i) components in the data storage system experiencing the traced workload activity and (ii) information type including response times and task events;using the categories for access to trace-related information for trace analysis by the computerized trace analysis process; andpresenting the categorized trace-related information on a user screen. 2. The method of claim 1, wherein a summary file summarizing the captured trace is created including the categories information and the summary file is used for the access to trace-related information by the computerized trace analysis process. 3. The method of claim 2, the method of claim 1, wherein the category of information type includes I/O activity. 4. The method of claim 2, wherein the category of components includes logical volume representation of the data storage volumes. 5. The method of claim 2, wherein the computerized trace analysis process includes communication to the trace capture process for being responsive to the act of a trace being captured. 6. The method of claim 2, wherein the computerized trace analysis process includes communication to the trace capture process for being responsive to the act of a trace being captured. 7. The method of claim 1, wherein the category of information type includes I/O activity. 8. The method of claim 1, wherein the category of components includes logical volume representation of the data storage volumes. 9. The method of claim 1, wherein the computerized trace analysis process includes communication to the trace capture process for being responsive to the act of a trace being captured. 10. The method of claim 9, wherein the category of components includes logical volume representation of the data storage volumes. 11. A computer program product available from computer readable medium for preparing a captured trace, the captured trace being stored for later use, of I/O workload activity experienced on one or more data storage volumes included with a data storage system for being analyzed by a computerized trace analysis process, the computer program product when loaded into a computer system, causing the computer system to execute the steps of:preparing the captured trace for being analyzed by categorizing information from the captured trace into categories related to (i) components in the data storage system experiencing the traced workload activity and (ii) information type including response times and task events;using the categories for access to trace-related information for trace analysis by the computerized trace analysis process; andpresenting the categorized trace-related information on a user screen. 12. The program product of claim 11, wherein a summary file summarizing the captured trace is created including the categories information and the summary file is used for the access to trace-related information by the computerized trace analysis process. 13. The program product of claim 11, wherein the category of information type includes I/O activity. 14. The program product of claim 11, wherein the category of components includes logical volume representation of the data storage volumes. 15. The program product of claim 11, wherein the computerized trace analysis process includes communication to the trace capture process for being responsive to the act of a trace being captured. 16. A system for preparing a captured trace, the captured trace being stored for later use, of I/O workload activity experienced on one or more data storage volumes included with a data storage system for being analyzed by a computerized trace analysis process, the system including:a data storage system including one or more data storage volumes;a computer in communication with the data storage system including program logic for carrying out the computer-executed steps of:preparing the captured trace for being analyzed by categorizing information from the captured trace into categories related to (i) components in the data storage system experiencing the traced workload activity and (ii) information type including response times and task events;using the categories for access to trace-related information for trace analysis by the computerized trace analysis process; andpresenting the categorized trace-related information on a user screen. 17. The system of claim 16, wherein a summary file summarizing the captured trace is created including the categories information and the summary file is used for the access to trace-related information by the computerized trace analysis process. 18. The system of claim 16, wherein the category of information type includes I/O activity. 19. The system of claim 16, wherein the category of components includes logical volume representation of the data storage volumes. 20. The system of claim 16, wherein the computerized trace analysis process includes communication to the trace capture process for being responsive to the act of a trace being captured.
	description	This is a continuation of U.S. patent Ser. No. 11/820,966, filed Jun. 21, 2007, which is a continuation of U.S. patent Ser. No. 10/715,069, filed Nov. 17, 2003, which is a continuation of International Application No. PCT/EP02/05274, filed May 14, 2002, all of which are incorporated by reference herein. The invention lies in the boiling water reactor technology field. More specifically, the invention relates to a method for protecting the components of the primary system of a boiling water reactor in particular from stress corrosion. In a boiling water reactor, the coolant which comes into contact with the reactor core is known as primary coolant, and the lines and components which are exposed to the primary coolant are known as the primary system. In addition to the reactor pressure vessel, the primary system of a boiling water reactor includes systems of lines as well as various internal fittings and pumps. The components generally consist of stainless steel, for example of a CrNi steel, or an Ni-base alloy, such as Inconel® 600 (Inco Alloys International, Inc.). Radiolysis of the primary coolant causes, inter alia, the reaction products hydrogen peroxide, oxygen, and hydrogen to form in the boiling water reactor. The oxidizing conditions which result from the excess of oxidizing agents promote corrosion, in particular stress corrosion cracking, of the components. To remedy this, it is known to admix hydrogen with the primary coolant. This bonds oxidizing agents contained in the primary coolant and shifts the electrochemical potential of the component surfaces toward negative values. A drawback of the conventional method is that relatively large quantities of hydrogen are required to ensure sufficient protection against corrosion. The high demand for hydrogen, which entails corresponding costs, is attributable not least to the fact that the electrochemical oxidation of the hydrogen on the component surfaces which are covered with an oxide layer is subject to considerable reaction inhibition, and this has to be compensated for by increased hydrogen concentrations. A further drawback is the outlay on apparatus for metering the gaseous hydrogen. European patent disclosure EP 0736878 describes a method in which the oxide layer of the component surfaces in the primary System is doped with precious metal, which makes it possible to use smaller quantities of hydrogen. German published patent application DE 100 30 726 A1 describes a method in which the quantities of hydrogen and precious metal are supposed to be reduced by coating the component surfaces with a film which includes a substance with a photocatalytic action. The substances with photocatalytic action that are used—preferably Ti02 and Zr02—are N-type semiconductors which are excited by the Cherenkov radiation which is present in the reactor, shifting the corrosion potential of the component surfaces toward negative values. Soviet Union patent disclosure SU 653953 describes a system having to do with what is referred to in the document as “boiling nuclear reactors.” There, an alcohol is added into the primary coolant instead of hydrazine. While relatively little information is presented in the document concerning the operational setup of the reactor, certain statements strongly suggest that the boiling reactor of the prior art publication is not a boiling water reactor according to the Western understanding. One such hint is that the publication states that, in its prior art, hydrazine had been introduced in such boiling nuclear reactors during the reactor operation for the purpose of providing corrosion protection. The addition of hydrazine, however, during the operation of a boiling water reactor would be entirely prohibited. The Soviet document discloses corrosion protection measures by way of the addition of alcohol in the coolant/moderator. The specific concentration disclosed is approximately 10 to 105 μmol/kg (≈0.32 to 3200 ppm for methanol) in order to completely prevent oxygen formation during the radiolysis of the coolant. In order to ensure this, the disclosed alcohol concentration must necessarily be present at those locations at which the radiolysis processes are the strongest, that is, at the fuel rods in the reactor core. A problem associated with very high alcohol concentration is that a relatively large portion of the alcohol remains unused, i.e., it is not oxidized by radiolysis oxygen or decomposed by the radiolysis, it subsequently passes through a phase change into the vapor phase and then reaches the steam turbine and the condenser downstream of the steam turbine. There, the alcohol is cooled to about 40° C. At this temperature, only a small proportion of the alcohol is dissolved in the liquefied condensate which is fed back into the reactor pressure vessel in the form feedwater. The by far largest proportion is contained in the vapor phase. The latter is not simply let go into the environment but it is transported via an off-gas path within which a catalytic recombination of hydrogen and oxygen to water is effected. An alcohol component in the vapor phase could, on the one hand, disturb the recombination. On the other hand, additional functional elements and processing steps would have to be provided in order to hold back the alcohol or to convert the same into a non-damaging form. High alcohol contents, furthermore, lead to radiolysis in the reactor due to the high radiation density, which results in products such as CO2, formaldehyde, and formic acid. These products, of course, are undesirable in the reactor pressure vessel itself and in the downstream vapor carrying systems such as the condenser. Besides an increase in the conductivity of the primary coolant, they can lead to a decrease in the pH which has a negative effect on the component corrosion. Yet, it is exactly the component corrosion which is to be avoided or reduced with the addition of alcohol. An object of the present invention is to provide a method for protecting the components of the primary system of a boiling water reactor which overcomes the above-mentioned disadvantages of the heretofore-known devices and methods of this general type and which ensures efficient protection against corrosion with little outlay on materials and time. With the foregoing and other objects in view there is provided, in accordance with the invention, a method for protecting components of a primary system of a boiling water reactor having a pressure vessel and a feedwater line opening out into the pressure vessel. The novel method, which is particularly suitable to protect the components against stress corrosion cracking, comprises the following steps: providing an alcohol that is oxidizable under operating conditions of the primary system; feeding the alcohol into a primary coolant to establish an alcohol concentration of from 0.1 to 300 μmol/kg (≈0.0032 to 9.6 ppm for methanol) in a downcomer, the downcomer extending downward at an opening of the feedwater line, with surfaces of the components still being bright or covered only by a native oxide layer. In accordance with an added feature of the invention, the alcohol concentration is adjusted to less than 10 μmol per kg (≈0.32 ppm for methanol). In other words, the objects of the invention are achieved by a method in which an alcohol that can be oxidized under the conditions prevailing in the reactor system, preferably in liquid phase, is fed into the primary coolant instead of hydrogen, with the component surfaces being bright or being covered only by a native oxide layer. In this context, a native oxide layer is to be understood as meaning an oxide layer which forms as a result of corrosion to the component material, if appropriate with the intercalation of foreign metals or foreign metal oxides, during reactor operation or during an oxidizing pretreatment. It has been found that the metering-in of an alcohol of the above type as the only measure is sufficient to reduce the corrosion potential of the component surfaces to values of lower than −230 mV, and it is possible to dispense with complex coatings in particular comprising substances with a photocatalytic action. The advantage of an alcohol over hydrogen as reducing agent is firstly that it can be metered in liquid form or as a solution. A liquid is more easy to feed into the primary coolant than a gaseous substance in terms of the apparatus required. Furthermore, the compounds mentioned offer advantages in terms of handling and storage. Finally, they are less expensive than hydrogen, with the result that the plant operating costs can also be reduced. In accordance with a concomitant feature of the invention, the component surfaces are doped with precious metal, for example with Pt, with the result that a lower concentration of alcohol is required in the primary coolant. The alcohol concentration is maintained at between 0.1 and 300 μmol per kg (≈0.0032 to 9.6 ppm for methanol) of the primary coolant and, in a preferred embodiment, it is maintained at less than 10 μmol/kg. It is expedient for the alcohol to be fed into the condensate or feedwater system. The quantity which is metered in is in this case such that the abovementioned concentration is established in the downcomer of the boiling water reactor. The downcomer is the area in the reactor pressure vessel which extends downward from the opening points of the feed tubes. It is preferable to use methanol, ethanol and propanol. However, formic acid, formaldehyde, and acetaldehyde are also eminently suitable. As noted above, the metering-in of alcohol may lead to several disadvantageous results. That is, it is in effect a balancing act between the positive and the negative effects thereof. The instantly claimed invention provides a successful compromise with highly improved corrosion protection while the negative effects of the alcohol are virtually unnoticeable. This is particularly so when the alcohol concentration is maintained at below 10 μmol/kg (≈0.32 ppm for methanol). 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 method for protecting the components of the primary system of a boiling water reactor in particular from stress corrosion cracking, it is nevertheless not intended to be limited to the details shown, since various modifications and structural changes may be made therein without departing from the Spirit of the invention and within the scope and range of equivalents of the claims. The construction and method of operation of the invention, however, together with additional objects and advantages thereof will be best understood from the following description of specific embodiments when read in connection with the accompanying drawings. Referring now to the figures of the drawing in detail and first, particularly, to FIG. 1 thereof, there is shown a highly simplified illustration of a boiling water reactor. A pressure vessel 1 of the reactor houses fuel assemblies 2 or fuel elements. An alcohol of the above-mentioned type, preferably methanol, is injected into a feedline 3, which continues inside the pressure vessel in the form of an annular distributor line, to protect against corrosion and in particular against stress corrosion cracking (IGSCC). The reactor is in an operating state in which the components of the reactor, i.e. for example the pressure vessel 1 and the non-illustrated core grid, which usually consist of CrNi steel or an Ni-base alloy, are bright or are covered only with a native oxide layer. The former case occurs, for example, if an oxide layer has been removed from the component surfaces during maintenance work. The quantity injected into the feedline 3 is such that a concentration of from 0.1 to 300 μmol/kg (≈0.0032 to 9.6 ppm for methanol) of alcohol, in particular methanol, is established in the downcomer 4 which adjoins the feedline 3 at the bottom. The optimum concentration of alcohol is dependent on various factors, such as the component material, the presence of precious metal doping, etc., and is therefore to be determined on a case-by-case basis for each individual reactor. In a specific embodiment, the concentration is set to less than 10 μmol/kg (≈0.32 ppm for methanol) which, in a given context, provides for an acceptable compromise with regard to good corrosion protection and virtually negligible disadvantages otherwise associated with the alcohol. Tests using Pt and CrNi steel electrodes were carried out in order to test the theoretical effectiveness of the proposed method. The electrodes made from CrNi steel were subjected to preliminary oxidation for 500 hours at 280° C. with a water chemistry that corresponds to the conditions of use in the reactor. The CrNi steel electrodes which have been pretreated in this way and the Pt electrodes were arranged in an autoclave through which hot water at a temperature of 280° C. was flowing. The chemistry of the circulated water was set so as to correspond to the conditions in a boiling water reactor. The oxygen content was kept at between 0.2 and 2 ppm. The reducing agents used were methanol and, for comparison purposes, hydrogen. The potentials of the electrodes were determined as a function of the methanol or hydrogen content and are plotted in the diagram shown in FIG. 2 against the methanol/oxygen or hydrogen/oxygen molar ratio. In FIG. 2, the indication “CrNi” indicates CrNi steel. It can be seen that metering of methanol results in a protective action which is comparable to that achieved by metering hydrogen. In both cases, the Pt potential is reduced to below the protection potential of −230 mV. In the case of the undoped CrNi steel electrode, it is likewise possible to observe similar electrochemical activities with both methanol and hydrogen. However, in order in this case to reduce the potential to below the protection potential, it is necessary to establish significantly higher molar ratios. Therefore, it is necessary to operate with a lower oxygen content or with an excess of reducing agent. A potential of −500 mV was measured for an oxygen content of less than 10 ppb and a methanol content of 2 ppm (62.5 μmol/kg). Although hydrogen and methanol or other alcohols, such as in particular ethanol or propanol, have similar electrochemical activities to hydrogen, their reactivity with respect to the strongly oxidizing OH radicals formed during the radiolysis of water is greater. A further advantage of the proposed method results from the significantly lower volatility of the alcohols in question. Whereas a large proportion of the hydrogen which is metered in is converted into the vapor phase, discharged with this phase and has to be catalytically oxidized as a non-condensable gas in the off-gas system of the reactor by adding stoichiometric quantities of oxygen, the proportion of alcohols which is converted into the vapor phase is lower. Moreover, the proportion of alcohol discharged with the vapor phase can be virtually completely condensed and thereby recycled to the reactor. Consequently, the outlay an chemicals, apparatus and control engineering measures is reduced compared to conventional methods.
	abstract	A dosimeter container comprising a housing portion and a shield portion that surrounds the housing portion is provided. The housing portion houses a radiation dosage measuring device for measuring a dosage of predetermined radiation other than neutron radiation. The shield portion is composed of a member made of a material that transmits predetermined radiation and shields neutron radiation. The shield portion is a LiF sintered body, in particular, a 6LiF sintered body. Further, the shield portion includes at least two or more shield portion components (a body portion and a lid portion), in which adjacent members can abutt against each other. The housing portion is same size as or larger than the size of the radiation dosage measuring device; and the housing portion extends over the entirety of the components. The dosimeter container is preferably used as a dosage measuring body having a radiation dosage measuring device stored in the housing portion.
042749204	description	For convenience of reference, similar structural elements are designated by the same reference numerals throughout the drawings. DETAILED DESCRIPTION Referring to the drawings, FIG. 1 is a schematic illustration of a nuclear reactor in accordance with the general concept of the invention comprising a reactor core 10 including a plurality of nuclear fuel elements (not shown in FIG. 1), a surrounding reactor vessel 11, including appropriate radiation, thermal and environmental shielding, fuel dispersal guide conduits (or thimbles) 12, a heat sink 13 of heat capacity adequate to absorb the residual heat contained in the reactor core at the time of dispersal and spontaneously generated thereafter, and an optional refueling station 14 to perform on-line refueling. Control rods (not shown) are conventionally provided for reactor power control and non-emergency shutdown. The fuel dispersal guides 12, only two of which are shown, provide guide paths for fuel elements from the reactor vessel 11 to heat sink 13. The guides generally diverge along these paths from a critical configuration within the reactor core to a non-critical dispersal configuration within the heat sink. Conveniently, the guides comprise hollow tubular members, in the form of pipes or thimbles, which fan out from the reactor pressure vessel into the heat sink. These members are closed with respect to the heat sink in order to maintain a pressure barrier between the reactor core and the heat sink. The portions of guides 12 immersed in water of the heat sink are preferably of sufficient length to accommodate the entire lengths of fuel stacks in their non-critical configuration. Advantageously, the guides 12 comprise vertical portions immersed in water which can accommodate the entire fuel content of respective fuel stacks. Inside the reactor vessel the fuel elements are conveniently supported and guided by fuel supporting guides 23. Heat sink 13 is conveniently a body of water such as a short, wide tunnel open to a large body of water such as a river, lake, or ocean. Advantageously, the water level in the tunnel is gravitationally maintained by the natural water level of the outside body of water. Preferably, the heat sink is accommodated within the seismic support region for the reactor. The reactor core 10 is disposed sufficiently adjacent to and preferably above the heat sink 13 so that the utilizable energy content of the pressure vessel and the core, e.g., the potential energy due to gravitational forces and the potential energy due to pressurized contents of the pressure vessel and its associated accumulator tanks, is adequate to drive the bulk of the core fuel content from the reactor vessel, along the dispersal guides 12 and into the heat sink in a non-critical dispersal configuration wherein the chain-reacting relationship is terminated and the dispersed fuel is gradually cooled down by conduction through the guide tube walls into the heat sink. As a safety feature in case of damage to the fuel dispersal guides, isolation means can be provided for isolating heat sink 13 from the general environment. Such isolation can take the form of an arrangement for permitting the flow of water into the tunnel but preventing it from leaving, such as a fast dropping gate in a throat connecting to the outside water. Fast dropping gate 16--which may be connected to a gate control--is provided for throat 17 in wall 18. The water level outside wall 18 is higher than the water level in the tunnel for replenishing water in the heat sink. Such isolation means would prevent widespread contamination of the outside body of water. Replenishing means is thus provided for replenishing the water in the tunnel which is evaporated by residual heat of the dispersed core in the thimbles by allowing for overflow feeding from the higher water level in the outside lake or river. Operation of the isolation and replenishing means should follow any detection of radioactivity in the water of heat sink 13, and decontamination of the isolated portion of the heat sink should follow. A refueling station 14 is conveniently coupled to a continuation of fuel dispersal guides 12 to permit on-line refueling after fuel elements are retrieved beyond the cooling zone of the heat sink. Isolation valves 15 are advantageously provided to control access between the reactor vessel and the dispersal guides and between the dispersal guides and the refueling station. FIG. 2(A) is a transverse cross section of a portion of a nuclear reactor having a passive fuel element dispersal arrangement in accordance with a preferred embodiment of the invention. In substance, nuclear fuel elements 21 are disposed inside a reactor pressure vessel 11 in a plurality of stacks or columnar arrangements 22 (only one of which is shown). Respective dispersal guide thimbles 12 penetrating the pressure vessel, but sealed with respect to the reactor vessel exterior, are provided for receiving fuel elements from one or more of the columnar arrangements while maintaining the pressure barrier. The fuel dispersal guide tubes can conveniently be made of stainless steel, such as type 304ss or Inconel (ASME-SB-163). Preferably, guide means are provided within the pressure vessel for constraining each column of fuel elements in alignment with the respective dispersal guide thimbles and for supporting the fuel elements inside the core. Such guide means can conveniently comprise a plurality of longitudinally straight-edged stainless steel members 23 radially positioned around each fuel column. In a preferred arrangement shown in FIG. 2(B), at least three such elements are spaced around each column approximately 120.degree. apart. In an alternative arrangement shown in FIG. 2(C), four elements are used and spaced approximately 90.degree. apart. These guides facilitate the up-and-down sliding of fuel elements essential for emergency shutdown and refueling operations. In addition, removable support means are provided for constraining the nuclear fuel elements within the operative region of the reactor core during the operating cycle and, advantageously, removable sealing means are provided for sealing the dispersal guide thimbles from the reactor core during normal core operation. In the embodiment of FIG. 2, both the support and sealing functions are performed by a sealing piston 25 which is held in position in guide thimble 12 against conical sealing shoulders 26 by hydraulic means, such as a source (not shown) of pressurized fluid 27 coupled to thimble 12. Condensate or demineralized water of reactor coolant quality may conveniently serve as pressurized fluid, since any leaks of such water past the piston (particularly during refueling) will not contaminate, or in any way degrade, the reactor coolant. Preferably, sealing piston 25 comprises two or more cylindrical surfaces 25A and 25B, equipped with sealing rings, and arranged in tandem by a relatively flexible link 25C in order that the piston can easily follow bends in thimble 12. The upper part of the piston has a matching conical surface to the sealing shoulders 26 to effectuate a tight seal. As long as fluid pressure P remains greater than counteracting pressure p, which is a sum of pressure of the reactor coolant and weight of the fuel column, the reactor remains sealed, and the fuel is positioned in the core. One or more dummy spacer elements 28 can be conveniently provided between the piston and the fuel elements in order to space the fuel elements away from the sealing area into the operative region of the reactor core. In the event of conditions requiring emergency shutdown, support for columnar arrangement 22 is removed as by a controlled decrease in the pressure P of pressurized fluid 27 supporting piston 25, and the fuel elements drop from the reactor core through dispersal thimble 12 into a non-critical dispersal configuration within a heat sink (not shown) in the manner described in connection with FIG. 1. This emergency dispersal is substantially completely passive in that the energy required for dispersal is provided by the potential energy due to gravitational pull on the fuel elements and the potential energy due to the pressurized reactor coolant in the pressure vessel. In case of a loss of coolant accident (LOCA), the surge drums or the accumulator and volume control tanks (not shown but typical features in present day reactor designs) provide sufficient pressure for a sufficient duration to effectuate a complete fuel dispersal into the thimbles. At the completion of dispersal, valves 15 of FIG. 1 near the reactor vessel can be closed; and the pressure in the thimbles can be passively maintained by an additional small tank (not shown) connected to the thimbles. The speed of movement of the sealing piston and the fuel elements can be easily controlled by controlling the pressure of the pressurized fluid P. Such control can be exercised simultaneously for all of the thimbles, as in emergency shutdown, or selectively with only one or a few thimbles, as in refueling or retrieving and replacing damaged fuel elements. In case refueling is desired, the stacks containing the elements to be replaced are selectively withdrawn from the core through the respective appropriate thimbles to refueling station 14. New fuel elements are substituted for the fuel elements to be replaced, and the process is reversed. Thus, the refueled stack is driven up through the fuel dispersal guides into the reactor vessel by operating fluid of a pressure P, greater than the counteracting pressure p until the sealing piston seals the reactor again at the conical piston-to-shoulder contact just below the reactor vessel. The pressure relationship P>p is maintained throughout the operation of the reactor and remains a relatively static feature. This capability for selective withdrawal has the important advantage of permitting on-line refueling or replacement of selected stacks without reactor shutdown and costly disassembly. Depending upon the proximity of the fuel dispersal guides to each other, in some areas, such as near the reactor vessel, it can be advantageous to supply a neutron-absorbing shield 28 around the thimbles in order to prevent criticality. Such a shield can be comprised of typical control rod materials of high neutron absorption cross section such as boron, cadmium, hafnium, or alloys thereof. Preferred shields comprise boron carbide or boron clad in or alloyed with stainless steel. As an additional safety feature, intended for an unlikely event of non-effective core evacuation combined with a concurrent failure of coolant circulation that may result in a core melt-down, the inside reactor bottom surface can be provided with a refractory bottom 29 shaped to funnel any molten material from fuel elements 21 into cooling thimbles 12 where it will cool and solidify. In the unlikely event of core melting, this feature would provide non-critical emergency dispersal of the molten core material. While the columnar arrangement 22 of FIG. 2 is an arrangement of individual fuel elements 21, it is clear that one could substitute for each individual fuel element a cluster of fuel elements, and the emergency dispersal system would operate in substantially the same manner on a correspondingly enlarged scale employing simultaneous dispersal of multiple fuel elements into fewer but larger cooling thimbles. A preferred embodiment of a hexagonal arrangement of fuel elements in cluster, as exemplified in FIG. 6, has geometrical advantages in that it can be easily patterned into core cross section and can be easily accommodated in tubular cooling thimbles after shutdown or during refueling. FIGS. 3(A) and 3(B) illustrate a preferred form of a nuclear fuel element 21 for use in the embodiment of FIG. 2, comprising a clad pellet 30 of nuclear fuel 31, such as enriched uranium dioxide, having a longitudinal dimension preferably not exceeding about three times its maximum axial direction. This compact geometry is desirable in order to permit the fuel element to readily pass through embodiments of the invention using curved dispersal thimbles. Each individual fuel element is provided with a fission gas plenum 32 at each end. The cladding 33 is preferably of Zircaloy and advantageously defines a plurality of longitudinally extending concave grooves, the spring action of which permits cladding expansion without reduction of cladding wall thickness upon swelling. This feature allows fuel elements to be fabricated with the cladding tightly compressed onto the nuclear fuel for a better heat transfer from fuel to cladding to coolant. Spring action of the grooves will maintain tight fuel-to-cladding contact throughout the life of the fuel element. Gained heat conductivity through tight fuel-to-cladding contact may be employed to strengthen the cladding walls for better isolation of radioactive decay components in the fuel from the coolant, thus lessening the likelihood of coolant contamination. Specifically, the fuel elements can be provided with cladding which is of reinforced (greater) thickness than that required to obtain comparable heat transfer in a non-compressed structure. FIG. 4(A) is a transverse cross section of a portion of a nuclear reactor having a passive fuel element dispersal system in accordance with a second embodiment of the invention. In substance, the arrangement of FIG. 4(A) is substantially the same as that of FIG. 2, except that the fuel elements 21 are provided with longitudinally extending hollow centers, and guide 23 is a guide rod member, such as a stainless steel rod or tube, inserted through the hollow centers of the respective fuel elements. In addition, the first (uppermost) dummy spacer 28 can conveniently be provided with an enlarged funnel-shaped receptacle to engage the guide rod 23 during the fuel loading operation. In a preferred arrangement, one or more intermediate dummy spacers 30 may be provided to be positioned at each level of one or more core support frames 31, as illustrated in a detail of FIG. 4(B). Such an arrangement limits horizontal movement and vibration of the fuel elements. In the event of emergency shutdown, fuel element support is removed and the fuel elements drop off guide rod 23 into the dispersal thimble 12. In all other respects, the operation of this embodiment is substantially the same as that described in connection with FIGS. 1 and 2. FIGS. 5(A) and 5(B) illustrate a preferred fuel element 21 for use in the arrangement of FIG. 4. In substance, this fuel element is the same as that described in connection with FIG. 3 except that it includes a longitudinally extending hollow center 50 for receiving guide rod 23. Advantageously the cladding in the hollow center region is provided with longitudinally extending concave grooves 51 for accommodating fuel swelling. Like the fuel element of FIG. 3, the longitudinal dimension of this fuel element preferably should not exceed about three times the maximum axial dimension so that the fuel element can be utilized in embodiments using a curved dispersal thimble. This arrangement offers an advantage over the design in FIG. 3 in that the hollow center eliminates the inherent problem of fuel melting in the center of a solid pellet. Enriched uranium dioxide has been cited in connection with FIGS. 3A, 3B, 5A, and 5B, however essentially the same specific embodiments can be employed for plutonium dixoide, PuO.sub.2, containing Pu.sup.239 as a fuel or as a mixture of PuO.sub.2 and UO.sub.2, containing Pu.sup.239 and U.sup.235. In such embodiment 316 stainless steel can be used as an alternate material for fuel element cladding. While the invention has been described in connection with only a small number of specific embodiments, it is clear that numerous and varied additional arrangements can be devised by those skilled in the art without departing from the spirit and scope of the invention.
	claims	1. A high voltage insulating radiation enclosure comprising:a first truncated cone section and a second truncated cone section;the two truncated cone sections secured together at their respective bases by an overlap joint;an interior space defined by the two truncated cones sections;the first and second truncated cone sections having walls, the walls made of a material comprising:a) a polymer matrix andb) barium sulfate within the polymer matrix in an approximate amount of at least 10% by volume;a first emission port passing through at least one wall;a second electrical port passing through at least one walls. 2. The high voltage insulating radiation enclosure of claim 1, further comprising an X-ray tube disposed within the hollow body. 3. The high voltage insulating radiation enclosure of claim 1, further comprising at least one oil port passing through the walls. 4. The high voltage insulating radiation enclosure of claim 1, wherein the polymer matrix comprises at least one member selected from the following group: epoxy, polyester, polyurethane, silicone rubber, bismaleimides, polyimides, vinylesters, urethane hybrids, polyurea elastomer, phenolics, cyanates, cellulose, flouro-polymer, ethylene inter-polymer alloy elastomer, ethylene vinyl acetate, nylon, polyetherimide, polyester elastomer, polyester sulfone, polyphenyl amide, polypropylene, polyvinylidene flouride, acrylic, homopolymers, acetates, copolymers, acrlonitrile-butadiene-stryene, flouropolymers, ionimers, polyamides, polyamide-imides, polyacrylates, polyether ketones, polyaryl-sulfones, polybenzimidazoles, polycarbonates, polybutylene, terephthalates, polyether sulfones, thermoplastic polyimides, thermoplastic polyurethanes, polyphenylene sulfides, polyethylene, polypropylene, polysulfones, polyvinylchlorides, stryrene acrylonitriles, polystyrenes, polyphenylene, ether blends, styrene maleic anhydrides, allyls, aminos, polyphenylene oxide, and combinations thereof. 5. The high voltage insulating radiation enclosure of claim 1, wherein the polymer matrix comprises epoxy resin is an approximate amount of 50% to 70% by volume. 6. The high voltage insulating radiation enclosure of claim 1, further comprising:c) a third material. 7. The high voltage insulating radiation enclosure of claim 6, wherein the third material comprises at least one member selected from the following group: electrically insulating materials, binders, high density materials and combinations thereof. 8. The high voltage insulating radiation enclosure of claim 6, wherein the third material comprises at least one member selected from the following group: tungsten, lead, platinum, gold, silver, tantalum, calcium carbonate, hydrated alumina, tabular alumina, silica, glass beads, glass fibers, magnesium oxide/sulfate, wollastonite, stainless steel fibers, copper, carbonyl iron, iron, molybdenum, nickel and combinations thereof. 9. An electrical insulator for an ion source, the insulator comprising:a generally annular body having a diameter of at least 6 inches;the body having at least one vacuum sealing surface dimensioned and configured to provide a tight seal;at least one alignment pin projecting from the vacuum sealing surface of the insulator;at least one metal insert secured to the body;the body made of a material comprising:a. a polymer matrix andb. barium sulfate within the polymer matrix in an approximate amount of at least 35% by volume.
	summary
059164972	abstract	A method of manufacturing a ceramic article includes the steps of: forming a body of particulate ceramic material having two end portions, a length and a middle portion, wherein a bulk particle density of each of the two end portions is less than a bulk particle density of the middle portion; compressing the body at opposing ends of the body; and thereafter sintering the body to form the ceramic article. The invention may be adapted for production of nuclear fuel pellets.
	abstract	A sealing device for a jet pump of a boiling water reactor is provided. The jet pump includes an inlet mixer and a diffuser receiving the inlet mixer at a slip joint such that an outer circumferential surface of the inlet mixer is received in an inner circumferential surface of the diffuser at the slip joint. The diffuser includes a plurality of guiding fins, each guiding fin including a radially inner surface, a radially outer surface and lateral surfaces extending radially between the inner and outer surfaces. The sealing device includes a seal configured for sealingly contacting the outer circumferential surface of the inlet mixer and a collar configured for holding the seal against the outer circumferential surface of the inlet mixer. The collar includes portions configured for being received radially between the radially inner surfaces of the guiding fins and the outer circumferential surface of the inlet mixer. The sealing device further includes a clamp configured for contacting the radially outer surfaces of the guiding fins to axially clamp the guiding fins. A method of mounting a sealing device onto a slip joint of a jet pump of a boiling water reactor is also provided.
	abstract	The present invention provides a remote-controlled in-pile creep test system used in in-pile creep tests for measuring and determining mechanical properties of nuclear materials irradiated in research reactors. The in-pile creep test system includes a creep tester vertically installed in a water pool of a nuclear reactor and used in a tensile, compressive or repeated loading or low cyclic fatigue creep test; a detecting unit electrically connected to the creep tester and used for detecting a temperature of the tester and creep strain of a specimen installed in the tester; a gas supply unit connected to the creep tester through gas supply tubes and controllably supplying helium gas from a helium gas reservoir tank to the tester or returning helium gas from the tester to the tank by an operation of an air compressor and a vacuum pump; and a control unit electrically connected to both the detecting unit and gas supply unit so as to control an operation of the creep test system in response to results of a comparison of input data from the detecting unit and the gas supply unit with stored data, whereby the creep tester has simple structure for the convenience of installing the specimen and assembling the capsule parts and also is easily cut and disassembled in a hot cell, and prevents damage or breakage of the specimen during a procedure of removing the specimen from the tester after a creep test, and is used for performing creep tests for specimens having a variety of shapes and sizes.
	claims	1. A device for inspecting welds in nozzles of a containment vessel filled with water in a nuclear power plant comprising:a main rail;buoyancy packs attached to said main rail;a plurality of radially expandable feet attached to said main rail near each end thereof, at least three of said expandable feet radially dispersed around said main rail, each foot of said expandable feet having a snubber for gripping said nozzle to hold said device in position, rollers adjacent said snubbers on said expandable feet aiding in positioning said device in said nozzle;a plurality of at least three transducer clusters attached to said main rail and being radially extendable outward therefrom, said transducer clusters having at least three transducers in each transducer cluster, said transducer clusters being radially located at equal angles around said main rail;thrusters attached to said main rail for moving said device in said water into one of said nozzles;a first cylinder for radially extending said expandable feet to secure said device in one of said nozzles on both sides of one of said welds;a first motor controller for moving said transducer clusters along said main rail until said transducer clusters are radially inside said weld;a second cylinder radially extending said transducer cluster adjacent to said weld;a second motor controller rotating said transducer clusters adjacent said weld so that said transducer clusters send test signals to said weld and receive reflected test signals from said weld to determine condition of said weld, said transducer clusters being rotated around said main rail while sending test signals to said weld and said receiving said reflected test signals from said weld. 2. The device for inspecting welds in nozzles of a containment vessel filled with water in a nuclear power plant as recited in claim 1 wherein said device is connected to a control processing unit located outside said water, said control processing unit communicating with motor controllers connected to said thrusters to cause a movement of said device in said water, inputs being received from said device and transmitted via said motor controllers to said control processing unit.
	claims	1. A sanitizing apparatus for writing utensils, comprising a housing assembly having a top wall, a base, and first, second, third, and fourth walls, said first and second walls are perpendicularly disposed to said third and fourth walls, and said third and fourth walls are lateral sidewalls that are in a parallel and spaced-apart relationship with respect to each other, said housing assembly also comprises an angled wall, said angled wall is in between said third and fourth walls, said angled wall protrudes outwardly beyond said top wall a first predetermined distance and protrudes outwardly beyond said first wall a second predetermined distance, said angled wall terminates at said base, said angled wall defines an elongated channeled slot through said top wall to receive at least one writing utensil, said second predetermined distance of said angled wall defining a tray that terminates with a lip, said lip to prevent said at least one writing utensil from falling off said tray, said housing assembly further comprises ultraviolet and ozone generating means for radiating said at least one writing utensil within said housing assembly with rays and ozone, to effectively sterilize bacteria and biological germs existing within said housing assembly and on said at least one writing utensil. 2. The sanitizing apparatus for writing utensils set forth in claim 1, further characterized in that said housing assembly further comprises electronic means to notify a user when said ultraviolet and ozone generating means is operating, wherein said electronic means comprises at least one visual indicator that illuminates to notify said user when said bacteria and biological germs are being sterilized from said at least one writing utensil. 3. The sanitizing apparatus for writing utensils set forth in claim 2, further characterized in that said housing assembly further comprises a battery compartment for a battery power source. 4. The sanitizing apparatus for writing utensils set forth in claim 2, further characterized in that said housing assembly further comprises an electrical plug to connect to an electrical outlet.
	description	This application is the U.S. National Phase application under 35 U.S.C. § 371 of International Application No. PCT/IB2015/056617, filed Sep. 1, 2015, published as WO 2016/038504 on Mar. 17, 2016, which claims the benefit of U.S. Provisional Patent Application No. 62/047,127 filed Sep. 8, 2014. These applications are hereby incorporated by reference herein. Technical Field The present disclosure relates to a method and an imaging system for generating spectrally different X-ray images with an X-ray source and an X-ray detector. More particularly, the present disclosure relates to including a filter providing different spectral filtration within an X-ray system in order to produce a spectrally modulated beam such that neighboring pixels of the X-ray detector receive different spectra, and using this spectral information to perform means of spectral X-ray imaging. Description of Related Art Computed tomography (CT) is the science of recovering a three-dimensional representation of a patient or object by utilizing projection views with different orientations. From this volume, e.g., two-dimensional cross-sectional images can be displayed. CT systems typically include an X-ray source collimated to form a cone beam directed through an object to be imaged, i.e., a patient, and received by an X-ray detector array. The X-ray source, the cone beam, and the detector array may be rotated together on a gantry within the imaging plane, around the imaged object. However, the X-ray radiation imposes unwanted effects. In the medical imaging domain, an unwanted effect may be the radiation dose that a patient receives, as it may induce damage to cells and genes. As a further unwanted effect, the interaction of X-ray radiation with matter imposes scattered X-ray radiation, which adds in the detector to the signal of interest, i.e., the signal of the primary radiation. As a most obvious method to reduce the unwanted effects, measures are taken to limit the amount of total X-ray exposure to a minimum, which is required to acquire images. To reduce the negative effects, three elements are used to form the cone beam. First, a collimator defines a cone shape such that the cone beam covers exactly the whole detector area in order that each detector imaging element (denoted herein as a “pixel”) is exposed to the beam, but the overlap to the non-detector area is reduced to a minimum. Second, a bowtie-shaped device, known as a “beam shaper,” “bow tie” or sometimes also as a “wedge,” is placed in the path of the X-ray beam. The wedge, functioning as an X-ray attenuation filter, is generally made of a light metal, such as aluminum, or a synthetic polymer, such as Teflon, having an X-ray absorption spectral characteristic near that of water, and, hence, the human body. The wedge is intended to compensate for the variation in thickness of the imaged body. The X-rays that pass through the center of the imaged body, normally the thickest part, are least attenuated by this filter, whereas the X-rays that pass through the periphery of the imaged body, normally the thinnest part, are more attenuated by this filter. The result of this selective attenuation is a better distribution of the X-ray dose. This allows, on one hand, for a total dose reduction for the scanned patient. On the other hand, the X-rays impinging on the detectors have a less spatially varying intensity profile. The wedge may therefore allow use of more sensitive X-ray detectors, thus reducing the total dynamic range of x-ray intensities to be detected. Finally, as a third element to reduce negative effects, a spatially homogeneous filter (typically in the form of a metal plate, e.g., made of copper) is induced to absorb mainly the low energy components of the spectrum. The low energy components of the plain X-ray spectrum are typically that strongly attenuated by an object or a patient that they do not significantly contribute to a measured signal. Thus, the filter reduces the total dose a patient is exposed to with an acceptable reduction of the acquired detector signal. Next, it is also desired to reduce scattered X-ray radiation to a minimum, as its intensity overlays to the primary intensity and therefore induces image artifacts due to a higher intensity measured. It is therefore desirable to develop methods to determine the amount of scattered radiation so as to correct the measured radiation for the scattered radiation signal. Typically the scattered radiation cannot easily be accessed as it is a priori not distinguishable from the primary radiation. Further, it is also difficult to determine it from the whole context of an acquired image as scattered radiation is related to the scanned patient geometry in a complex manner. Special aspects of CT are spectral methods commonly termed as dual energy CT, multi-energy CT, or spectral CT. The common characteristic of all these methods is that they take use of the fact that different materials attenuate X-rays differently with respect to the energy of the X-ray photons. Consequently, the acquisition of CT projections with different weightings put on the X-ray photon energies provides additional (3D) information, not only of the material density, but also of the chemical composition. In other words, if it is possible to scan an object volume with data sets representing different spectral weightings, it becomes possible to apply mathematical methods to generate a 3D data volume representing different physical or chemical properties. Commonly known examples for such properties are the ratio of bone mineral density to soft tissue density, or the visualization of the presence of contrast agent content like that of iodine, barium, gadolinium, gold or other chemical elements. Other examples are the generation of separate 3D volumes of material densities containing water-like tissue, bone mineral, and/or K-edge contrast material. All these methods work better the stronger the spectral separation of the acquired projections is. Dependent on the chosen methods, the number of separated physical/chemical properties or the number of distinguishable materials also depend on the number of different spectra used for the projection generation. Aspects of the present application address the above-referenced matters and others. In accordance with aspects of the present disclosure, an X-ray imaging system is presented. The X-ray imaging system includes an X-ray device having a single X-ray source for forming a plurality of X-ray beams, a filter positioned within the plurality of X-ray beams, an object space where the object to be imaged is accommodated, and an X-ray detector including an array of a plurality of pixels. The X-ray device, the filter, and the plurality of pixels are configured such that at least one pixel is exposed to the plurality of X-ray beams. X-ray radiation received by a particular pixel undergoes a same spectral filtration by the filter. Pixels receiving the X-ray radiation undergoing the same spectral filtration are summarized to a pixel subset. At least two subsets of pixels exist. According to an aspect of the present disclosure, the X-ray device includes a collimator positioned between the X-ray device and the filter, the collimator having a plurality of openings for directing the plurality of X-ray beams generated by the X-ray source. According to a further aspect of the present disclosure, the X-ray source includes an X-ray emission area with a spatially modulated X-ray intensity profile such that the plurality of X-ray beams originate from one or more pronounced intensity maxima of the X-ray emission area. According to another aspect of the present disclosure, the plurality of pixels have X-ray insensitive regions therebetween. The X-ray imaging system and the collimator are configured to reduce X-ray intensity in the X-ray insensitive regions between the plurality of pixels. According to yet another aspect of the disclosure, the filter includes at least two different materials. In one exemplary embodiment, one of the filter materials is air. According to yet another aspect of the disclosure, the filter includes one material having a spatial modulation. In one exemplary embodiment, the filter is a combination of at least two spatially separated filters. According to yet another aspect of the disclosure, the filter has a spatially alternating pattern of spectral filtration. The filter is a grating having grating lines or a pattern of tiles representing different spectral filtration. In one exemplary embodiment, the filter is replaceable and can be chosen from a set of a plurality of different filters. According to yet another aspect of the disclosure, the subsets of pixels of the X-ray detector form an interlacing and alternating pattern of rows, columns, or tiles. A smallest effective size of a row, a column, or a tile of the alternating pattern of a subset of pixels of the X-ray detector corresponds to the effective size of one pixel. According to yet another aspect of the disclosure, the filter is configured such that at least one pixel subset represents an opaque filtration of X-rays such that at least one pixel subset of the plurality of pixels of the X-ray detector is shadowed from any direct X-ray radiation from the X-ray source of the X-ray device. According to yet a further aspect of the disclosure, a method for measuring an intensity of scattered X-ray radiation for at least one pixel subset of an X-ray imaging system as described above is presented, the method including generating a plurality of X-ray beams via the X-ray device, transmitting the plurality of X-ray beams through a combination of one or more filters and collimators, as well as an object included in the X-ray imaging system, and detecting the scattered X-ray intensity for at least one pixel subset representing an opaque filtration of direct X-ray radiation from the X-ray device. According to yet a further aspect of the disclosure, a method for generating at least one X-ray projection data set including at least two subsets of spectrally different X-ray projections with an X-ray imaging system is presented, the method including generating a plurality of X-ray beams via the X-ray device, transmitting the plurality of X-ray beams through a combination of one or more filters and collimators, as well as an object included in the X-ray imaging system, detecting the X-ray beams via the X-ray detector of the X-ray imaging system, and logically assigning the acquired data of the pixel subsets of the plurality of pixels of the X-ray detector to subsets of spectrally different X-ray projections. Further scope of applicability of the present disclosure will become apparent from the detailed description given hereinafter. However, it should be understood that the detailed description and specific examples, while indicating preferred embodiments of the present disclosure, are given by way of illustration only, since various changes and modifications within the spirit and scope of the present disclosure will become apparent to those skilled in the art from this detailed description. Although the present disclosure will be described in terms of a specific embodiment, it will be readily apparent to those skilled in this art that various modifications, rearrangements and substitutions may be made without departing from the spirit of the present disclosure. The scope of the present disclosure is defined by the claims appended hereto. Collecting many projections of an object and filtration of the x-ray beams are factors used in CT image formation. The present disclosure relates to an x-ray device, particularly in the form of a Computed Tomography (CT) scanner, which includes at least a radiation source, a beam filter, and a radiation-sensitive detector array, described below. Special usage of spectral CT imaging may be seen in a device which provides different X-ray spectra for particular pixels. One might think that simply putting a spatially modulated filter into the X-ray beam might already provide such functionality by directly projecting the filter structures onto the image detector. Indeed, for a CT system with an ideal point-like X-ray source, such a filter may provide a desired spatially modulated X-ray spectrum. However, one skilled in the art understands that this simple method is very often not suitable for the reason that a finite extension of the X-ray spot is present. In fact the spatial extension of the X-ray emitting area blurs the projected structures, i.e., projected filter structures are convolved by a penumbra, which depends on the geometric dimensions of the system. In most cases, the filter needs to be placed close to the X-ray source such that a desired spectrum for a particular direction cannot be limited to the size of a pixel within the detector without a massive overlap with neighboring pixels. To overcome the penumbra problem, an approach is proposed which uses an array of almost point-like X-ray sources, in the sense that the spatial extension of these “point emitters” is so small that the penumbra effects in the detected image are limited to a broadening in the extension of a pixel size. In combination with a periodic array of filter materials, each point emitter projects the filter array onto the detector, such that the overlay of each particular image (using the periodicity of the arrays) produces a congruently superposed image of the filtered array. Thus, a system and method is suggested that includes introducing a filter which spatially and spectrally modulates the X-ray beam. This filter may be, for example, constructed from two different materials within an X-ray system in order to produce a spectrally modulated beam such that, for example, neighboring pixels of the X-ray detector receive different spectra. Reference will now be made in detail to embodiments of the present disclosure. While certain embodiments of the present disclosure will be described, it will be understood that it is not intended to limit the embodiments of the present disclosure to those described embodiments. To the contrary, reference to embodiments of the present disclosure is intended to cover alternatives, modifications, and equivalents as may be included within the spirit and scope of the embodiments of the present disclosure as defined by the appended claims. Referring to FIG. 1, an imaging geometry of an imaging system having at least one X-ray source, according to the present disclosure is presented. The X-ray imaging system 100 includes at least one X-ray device 110 emitting X-rays from a number of locations 106, a filter grating 120, an optional collimator grating 130, an object space 140, and an X-ray detector 150 including an array of a plurality of pixels 151 to 155, which may be separated by X-ray insensitive gaps 170. The X-ray device 110 generates a plurality of X-ray beams 104, each beam 104 characterized by connecting one of the locations 106 with one of the pixels 151 to 155, respectively. The X-ray beams 104 pass through a filter 120. The filter 120 may be referred to as a “filter grating” configured to apply a specific filtration to each of the X-ray beams 104. Locations 106, filter grating 120, and detector pixels 151 to 155 are configured such that all X-ray beams 104 connected to a particular single pixel undergo the same spectral filtration by filter grating 120. The spectral and also spatial separation of X-ray beams 104 may be supported by an optional collimator grating 130 which includes a plurality of openings 108. The openings 108 are configured such that they allow for the passage of photons propagating along the center of each of the X-ray beams 104, or stated differently, each photon propagating from the center of a location 106 to the center of any pixel 151 to 155 passes an opening 108 of the optional collimator grating 130. Furthermore, blockings 131 of the optional collimator 130 are configured to suppress to a maximum amount of X-ray photons, which propagate from one of the locations 106 towards the gaps 170 in between the pixels 151 to 155. Not illuminating the pixel gaps 170 means that an object or patient placed in the object space 140 receives less dose compared to a configuration in which the optional collimator grating 130 is not present. For ideal opaque gratings, even a total shadowing of the pixel gaps 170 can be achieved. It can be shown that the dose saving is about 20% for current CT geometries. It is contemplated that the collimator grating 130 is constructed or formed from highly opaque materials, such as, for example, Tungsten or Lead of appropriate thickness, such that the amount of transmitted radiation through the blockings 131 is reduced to a minimum. An object to be imaged is positioned in the object space 140 located (from the viewpoint of the X-ray source) behind the filter grating 120 and the optional collimator 130, but before the X-ray detector 150. In this exemplary embodiment, five pixels 151 to 155 are shown. However, one skilled in the art may envision several more pixels forming the array of the X-ray detector 150. FIG. 2a and FIG. 2b show two possible configurations of the X-ray device 110. Both configurations provide locations 106 from which X-rays emerge. In FIG. 2a, the X-ray device 110 contains a single area 112 from which X-ray photons emerge. This area may be the common focal spot of a common X-ray tube. The emitted X-rays are collimated by a grating 114, which is configured to transmit X-rays only through its openings 106, which are therefore identical to the locations 106 of the X-ray device 110. It is contemplated that the grating 114 is constructed or formed from highly opaque materials, such as, for example, Tungsten or Lead of appropriate thickness, such that the amount of transmitted radiation through the blockings 231 is reduced to a minimum. It is noted that the X-ray beams 104 received by a particular pixel of detector 150 may pass only a particular number of locations 106, i.e., those locations 106 which are in the line-of-sight between a corresponding pixel of detector 150 and the X-ray emission area 112. Also it may be that X-ray beams 104 assigned to different pixels of detector 150 are assigned to completely different locations 106 of the X-ray device 110. In an alternative embodiment shown in FIG. 2b, the X-ray device 110 includes a single area 116 from which X-ray photons emerge in a spatially modulated intensity such that the locations 106 are represented by local intensity maxima of the X-ray area 116. The area 116 may be the focal spot of an X-ray tube for which the modulation of X-ray intensity is performed by a correspondingly varying density of electrons hitting the metal anode. For example, the electrons may be generated from a spatially modulated electron source, and a common electron lens optics produces an “image” of the electron source, such that the focal spot displays the same spatial X-ray intensity pattern as the electron source area. The filter 120 forms a characteristic pattern of different X-ray filtration. For example, referring to FIGS. 3a and 3b, this pattern may be in the form of alternating stripes that extend a horizontal (or vertical) length of the filter 120. The filter 120 is shown in a side view in FIG. 3a, where a first grating 210 and a second grating 220 are shown both differing from each other by different X-ray filtration properties. The filter 120 is shown in a top view in FIG. 3b, where the first grating 210 and the second grating 220 are shown extending a length of the filter 120. This alternating pattern design of the filter 120 allows for alternating pixels of the plurality of pixels 151 to 155 to be illuminated with different X-ray spectra. For example, as shown in all of FIG. 1, FIG. 4, and FIG. 5, the first pixel 151 receives a first spectra 180 and the second pixel 152 receives a second spectra 182. Additionally, the third pixel 153 and the fifth pixel 155 receive the first spectra 180, whereas the fourth pixel 154 receives the second spectra 182. Stated differently, the odd pixels (i.e., pixels 151, 153, and 155) receive the first spectra 180, whereas the even pixels (i.e., pixels 152 and 154) receive the second spectra 182. Thus, each adjacent or neighboring pixel may receive different spectra (i.e., creation of an alternating configuration of spectra). Stated differently, spectrum separation may be achieved. Moreover, the first spectra 180 may have a strong weight on high energy photons, whereas the second spectra 182 may have a strong weight on low energy photons, and vice versa. As a result, by incorporating filter 120 into the imaging system 100, alternating pixels of the array of the X-ray detector 150 may receive different spectra. Thus, together with the arrangement of the X-ray emitting locations 106 of the X-ray device 110, the filter 120 causes the generation of at least two spectra, such that no spatial overlap between the spectrum occurs, or the spatial spectrum overlap is reduced to a minimum for particular locations on the X-ray detector 150. It is noted that a pattern used for a filter grating 120 needs to be aligned to the pattern of the X-ray emission locations 106 of the X-ray device 110, to the pattern of the optional collimator grating 130, and the geometry of the pixel array of the X-ray detector 150. Thus, subsets of pixels are assigned for the X-ray detector 150, which are aligned to the pattern of different X-ray spectra. For example, the grating-line pattern of the filter grating 120 reported in FIGS. 3a and 3b would be used together with a similar grating-line pattern of the X-ray emission locations 106 of the X-ray device, optionally a grating-line pattern of the collimator grating 130, and a rectangular pattern of detector pixels of the X-ray detector 150 where subsets of pixels form an interlacing grating-line pattern. One skilled in the art will recognize that there are many more possibilities of pattern arrangement than those displayed in FIGS. 3a and 3b. An alternative arrangement is, for example, shown in FIG. 3c where the filter grating 120 is formed by a two-dimensional pattern of rectangular tiles 230, 240. This tile pattern is used together with an X-ray device 110 having X-ray emission locations 106 arranged in rectangular array of almost point emitters, optionally with a collimator grating 130 having a rectangular array of openings 108, and an X-ray detector 150 with a rectangular pixel matrix with pixel subsets forming interlacing tile patterns. Referring back to FIG. 1, a detector pixel array is shown configured to detect the alternating pattern of spectral filtration, i.e., each pixel detects a spectrum different from its direct neighboring pixel. However one skilled in the art will recognize that other configurations are possible. For example, the X-ray device 110, the filter grating 120, and the optional collimator grating 130 may be configured in a way that the first spectra 180 and second spectra 182 form an interlacing pattern with a larger periodicity such that a column, a row, or a tile of the pattern covers an detector area with size dimensions larger than that of a single detector pixel. This means that the subsets of pixels may contain sequences of directly neighboring pixels. In other words, the size of a column, a row or a tile of the pattern represented by the subsets of pixels may correspond to the size of one pixel, but it is not limited to the size of one pixel and may therefore have a larger size. Further, while only means for the first spectra 180 and the second spectra 182 are shown in FIG. 1, it is understood for one skilled in the art that the X-ray device 110, the filter grating 120, and the optional collimator grating 130 and the detector 150 may be configured in a way that interlacing patterns with more than two spectra can be generated. For example, by choosing a filter grating providing means for three or more different X-ray filtrations and having appropriate grating pitches, it is possible to generate an alternating and interlacing sequence of three or more different spectra on the detector 150. It is noted that a filter 120 fulfilling the configuration (i.e., by choosing an appropriate geometry) may be placed at different positions between X-ray device 110 and detector 150. For example, FIG. 4 shows an alternative positioning of the filter grating 120 (it is noted that the grating pitch needs to be adapted according to the total system geometry). Also, there is no fixed order in which, for example, the filter grating 120 and the optional collimator grating 130 have to be mounted into the system. As an alternative embodiment, the filter grating 120 may be replaced by a combination of two or more filters 122 and 124 as shown, for example, in FIG. 5. One skilled in the art may envision several more arrangements and patterning of an effective grating 120. The filter 120 may include or be formed of two or more different materials. Alternatively, the filter 120 may be formed from a single material providing different filtration by a modulation of the material thickness. One of the filter materials may be air or another weakly attenuating material. Effectively, this weakly attenuating material represents an effective zero attenuation of X-rays such that one of the subsets of pixels is assigned to an effectively unfiltered spectrum. Further, one or more of the materials of the filter grating 120 may be formed of materials with relatively large K-edge energies, such as tantalum, tungsten, or lead which results in spectra filtration with a relatively enhanced transmission of photons with energies below the K-edge. The K-edge may be chosen high enough such that photons with energies below the K-edge may still pass the object placed in the object space 140. Using materials with a relatively low K-edge such as aluminum, copper or tin are also contemplated to create spectra which put more relative weight to “high” energies, since it is assumed that their K-edge energy is that low that photon energies below their K-edge will effectively not be able to transmit the object placed into the object space 140. There are many more conceivable embodiments that enable access to spectral CT capabilities without the need of highly specialized detectors (e.g., photon-counting or multi-layer detectors) or tubes with fast switching capabilities. Additionally, one or more materials (or their thickness) of filter grating 120 may be made opaque for X-rays, such that at least one subset of pixels does not get illuminated by primary radiation. Consequently, radiation received by these pixels consists of scattered radiation only. By interpolation, the intensity received by the fully illuminated pixels may be corrected for by its scatter content. The foregoing examples illustrate various aspects of the present disclosure and practice of the methods of the present disclosure. The examples are not intended to provide an exhaustive description of the many different embodiments of the present disclosure. Thus, although the foregoing present disclosure has been described in some detail by way of illustration and example for purposes of clarity and understanding, those of ordinary skill in the art will realize readily that many changes and modifications may be made thereto without departing form the spirit or scope of the present disclosure. Therefore, the above description should not be construed as limiting, but merely as exemplifications of particular embodiments. Those skilled in the art will envision other modifications within the scope and spirit of the claims appended hereto.
	description	This application claims the benefit of priority to U.S. Provisional Application No. 61/492,697, filed on Jun. 2, 2011, which is incorporated herein by reference in its entirety. The present disclosure generally relates to methods of consolidating radioactive materials by hot isostatic pressing (HIP). In particular, the present disclosure relates to a ubiquitous consolidation technique for radioactive material that is based on a novel, bottom-loading HIP method. Beginning in 1953, Spent Nuclear Fuel (SNF) was reprocessed by the Department of Energy to recover highly enriched uranium and other nuclear related products. Processing operations involved multiple cycles of solvent extraction to recover uranium-235 and other defense-related materials from SNF. The end of the Cold War also ended the program to reprocess SNF, with the last reprocessing cycle ending in 1994. These reprocessing activities, as well as other ancillary facility activities and operations, generated millions of gallons of liquid radioactive wastes, which were stored in underground storage tanks. To mitigate the dangers associated with leakage of these storage tanks, a fluidized bed calcination process was put in operation in the early 1960's to convert the liquid tank waste into a small, granular solid calcine generally having consistency similar to laundry detergent. The calcination process produced a safer product for storage while reducing the volume of stored waste by an average factor of seven. Approximately 8 million gallons (30,300 m3) of liquid tank waste were converted to 4,400 m3 of calcine, which is now being stored while awaiting future disposition. The disposition of this stored calcined waste is driven by the waste form itself. The waste form determines how well the waste is locked up (chemical durability), as well as the waste loading efficiency, i.e., a higher efficiency requires fewer containers, which reduces disposal cost. The use of glass-ceramic waste forms for problematic wastes such as the calcines, which are difficult to vitrify, offers significant performance improvements and efficiency savings, principally via higher waste loadings. Integral to the design of the waste form is the selection of the appropriate process technology used to treat the calcine. A key consideration is to select a flexible process that does not constrain the waste form chemistry. Constraints imposed by the consolidation technology on the waste form chemistry will result in a reduction in waste loading efficiency and process flexibility. For instance, Joule-heated melters (JHM) not only have a restricted maximum operating temperature but also require the glass to have specific electrical resistivity and viscosity characteristics. Similar considerations apply to cold-crucible melters. Therefore, the glass cannot be designed solely to suit the waste stream. Additional components need to be added to ensure the glass chemistry is such that it can be melted at the melter operating temperature and poured safely into a canister. These constraints significantly reduce the maximum achievable waste loading efficiency and/or process flexibility and therefore substantially increase the number of waste canisters required. The Inventors have discovered that by using a novel HIP technology, significant performance enhancements can be realized. These relate to higher waste loadings, enhanced process flexibility, reduced off-gas emissions, competitive production rates and reduction in secondary wastes, while readily complying with the required waste form acceptance criteria outlined by the DOE. Thus, there is disclosed a method of consolidating a calcined material comprising radioactive material, the method comprising: mixing a radionuclide containing calcine with at least one additive inside a mixing vessel to form a pre-HIP powder; loading the pre-HIP powder into a can and sealing the can, such as by welding. The Inventors have found that by loading the sealed can through the bottom of a HIP vessel using a fully automated system, the cans can be pre-heated and loaded while hot. This allows for a decrease in process time by as much as ⅓ or even ½. In one embodiment, the cans can be pre-heated and loaded while at temperatures up to 600° C. In accordance with the present teachings and as described in one exemplary embodiment, the sealed can is loaded into the bottom of a HIP vessel, wherein the can undergoes hot-isostatic pressing at a temperature ranging from 1000° C. to 1250° C., such as 1200 to 1250° C., and a pressure ranging from 30 to 100 MPa for a time ranging from 10-14 hours. In accordance with one exemplary embodiment, the HIP can is encapsulated in an additional containment vessel prior to being loaded into the HIP vessel. The additional containment vessel with the can contained therein, is generally positioned on the HIP vessel bottom closure, which is then raised and secured to seal the HIP vessel. In one embodiment, the pre-HIP powder is subjected to at least one pre-heat process prior to being loaded into the HIP vessel, such as heating to a temperature necessary to remove excess moisture present in the pre-HIP powder. In this embodiment, pre-heating comprises heating to a temperature ranging from 100° C. to 400° C. This pre-heat process is typically done while the powder is in the HIP can. In an alternative embodiment, the at least one pre-heat process comprises heating to a temperature sufficient to drive off unwanted constituents present in the pre-HIP powder, but not high enough to volatilize any radionuclides present in the powder. In this embodiment, pre-heating occurs at temperatures ranging from 400° C. to 900° C. In one embodiment, the pre-HIP powder is heated prior to loading in the HIP can. A subsequent evacuation step will also be performed on the filled can prior to sealing it. One benefit of the inventive method is the high waste loading capacity. For example, the pre-HIP powder may comprise 60-80% radionuclide containing calcine. In one embodiment, the ratio of radionuclide containing calcine to additives is about 80:20 by weight, wherein the non-radioactive additives such as such as BaO, CaO, Al2O3, TiO2, SiO2 combine with the waste elements and compounds to form a ceramic mineral or glass/ceramic material. Non-limiting examples of the resulting mineral phases that may form are: hollandite (BaAl2Ti6O16), zirconolite (CaZrThO7), and perovskite (CaTiO3). The non-radioactive additives are selected based on the type(s) of radionuclide presented in the calcine, such as if the calcine contains spent nuclear fuel, sodium containing waste, or heavy metals. Aside from the subject matter discussed above, the present disclosure includes a number of other exemplary features such as those explained hereinafter. It is to be understood that both the foregoing description and the following description are exemplary only. As used herein “calcine” is the solidified liquid waste stream remaining after first-cycle solvent extraction of uranium from SNF and the concentrated wastes from second- and third-cycle extraction of the fuel. “Spent Nuclear Fuel” (SNF), may also be referred to as “Used Nuclear Fuel,” and is nuclear fuel that has been irradiated in a nuclear reactor (usually at a nuclear power plant) but is no longer useful in sustaining a nuclear reaction in an ordinary thermal reactor. “Raffinate” is a product which has had a component or components removed. The product containing the removed materials is referred to as the extract. For example, in solvent extraction, the raffinate is the liquid stream which remains after solutes from the original liquid are removed through contact with an immiscible liquid. In metallurgy, raffinating refers to a process in which impurities are removed from liquid material. “Fully Automated System” refers to the ability to load and unload HIP cans from the HIP system using machines and control systems, including robotics, without any direct human contact. “RCRA” refers to the “Resource Conservation and Recovery Act” of 1976 (42 USC 6901), which is the principal Federal law in the United States governing the disposal of solid waste and hazardous waste. “ACOP”—(Active Containment OverPack) refers to an additional container, such as a secondary can, encapsulating the first HIP can. “CRR”—(Carbon Reduction Reformer) refers to a fluidized bed steam reformer. “CCIM”—(Cold Crucible Induction Metter) refers to a water cooled crucible that heats the feed or melt by an inductive coil that surrounds the crucible. “DMR”—(Denitration/Mineralization Reformer) refers to fluidized bed steam reformer where additives are introduced to form a mineralized product. “HEPA” refers to a high-efficiency particulate air. “HLW”—(high-level waste) are the highly radioactive materials produced as a byproduct of the reactions that occur inside nuclear reactors. High-level wastes take one of two forms: (1) spent (or Used) reactor fuel when it is accepted for disposal, and (2) waste materials remaining after spent fuel is reprocessed “JHM”—(Joule Heat Metter) refers to a melter that relies on Joule heating, also known as ohmic heating and resistive heating, which relies on the passage of an electric current through a conductor to release heat. “LLW” (low-level waste) refers to items that have become contaminated with radioactive material or have become radioactive through exposure to neutron radiation. This waste typically consists of contaminated protective shoe covers and clothing, wiping rags, mops, filters, reactor water treatment residues, equipment and tools, luminous dials, medical tubes, swabs, injection needles, syringes, and laboratory animal carcasses and tissues. “MGR” refers to monitored geological repository. “MTHM” refers to metric tons of heavy metal. “SBW” (Sodium-Bearing Waste)—derives its name from the relatively high concentration of sodium ions (1-2 molars) present in the waste. The sodium came from processes and activities that made use of sodium containing chemicals, such as sodium hydroxide, sodium permanganate, and sodium carbonate. SBWs typically have much lower levels of fission product activity than first cycle raffinates. The present disclosure relates to a method for disposing the large amount of stored radioactive calcine that relies on a novel consolidation method. In particular, the Inventors have shown that their unique Hot-isostatic Pressing (HIP) process can deliver at least the following life-cycle savings to either direct disposal or treatment of the calcine: Higher treatment waste loadings (fewer disposal canisters); Maximum volume reduction (repository cost savings for treatment or direct disposal); Enhanced treatment chemical durability (lower environmental risk); Greater processing flexibility (one additive composition for all calcines); Low off-gas emissions; High degree of contamination control; No liquid waste generation; and Reuse of existing facilities with minimum modification. As mentioned, the principle source of the liquid waste that was calcined was raffinate (waste solution) from spent nuclear fuel dissolution and subsequent uranium extraction. Other waste sources included equipment decontamination, uranium purification (second- and third-cycle raffinates), and support operations including ion exchange water treatment systems and off-gas treatment systems, and laboratory analyses of radioactive materials. A variety of spent nuclear fuels from numerous reactors were reprocessed by the DOE during the nearly 40 years it ran the SNF Reprocessing program. Differences in the fuel configuration, especially the fuel-cladding material, dictated the use of different chemicals to reprocess the various types of fuel. These chemically differing processes generated chemically different liquid wastes and, consequently, chemically different calcine. Two general categories of liquid wastes were generated and calcined: (1) first-cycle raffinate and (2) sodium bearing waste (SBW). First-cycle raffinate, produced by dissolution of spent nuclear fuel and then extraction of uranium, contained dissolved fuel cladding and the bulk of the fission products originally in the spent fuel. Sodium-bearing waste (SBW) is the product of site operations such as decontamination activities, some of which use dilute sodium hydroxide to wash surfaces and solubilize residues. As a result, significant sodium nitrate salts are present in the SBW solutions. The relatively high sodium content makes these solutions unsuitable for direct calcination in their present form because sodium nitrate melts at low temperatures and will not produce a granular, free-flowing calcination product. The SBW can be calcined upon addition of aluminum nitrate. This calcine product can be treated according to the present invention. However, because SBW does not come from spent nuclear fuel (SNF) reprocessing, it is not initially a high level waste (HLW) but rather a mixed low-level transuranic (TRU) waste that becomes HLW by this mixing with a HLW stream. For example, in one embodiment, the radioactive content of SBW is about 0.2 curies (Ci) per liter each of 90Sr and 137Cs, and the actinide activity is approximately 500 microcuries per liter (μCi/l), composed of approximately 350 μCi/l from 238pu and 125 μCi/l from 239pu. Conventional processing of SNF uses a glass melting process in which SNFs are melted and thus sequestered in a traditional borosilicate glass. However, at least one type of calcine treated using the inventive method is heterogeneous and contains significant proportions of components problematic to conventional glass melting processes used to sequester radionuclides. For example, calcines may contain alumina, zirconia, and calcium fluoride, as well as heavy metals. Any or all of these components are problematic and difficult to incorporate in conventional melting routes at cost competitive waste loadings. They are either refractory with low solubility in glass (zirconia and alumina) or can have a dramatic impact on glass viscosity (pourability) and melter corrosion in the case of calcium fluoride. However, the limitations associated with these materials are irrelevant and overcome by utilizing the HIP technology disclosed herein, since HIP is not sensitive to the viscosity of the waste form and not susceptible to melter corrosion. HIP in combination with a hybrid glass/ceramic formulations, require lower operating temperatures to produce dense waste forms. Glass/ceramic waste forms overcome the solubility limitations of glass, by allowing the controlled crystallization of components that enhance rather than detract from the chemical durability of the system. By utilizing glass-ceramic waste forms, waste loading efficiency can be increased by at least three times traditional borosilicate glass processes used to sequester SNF while maintaining chemical durability far superior to the Environmental Assessment (EA) glass standard. This can be achieved while also delivering well in excess of a 35% volume reduction compared to direct calcine disposal, due to the increased density of the consolidated waste form. Thus delivering billion dollar savings in shipping and repository disposal costs, while using currently qualified repository disposal canisters. These benefits are attainable by using the disclosed HIP technology. The glass-ceramic waste form is also exceedingly robust in regard to process chemistry and waste variation. In the inventive process one single, chemically flexible, glass-ceramic formulation can treat the entire suite of calcine with a waste loading of 80 wt %. The breadth of the processing window minimizes risk from compositional uncertainty of the waste feed stream, while still maintaining significantly higher waste loading efficiencies and chemical durability than borosilicate glass. The HIP process produces a glass-ceramic waste form made from several natural minerals that together incorporate into theft crystal structures nearly all of the elements present in high level waste calcine. By combining traditional oxides, such as BaO, CaO, Al2O3, TiO2, SiO2 with the waste elements and compounds, stable, monoliths of a ceramic mineral or glass/ceramic material are formed. Non-limiting examples of the resulting mineral phases that may form include hollandite (BaAl2Ti6O16), zirconolite (CaZrThO7), and perovskite (CaTiO3). Zirconolite and perovskite are the major hosts for long-lived actinides, such as plutonium, though perovskite principally immobilizes strontium and barium. Hollandite principally immobilizes cesium, along with potassium, rubidium, and barium. In more general terms, HIP consists of a pressure vessel surrounding an insulated resistance-heated furnace. Treating radioactive calcine with the HIP involves filling a stainless steel can with the calcine and additives. The can is evacuated and placed into the HIP furnace and the vessel is closed, heated, and pressurized. The pressure is typically provided via argon gas, which, at pressure, also is an efficient conductor of heat. The combination of heat and pressure consolidates and immobilizes the waste into a dense monolith. Calcine treatment begins with the mixing of retrieved calcine with treatment additives in a ratio of ˜80:20 by weight, typically inside a dedicated mixing vessel. The mixture is then loaded into a HIP can, and optionally preheated to a temperature of ˜600° C. and evacuated. In one exemplary embodiment, the mixture of calcine and additive is preheated prior to the can loading, and preferably loaded into the HIP vessel while still hot, even at temperatures up to 600° C., which allows for a significant decrease in process time, such as by a third or even half the normal HIP process time. The loaded can is then sealed, by welding for example, and loaded into the HIP vessel. The HIP will process one can at a time to a temperature, such as a temperature ranging from of about 1000° C. to 1250° C., more particularly about 1200° C. at a processing pressure ranging from 30-100 MPa. The cycle time to process a HIP can ranges from about 10-16 hours, such as about 12 hours. Once removed from the HIP, the can will be allowed to cool to ambient temperature prior to being loaded into a disposal canister. Direct HIP of the calcine could be achieved on the same process line. In this case while no treatment additives would be required, some benefits to the consolidation process may gained by adding a very small amount of processing aids. The HIP temperature may also be modified depending on the waste. Various changes in HIP conditions such as temperatures, pressures, and atmospheres depending on the material being consolidated are discussed in U.S. Pat. Nos. 5,997,273 and 5,139,720, which are herein incorporated by reference. A more specific discussion of the entire process, from calcine loading to finished monolith product is described below, with reference to conventional elements, such as feeders, hoppers, blenders, filters, ventilation systems, and the like, all adapted to work in the inventive process. In one exemplary embodiment of the inventive method, the calcine is transferred through a surge tank discharge rotary valve into the calcine feed blender. The calcine feed blender weigh cells control the total transfer amount. During transfer, the calcine feed blender is vented through sintered metal blowback filters to a central ventilation system. HIP additive is added to the calcine feed blender through a line penetrating the Can filling cell roof from an additive feed hopper and metered using an additive feed screw. The additive feed hopper weigh cells control the amount of material added to the calcine and serve as a weigh cell check for the calcine feed blender. The calcine feed blender is actuated to mix the calcine with the additive. The calcine feed blender rotary discharge valve transfers the mix to the HIP can feed hopper that provides volumetric control to prevent overfilling of the HIP can. The calcine feed blender may be provided with air pads near the discharge point to assure solids movement. The HIP can feed hopper may use de-aeration techniques to assure complete filling of the HIP cans. The HIP can feed hopper is vented through sintered metal blowback filters to the central ventilation system. The HIP can feed hopper is mounted on weigh cells to verify the amount of feed transferred to the HIP cans and to verify the calcine blender weigh cells. The HIP can feed hopper rotary discharge valve transfers feed to the HIP cans through a gravity feed connection to the HIP can. The HIP can feed hopper may be provided with air pads on the discharge area to assure solids movement. The HIP can has two ports: the first is the feed port which is connected to the HIP can feed hopper discharge piping, and the second is a vent line that is connected to the central ventilation system or a subsystem that discharges into the central ventilation system. The HIP can contains a built-in sintered metal filter that prevents solids escaping into the vent line. The HIP can fill port is mated to a special fill nozzle that is designed to minimize solids contamination on the exterior of the HIP can. The HIP can loading station may use de-aeration techniques to assure complete filling of all material from the HIP can feed hopper. A collimated gamma detector system will be used to verify the HIP can is filled to the proper level. Additional measures will be taken to assure minimum contamination of the HIP can fill area and the HIP can outer surfaces when the fill port is detached. The HIP can will be located below a contamination control table (either a flat table or a downdraft table) with just the fill port exposed to the can filling cell above the table surface. A circular slot hood surrounds the HIP can fill port, and a flow will be maintained through the HIP can to keep dust within the HIP can. The HIP can fill nozzle will have special features that prevents dust migration. Various methods like vacuum pickup and wiping may also be employed to clean exposed surfaces of the HIP can prior to moving the HIP can. Once the HIP can is filled and the solids level verified, the HIP can feed nozzle is detached and a fill port plug is inserted into the HIP can to minimize the potential for down stream contamination of equipment. The fill port plug is also necessary for maintaining structural integrity of the HIP can during the HIP process. The plug surface and exposed port surfaces are cleaned in place and a cover put over the fill port area on the table. The HIP can is lowered away from the filling station and the process vent line detached and capped. The HIP can is moved into an air lock where swabs can be made and where decontamination activities can be carried-out. When decontaminated, the HIP can is moved out of the air lock to the welding station where the HIP can fill port cap is inserted into the port, seal welded, and the weld leak checked. The HIP can is moved to the HIP can bake-out station where the can is lowered into a secondary containment vessel—Active Containment OverPack (ACOP)—and a vent line attached to the HIP can bake-out off-gas system (described in more detail below). The HIP can bake-out furnace insulating lid is replaced and the HIP can and ACOP are heated to approximately 700° C. over the course of several hours. Any bake-out off-gas is routed through filters and traps to remove any particulates or gaseous components. For example, mercury is captured using sulfur impregnated carbon bed traps that cannot be vented through the central ventilation system to the environment. When the bake-out is complete, a vacuum is pulled on the HIP can through the vent line. When the vacuum reaches the set-point, the vacuum is verified. Once the vacuum is verified, the vent port is closed and the vacuum line removed. The ACOP is removed from the HIP can bake-out furnace and placed in an insulated container attached to a rail-guided cart. The cart is moved to a weld station where the closed vent port is seal welded and leak checked. An ACOP lid is fastened to the ACOP body. A shielded isolation door is opened between an isolated portion of the Can Filling hot cell and the HIP hot cell. The cart is moved into the HIP hot cell and the shielded isolation door closed. The HIP vessel bottom closure is in the open/lowered position. The ACOP is moved into position on the HIP vessel bottom closure. The HIP vessel bottom closure is raised and secured to seal the HIP vessel. The HIP process is initiated by heating the ACOP with the HIP can inside up to 1000-1250° C. while controlling the pressure within the HIP vessel. Compressors outside the hot cells protected by in-line filtration control the argon atmosphere within the HIP vessel. Precise control of pressure and temperature are done to HIP the can within the ACOP. After the HIP process is complete, the ACOP and HIP can are cooled within the HIP vessel to a temperature sufficient for removal. The HIP vessel bottom closure is lowered and the ACOP is moved to an insulated HIP can cooling cabinet, where the ACOP and HIP can are cooled to ambient temperatures. Once cooled, the ACOP is removed from the HIP can cooling cabinet and placed on a rail-guided cart. A shielded isolation door between the HIP hot cell and the first canister fill/decontamination hot cell is opened. The cart is moved into the hot cell and the door closed. The ACOP is opened and the HIP can is swabbed for contamination. If contamination is found, the ACOP and HIP can are decontaminated. The clean ACOP is returned to the Can filling Hot cell. The shielded bell crane removes the hot cell roof plug and retrieves the HIP can from within the ACOP if not already removed for decontamination. Metal containing calcined products, such as Al and Zr, need to be baked-out prior to HIP'ing to remove unwanted constituents, such as heavy metals, and excess moisture present in the pre-HIP powder. For example, in one embodiment, Al calcine was baked to 900° C. (1,652° F.) prior to the HIP tests. The purpose of the bakeout was to drive off the Hg and complete calcination before HIP treating. For the average RCRA metal tests, the Hg was reduced from 1.04 to 0.54 wt %. For the maximum RCRA metal tests, the Hg was reduced from 2.25 to 0.36 wt %. Actual Al calcine was produced at 400° C. (752° F.). The preferred design approach is to bakeout the calcine to 100° C. (212° F.) to remove excess water. Baking-out to 900° C. (1652° F.) will volatilize Cs-137 into the off-gas system. Similarly, for Zr-containing calcine bakeout can be accomplished at 100 to 200° C. (212 to 392° F.) to remove excess moisture. However, no bakeout will be performed on surrogate Zr calcine at elevated temperatures, since this could volatilize Cd and Cs-137 located in the calcine. Unless otherwise indicated, all numbers expressing quantities of ingredients, reaction conditions, and so forth used in the specification and claims are to be understood as being modified in all instances by the term “about.” Accordingly, unless indicated to the contrary, the numerical parameters set forth in the following specification and attached claims are approximations that may vary depending upon the desired properties sought to be obtained by the present teachings. Other embodiments of the present teachings will be apparent to those skilled in the art from consideration of the specification and practice of the subject matter disclosed herein. It is intended that the specification and examples be considered as exemplary only, with a true scope and spirit of the disclosure being indicated by the following claims.
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047524324	summary	DESCRIPTION 1. Technical Field This invention relates to a device and process for the direct production of nitrogen-13 ammonium ion in an aqueous or other fluid solution from a carbon-13 fluid slurry target. 2. Background Art Nitrogen-13 is commonly used in scanning operations where it is introduced into the body and monitored by state-of-the-art techiques. It is desirable to produce nitrogen-13 by a relatively simple process. Known prior art methods teach the use of natural water in a batch or recirculating mode to produce predominantly nitrogen-13 oxides. These oxides must be chemically reduced in a basic solution to ammonia which is then distilled and collected. Prior devices and methods employing this approach produce added complexity, chemical losses, and processing time with concomitant, crucial radioactive decay loss. In addition, the p,.alpha. nuclear reaction on natural water has a much lower probability of occurrence for low energy protons than the p,n reaction on carbon-13 in the target original employed in the present invention. Accordingly, it is an object of the present invention to provide a slurry target capable of generating a high yield of nitrogen-13, and the direct production of the desired chemical form in a simple continuous flow collection which precludes complex chemical processing and radioactive decay losses. It is another object of the invention, with the utilization of 10.2 MeV protons entering the target at a beam current of 20 .mu.A, to produce about 175 mCi of nitrogen-13 ammonium ion in a time period of 10 minutes. In the prior art, a typical larger cyclotron (16 MeV) produces nitrogen-13 using 20 .mu.A of protons on natural water; and after chemical reduction, about 175 mCi of ammonium ion is available in a time period of about 25 minutes after the initiation of bombardment. Therefore, the slurry target of the present invention produces in one embodiment about the same activity in about half the time using two-thirds of the proton energy. It is a further advantage of the present invention that the enriched carbon-13 inventory employed as a constitutent part of the fluid slurry target is not expended since it remains fixed in the target for subsequent production runs.
051046121	summary	FIELD OF THE INVENTION This invention relates to an improvement in the mechanical means for refueling power generating nuclear reactor plants such as the conventional water cooled and moderated boiling water and pressure water systems. The invention comprises a unique apparatus for handling nuclear fuel bundle units underwater within the nuclear reactor vessel, including a remotely operable grapple provided with a viewing means for grasping and transfer fuel bundles while submerged underwater. BACKGROUND OF THE INVENTIONS Typical water cooled and moderated nuclear reactor plants for power generation comprise a large pressure vessel containing cooling and neutron moderating water, and have a heat generating core of fissionable fuel submerged a substantial distance beneath the surface of the cooling and moderating water. The submerged fissionable fuel of the core must be periodically replaced, including the removal of spent fuel and replacement with new fuel, as well as rearranging partially spent fuel within the core. Due to the high levels of radioactivity within the nuclear reactor pressure vessel, the means for handling the water submerged fuel must be remotely controlled by an operator from out beyond the water containing reactor pressure vessel. Conventional fuel handling systems comprise a fuel handling mast or pole extending down from above an open top of the water containing reactor pressure vessel with a grapple head affixed to the lower end of the mast. The system is designed for attachment to fuel bundles and their transfer while submerged in the reactor vessel to remove spent fuel and introduce new fuel, and rearrange fuel bundles within the core. The fuel handling mast is frequently supported on and operated from a movable platform which can travel back and forth over an open top of the water containing reactor vessel above the fuel core. Typically the fuel handling mast is mechanically telescoping downward from the supporting movable platform to facilitate reciprocal travel of the grapple head affixed to the lower end of the mast down into and back up from the interior of the reactor vessel. This arrangement provides greater versatility for transferring fuel bundles within and about the reactor vessel. To facilitate operating personnel in manipulating such fuel handling systems with the grappling devices submerged a substantial depth below the surface of the water containing reactor vessel from a safe position above the open top of the reactor vessel, underwater viewing means are commonly employed. For example, underwater periscopes or television cameras suspended on a pole and connected to an above surface monitoring screen have been utilized for enabling remotely located operators to more accurately and clearly observe their underwater manipulation and relative location of the grappling head on the mast with respect to fuel assemblies to be transferred, and its application to fuel bundles. However, controlling the manipulation of two distinct underwater units and their coordination by a remotely located operator is cumbersome and slow, and space limitations sometimes impede positioning of such underwater viewing mechanisms in conjunction with the fuel handling means. Fuel bundles for typical water cooled and moderated nuclear reactor plants used to generate power commonly consist of a multiplicity of small diameter sealed tubes elements enclosing fissionable fuel which are grouped, spaced apart, into an assembled unit. Each assembled unit of the grouped tube elements is provided with an upper and lower end piece having sockets to receive and secure the end portions of the grouped tube elements, and the overall assembled unit is substantially surrounded with an open ended housing or channel unit. A handle or bail is provided on the upper end piece of the assembled units for convenient and effective graphing and secure attachment of a transferring means such as a grapple device. The identification number of each assembled unit or fuel bundle is stamped on the top of its bail. The grouping of a multiplicity of the fuel containing tube elements in assembled units greatly facilitates the transfer of fuel in reloading operations, among other benefits. SUMMARY OF THE INVENTION This invention comprises an improved means for handling fuel bundles within the water containing vessel of a nuclear reactor plant. The invention comprises a composite arrangement of a mast with a grapple head mounted on the lower end thereof and having a viewing camera associated with the grapple head for observing submerged fuel bundles and transferring the bundles underwater. OBJECTS OF THE INVENTION It is a primary object of this invention to provide an improved fuel handling system for nuclear reactor plants. It is another object of this invention to provide a unique fuel handling system for transferring fuel bundles underwater within the enclosing pressure vessel of a water cooled and moderated nuclear reactor plant. It is a further object of this invention to provide a fuel handling system for water cooled and moderated nuclear reactor plants employing improved means for viewing underwater when carrying out the transfer of fuel bundles submerged within the water containing pressure vessel of the nuclear reactor. It is a still further object of this invention to provide a composite system for handling fuel bundles underwater within a nuclear reactor including an underwater grappling means combined with a viewing camera and improved viewing and/or illuminating measures. It is also an object of this invention to provide an underwater fuel handling system for nuclear reactors having a camera associated with grappling means for close, distortion free viewing of all underwater work performed with the handling system.
043839690	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a method for removing the small amounts of .sup.14 CO.sub.2, .sup.14 CO and corresponding alkanes produced in nuclear power stations from the exhaust gases of the purification plants. 2. Description of the Prior Art In most nuclear power stations, water is used as the coolant. It is unavoidable in pressurized-water reactors and in boiling-water reactors that radioactive impurities, which may also be of a gaseous nature, get into the cooling water loop or are formed there. It is therefore customary to always branch off part of the circulating water from the main coolant loop and to conduct it through a purification plant, to remove the radioactive impurities there, to degas the water and then return it to the main coolant loop. This known technique is schematically shown for a pressurized-water reactor in FIG. 1 and for a boiling-water reactor in FIG. 2. By the extremely high radiation density in the reactor core, a very small amount of water is furthermore dissociated radiolytically into hydrogen and oxygen. In the degassification station of the purification plant, these gases are likewise liberated and changed catalytically into water again in a recombination arrangement. In this manner, the development of an ignitable hydrogen-oxygen mixture is prevented from the start. The remaining exhaust gases are customarily transported over a bed of activated carbon, where they are adsorbed, lose most of their radioactivity during the storage time and are discharged after delay into the outside air via the exhaust air stack. The traces of radioactive carbon .sup.14 C which is contained in the exhaust gases and has a half-life of more than 5000 years, are discharged to the outside via the stack practically unchanged. The formation of this radioactive carbon isotope is derived from the (n,.alpha.) reaction with the oxygen isotope of the water, .sup.17 0, and also from the (n,p) reaction with possible nitrogen contaminations. This radioactive carbon is present substantially as monoxide, dioxide and as alkane. Although only small amounts of this radioactive carbon are formed, it might become necessary with the expected increased energy production via nuclear power plants, because of the biochemical importance of this carbon isotope, to no longer discharge the latter into the free atmosphere but to collect it and to add it to the radioactive wastes. SUMMARY OF THE INVENTION An object of the present invention is to provide a method for separating these small shares of radioactive carbon from the exhaust gases of a nuclear reactor plant. It should be possible to retrofit the apparatus required for implementing it easily later in already existing purification and exhaust gas systems. With the foregoing and other objects in view, there is provided in accordance with the invention a method for removing .sup.14 CO.sub.2, .sup.14 CO and alkanes having radioactive .sup.14 C produced in nuclear power plants from exhaust gases of the nuclear power purification plants containing small amounts of radioactive carbon compounds, which comprises treating said exhaust gases to oxidize the radioactive carbon compounds contained in the exhaust gas to .sup.14 CO.sub.2 and subsequently passing the oxidized exhaust gas containing .sup.14 CO.sub.2 in contact with an absorption medium to effect removal of the .sup.14 CO.sub.2 from the exhaust gas. 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 method for removing radioactive carbon produced in nuclear power plants, it is nevertheless not intended to be limited to the details shown, since various modifications may be made therein without departing from the spirit of the invention and within the scope and range of equivalents of the claims.
039322149	description
	abstract	The present disclosure relates to a device for closing and opening a beam path of electromagnetic and/or ionizing radiation, comprising at least one part of a shutter body which is permanently situated in the beam path and rotatable about a longitudinal axis situated essentially transversely with respect to the beam path, and which contains a material that is opaque to the radiation and blocks the beam path when the shutter body is in a closed rotary position, and which defines a passage that is transparent to the radiation when in an open rotary position; and comprising a magnetic drive which is coupled to the shutter body for rotation of same about the longitudinal axis between the rotary positions. The magnetic drive is an electromagnetic drive, and is configured for moving the shutter body between the rotary positions, wherein at least one of the rotary positions corresponds to a stable position of the magnetic drive which maintains the magnetic drive without current.
	abstract	A method (and structure) for controlling a beam used to generate a pattern on a target surface includes generating a beam of charged particles and directing the beam to a mask surface and causing the beam to be either absorbed by or reflected from the mask surface, thereby either precluding or allowing the beam to strike the target surface, based on a reflection characteristic of the mask surface.
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	description	The invention relates to a lead-free radiation protection material in the energy range of an X-ray tube having a voltage of from 60 to 125 kV. Conventional radiation protection clothing in X-ray diagnostics mostly contains lead or lead oxide as the protective material. Because of its toxicity, lead and the processing thereof result in considerable damage to the environment. Because of lead's very great weight, protective clothing of lead is unusually heavy, which means a considerable physical strain on the user. When wearing protective clothing, for example during medical operations, the weight is of great importance in terms of wear comfort and the physical strain on the medical staff. Lead substitute materials for use in radiation protection are already known. DE 199 55 192 A1 describes a process for the production of a radiation protection material from a polymer as matrix material and the powder of a metal having a high atomic number. DE 201 00 267 U1 describes a highly resilient, lightweight, flexible, rubber-like radiation protection material, wherein chemical elements and oxides thereof having an atomic number greater than or equal to 50 are added to a specific polymer. In order to reduce the weight as compared with conventional lead aprons, EP 0 371 699 A1 proposes a material that likewise contains, in addition to a polymer as the matrix, elements having a higher atomic number. A large number of metals is mentioned therein. DE 102 34 159 A1 describes a lead substitute material for radiation protection purposes in the energy range of an X-ray tube having a voltage of from 60 to 125 kV. Depending on the elements used, the degree of attenuation or the lead equivalent (International Standard IEC 61331-1, Protective devices against diagnostic medical X-radiation) of the material in question in some cases exhibits a pronounced dependency on the radiation energy, which is a function of the voltage of the X-ray tube. Compared with lead, the absorption behaviour of lead-free materials in some cases differs considerably depending on the X-ray energy. For this reason, an advantageous combination of different elements is required in order to imitate the absorption behaviour of lead while at the same time maximising the saving in terms of weight. For this reason, the field of application of commercial lead-free radiation protection clothing is generally limited. In order to be able to replace lead for radiation protection purposes, an absorption behaviour is required, in relation to lead, that is as uniform as possible over a relatively large energy range, because radiation protection materials are conventionally classified according to the lead equivalent, and radiation protection calculations are frequently based on lead equivalents. In the case of a lead substitute material composed of protective layers, the overall lead equivalent is understood as being the lead equivalent of the sum of all the protective layers. The overall nominal lead equivalent is understood as being the lead equivalent to be indicated by the manufacturer according to DIN EN 61331-3 for personal protective equipment. During measurements of the lead equivalents and the attenuation factors in dependence on the tube voltage it has been found that the protective effect of lead-free materials in particular at an X-ray tube voltage of from 60 to 80 kV is considerably lower compared with lead than in the energy range of from 80 to 100 kV. There are substantially two reasons for this. On the one hand, the mass attenuation coefficient of lead-free materials such as tin, at the middle energy of the 60 kV spectrum, i.e. at about 25 keV, is lower than that of lead. On the other hand, there is a particularly great dose build-up effect in this low energy range. In other words, the protective effect of the material is reduced by the formation of secondary radiation on the radiation outlet side. In order to achieve a high protective effect, the dose build-up in the lead-free material should remain as low as possible. As already mentioned, a secondary radiation is excited in the material, which in large radiation fields acts to diminish the shielding effect of the material. In most cases, the excited fluorescent radiation is responsible for the dose build-up. The dose build-up is expressed numerically by the so-called build-up factor according to IEC 61331-1. The object of the present invention is to provide a lead-free radiation protection material that exhibits low or only negligible amounts of secondary radiation over the energy range of an X-ray tube having a voltage of from 60 to 125 kV and that accordingly ensures an optimum shielding effect. The object of the present invention is achieved by a lead-free radiation protection material according to patent claim 1. The present invention relates to a lead-free radiation protection material in the energy range of an X-ray tube having a voltage of from 60 to 125 kV, having a layer structure of at least two layers with different shielding properties. The invention relates further to radiation protection clothing made from the lead-free radiation protection material according to the invention. It is important according to the invention that the lead-free radiation protection material has at least two layers with different shielding properties. In this two-layer structure, the composition of the protective materials in one layer is such that one layer alone does not achieve the desired properties in respect of the shielding effect, in particular over a larger energy range of from 60 to 125 kV. Only the two layers together give optimum shielding properties. The layer structure, comprising at least two layers with different shielding properties, of the lead-free radiation protection material according to the invention is preferably composed of a secondary radiation layer and a barrier layer. The secondary radiation layer converts a large part of the incident X-rays into secondary radiation, i.e. fluorescent radiation. The barrier layer blocks the fluorescent radiation produced in the secondary radiation layer and itself develops only slight secondary radiation. The secondary radiation layer and the barrier layer, as a layer structure, exhibit very good shielding properties when the lead-free radiation protection material according to the invention is processed to form protective clothing. The secondary radiation layer is then provided as the layer of the protective clothing that is remote from the body. The barrier layer, which is arranged in the protective clothing as the layer that is close to the body, effectively blocks the fluorescent radiation produced in the secondary radiation layer in the direction of the body. This ensures optimum shielding efficiency against X-radiation. The lead-free radiation protection material is suitable in particular for the energy range of an X-ray tube having a voltage of from 60 to 125 kV, preferably from 60 to 100 kV, especially from 60 to 80 kV. The secondary radiation layer comprises at least one element of atomic numbers 39 to 60 or a compound thereof. Examples of a suitable element are tin, iodine, caesium, barium, lanthanum, cerium, praseodymium, neodymium and compounds thereof. Particular preference is given to tin or a mixture of tin and caesium. The secondary radiation layer can comprise, for example, tin in an amount of from 50 to 100 wt. %. In a preferred embodiment of the invention, the secondary radiation layer comprises tin in an amount of from 50 to 90 wt. % and at least one further element and/or compound(s) thereof of atomic numbers 39 to 60, in an amount of from 10 to 50 wt. %. The barrier layer of the lead-free radiation protection material according to the invention comprises at least one element of atomic numbers greater than 71 (with the exception of lead) or a compound thereof. In a preferred embodiment, the element is selected from bismuth, tungsten and compounds thereof. The use of bismuth is preferred. It has proved advantageous for the barrier layer to comprise tungsten in an amount of from 0 to 30 wt. % and/or bismuth in an amount of at least 30 wt. %. It has been shown that the barrier layer exhibits an even better barrier effect against secondary radiation of the secondary radiation layer when it further comprises at least one element of atomic numbers 61 to 71 or compounds thereof. In a preferred embodiment of the present invention, the element is selected from the group erbium, holmium, dysprosium, terbium, gadolinium, europium, samarium, lutetium, ytterbium, thulium and compounds thereof. Particular preference is given to gadolinium or a compound thereof. It has further proved advantageous for the barrier layer additionally to comprise at least one element from the group tantalum, hafnium, thorium, uranium and compounds thereof. The proportion by weight of the further elements and/or their compounds present in the barrier layer may be up to 80 wt. %. The amount of the further element(s) and/or compounds thereof is preferably in a range of from 20 to 70 wt. %. The at least two layers of the lead-free material according to the invention comprise a matrix material in an amount of from 0 to 12 wt. %, preferably from 2 to 10 wt. %, especially from 4 to 8 wt. %. The matrix material forms almost a carrier layer for the protective materials, in which the latter are dispersed in powder form; Examples of a matrix material are rubber, latex, synthetic flexible or rigid polymers and silicone materials. It has accordingly been found, surprisingly, that the dose build-up, or the secondary radiation yield, in the lead-free radiation protection material according to the invention is considerably lower than in commercial lead-free materials as a result of its separation into a layer having low secondary radiation and a layer having high secondary radiation. Reference is made in this connection to FIG. 1. In FIG. 1, YM denotes the curve of the lead-free material according to the invention, and the curves A and B are based on commercial lead-free materials, which represent a powder mixture without a layer structure. It will readily be seen that the YM curve comes very close to the Pb curve, which means that the lead-free radiation protection material according to the invention has similarly good shielding properties to the lead material. The secondary radiation layer and/or the barrier layer of the lead-free radiation protection material according to the invention may preferably comprise at least one pure-material layer. The expression “pure-material layer” means a layer that comprises, in addition to matrix material, in each case only one of the above-mentioned elements and compounds thereof, i.e. one protective substance. In a preferred embodiment, these pure-material layers comprise less than 5 wt. % matrix material. It has further been found, surprisingly, that a protective substance or a combination of protective substances provided in separate pure-material layers possesses a substantially better protective effect, i.e. shielding effect, than a material in which all the materials are mixed, for example in the form of a powder. It has been found in practice that the pure-material layers provide a particularly good shielding effect when they are greatly compressed, i.e. when gaps that are as small as possible are present between the particles of the shielding material, so that a layer having as high a density as possible is present. Compression of the layer is effected, for example, by way of a suitable particle size distribution and/or by mechanical compression by known processes. In a preferred embodiment, the pure-material layers should be compressed to more than 75 vol. %. Compression of the pure-material layers to more than 90 vol. % is particularly preferred. In a preferred embodiment of the lead-free radiation protection material according to the invention, the secondary radiation layer and/or the barrier layer comprise(s) at least one pure-material layer. The secondary radiation layer is in such a form that it comprises elements of atomic numbers 39 to 60 or their compounds. It is also possible to provide a plurality of pure-material layers comprising these elements and/or their compounds. In a further preferred embodiment of the lead-free radiation protection material according to the invention, the barrier layer comprises one or more pure-material layers of elements of atomic numbers greater than 71 and/or compounds thereof. The barrier layer may additionally also comprise one or more pure-material layers of elements of atomic numbers 61 to 71 or compounds thereof. The elements having atomic numbers from 61 to 71 and/or their compounds may also be present in a separate layer in the form of a so-called intermediate layer arranged between the secondary radiation layer and the barrier layer. In some cases practice has shown that the best shielding results are obtained when the highly compressed pure-material layers are present in the form of metal foils, such as, for example, in the form of foil strips or foil plates. The metal foils generally have a thickness of from 0.005 to 0.25 mm. The foils are normally located one above the other without being joined together. However, if a bond is to be produced between the foils for practical or technical reasons, such bonds can be produced according to conventional processes. It is shown in the following that the lead-free radiation protection material according to the invention, in comparison with already known lead-free radiation protection materials, exhibits very good results in respect of the shielding effect, especially at 60 kV. The following materials were produced from the following constituents and were tested: Constituents: 40 wt. % tin, 10 wt. % cerium oxide, 20 wt. % gadolinium oxide, 20 wt. % bismuth, 10 wt. % tungsten. The radiation protection materials were processed as follows: Material 1: The above constituents are uniformly mixed in powder form in a polymer matrix; Material 2: Layering of the individual constituents in pure-material layers, in powder form; Material 3: Layering of the above constituents individually in pure-material foils. The weight per unit area was 4.7 kg/m2 in all cases. In the narrow beam cluster of an X-ray tube, the following attenuation factors were obtained according to Table 1 below: TABLE 1Tube voltage(kV)Material 1Material 2Material 3603484977461259.8511.2711.89 As will be seen from the values of the attenuation factors, the lead-free radiation protection material according to the invention arranged in layers (material 2 and material 3) exhibits a better shielding effect than the powder mixture of material 1. In particular, a very good shielding effect is found at 60 kV. It is important that the pure-material layers in the radiation protection material are layered in such a manner that the layers are arranged with increasing secondary radiation. Accordingly, when the material is processed to form radiation protection clothing, the layer having the highest secondary radiation yield is remote from the body, while the layer having the lowest secondary radiation is arranged close to the body. In a further preferred embodiment, the at least one pure-material layer of the secondary radiation layer and of the barrier layer of the lead-free radiation protection material according to the invention may be present in a so-called sandwich structure. A sandwich structure is understood as being a structure in which further layers are provided between the pure-material layers. In a particular embodiment, the at least one pure-material layer has a carrier layer on one side in each case. Alternatively, the at least one pure-material layer may have a carrier layer on both sides. The carrier layers are preferably formed by a polymer. The polymer may be one that is also used as the matrix material. The polymer is usually a latex or elastomer polymer. It has proved advantageous in practice for the one or more carrier layer(s) in the layer structure of the lead-free radiation protection material according to the invention to have a thickness of from 0.01 to 0.4 mm. If necessary, the carrier layer or layers may also comprise small amounts of protective substances, as described above. However, they are generally free of protective substances. The carrier layers on one side or on both sides of the pure-material layers contribute towards increasing the mechanical stability of the “internal”, highly compressed material layer, whether it be the secondary radiation layer or the barrier layer, while the radiation-shielding effect of the individual protective layers is improved. FIG. 2 shows a sandwich structure of the lead-free radiation protection material according to the invention. The highly compressed layer of protective substance 2 is surrounded on both sides by a carrier layer 1, which increases the mechanical stability of the structure. It is also possible to form an alternative sandwich structure by providing a layer having low secondary radiation on both sides of each layer having high secondary radiation. In this manner, the barrier layer effect of the barrier layers having low secondary radiation can contribute towards the provision of a direct barrier effect, i.e. on both sides, for layers having high secondary radiation. In general, the radiation protection materials in the individual layers are in the form of metal powders having particle sizes of from 2 to 75 μm. It is important that there should be as little matrix material as possible in the gaps. It has been found that, in a layer system having an even number of layers, the mass loading (weight per unit area) is 1:1. For example, for a nominal lead equivalent of 0.5 mm (Pb), a weight per unit area of 2.6 kg/m2 per layer is obtained in the case of two layers, which may each in turn be divided into two layers. In a layer structure having an odd number, it has proved advantageous to divide the weights per unit area 2:1 (secondary radiation layer:barrier layer). In a preferred embodiment of the present invention, the division of the weights per unit area in the case of a layer structure of three layers is 1:1:1. This division is particularly advantageous in the case of a layer structure comprising secondary radiation layer:intermediate layer:barrier layer. The intermediate layer comprises predominantly at least one element of atomic numbers 61 to 71 or their compounds. The lead-free radiation protection material according to the invention is suitable for the production of radiation protective clothing such as, for example, a radiation protection apron. In addition, the material according to the invention can advantageously be used, for example, in protective gloves, patient coverings, gonad protection, ovary protection, protective dental shields, fixed lower-body protection, table attachments, fixed or movable radiation protection walls or radiation protection curtains. The invention is explained in greater detail hereinbelow by means of examples. A lead-free radiation protection material according to the invention is produced having a layer (A), which corresponds to the secondary radiation layer, and a layer (B), which corresponds to the barrier layer. Layer (A) comprises 54 wt. % tin, 36 wt. % cerium and 10 wt. % matrix material. Layer (B) comprises 36 wt. % gadolinium, 36 wt. % bismuth, 18 wt. % tungsten and 10% matrix. A lead-free radiation protection material according to the invention is produced. Layer (A) comprises 90 wt. % tin and 10 wt. % matrix, while layer (B) comprises 54 wt. % gadolinium, 36 wt. % bismuth and 10 wt. % matrix material. A radiation protection material according to the invention comprising a layer (A) as in Example 1 and a layer (B) as in Example 2 is produced. A radiation protection material according to the invention having a layer (A) as in Example 2 and a layer (B) as in Example 1 is produced. The measurement results for the lead equivalents (LE) of the radiation protection materials produced in Examples 1 to 4 for tube voltages of 60, 80, 100 and 120 kV are shown in Table 2 hereinbelow. The weight per unit area of the protective substances is 4.7 kg/m2 in each case. TABLE 2Tube voltageExample 1Example 2Example 3Example 4(kV)mm LEmm LEmm LEmm LE600.510.570.580.55800.620.680.710.661000.600.650.660.631250.490.510.530.50
	summary
	abstract	A system for refueling a nuclear reactor is provided. The system includes a lower reactor vessel with a plurality of fuel rods and a plurality of control rods disposed therein, the lower reactor vessel further comprising an upper flange. An upper reactor vessel is provided which encloses a steam generator and a pressurizer, the upper reactor vessel further comprising a lower flange that matingly engages the upper flange of the lower reactor vessel. A transporter surrounds an outer surface of the upper reactor vessel, wherein the transporter is configured to translate the upper reactor vessel vertically toward and away from the lower reactor vessel and also to translate the upper reactor vessel horizontally toward or away from alignment with the lower reactor vessel.
	abstract	A concrete storage module (26) is adapted to slideably receive a cylindrical canister assembly (12) therein. Heat dissipation fins (62) and a tubular heat shield (96) are disposed within the module to help dissipate heat emitted from the nuclear fuel assemblies stored in the canister to air flowing through the module. The canister assembly (12) is composed of a basket assembly (70) constructed from multi-layer structural plates disposed in cross-cross or egg carton configuration. A single port tool (106) is provided for draining water from the canister (12) and replacing the drain water with make-up gas. The single port tool is mounted in the cover (100) of the canister and is in fluid flow communication with the interior of the canister.
047073250	abstract	A gauge plate for use in customizing replacement upper core plate inserts of a nuclear reactor which comprises: a circular metal plate of a known diameter corresponding to that of the upper core plate of the nuclear reactor to be gauged; a plurality of U-shaped gauging slots of known size formed in the peripheral surface of the gauge plate at locations corresponding to the locations of the guide pins for the lower internals of the reactor; an arrangement for positioning the gauge plate within the core barrel at the normal elevation of the upper core plate inserts and the guide pins; and devices, disposed on the gauge plate, for positioning the gauge plate relative to the reactor baffle plate arrangement; and, for determining the actual position of the gauge plate relative to the baffle plate arrangement. The gauge plate is positioned in a core barrel containing a baffle plate arrangement but no upper internals so that the gauge plate rests on the upper end of the baffle plates with the guide pins extending into the gauging slots and with the gauge plate properly positioned relative to the baffle plates; the clearances between the gauge plate and the inner surface of the core barrel, and between the three sides of a gauging slot and the adjacent sides of the associated guide pin are measured at each gauging slot; and, the actual position of the gauge plate relative to the baffle plates is determined by gauging devices.
	summary
	description	The present specification is a continuation application of U.S. patent application Ser. No. 15/648,724, entitled “Systems and Methods for Improving Penetration of Radiographic Scanners” and filed on Jul. 13, 2017, which relies on, for priority, U.S. Patent Provisional Application No. 62/362,585, of the same title and filed on Jul. 14, 2016. The above-mentioned applications are herein incorporated by reference in their entirety. The present specification is related to radiographic systems. More specifically the present specification is related to a method of increasing penetration of radiographic systems and reducing exclusion zones. X-ray imaging is one of the most common methods used for detecting contraband in cargo. However, during the inspection of large containers, as a result of inadequate penetration by the radiation, it is common for traditional X-ray systems to produce images with dark areas. These dark areas might be indicative of the presence of threat materials; however, they yield little information about the exact nature of threat. Typical penetration depths of existing cargo inspection systems range between 200 and 400 mm of iron. While it is known that systems with higher penetration can be obtained with high-power sources, using a higher power source increases the size and footprint of the radiation exclusion zone, limiting wide deployment of such systems. Thus, the use of high-energy X-rays for cargo inspection is not without some tradeoff. On one hand, the source needs to produce high-intensity, high-energy X-ray beams in order to provide high imaging penetration of the cargo. On the other hand, higher X-ray intensities/energies lead to larger radiation footprint, requiring a larger controlled area (exclusion zone), or more shielding around the system. This may also lead to higher radiation dosage to cargo, and in the case of portal systems, to the driver of the cargo as well. When the exclusion zone is not limited or a shielded building is provided to limit the size of the system, the increase of penetration depth begins to taper down as the source intensity is increased, until it reaches a point when larger intensities of the X-ray source do not cause an increase in the penetration depth of the X-rays. The main effect that limits the highest achievable penetration depth is scatter, which represents a background added to the transmitted signal. X-rays from the shaped fan beam scatter from the container walls and cargo and produce a low-frequency background that adds to the transmitted image, effectively reducing contrast, thereby limiting penetration. The intensity of the scatter depends on the number of X-rays impinging on the object being scanned. Longer and wider fan beams produce more scatter than shorter and narrower fans, approximately proportional to the ratio of the irradiation areas. The transmitted signal received at the detectors is thus polluted from X-rays scattering from other parts of the object being inspected. Hence, there is a need to reduce the scatter further to increase X-ray penetration. The most common approach to reduce scatter is to use collimators in conjunction with the detectors. However, deep, heavy and expensive collimators are needed for obtaining desired penetration. In addition, the scatter rejection is only reduced partially, as a collimator itself becomes a source of scatter. Other existing methods to reduce the measured scatter radiation consist of employing Cerenkov detectors that intrinsically are not sensitive to low-energy X-rays, which is characteristic of the scatter radiation. However, these Cerenkov and energy-sensitive detectors are more complex and expensive than standard X-ray detectors and typically do not enable improved intensity modulation. Also, when the source intensity is increased, these detectors start saturating due to the very high count rate. Still other methods are based on measuring the energy spectrum of the radiation and removing the low-energy signals. Currently available X-ray sources usually have a single fixed intensity setting that is set to the output level requested by the customer, which is typically the highest setting that still complies with a required radiation footprint. Moreover, during a typical scan, source output is often much higher than needed to achieve sufficient imaging penetration; not just from one vehicle or container to the next, but also within the cargo of the same vehicle or container. Hence, there is a need to increase X-ray intensity in order to increase penetration without increasing the exclusion zone and/or radiation dosage. Current methods for increasing penetration are based on beam-modulating intensity based on the highest attenuation measured in the previous slice. However, the beam intensity along the slice may be higher than required due to the high attenuation of a small area of the object. The higher intensity results in a larger exclusion zone, or if limited, in a reduction of the source strength that results in lower penetration. PCT Publication Number WO2011095810A3, assigned to the Applicant of the present specification discloses “[a] scanner system comprising a radiation generator arranged to generate radiation to irradiate an object, detection means arranged to detect the radiation after it has interacted with the object and generate a sequence of detector data sets as the object is moved relative to the generator, and processing means arranged to process each of the detector data sets thereby to generate a control output arranged to control the radiation generator to vary its radiation output as the object is scanned.” There is still a need, however, for more fine control to modulate the intensity as a function of vertical positions within the slice to further optimize the intensity imparted to the object. The WO2011095810 publication is incorporated herein by reference in its entirety. In addition, U.S. Pat. No. 9,218,933, also assigned to the Applicant of the present specification, discloses “[a]n X-ray source for scanning an object comprising: an electron beam generator, wherein said electron beam generator generates an electron beam; an accelerator for accelerating said electron beam in a first direction; and, a first set of magnetic elements for transporting said electron beam into a magnetic field created by a second set of magnetic elements, wherein the magnetic field created by said second set of magnetic elements causes said electron beam to strike a target such that the target substantially only generates X-rays focused toward a high density area in the scanned object”. What is still needed, however, is a system that does not require complex electron-transport components. The '933 patent is incorporated herein by reference in its entirety. Even when a system has very high penetration, there may be dark alarms that require labor-intensive manual inspection for clearing. There is a need for reducing the dark alarm rate further to reduce manual inspections. Therefore, there is a need for scanning systems with increased penetration and smaller exclusion zones, resulting in improved performance and lower alarm rates and easy deployment in a wide range of environments. In some embodiments, the present specification discloses an X-ray detection system with increased penetration comprising: an X-ray source for projecting an X-ray beam towards an object; a mechanism for producing one or more fanlets from the X-ray beam, each fanlet comprising a vertically moving fan beam having an angular range smaller than the angular coverage of the object; a detector array for detecting the fanlets projected on the object; a controller for synchronizing the X-ray source and the mechanism, and collecting image slices from the detector array corresponding to the fanlets; and a processing unit for combining the image slices collected into a composite image. In some embodiments, the present specification discloses an X-ray detection system configured to provide for increased penetration of an object, comprising: an X-ray source for generating an X-ray beam in an inspection volume; a conveyor for moving the object through the inspection volume; a collimator positioned between the X-ray source and the object, wherein the collimator is configured to receive the X-ray beam and produce one or more fanlets from the X-ray beam, wherein each fanlet comprises a vertically moving fan beam having an angular range greater than 1 degree but smaller than the angular coverage of the object; a detector array opposing said X-ray source and positioned within the inspection volume for detecting the one or more fanlets projected on the object; a controller configured to synchronize the X-ray source and the collimator and collect image slices from the detector array corresponding to each of the one more fanlets; and a processing unit for combining the image slices collected into a composite image. Optionally, the X-ray source is a pulsed X-ray source. Optionally, the X-ray source produces dual-energy beams. Still optionally, the dual-energy beams are interlaced. Optionally, the X-ray source produces X-ray pulses comprising low and high energy X-ray beams separated in time. Optionally, the controller is configured to control the conveyor such that a total time for the one or more fanlets multiplied by a rate of speed of the conveyor is equal to or less than a width of a detector in the detector array. Optionally, the collimator is configured to generate an overlap between the one or more fanlets of approximately 1 degree with respect to the object. Optionally, the X-ray source is a CW X-ray source. Optionally, the collimator for producing the one or more fanlets comprises a plurality of controlled fast actuators coupled with beam attenuators to shape the X-ray beam. Optionally, the collimator for producing the one or more fanlets comprises a beam chopper. Optionally, the collimator for producing the one or more fanlets comprises a rotating wheel with slits designed to produce the vertically moving one or more fanlets. In some embodiments, the present specification is directed toward an X-ray detection method comprising: irradiating an object with more than one X-ray fanlet, wherein each X-ray fanlet comprises a vertically moving fan beam having an angular range greater than 1 degree but smaller than the angular coverage of the object and wherein each X-ray fanlet is produced by using a collimator for collimating an X-ray beam generated by an X-ray source; synchronizing the X-ray beam and the more than one fanlet; detecting the more than one fanlet irradiating the object; collecting image slices from the detector array corresponding to a complete scan cycle of the more than one fanlet; and processing the image slices and combining the image slices into a composite image. Optionally, the method further comprises adjusting a beam intensity and energy of each of the more than one fanlets based on signals detected from a previous fanlet at a same vertical position with respect to the object to generate a control output, wherein the control output is used to control the X-ray detection method. Optionally, the X-ray source is a pulsed X-ray source. Optionally, the X-ray source produces dual-energy beams. Optionally, the dual-energy beams are interlaced. Optionally, the X-ray source produces X-ray pulses comprising low and high energy X-ray beams separated in time. Optionally, the collimator is configured to generate an overlap between the one or more fanlets at every position with respect to a surface area of the object. Optionally, the collimator comprises a spinning cylinder with a helical aperture. Optionally, the collimator comprises a plurality of controlled fast actuators coupled with beam attenuators to shape the X-ray beam. Optionally, an energy of each of the more than one fanlet is adjusted at a same fanlet location in a following cycle to allow for interlaced dual-energy scanning of every vertical position. Optionally, the X-ray source is a CW X-ray source. In some embodiments, the present specification discloses a method for operating a scanning system, wherein said scanning system comprises an X-ray source, an array of detectors, and a processor to process and analyze image data, the method comprising: generating a first X-ray beam in order to conduct a first scan to produce an image of the object being scanned; determining areas in said image data that require a more detailed inspection; configuring a collimator to limit a second X-ray beam such that, upon emission of the second X-ray beam, the collimator emits a plurality of fanlets, wherein each fanlet has an angular range that is less than an angular range covering an object but greater than 1 degree; and moving the object relative to the X-ray source and the array of detectors to perform a second scan on the areas. Optionally, said areas represent a lack of penetration by the first X-ray beam during said first scan. Optionally, said areas represent items of interest or alarm such as explosive, firearms, drugs or contraband. Optionally, the X-ray source and array of detectors are mounted on a gantry. Optionally, the collimator comprises a plurality of controlled actuators coupled with beam attenuators to shape the second X-ray beam. Optionally, the collimator comprises two vertically controlled attenuators to inspect only said areas. Optionally, a scan of said areas using said plurality of fanlets is performed at a lower speed compared to a speed of a scan using the first X-ray beam. Optionally, the method further comprises replacing the areas generated by a scan using a first X-ray beam with images of the areas generated by a scan using the plurality of fanlets. The aforementioned and other embodiments of the present shall be described in greater depth in the drawings and detailed description provided below. The present specification describes scanning systems having increased penetration capability and smaller exclusion zones, resulting in improved performance and easy deployment in a wide range of environments. Embodiments of the present specification are well-suited for applications in environments including, but not limited to, container, truck and railcar inspection. Some embodiments of the present specification are particularly well-suited for use in inspecting slow-moving vehicles. The present specification is directed towards systems and methods for both reducing the exclusion zone and increasing the penetration capability of radiographic systems, such as X-ray scanners. In an embodiment, the imaging system described in the present specification enables the scanning of high density cargo with a sufficient penetration depth for the detection of contraband resulting in a low probability of dark alarms that may require a secondary inspection. The present specification also describes an imaging system having a lower impact from scatter radiation that is observed in conventional X-ray scanners and that can be used for inspecting high-density cargo. The present specification also describes a novel method that allows for optimization of the radiation intensity imparted to cargo and environment, which further increases penetration. In an embodiment, the present specification describes a novel mechanism for reducing scatter by producing a vertically moving fan beam with an angular range smaller than the angular coverage of the object being scanned. The present specification provides a vertically moving fan beam or “fanlet” synchronized with a pulsed X-ray source and a data acquisition system. In an embodiment, the “fanlet” represents a portion of the total overall fan beam, and is vertically translated to cover the extent of the object. In an embodiment, a vertical collimator projects a fanlet having an angular range smaller than the angular coverage of the object being scanned. In an embodiment, the angular range is achieved by using a collimator having dimensional characteristics that are independent of the object, but that are tailored to insure the highest and widest possible object dimensions are accounted for. In an embodiment, the collimator is designed to provide collimation for a predefined object height and object width, which are larger than a standard object height and width, thereby insuring no portion of the object remains unscanned. The fanlet, via collimator mechanics, is translated vertically to cover the angular spread of the object. A pulsed linac X-ray source and a data acquisition system are synchronized with the moving collimator in such a way that the image of the object is acquired at intervals, where in one cycle the fanlets cover a slice of the object with no gaps and, optionally, a minimal overlap. The image from each fanlet is then combined to produce a slice image. In one embodiment, to minimize the effect of object motion, the source pulsing frequency is increased by the number of fanlets. The advantage of this embodiment is that the scatter is reduced as the irradiated area is reduced in each acquisition. The present specification is also directed towards reducing the radiation exclusion zone. In additional embodiments, the signals from each fanlet are used to control the intensity of the fanlet for the following cycle, to optimize the source intensity. In an embodiment, the beam intensity and/or energy is modulated based on the transmission observed in each fanlet to expose the object to the minimum intensity required for penetration, while at the same time reducing the dose to cargo and the environment resulting in a smaller exclusion zone. This is similar to the intensity modulation described in PCT Publication Number WO 2011095810A, incorporated herein by reference in its entirety, which is applied to the full fan beam. The embodiments described herein may be employed for dual-energy scanning as well, since the time between pulses at the same vertical location is the same as in a standard system because the pulsing rate is increased accordingly. However, for fast moving objects, the pulsing frequency is high and it might not be possible to increase the pulsing frequency by a factor of two or three. In these applications, the preferred embodiment is to use a pulsed source, where each pulse contains dual energies separated by a short time. In another embodiment, a Continuous Wave (CW) source is used. In this embodiment, the data acquisition system collects data continuously at a plurality of time intervals with times shorter than the time it takes for the collimator to move from the top position to the bottom position to cover the slice. The present specification is directed towards multiple embodiments. The following disclosure is provided in order to enable a person having ordinary skill in the art to practice the specification. Language used in this specification should not be interpreted as a general disavowal of any one specific embodiment or used to limit the claims beyond the meaning of the terms used therein. The general principles defined herein may be applied to other embodiments and applications without departing from the spirit and scope of the specification. Also, the terminology and phraseology used is for the purpose of describing exemplary embodiments and should not be considered limiting. Thus, the present specification is to be accorded the widest scope encompassing numerous alternatives, modifications and equivalents consistent with the principles and features disclosed. For purpose of clarity, details relating to technical material that is known in the technical fields related to the specification have not been described in detail so as not to unnecessarily obscure the present specification. It should be noted herein that any feature or component described in association with a specific embodiment may be used and implemented with any other embodiment unless clearly indicated otherwise. FIG. 1A illustrates an X-ray system comprising an X-ray source 110 and a detector array 120 scanning a railcar 130 containing cargo 140. X-ray path 150 represents the non-interacting X-rays that are transmitted through the cargo 140. In an ideal system, these would be the only X-rays that would be detected. X-ray paths 160 represent X-rays scattered by the walls of the railcar container 130, and X-ray paths 170 represent X-rays scattered within the cargo 140. The scattered X-rays represented by paths 170 constitute background noise for the X-ray system. In various embodiments, the present specification provides systems and methods to reduce the background noise. FIG. 1B illustrates a collimator coupled with a detector array for reducing the X-ray scatter signal. In FIG. 1B a detector collimator 180 is coupled with the array of X-ray detectors 120 for reducing the scattered X-rays (such as X-rays 170 shown in FIG. 1A). As shown, the path of the primary X-ray beam 190 does not interact with the collimator 180 and is detected by the detector array 120, while X-rays following path 192 are absorbed in the collimator 180 and not detected. Also, X-rays following path 194 go through collimator 180 and are detected by detector array 120, while X-rays following path 196 scatter in collimator 180 into the detector array 120 and are also detected. These effects show that collimators reduce scatter, however, deeper collimators, or collimators that have a longer source to detector distance, result in higher rejection. The performance of the collimator is affected by the ratio of length to width of the individual collimator openings. The higher the ratio of length to width, the better the scatter rejection of the collimator; however, such an embodiment is more expensive to manufacture. Further, as the collimator is made deeper, scatter in the collimator limits the rejection. Thus, there is a trade-off between using a deep collimator and achieving scatter reduction as the X-rays that scatter in the collimator (which is used to reduce scatter from the cargo) may become greater in number than the left-over scatter from the cargo. In an embodiment, collimator depth is maximized at 300 mm, after which depth, gain is minimized. It should be noted that the collimator wall thickness cannot be made too thick as it would reduce the number of unscattered X-rays. Thus, in order to reduce X-ray scatter, a greater number of collimator panes is employed. The present specification, in an embodiment, provides a method of reducing X-ray scatter signal by generation of a vertically moving X-ray beam or fanlets. FIG. 2 illustrates a system comprising a pulsed source projecting vertically-moving fanlets to scan a cargo with reduced scatter, in accordance with an embodiment of the present specification. The system comprises a pulsed X-ray source 210 for scanning a railcar (or other object) 230 and a detector array 220. Examples of suitable X-ray sources include, but are not limited to, electron linac hitting a tungsten target and CW sources such as Rhodotron and superconducting linac. One of ordinary skill in the art would appreciate that any pulsed X-ray source known in the art may be employed. Collimator 240 represents a mechanism that produces a vertically moving fan beam or fanlets 250, 260 and 270, with an angular range smaller than the angular coverage of the railcar 230. Referring back to FIG. 2, the signal produced by fanlet 260 has reduced scatter compared to the full fan-shaped X-ray beam that is generally used to inspect cargo in conventional systems. In an embodiment, the X-ray pulses and the scanning mechanism are synchronized to collect data when the fan beam(s) are projected to fanlet positions 250, 260 and 270 to cover the vertical extent of the cargo railcar 230 in one cycle. A processing unit combines the data from the fanlets 250, 260, 270 to form an image of a slice of the cargo railcar 230. As the collimator defines the fanlet and tends to produce a beam with fuzzy edges, a small overlap between the fanlets 250, 260, 270 is preferred to allow for better “stitching” of the fanlets 250, 260, 270 into a slice image to eliminate or minimize edge effects. In an embodiment, an overlap of approximately 1 degree is employed. It may be noted that any suitable approach known in the art may be employed for stitching together the image slices. In an embodiment, in order to reduce the effect of cargo motion, the source pulsing frequency is increased approximately in proportion to the number fanlets. For example, in a mobile application, the pulsing frequency is about 100 Hz. If the number of fanlets is 3, the frequency would be increased to 300 Hz. In an embodiment, the smallest number of fanlets is produced by dividing the corresponding fan beam in half; however this does not provide a significant reduction in scatter. By increasing the number of fanlets, which is achieved by decreasing the angular range of each fanlet, scatter radiation is decreased. However, an increased number of fanlets can only be obtained by proportionately increasing the pulsing frequency for a pulsed Linac source. In an embodiment, a typical angular range for a fan beam for a scanner is approximately 60 degrees. In an embodiment, the angular range of a fanlet ranges from 1 degree to 30 degrees. In an embodiment, ten fanlets are employed, each having an angular range of 5 degrees. One of ordinary skill in the art would appreciate that a fanlet has a considerably larger angular range than a conventional pencil beam, which is on the order of a fraction of a degree. The X-ray dose to cargo and the environment does not increase, because the total number of X-rays is the same as compared to a standard X-ray scan. However, the scatter is reduced as there are fewer X-rays inspecting the cargo at any acquisition time relative to the primary beam incident on the detectors. For dual-energy scanning, the source may be either interlaced (meaning at a first pulse, a first energy, at a second pulse a second energy, and at an nth pulse an nth energy) or may contain both energies in the same pulse separated by a small time gap (>˜100 ns). In this way, the frequency is effectively increased by a factor of two. For example, in a standard system operating at 250 Hz, the source emission frequency maybe increased to 375 Hz with a dual-energy per pulse, resulting in an effective frequency of 750 Hz, enabling the use of three fanlets with small cargo motion effects. In an embodiment, for interlaced dual-energy scanning, an odd number of fanlets are generated so that the second energy is at the same fanlet location in the following cycle to allow for dual-energy scanning of every vertical position. For example, in the case of three fanlets, in the first cycle, the following pattern would be seen: Top Fanlet having High Energy (HE), Center Fanlet having Low Energy (LE), and Bottom Fanlet having High Energy (HE). In the subsequent cycle, the following pattern would be seen: Top Fanlet having Low Energy (LE), Center Fanlet having High Energy (HE), and Bottom Fanlet having Low Energy (LE). Thus, in an embodiment, the first cycle is HE-LE-HE and the following cycle is LE-HE-LE, thereby allowing interlacing energy for the corresponding fanlet positions for consecutive cycles. It may be noted that if the number of fanlets is even, then the energy at each position would be either LE or HE, and arrangements of LE-HE or HE-LE for the same vertical position will not be possible. FIG. 3 illustrates a system comprising a CW source projecting continuously-moving fanlets, in a vertical motion, to scan cargo with reduced scatter, in accordance with another embodiment of the present specification. FIG. 3 illustrates an X-ray system comprising a CW X-ray source 310 and a detector array 320 scanning a cargo railcar 330. Collimator 340 represents a mechanism that produces a vertically continuously moving fan beam with an angular range smaller than the angular coverage of the railcar 330. The scanning mechanism is synchronized with a data acquisition module to start data collection at the detector array 320 in position 350 and end data collection at position 360 to cover an angular range of fanlet 370. In FIG. 3, the end position 360 constitutes the start position of the next acquisition cycle. The data collection continues in similar fashion until the full vertical extent of the cargo is covered by the “individual” fanlets. As in the pulse-source embodiment shown in FIG. 2, the scatter is reduced by using the CW source 310. It may be noted that the operation of the system remains the same regardless of whether the source is pulsed or CW. While a pulsed high energy x-ray source produces a pulse of a few microseconds separated by few milliseconds, a CW source continuously produces X-rays. FIG. 4 is an exemplary illustration in which the imaging system of the present specification is used for scanning an ANSI 42.46 standard penetration phantom object. As shown in FIG. 4, an ANSI 42.46 penetration phantom object 401 is placed inside a rail-cargo 405. The ANSI 42.46 standard penetration phantom object 401 is used for assessing the penetration capability of high-energy radiographic systems. Said object 401 comprises a rectilinear iron block 406 having a length and a width of at least 60 cm each; and an iron block 404 of an approximate rhomboidal shape placed behind the rectilinear block 406. The thickness of the rhomboidal block 406 is approximately 20% of the thickness of the rectilinear block 406. In the testing procedure shown in FIG. 4, the phantom object 401 is placed at the center of a rail-cargo container 405 tilted towards the X-ray source 402. An array of X-ray detectors 403 is set up to detect the X-rays transmitted through the object 401. A successful ANSI test of penetration for an X-ray system is based on assessing the capability of that X-ray system in determining the direction in which a tip 407 of the rhomboidal object 406 points in the captured image. FIG. 5 illustrates exemplary simulated images for ANSI 42.46 penetration phantom objects obtained with a full fan beam of X-rays and with the use of multiple fanlets via the imaging system described in FIG. 4, in accordance with an embodiment of the present specification. Image 510 is formed by irradiating the phantom object (such as object 401 shown in FIG. 4) comprising a rectilinear object coupled with a rhomboidal shaped object, with a full fan beam. As can be seen, the image quality of image 510 is poor as it is difficult to distinguish the rhomboidal shaped object 502 within rectilinear object 501 in this image. Image 520 is obtained by irradiating the phantom object (such as object 401 shown in FIG. 4) by using multiple fanlets of X-rays such as described with reference to FIG. 4. Using multiple fanlets, the image contrast is improved as less scatter is measured. As can be seen, the image quality of image 520 is better as the rhomboidal shaped object 502 within rectilinear object 501 is better visible as compared to the image 510. Image 530 is obtained by irradiating the phantom object (such as object 401 shown in FIG. 4) by using a larger number of fanlets of X-rays than used to obtain image 520. By using a larger number of fanlets, even a lower number of scattered X-rays are detected. As can be seen from the figure, the quality of image 530 is better than that of image 520 as the rhomboidal shaped object 502 within rectilinear object 501 is most clearly visible in image 530. The production of vertically moving fanlets of X-rays requires a system for projecting an X-ray beam with an angular range smaller than the angular coverage of the object being inspected. In one embodiment, the system comprises a radiation source that emits radiation at an emission rate (Re) and a conveyor that moves an object through the system at a conveyor rate (Rc), where the time (Tf) for a fanlet to traverse the object is preferably equal to the time for a single radiation pulse. In such a case, the total amount of time for a set of fanlets (which, when combined, cover the entire angular range encompassing the object) to be emitted is equal to times the total number of fanlets (Nf): TfNf. That total time, when multiplied by the conveyor rate (Rc), should preferably be equal to or less than a detector width (Dw), thereby insuring no portion of the object is missed. Therefore: TfNf*Rc≤Dw, where Tf is the time for one fanlet, Nf is the total number of fanlets, Rc is the conveyor speed, and Dw is the detector width. Various embodiments for producing vertically-translated fanlets are described below. FIG. 6A illustrates a mechanism comprising multiple actuators connected to beam attenuators to produce vertically-moved fanlets, in accordance with a preferred embodiment of the present specification. A plurality of actuators 610 connect to a plurality of beam attenuators 630 through steel push/pull drive rods 620. The actuators 610 are computer-controlled to move the beam attenuators 630 to attenuate the beam to project vertically moved fanlets, as described in more detail in FIG. 6B. In an embodiment, the actuators 610 are rotary actuators for obtaining a fast response time for scanning fast moving objects. In alternate embodiments for deep scanning which includes scanning slow moving or stationary objects, other types of actuators such as pneumatic actuators may be used. In an embodiment, for performing a deep scan, a single fanlet having an angular range sufficient to cover the object's area of interest is used. In cases where a large part of a cargo being scanned is highly attenuating, and scanning the same at a low speed is possible, X-ray fanlets such as described above are used to scan the cargo. However, the speed of scan is maintained lower than that used for scanning a fast moving cargo. In an embodiment, the number of fanlets used for scanning the cargo at a slow speed is greater than that used for scanning a fast moving cargo. For example, and by way of example only, at a pulsing frequency of 1 KHz, a Linac source produces 1 X-ray pulse every 1 millisecond ( 1/1000 Hz=1 ms). While scanning an object moving at 3.6 km/h (or 1 mm in 1 ms or 1 mm per pulse), by using a detector having a width of 10 mm, the entire object is covered by the X-rays because the detector is wider than the distance moved by the object per pulse. Hence, the maximum number of fanlets that can be used to scan the object without missing any part of the object is 10, as it takes 1 ms per fanlet, which if multiplied by 10 fanlets=10 ms, meaning 10 mm of distance travelled by the object, which is equal to the detector width. However, if the number of fanlets is increased, for example to 20 fanlets, the time it would take the fanlets to cover the object would be 20 ms, which means the object also moves by 20 mm. Since the detector width is only 10 mm, a part of the object would be missed by the X-rays. However, if the speed of the object is lowered to 1.8 km/h, the object moves 10 mm in 20 ms, thereby allowing every part of the object to be scanned. Accordingly, in one embodiment, the system monitors whether the total fanlet time, when multiplied by the conveyor rate (Rc), is greater than a detector width (Dw). If the system determines that it is, the conveyor rate is (Rc) is decreased to a rate sufficient to insure that the total time, when multiplied by the conveyor rate (Rc), is equal to or less than a detector width (Dw). FIG. 6B is a block diagram illustrating various attenuator configurations in the mechanism to produce vertically-moved fanlets shown in FIG. 6A. As shown in FIG. 6B, a vertical collimator 640 is coupled with a plurality of beam attenuators 630a, 630b, . . . , 630n, which in turn are connected to a plurality of actuators (not shown in FIG. 6B) as shown in FIG. 6A. The vertical collimator 640 projects a fan beam that covers the complete vertical extent of the object being scanned. The plurality of attenuators 630a, 630b, . . . , 630n may be controlled by means of the rods 620 coupled with actuators 610, to move in and out of the projected beam to project X-ray fanlets that move vertically with respect to the object being scanned. In the configuration 650, attenuators 630b, 630c and 630d are moved into the beam to attenuate the beam, while attenuator 630a stays out of the beam to project a fanlet over an upper part of the object being scanned. In the configuration 660 attenuators 630a, 630c and 630d are moved into the beam to attenuate the beam, while attenuator 630b stays out of the beam to project a fanlet over an upper middle part of the object being scanned. In the configuration 670 attenuators 630a, 630b and 630d are moved into the beam to attenuate the beam, while attenuator 630c stays out of the beam to project a fanlet over a lower middle part of the object being scanned. In the configuration 680 attenuators 630a, 630b and 630c are moved into the beam to attenuate the beam, while attenuator 630d stays out of the beam to project a fanlet over a lower part of the object being scanned. Hence, the fanlet is moved to project X-rays over different parts of the object being scanned by moving an attenuator out of the X-ray beam being projected. The movement of the attenuators as described provides vertically moving X-ray fanlets. In various embodiments, the beam attenuators 630a, 630b, . . . , 630n are made of high-density materials such as but not limited to lead or tungsten. In another embodiment X-ray fanlets may be moved vertically with respect to an object being scanned by means of a helical profile aperture formed on a rotating cylinder. FIG. 7 illustrates an exemplary design of a spin-roll chopper being used for moving X-ray fanlets vertically with respect to an object being scanned, in accordance with an alternate embodiment of the present specification. The spin-roll chopper is described in U.S. Pat. No. 9,058,909 B2, which is incorporated herein by reference in its entirety. The rotation of the spin roll/beam chopper provides a vertically moving fanlet of constant size and velocity. Beam chopper 702 is, in one embodiment, fabricated in the form of a cylinder made of a material that highly attenuates X-rays. Beam chopper 702 comprises helical chopper slits 704. The cylindrical shape enables the beam chopper 702 to rotate about a Z-axis 703 and along with the helical apertures 704, create a spin-roll motion, which provides an effective vertically moving aperture 704 that may project a vertically-moving fanlet of X-rays onto an object being scanned. In one embodiment, slits 704 are wide enough to allow a fanlet beam to be projected, as required by the system of present specification. It may be noted that narrow slits would produce a pencil beam and not a fan or fanlet beam. FIG. 8a shows an exemplary mechanism for generating moving fanlets, according to another alternate embodiment of the present specification. Referring to FIG. 8a rotating mechanism 800 comprises a wheel 801 with three slits 802, 803 and 804, which are in the shape of an arc or a partial circle. In one embodiment, the wheel is made of a material highly attenuating for X-rays, such as lead or tungsten. Wheel 801 further comprises a vertical collimator 805. In operation, as the wheel is rotated, the intersection of a slit 802 and the vertical collimator 805 results in the blocking of the radiation from the slit, except for a section 806a that projects a fanlet. In one embodiment, the width of the slit is configured to produce the desired fanlet angular extent. In one embodiment, the rotating frequency of the wheel is determined based on the fanlet width and linac pulsing frequency. The wheel rotation is synchronized with the linac pulsing frequency to generate fanlets with little overlap and cover the cargo extent in one cycle. FIGS. 8b, 8c and 8d are a series of figures illustrating various positions of the wheel to indicate how the fanlets are produced and move to cover the extent of an object being scanned. Referring to FIGS. 8b, 8c and 8d, along with FIG. 8a, position 810 shows the fanlet 806a in the upper most location. When wheel 801 is rotated in a counterclockwise direction, the fanlet 806b moves downwards as shown by position 820 in FIG. 8b. One of ordinary skill in the art would appreciate that the wheel may be rotated in clockwise direction as well. Thus, with further rotation after position 820, the fanlet 806c moves further down as shown in position 830 in FIG. 8c. When the fanlet exits the lowest position, the next slit 803 in the wheel projects the upper fanlet 807. This is shown as position 840 in FIG. 8d. The cycles of rotation are repeated until the complete object is scanned. It may be noted that while the utilization of fanlets for scanning reduce the scatter, but there is still some scatter produced by the cargo interacting with the x-ray beam within the fanlet. Therefore in one embodiment, the system of the present specification measures the scatter with the detectors outside the fanlet and uses this measurement to estimate the scatter in the fanlet. The estimated scatter is then subtracted from the transmission image data to increase contrast of the resultant image. One of ordinary skill in the art would appreciate that even with the increased penetration provided by the embodiments of the present specification, there would be dark alarms that may require manual inspection which is labor intensive. Therefore, in another embodiment, the present specification describes a method for scanning an object that employs a two-step process to further reduce dark alarms. This process is illustrated by means of a flow chart in FIG. 9. Referring to FIG. 9, in the primary scan 901, a truck or cargo container is scanned with a standard fan beam or fanlets of single or multi-energy high-energy radiation, where the transmitted radiation is measured with an array of detectors. In an embodiment, the truck or cargo container is scanned through a complete cycle, wherein a complete cycle is a scan of the vertical extent of the object under inspection using a standard fan beam having an angular range or a plurality of fanlets having a total angular range of a standard fan beam, as described above. Thus, in an embodiment, the fanlet, via collimator mechanics, is translated vertically to cover the angular spread of the object in a complete cycle. A pulsed linac X-ray source and a data acquisition system are synchronized with the moving collimator in such a way that the image of the object is acquired at intervals, where in one cycle the fanlets cover a slice of the object with no gaps and, optionally, a minimal overlap. The image from each fanlet is then combined to produce a slice image. The transmission information is analyzed in step 902 to determine areas of dark alarm. If no areas of dark alarm are found (903), then the transmission image is analyzed to determine the presence of contraband and other items of interest, as shown in 909. If one or more areas of the image are not penetrated by the beam (dark alarm), the areas are subjected to a secondary scan, as shown in step 903. In the secondary scan, a horizontal collimator is adjusted to only cover the vertical extent of the dark area, and suspect areas, if any. This is shown in 904. The container is then repositioned to allow the location of suspect area to be rescanned. In one embodiment, the radiation source is tilted to align with the center of dark area, as shown in 905. In one embodiment, the rescan is preferably performed at a lower speed than the primary scan, such as for example at 1/40th of the standard scanning speed. This is shown in 906. In one embodiment of the system, the source and detectors are mounted on a gantry that allows repositioning the system and scanning any part of the object with a wide range of speeds. Optionally, the source is tilted in such a way that the beam center line is aligned with the center of the dark areas to increase the beam intensity, since the Bremsstrahlung x-rays are more intense. The reduction of the vertical extent by suitably using a collimator prevents scatter from other areas of the container and increases penetration. It may be noted that scatter reduction also helps improving material separation with dual-energy beams as the single-energy images are cleaner from the scatter that distorts the x-ray spectra. The lower scanning speed further allows for improved statistical accuracy and also increases penetration. Thereafter, the scanning system examines the transmission image again to check if there are any more dark alarms, as shown in 907. If more dark alarms are found in the scan image, a rescan is performed again, by repeating the steps 904, 905 and 906. This process continues until all dark alarms are resolved. When there are no more dark alarms, the rescanned sections of the image are integrated into the original image of the object, as shown in 908. This is done, in one embodiment, by replacing the original sections of the image with corresponding rescanned sections. The transmission image is then analyzed to determine the presence of contraband and other items of interest, as shown in 909. Another motivation for the secondary scan, in addition to clear dark alarms, is to clear automated high-Z alarms. It may be noted that the system of present specification uses automated programs to generate alarms when a high Z material is detected. This system and method of automatically generating alarms when a high Z material is detected is described in U.S. patent application Ser. No. 14/104,625, entitled “Systems and Methods for Automated, Rapid Detection of High Atomic Number Materials” and filed by the applicant of the present specification, which is incorporated herein by reference in its entirety. It may be noted that the method for automatically detecting high Z materials employs attenuation information from the segmented objects and surrounding background. Therefore, rescanning suspect objects with lower scatter can resolve the alarm, as there is an improved single- and dual-energy contrast to reduce the need for active interrogation. Thus, in one embodiment, the system of present specification employs the rescan approach described above with reference to FIG. 9, to clear automated high Z alarms in a manner similar to clearance of dark alarms. In one embodiment, additional improvement is obtained by another scan performed at a 10-20° angle to allow for a different view of the cargo that would have a different set of superimposing objects. One of ordinary skill in the art would appreciate that the requirement of confirming an alarm in all stages of scan would result in an even lower false-alarm rate. Those skilled in the art would also appreciate that secondary inspection may be applied not only to high Z materials, but may be extended to other objects of interest as well, such as suspected contraband including explosives, firearms, drugs, etc. In one embodiment, the X-ray source may be replaced with a neutron source. It may be noted that when the x-ray source is replaced with a neutron source, the detectors are replaced with neutron detectors and the collimators are replaced with neutron-attenuating materials instead of lead. However, the operation of the system remains the same. In the description and claims of the application, each of the words “comprise” “include” and “have”, and forms thereof, are not necessarily limited to members in a list with which the words may be associated. The above examples are merely illustrative of the many applications of the system and method of present specification. Although only a few embodiments of the present specification have been described herein, it should be understood that the present specification might be embodied in many other specific forms without departing from the spirit or scope of the specification. Therefore, the present examples and embodiments are to be considered as illustrative and not restrictive, and the specification may be modified within the scope of the appended claims.
048083372	description	DETAILED DESCRIPTION OF THE DRAWINGS Referring firstly to FIGS. 1 and 2 of the drawings, there is shown a compressible, bellows-type metal canister 1, for use in a hot pressing process for immobilising high level radioactive nuclear waste material in the form of a synthetic rock. The canister typically is generally as described in U.S. Pat. No. 4,645,624. The canister includes a gas filter and discharge arrangement constituting one embodiment of the invention. The canister 1 comprises a base wall 2 and a corrugated bellows like side wall 3 of generally circular cross-section. Concentrically arranged within the corrugated side wall 3 is a cylindrical liner 4. In the centre of the base wall 2 is located a conically-tapered aperture 5 provided with a filter plug shown diagrammatically at 6. Between the corrugated side wall 3 and inner liner 4 of the canister are provided two further, diametrically-opposed apertures 7. All three apertures 5, 7 are connected by an outlet pipe 8 extending diametrically across the base wall 2 and exteriorly of the canister. This outlet pipe 8 is connectable to any suitable waste disposal system, as will be described hereinafter with respect to a preferred embodiment. Referring now to FIGS. 3A and 3B, there are shown two alternative embodiments of filter plug 6 which may be used in association with the central aperture 5 in the base wall 2 of the compressible canister 1. The filter plug 6 in FIG. 3A comprises an inverted castellated cap 9 with which is associated a filter mass 10 made of alumina or titania fibre. This filter material is packed into the conically-tapered aperture 5 and into the gaps between the castellations of the cap 9. The projecting lugs of the castellated cap 9 rest on the upper surface of the base wall 2 around the periphery of the conical aperture 5 and thus compressive forces in the axial direction of the canister are absorbed and ingress of synthetic rock forming components into the filter structure are substantially avoided. The filter plug 6 shown in FIG. 3B differs from that of FIG. 3A only in that it has a filter disc 10' made of Hastalloy in place of the mass of alumina or titania fibre. The filter disc 10' is welded around its periphery as shown at 16 to the conical-aperture 5. Furthermore, in the embodiment of FIG. 3B the outlet duct 8 is formed by the co-operation of a slot in the underside of the base wall 2, the duct being closed on its lower side by co-operation with the uper face of pressure pad 12 resting on a hydraulic ram. The discharge of gases through the outlet duct 8 can be to a gas processing system of the type described below with reference to FIGS. 4A and 4B. The gases will comprise the gas in the interstices of the particulate material in the canister and any volatile components produced from the particulate material during the heating stage. As shown in FIGS. 4A and 4B, the outlet pipe 8 (or outlet duct) is connected to an outlet tube 11. In FIG. 4A, the compressible canister 1 is shown in a free-standing position upon a lower pressure pad 12 of a hydraulic press associated with an induction furnace (not shown) in which the canister is to be heated to a high temperature and then compressed axially. In this arrangement, the outlet tube 11 is L-shaped and has its horizontal limb rotatably but sealingly mounted in a side of the base wall 2; the terminal limb in the illustrated loading position extends upwardly, with its open end free to the atmosphere. In the process, as shown in FIG. 4B, the compressible bellows-type canister 1 is raised by the hydraulic ram to place the upper wall 17 of the canister against a fixed refractory abutment pad 13. The canister is thus positioned so as to be heated in the induction furnace (not shown) which surrounds the canister. However, before heating can commence, the outlet tube 11 is rotated through 180.degree. into a downwardly extending position, such that the terminal limb extends into a manifold arrangement 14 communicating with an exhaust tube 15, which is connected to a low pressure gas filtration system. It is to be noted that the manifold arrangement 14 and associated down pipe 15 are mounted on the lower pressure pad 12, so that they can move in unison with the exhaust tube 11 and canister 1 supported on that pad. Although the high level radioactive nuclear waste incorporated into the synthetic rock materials includes elements volatile at the typical temperatures to which the material is heated (about 1150.degree. C.) it has been fund that little, if any of these components are infact exhausted from the canister; it is thought these volatile components are absorbed into the synthetic rock materials. However, in order to maximize safety aspects it is proposed to collect all gases discharged through the outlet duct 8. The filter structure has a filter material for preventing the ejection of any particulate matter from the canister which might be entrained with the gases. Due to the gae collection system shown in FIGS. 4A and 4B the gaseous stream can be filtered and any radioactive components removed. FIG. 4A shows the loading postion. For transportation the terminal limb 11 of the outlet duct is directed upwardly to prevent damage or catching on any objects. After positioning of the canister 1 on the pressure pad 12, the limb is rotated downwardly to engage in the slotted open end of manifold 14 which together with discharge pipe 15 are fixed to the side of the pressure pad 12. Other configurations for discharge pipe connections could be utilised. Simply, reliable connections are important and one uesful alternative is to provide a V-shaped slot in opposite walls at the end of manifold 14 and to raise the manifold and orientate it so that it engages a side wall of a fixed discharge tube 11 and bridges across a portion of the side wall of the discharge tube having a gas discharge aperture.
	abstract	A nuclear reactor having a liquid metal or molten salt coolant in a riser space 130′, has a cylindrical containment vessel 134 with a reactor vessel 120′, at least two lobes 121, preferably three to nine lobes 121, each lobe 121 interconnected with the other lobe(s) and each containing a fast reactor core, 116′, 116″, 116″ and 116″″.
	summary
	summary
043483526	abstract	After their removal from the reactor, fuel element rods of nuclear reactors are stored in bundles in a water tank before they are transported away. The invention relates to a rack, to be installed in the water tank, which according to safety regulations is designed earthquake proof and receives the fuel element bundles in close packing. The rack has square receiving tubes for one fuel element bundle each; the receiving tubes are arranged vertically and connected with a bottom plate resistant to bending.
050770004	abstract	An auxiliary flexible vacuum seal for preparing a reactor coolant pump for vacuum degasification of the reactor coolant system employs a flexible boot member having a pair of longitudinally-displaced opposite open end portions and a pair of side-by-side longitudinally-extending side portions in the form of flanges defining a split in the boot member along a side thereof and extending between the open end portions for allowing flexing of the boot member between open and closed side configurations to permit its installation and removal on and from the pump. The seal also employs clamping brackets and fasteners for releasably and sealably clamping together the flanges of the boot member at the split to retain the boot member in its closed configuration. The seal further includes a pair of circumferentially-extending sealing portions formed integrally on the interior of the boot member at opposite open end portions thereof for sealably engaging the pump when the boot member is flexed to its closed configuration to thereby permit generation of a vacuum seal condition between the boot member and the pump. Preferably, a boot support member is disposed within the boot member between the boot member and the pump for supporting the boot member when in its closed configuration.
	summary
049833537	claims	1. A steam generator for receiving nonradioactive liquid sodium from a sodium cooled reactor at an intermediate heat exchanger and generating steam for producing power, said steam generator comprising: first and second upstanding cylindrical vessels; said first upstanding cylindrical vessel being exterior, and larger than said second cylindrical vessel, said first vessel closed at the top to define an inert gas plenum and closed at the bottom to define a sodium plenum; a sodium inlet defined at the top of said first vessel for supplying sodium from said intermediate heat exchanger into an interstitial volume between the inside of said first vessel and the outside of said second vessel; a sodium outlet communicated to said sodium plenum at the bottom for returning sodium to said intermediate heat exchanger in said reactor; said second cylindrical vessel being interior and smaller than said first cylindrical vessel, said second vessel open to the bottom of said first cylindrical vessel at said sodium plenum and open at the top to said inert gas plenum within said first cylindrical vessel; at least one inlet feedwater plenum at the bottom of said first cylindrical vessel communicated to the interstitial volume between said first cylindrical vessel and said second cylindrical vessel; at least one steam outlet plenum at the top of said first vessel communicated to the interstitial volume between said first cylindrical vessel and said second cylindrical vessel; a plurality of tubes communicated to said feed water inlet plenum at the bottom and said steam outlet plenum at the top, said tubes being coiled in the interstitial volume between said first and second upstanding cylindrical vessels; a standing head of sodium in the interstitial volume in said second cylindrical vessel, said standing head of sodium supported by sodium pressure at said sodium plenum; and a rupture diaphragm defined at the bottom plenum of said steam generator, said rupture diaphragm opening responsive to a sodium-water reaction on a casualty involving tube breakage; said second cylindrical vessel defining means for relieving pressure and flow from an explosive reaction at said tubes through said second cylindrical vessel for opening and rupturing said rupture diaphragm along a path independent of said tubes whereby reactives and the continuing steam and feedwater flow through the broken tubes can pass from said inert gas plenum through said volume interior of said second cylindrical vessel to said diaphragm without causing a sufficiently high pressure drop to cause the sodium/steam interface to be forced back into the intermediate heat exchanger. a sodium reactor having a core for said reactor defining a primary sodium loop and a secondary sodium loop; an intermediate heat exchanger located within said sodium reactor; first pumping means for pumping sodium in said primary sodium loop through said core through said intermediate heat exchanger and through said pump in an endless loop for carrying the heat of said core to said intermediate heat exchanger; a steam generator, said steam generator for receiving nonradioactive liquid sodium from the intermediate heat exchanger of said sodium cooled reactor and generating steam for producing power; second pumping means for pumping sodium in said secondary loop from said intermediate heat exchanger and through said steam generator; said steam generator including first and second concentric upstanding cylindrical vessels; said first upstanding cylindrical vessel being exterior, and larger than said second cylindrical vessel, said first cylindrical vessel closed at the top to define a gas plenum overlying the top of said sodium and closed at the bottom to provide a sodium plenum; a sodium inlet defined at the top of said first vessel for supplying heated sodium from the secondary loop in said sodium reactor into an interstitial volume between the inside of said first vessel and the outside of said second vessel; a sodium outlet communicated from the bottom of said first cylindrical vessel for the discharge of sodium back to said intermediate heat exchanger in said secondary sodium loop; said second cylindrical vessel being interior of and smaller than said first cylindrical vessel, said second cylindrical vessel open to the bottom of said first cylindrical vessel at said sodium plenum and defining at the top an opening to gas plenum within said first cylindrical vessel; at least one inlet feedwater plenum at the bottom of said first cylindrical vessel communicated to the interstitial volume between said first cylindrical vessel and said second cylindrical vessel; at least one steam outlet plenum at the top of said first cylindrical vessel communicated to the interstitial volume between said first cylindrical vessel and said second cylindrical vessel; a plurality of tubes communicated to said feedwater inlet plenum at the bottom of said steam outlet plenum, said tubes being coiled in the interstitial volume between said first and second upstanding cylindrical vessels; a standing head of sodium in the interior of said second upstanding cylindrical vessel, said standing head of sodium supported by sodium pressure at said lower plenum; and a rupture diaphragm defined at the bottom plenum of said steam generator means for relieving pressure and flow from an explosive reaction at said tubes through said second vessel for opening and rupturing said rupture diaphragm along a path independent of said tubes whereby reactives and the continuing steam and feedwater flow from broken tubes in said steam generator can pass through the volume interior of said second cylindrical vessel to said rupture diaphragm without causing the sodium-steam interface to be forced back into said intermediate heat exchanger of said reactor. said third upstanding cylindrical vessel defining said outlet for said sodium to said intermediate heat exchanger, said third upstanding cylindrical vessel defining in the interstices to said second cylindrical vessel a path for reactants to pass out from the top of said first upstanding cylindrical vessel through said rupture diaphragm without causing a sufficiently high pressure drop to cause the steam sodium interface to be forced back into said intermediate heat exchanger. 2. The invention of claim 1 and including a third upstanding cylindrical vessel, said third upstanding cylindrical vessel open at the bottom to receive sodium and communicated at the top to said sodium outlet at the top, said third cylindrical vessel being interior of and smaller than said second cylindrical vessel. 3. The invention of claim 2 and wherein said third cylindrical vessel includes a sodium pump, said pump disposed to pump sodium from the bottom and open end of said third cylindrical vessel to the top of cylindrical vessel for recirculation to said intermediate heat exchanger of said reactor heating sodium. 4. In combination: 5. The apparatus of claim 4 and including a third upstanding cylindrical vessel, said third upstanding cylindrical vessel open at the bottom to communicate to said sodium plenum; and, 6. The invention of claim 5 and including at least one pump disposed interior of said third upstanding cylindrical vessel for pumping liquid sodium from said lower plenum into and out of the top of said third upstanding cylindrical vessel.
	summary
	description	This application claims the benefit of U.S. Provisional Patent Application No. 60/840,135, filed Aug. 24, 2006, the disclosure of which is hereby expressly incorporated by reference. Embodiments of the present disclosure relate generally to transportation containers and assemblies and, more specifically, to transportation containers and assemblies for containing and transporting radioactive material. This summary is provided to introduce a selection of concepts in a simplified form that are further described below in the Detailed Description. This summary is not intended to identify key features of the claimed subject matter, nor is it intended to be used as an aid in determining the scope of the claimed subject matter. In accordance with embodiments of the present disclosure, a transportation assembly for transporting radioactive material is provided. The transportation assembly generally includes an outer container defining an inner cavity, the outer container having an inner shell, wherein at least a portion of the inner shell includes a plurality of layers including at least one layer of chopped fiberglass mat and at least one layer of aramid fabric. The transportation assembly further includes an inner container disposed within the inner cavity of the outer container. In accordance with other embodiments of the present disclosure, an outer container to provide protection for an inner container for transporting radioactive material is provided. The outer container generally includes first and second portions defining an inner cavity, the first and second portions both having an inner shell, wherein at least a portion of the inner shell includes a plurality of layers including at least one layer of chopped fiberglass mat and at least one layer of aramid fabric. In accordance with other embodiments of the present disclosure, an outer container to provide protection for an inner container for transporting radioactive material is provided. The outer container generally includes first and second portions coupled to one another at an interface, wherein the first and second portions define an inner cavity. The outer container further includes a closure system for securing the first and second portions to one another, wherein the closure system includes a plurality of latches and a plurality of fasteners. In accordance with other embodiments of the present disclosure, a method of transporting radioactive material is provided. The method generally includes placing an inner container into an outer container, wherein the inner container contains radioactive material. The outer container includes first and second portions defining an inner cavity, the first and second portions both having an inner shell, wherein at least a portion of the inner shell includes a plurality of layers including at least one layer of chopped fiberglass mat and at least one layer of aramid fabric. The method further includes securing the first and second portions of the outer container using a closure system, wherein the closure system includes a plurality of latches and a plurality of fasteners. Embodiments of the present disclosure are generally directed to transportation containers and assemblies for radioactive material. Referring to FIGS. 1-3, there is shown a transportation assembly, generally indicated 20, constructed in accordance with the one embodiment of the present disclosure. The assembly 20 generally includes an outer container 22 defining an inner cavity 23, and an inner container 24 disposed within the inner cavity 23 of the outer container 22 (see FIGS. 2 and 3). As will be described in detail below, the outer container 22 and the inner container 24 are cooperatively configured and arranged such that the outer container 22 provides insulation and protection to the inner container 24 during the normal conditions of transport, as well as in hypothetical accident conditions. Embodiments of the assembly 20 described herein are designed and configured for the transportation of radioactive material including fissile material in the form of dry solids, such as enriched uranium oxide. As a non-limiting example, the enriched uranium oxide may be a powder enriched to a maximum of 1.2%. In that regard, embodiments of the assembly 20 are minimally designed to protect the transport staff, other people, and the environment from the potentially hazardous material as a result of fire, submersion, impact, or damage to the assembly 20. However, it should be appreciated that embodiments of the assembly 20 described herein can also be used to transport other radioactive or nonradioactive material. Embodiments of the assembly 20 are generally designed to contain the radioactive material without release to the environment when subjected to standard crush, drop, puncture, hypothetical fire, and water immersion tests required for their certification. Further, embodiments of the assembly 20 of the present disclosure are generally sized and configured to be transportable, for example, to be carried by a suitable transportation means, such as truck or rail. However, it should be appreciated that non-portable or stationary assemblies are also within the scope of the present disclosure. While embodiments of the transportation assembly 20 described herein generally include an outer container 22 having an inner container 24 disposed within the inner cavity 23 of the outer container 22, it should be appreciated that embodiments of the present disclosure are also directed to a discrete outer container 22, i.e., without an inner container. Referring to FIG. 3, the outer container 22 will now be described in greater detail. The outer container 22 is not designed as the containment boundary for the radioactive material. Rather, it is an “overpack” device designed to protect the inner container 24 (which is designed to contain radioactive material) and reduce the severity in a hypothetical accident condition by preventing any loss of contents from the inner container 24. In the illustrated embodiment, as best seen in FIG. 3, the outer container 22 is a substantially cylindrical container having an outer wall 26 and first and second ends 28 and 30, shown as top and bottom ends 28 and 30 in FIG. 3. While the illustrated embodiment is shown as a cylindrical container, it should be appreciated that other shapes are also within the scope of the present disclosure. The outer container 22 includes two couplable portions, a first portion 32 and a second portion 34. The first portion 32 is substantially a lower portion when the outer container 22 is oriented in its upright position, as best seen in FIGS. 1 and 3. In that regard, the first portion 32 generally includes the bottom end 30 of the outer container 22 and a portion of the wall 26. The second portion 34 is substantially an upper portion when the outer container 22 is oriented in its upright position, generally including the top end 28 of the outer container 22 and a portion of the wall 26. As described in greater detail below, the lower and upper portions 32 and 34 are couplable to one another at a joint or interface 36 along the wall 26, and are securably attachable by a closure system 38 (for example, including latches 120 and fasteners 122 shown in FIGS. 6 and 7, respectively) located along the outer perimeter of the wall 26 at the interface 36. The lower and upper portions 32 and 34, when coupled together, define the inner cavity 23, which is designed and configured to receive the inner container 24. In that regard, FIG. 2 is a top view of the outer container 22, showing the inner container 24 and inner cavity 23 in phantom lines. Referring to FIG. 2, two recesses or keyways 40 in the inner cavity 23 are shown. The recess or keyways 40 in the inner cavity 23 are designed to accommodate a closure system 42 on the inner container 24, which is described in greater detail below. As best seen in FIG. 3, when in use, the inner container 24 can be received within the inner cavity 23 of the lower portion 32 of the outer container 22. The upper portion 34 of the outer container 22 is placed on top of the lower portion 32, such that the two portions 32 and 34 are coupled and in alignment at their interface 36. As best seen in FIG. 1, the outer container 22 is then secured in the closed position by its closure system 38. As mentioned above, the outer container 22 is designed to protect and insulate the inner container 24. In that regard, the ends 28 and 30 and walls 26 of each of the lower and upper portions 32 and 34 of the outer container 22 are made up of a plurality of materials, configured as layers in a sandwich lay-up, as best seen in FIG. 3. In the illustrated embodiment, each of the lower and upper portions 32 and 34 have three layers: an outer shell 50, an intermediate liner 52, and inner shell 54, each of which provide individual protective and insulative properties that make up the properties of the outer container 22 as a whole. It should be appreciated that the outer shell 50, the intermediate liner 52, and the inner shell 54 may be of different lay-up configurations, for example, the lower and upper portions 32 and 34 may have unique lay-up configurations. However, each layer will be described generally below for application in any of the outer container 22 portions, e.g., either of the lower and upper portions 32 and 34. Moreover, while in the illustrated embodiment, the outer container 22 is shown as generally having three layers, it should be appreciated that more than three layers are within the scope of the present disclosure. The outer shell 50 is designed and configured to provide a rigid, protective, external surface for the outer container 22, for example, to provide durability and prevent degradation of the outer container 22 during use. In that regard, the outer shell 50 may be configured from a weldable sheet metal, so as to provide ease of manufacturing by being weldable. As a non-limiting example, the outer shell 50 is made from 18 gauge galvanized carbon steel or stainless steel sheet metal; however, it should be appreciated that other materials, whether metal or non-metal are also within the scope of the present disclosure. It should further be appreciated that the outer shell 50 may include more than one layer of material, for example, at a particular location for additional strength or reinforcement purposes. In the illustrated embodiment, the outer shell 50 has continuous welded seams on the exterior side and stitch welding on the interior side of the lap joints and for attaching structural angles 100, 104, and 108 (described in greater detail below with reference to FIGS. 6-8). The intermediate liner 52 is designed to provide both impact and thermal protection for the material being contained within the inner container 24, and is suitably configured as a light weight material compared to the outer and inner shells 50 and 54. As such, the intermediate liner 52 may have certain density and compressive strength properties, as well as flame retardant and intumescent properties. In one embodiment, the intermediate liner 52 is formed from polyurethane foam, having a density of about 3 lb/ft3+/−15%. However, it should be appreciated that other light weight, energy-absorbing, thermal-insulative materials having similar densities and compressive strength properties are also within the scope of the present disclosure. The intermediate liner 52 may have suitable compressive strength, such that when loaded parallel-to-rise in a compression strength test, under strains of about 10%, 40%, and 70%, the intermediate liner 52 may have strain values of about +/−15% of 67, 56, and 87 psi, respectively. In addition, when loaded perpendicular-to-rise in a compression strength test, under strains of about 10%, 40%, and 70%, the intermediate liner 52 may have strain values of about +/−15% of 41, 41, and 75 psi, respectively. In one embodiment, a foam intermediate liner 52 is preferably installed such that the rise of the foam is parallel with the axial direction. In another embodiment, a liquid foam can be poured into the cavity between the inner and outer shells 54 and 50 and allowed to expand therein, completely filling the void. Regarding the flame retardant properties, the intermediate liner may have the following flame extinguishment results when subjected to a 1500° F. flame: fire extinguishment of the sample in less than about 15 seconds; flame extinguishment of any drips from the test sample in less than about 3 seconds; and an average burn length of the sample of less than about 6 inches. In addition, the intermediate liner may have an intumescence result of greater than about zero. As a non-limiting example, the foam thickness of the lower portion 32 of the outer container 22 may be in the range of about 3½ inches to about 2½ inches. It should be appreciated, however, that the foam thickness may be greater on the top and bottom ends 28 and 30 of the outer container 22 for greater impact and thermal insulation protection. In that regard, as a non-limiting example, the foam thickness of the top and bottom ends 28 and 30 of the outer container 22 may be in the range of about 5⅛ inches to about 6⅞ inches. The inner shell 54 is designed and configured to provide fire resistance or retardance, resistance to corrosion, resistance to abrasion, impact resistance, toughness, and strength to the outer container 22, during both normal conditions of transport and hypothetical accident conditions. In that regard, the inner shell 54 is suitably designed to prevent any penetration into the inner cavity of the outer container 22, for example, by fire or by any materials from the outer shell 50 or intermediate liner 52 if damage occurs to the outer container 22 as a result of, for example, crushing, dropping, or puncturing the assembly 20. A suitable inner shell 54 is flame retardant such that when subjected to a 1500° F. flame for 60 seconds, the flame extinguishment time does not exceed 30 seconds and the extinguishment time of drips from the test sample do not exceed 10 seconds. In one embodiment, the inner shell 54 includes a double bias glass fabric, for example, fabric style DBM1708, manufactured by OWENS CORNING®, which combines a glass mat and equal amounts of continuous knitted biaxial glass fiber oriented in the +45° and −45° directions into a single fabric. In another embodiment of the present disclosure, the inner shell 54 comprises a plurality of layers in a lay-up design, including at least one layer of aramid fabric, commonly known as KEVLAR® fabric, and at least one layer of chopped fiberglass. It should be appreciated that other layers may be included in the lay-up design, including, but not limited to, double bias glass fabric material, as well as multiple layers or aramid fabric, chopped fiberglass, and/or double bias glass fabric material. Aramid fabric provides strength to the inner shell 54. Double bias glass fabric provides improved tear resistance, penetration resistance, and strength to the inner shell 54. Chopped fiberglass adds spacing between the stronger double bias glass fabric and aramid layers to allow proper bonding between the layers of the lay-up and create a combination high strength, minimum weight inner shell 54. It should further be appreciated that fire retardant resins may also be added to the fabric, aramid, and fiberglass layers. In another embodiment, the inner shell 54 comprises a plurality of layers in a lay-up design, including at least one layer of double bias glass fabric material and at least one layer of aramid fabric. In yet another embodiment, the inner shell 54 comprises a plurality of layers in a lay-up design, including at least one layer of double bias glass fabric material, at least one layer of aramid fabric, and at least one layer of chopped fiberglass. It should be appreciated that the double bias glass fabric in the inner shell 54 can be oriented such the fibers run 45° offset from an axis line running along the wall 26 from the top end 28 to the bottom end 30 of the outer container 22. In addition, it should be appreciated that the aramid fabric may be oriented such that the fibers run at a different angle than the double bias glass fabric. It should be further appreciated that the inner shell 54 may further include an optional inner gel coat on the inner surfaces of the lay-ups at the top and bottom ends 28 and 30 as well as the wall 26 of the outer container 22 for an added layer of protection to the inner surfaces of the inner shell 54. As a non-limiting example, referring to FIG. 4, the inner shell 54 may include at least seven layers in a lay-up order as follows from right to left: double bias glass fabric 60, aramid fabric 62, chopped fiberglass 64, double bias glass fabric 60, aramid fabric 62, chopped fiberglass 64, and double bias glass fabric 60. An optional gel coat 66 is the eighth layer in the illustrated embodiment of FIG. 4. Such a lay-up has a thickness of about ⅛ inch. As another non-limiting example, referring to FIG. 5, the inner shell 54 includes at least ten layers in a lay-up order as follows from right to left: double bias glass fabric 60, aramid fabric 62, chopped fiberglass 64, four layers of double bias glass fabric 60, aramid fabric 62, chopped fiberglass 64, and double bias glass fabric 60. An optional gel coat 66 is the eleventh layer in the illustrated embodiment of FIG. 5. Such a lay-up has a thickness of about ¼ inch. However, it should be appreciated that any number of lay-up layers that meet the desired strength and weight properties for the inner shell 54 are within the scope of the present disclosure. As best seen in the illustrated embodiment of FIG. 3, the inner shell 54 of the upper portion 34 of the outer container 22 is a thicker lay-up, for example, a ten layer lay-up in the exemplary lay-up order described above, and the inner shell 54 of the lower portion 32 of the outer container 22 is a thinner lay-up, for example, a seven layer lay-up in the exemplary lay-up order described above. In addition to the layers, the inner shell 54 at that top and bottom ends 28 and 30 of the outer container 22 may include an optional stiffening member 56 (see FIG. 3) to stiffen the inner shell 54 and provide additional crush protection at the top and bottom ends 28 and 30 of the outer container 22. It should be appreciated that the stiffening member 56 may be sandwiched between lay-up layers to help the stiffening member 56 resist buckling and shattering under load or when subjected to dropping, crushing, or puncture forces. In one embodiment, the stiffening member 56 is a plywood sheet. It should be appreciated, however, that other stiffening materials besides plywood are also within the scope of the present disclosure, including other wood, plastic, metal, and honeycomb stiffening members. As mentioned above, the outer container 22 includes a lower portion 32 and an upper portion 34, which are couplable to one another at an interface 36. The interface 36 is suitably designed to resist spillage or leakage of any contents from the assembly 20 and also, in the case of a fire, to prevent any flames from entering the outer container 22 at the interface 36. Referring to FIGS. 3 and 6-8, the interface 36 between the lower portion 32 and the upper portion 34 is a stepped joint 36. The stepped joint 36 makes it difficult for the upper portion 34 to be removed or knocked from the lower portion 32, for example, when the outer container 22 is standing in its upright position, but not secured by its closure system 38. In addition, the stepped joint 36 reduces the risk of flame impingement into the outer container 22 at the interface 36 by blocking the direct path for a flame into the outer container 22. Briefly described, FIGS. 6-8 are partial, close-up, cross-sectional views of the interface 36 between the lower and upper portions 32 and 34 of the outer container 22, taken through three different longitudinal planes of the container. In that regard, FIG. 6 also shows a latch 120 in cross section, FIG. 7 shows a fastener 122 in cross section, and FIG. 8 shows a recess 40 in the upper portion 34 in cross section, all of which are described in greater detail below. As best seen in FIGS. 6-8, in the stepped joint 36, the lower portion 32 includes a first rim portion 80 that is couplable with a corresponding second rim portion 82 on the upper portion 34. The first rim portion 80 includes a lower annular lip 84 and an upper annular lip 86, both of which are substantially horizontally oriented when the outer container 22 is in its upright, standing position, as shown in FIGS. 1 and 3. The first rim portion 80 further includes a beveled portion 88, which extends outwardly from the lower annular lip 84 to the upper annular lip 86. The second rim portion 82 is designed to correspondingly interface with the first rim portion 80. In that regard, the second rim portion 82 also includes a lower annular lip 94 and an upper annular lip 96, both of which are substantially horizontally oriented when the outer container 22 is in its upright, standing position, as shown in FIGS. 1 and 3. The second rim portion 82 further includes a beveled portion 98, which extends inwardly from the upper annular lip 96 to the lower annular lip 94. When the lower and upper portions 32 and 34 of the outer container 22 are joined with one another at the interface 36, the beveled portions 88 and 98 of the respective first and second rim portions 80 and 82 align with one another, such that the upper annular lip 96 of the second rim portion 82 and the upper annular lip 86 of the first rim portion 80 compress a sealing element, such as a gasket 110, as seen in the illustrated embodiment of FIGS. 6-8. When aligned, the lower annular lip 94 of the second rim portion 82 is in contact with the lower annular lip 84 of the first rim portion 80. Therefore, when the outer container 22 is in its upright, standing position, as shown in FIGS. 1 and 3, the upper portion 34 of the outer container 22 is supported by the lower portion 32 along the interface 36. Referring to FIGS. 6-8, the respective inner shells 54 of the lower and upper portions 32 and 34 of the outer container 22 may extend along the inner surfaces of the lower and upper portions 32 and 34 to the first and second rim portions 80 and 82 to provide additional impact resistance, toughness, and strength reinforcement at the interface 36 between the lower and upper portions 32 and 34. In addition, the first and second rim portions 80 and 82 may further include reinforcing structural angles 100 and 104 at the interface 36 to provide improved structural integrity at the joint. The structural angles 100 and 104 add structural strength to the lower and upper portions 32 and 34 of the outer container 22 by distributing loads placed on the outer container 22. It should be appreciated that the structural angles 100 and 104 may include a plurality of discreet L-shaped structural angles positioned, for example, at the locations of the coupling devices, such as latches and fasteners 120 and 122 described below, or may include continuous angles, for example, extending along the entirety of the perimeter of the lower and upper portions 32 and 34 of the outer container 22. As best seen in FIGS. 6-8, the first rim portion 80 includes an annular structural angle 100 extending downwardly around the perimeter of the outer corner of the first rim portion 80. In that regard, the structural angle 100 has a first, substantially horizontal portion that is attached to the inner surface of the inner shell 54 of the upper annular lip 86 of the first rim portion 80 and a second, substantially vertical portion that is attached to an inner surface of the outer shell 50 of the first rim portion 80. In the illustrated embodiment, the first structural angle 100 is secured to the first rim portion 80 at the upper annular lip 86 by rivet 102. However, it should be appreciated that the structural angle 100 may be secured to the outer container 22 by any suitable attachment means, including but not limited to, one or more pins, screws, bolts, welding, adhesive, or any other suitable fastening means. Still referring to FIGS. 6-8, the second rim portion 82 also includes an annular structural angle 104 extending downwardly around the perimeter of the outer corner of the second rim portion 82. In that regard, the structural angle 104 has a first, substantially horizontal portion that is attached to the inner surface of the inner shell 54 of the upper annular lip 96 of the second rim portion 82 and a second, substantially vertical portion that is attached to an inner surface of the outer shell 50 of the second rim portion 82. In the illustrated embodiment, structural angle 104 extends from the second rim portion 82 as an downwardly depending flange to provide a cover to both the interface 36 and a portion of the first rim portion 80. Like the first structural angle 100, the second structural angle 104 may be secured to the second rim portion 82 by any suitable attachment means, including but not limited to, a rivet 102, as seen in the illustrated embodiment, one or more pins, screws, bolts, welding, adhesive, or any other suitable fastening means. Now referring to FIG. 6, the second rim portion 82 includes a third type of structural angle, a discreet L-shaped structural angle 108 to provide additional structure to the outer container 22 at the attachment point of one of the plurality of latches 120 and lift assemblies 124, as described in greater detail below. It should be appreciated that individual structural angles 108 can be used at each of the attachment points for each of the plurality of latches 120. As seen in FIG. 6, structural angle 108 extends upwardly around the perimeter of the outer corner of the second rim portion 82. In that regard, the structural angle 108 has a first, substantially horizontal portion that is attached to the inner surface of the inner shell 54 of the upper annular lip 96 of the second rim portion 82, interfacing with the substantially horizontal portion of structural angle 104. The structural angle 108 further includes a second, substantially vertical portion that is attached to an inner surface of the outer shell 50 of the second rim portion 82. Like the other structural angles 100 and 104, the third structural angle 108 may also be secured to the second rim portion 82 by any suitable attachment means, including but not limited to, rivets 102, as seen in the illustrated embodiment, one or more pins, screws, bolts, welding, adhesive, or any other suitable fastening means. Returning to FIGS. 6-8, at the interface 36 between the lower and upper portions 32 and 34 of the outer container 22, a gasket 110 is positioned to seal the interface 36, for example, to resist spillage or leakage of material being carried by the assembly 20 and to further reduce the risk of flame impingement into the outer container 22 at the interface 36. In the illustrated embodiment, the gasket 110 is positioned between the upper annular lip 86 of the first rim portion 80 and the upper annular lip 96 of the second rim portion 82. However, it should be appreciated that the gasket may be positioned in other suitable locations, for example, between the lower annular lips 84 and 94 or between the beveled portions 88 and 98 of the respective first and second rim portions 80 and 82. While the gasket 110 is suitably configured to resist spillage or leakage of material being carried by the assembly 20, the gasket 110 can be configured to allow gases to pass from the inner cavity 23 of the inner container 22 to the exterior environment and prevent over-pressurization of the inner cavity 23. For additional venting purposes, the outer container 22 may include a plurality of vents 116 on the outer surface of the outer container 22 to release any gases generated by the intermediate liner 52, for example, generated by a polyurethane foam. The gasket 110 is preferably a high temperature ceramic gasket, as a non-limiting example, heat resistant up to 2100° F. In one embodiment, the ceramic gasket is made from alumina silicate fibers formed into a yarn, which are then braided and formed into ¼ inch square braided ceramic rope encased within a 1-inch diameter braided ceramic sleeve. In one embodiment, the ceramic gasket has a silicone coating, such as a room temperature vulcanizing (RTV) silicone coating, to prevent fraying of the ceramic gasket. The silicone coating is designed so that no fibers from the ceramic gasket can enter the outer container 22 or the inner container 24 and contaminate the uranium oxide powder. As described in greater detail below, a similar gasket can also be used to seal the closure system 42 of the inner container 24. Returning to FIG. 1, the lower and upper portions 32 and 34 of the outer container 22, once coupled to one another, are securable in a closed configuration by a closure system 38 located along the outer perimeter of the outer container 22 at the interface 36 between the lower and upper portions 32 and 34. In that regard, the closure system 38 includes a plurality of latches 120 and fasteners 122, as seen in the close-up views of FIGS. 6 and 7. In the illustrated embodiment, the closure system 38 includes four heavy duty latches 120 and eight fasteners 122; however, it should be appreciated that more or less latches 120 and fasteners 122 are within the scope of the present disclosure. As best seen in FIGS. 1 and 6, the latches 120 secure the lower and upper portions 32 and 34 of the outer container 22 to one another. In one embodiment of the present disclosure, the latches 120 are high capacity, over-center locking latch devices, such as latches have a breaking strength of 4400 lbs, for example, latch 41-1292WB manufactured by Protex Fasteners Ltd. As a non-limiting example, the latches 120 and their respective catch plates may be made of steel, such as stainless steel, and may have a zinc finish. It should be appreciated that the latches 120 may include a safety catch preventing the accidental release of the latch, for example, by being locked by a sealing pin or tamper-indicating wire secured in the latch handles. It should further be appreciated that the latches may also be adjustable to provide alignment adjustment when the lower and upper portions 32 and 34 of the outer container 22 are coupled to one another. As described above, structural angles 108 or other structural components can provide structural attachment points for at least a portion of the latch 120. The latches 120 and/or any structural angles 108 proving structural support for latch attachment may be secured to the outer container 22 by any suitable attachment means, including but not limited to, one or more rivets, pins, screws, bolts, welding, adhesive, or any other suitable fastening means. In addition to the plurality of latches 120, the closure system 38 further includes a plurality of fasteners 122, including, but not limited to, screws and nuts 130 and 132, located around the exterior perimeter of the interface 36 between the lower and upper portions 32 and 34 of the outer container 22, as best seen in FIGS. 1 and 7. In the illustrated embodiment, the screws 130 enter through the outer shell 50, reinforced by structural angle 104, of the downwardly depending flange of the upper portion 34. The screws 130 engage with nuts 132 embedded in the intermediate liner 52 of the lower portion 32 of the outer container 22, also reinforced by a structural angle 100. In another embodiment, in place of nut 132, a helicoil insert and tapped bar may be used to received screws 130. These fasteners 122 provide added securement points for maintaining the integrity of the connection between the lower and upper portions 32 and 34 of the outer container 22, thus decreasing the chance that the outer container 22 will open upon impact, for example, if the assembly 20 is crushed or dropped. It should be appreciated that the screws 130 may be designed to be cold temperature fracture resistant to further prevent failure upon impact, for example, if the assembly 20 is crushed or dropped in cold temperatures. It should be appreciated that the plurality of latches 120 and fasteners 122 are suitably alternatingly oriented such that adjacent assemblies 22, when positioned along side one another for storage, can be closely packed next to one another without latches 120 of adjacent assemblies 20 aligning to interfere with one another resulting in a puncture or preventing close packing next to one another. Returning to FIG. 1, the assembly 20 also includes a plurality of lift assemblies 124 suitably located along the outer surface of the outer container 22. The lift assemblies 124 suitably include a structural tee with a hole to attach a shackle. As is well known in the art, such lift assemblies 124 can be used to lift and transport the assembly 20 when the assembly is in its upright orientation, as shown in FIGS. 1 and 3. The lift assemblies 124 and/or any structural angles 108 providing structural support for lift assembly attachment may be secured to the outer container 22 by any suitable attachment means, including but not limited to, one or more rivets, pins, screws, bolts, welding, adhesive, or any other suitable fastening means. Referring now to FIGS. 1 and 9, a forklift assembly 140 is suitably provided on the bottom end or base 30 of the outer container 22. In the illustrated embodiment, the forklift assembly 140 includes a plurality of pockets 142 designed and configured to receive forklift forks. As best seen in FIG. 9, the pockets 142 can be oriented such that the latches 120 and lift assemblies 124 on the outer container 22 are at a 45 degree angle relative to the pockets 142 to facilitate close stacking of adjacent assemblies 20 and prevent possible punctures to adjacent assemblies. The forklift assembly 140 also provides additional structural support to the outer container 22 for damage resistance when the assembly 20 is either crushed or dropped. In that regard, the forklift assembly 140 is designed to be crush absorbing. For example, in one embodiment, the forklift pockets 142 are configured from folded 12 gauge galvanized carbon steel or stainless steel sheet, with bracing from 14 gauge galvanized carbon steel or stainless steel sheet. The forklift assembly 140 is therefore configured to collapse when the assembly 20 is crushed or dropped to absorb the impact of the crush or drop forces. Returning to FIG. 3, the inner container 24 of the assembly will now be described in greater detail. The inner container 24 is designed and configured to support and contain radioactive material. In that regard, the inner container 24 includes a body portion 150, a bottom portion 152, and a lid 154. In one embodiment, the inner container 24 is a 55-gallon rolled steel cylindrical drum having a single welded seam, a closed bottom end, and an open top end, closeable by a lid. However, it should be appreciated that the inner container may be any suitable design or configuration so as to be cooperatively received within the inner cavity 23 of the outer container 22. The inner container 24 may be made from any suitable materials to provide strength and resist leakage or spillage of the contained material into the inner cavity 23 of the outer container 22. While it should be appreciated that other materials are within the scope of the present disclosure, in one embodiment, the inner container 24 is made from 16 gauge carbon steel, stainless steel, or an equivalent material. In yet another embodiment, the inner container 24 has 7A Type A and UN specification ratings. The lid 154 of the inner container 24 is designed to be received at an upper rim 156 of the body portion 150 of the inner container 24. The lid 154 is designed to be removable to receive or remove the contained material. When closed, the lid 154 includes a reinforced closure system 42 to ensure containment of the radioactive material, particularly when the assembly 20 is subjected to normal conditions of transport and hypothetical accident conditions, for example, immersion in water. In the illustrated embodiment, the closure system 42 includes a reinforced closure ring 158 having a flange 160 that is attachable to the upper rim 156 and body portion 150 of the inner container 24, for example, a clamshell closure as described in U.S. Patent Application Publication No. U.S. 2005/0269331 A1, published on Dec. 8, 2005, the disclosure of which is hereby incorporated by reference. The clamshell closure is generally a modified two-piece C-ring including a two-bolt closure system. The clamshell closure system 42 may further include a gasket (not shown) between the lid 154 and the upper rim 156 of the inner container 22 to seal the closure, for example, to resist spillage or leakage of material being carried by the inner container 24 and to further reduce the risk of flame impingement into the inner container 24 at the lid 154. It should be appreciated that the gasket may be a ceramic gasket, for example, similar to ceramic gasket 110 described above, and may have an optional silicone coating. As mentioned above, and as best seen in FIG. 2 showing the top view of the assembly 20, the upper portion 34 of the outer container 22 includes recesses or keyways 40 in the inner cavity 23 designed to accommodate the bolts of the two-bolt closure system 42 on the inner container 24, for example, the two-bolt closure system used to secure the clamshell closure described above. While illustrative embodiments have been illustrated and described, it will be appreciated that various changes can be made therein without departing from the spirit and scope of the disclosure.
041939530	summary	This invention concerns the manufacture of fuel and/or breeder kernels for fuel elements of nuclear reactors. In such manufactures, drops of a breeder or fuel and breeder-containing hydrosol are caused to fall through a gas phase containing gaseous ammonia into a precipitation bath containing ammonium hydroxide, the hydrosol being injected substantially horizontally into the gas phase above the precipitation bath in droplet form, so that it enters into the precipitation bath under the influence of gravity. In the fuel elements of nuclear reactors, the fissionable material uranium and the breeder material thorium is provided in the form of spherical grains or kernels having a diameter of from 0.2 to 0.6 mm. For high temperature reactors, breeder cores of thorium oxide, ThO.sub.2, or mixed oxide cores containing thorium and uranium (Th,U)O.sub.2, are used. For the manufacture of fuel and/or breeder kernels, a series of processes are known. The wet chemical processes have succeeded in getting into use because of their superior product quality. In these processes a water solution or an aqueous collodial suspension designated as a hydrosol, in whicn the fuel and/or breeder material is contained is dispersed into individual drops. These drops are then solidified by a chemical reaction. The greenware particles thus produced--if necessary or desirable after a washing step--are dried and then sintered to produce ceramic kernels or granules. In this manner, for example, processes have been described in which a partially neutralized solution of thorium nitrate or of thorium uranyl nitrate is dripped into a water solution of ammonia (Energia Nucleare, 1970, pp. 217-224; Kerntechnik, 1970, pp. 159-164). In these processes the drops are solidified by a gelification reaction. In order that the drops should survive the impact onto the surface of the ammonia solution without distortion, two precautions have been regarded as necessary. In the first place, the drops must react with gaseous ammonia before falling into the ammonia solution in order that at least a surface hardening of the drops may be obtained. Furthermore, the addition of substantial quantities of a polymeric water soluble material has been required in order to increase the viscosity of the drops. As a thickener for this purpose, methyl cellulose and polyvinyl alcohol have been found effective. The addition of these organic polymers as thickeners, however, brings about substantial disadvantages for the carrying out of the process of which a few of the more important are named in the following discussion. The concentration of Th(NO.sub.3).sub.4 an UO.sub.2 (NO.sub.3).sub.2 in the solution is limited a relatively small value (.ltoreq.0.7 mole/l) by the addition of the polymers, since the solution of the required quantity of the thickener is no longer possible at higher concentrations of the nitrates. This limits the heavy metal throughput in the pouring (casting) apparatus since at lower concentrations it is necessary to produce larger drop diameters. Furthermore, the thickeners cannot be washed out of the greenware spheres by water or solvents. After they are dried, a calcination stage is necessary, in which the organic material is decomposed and burned up. This reaction is exothermic and damage or destruction of the kernels can be prevented only with slow progress of the reaction and careful control of the reaction conditions. In addition, it is important to wash the greenware cores previous to the calcination step with a lower alcohol, for example, isopropanol. Finally, the macromolecules of the thickener are partly destroyed by the high radioactivity of these solutions upon their addition to solutions that are produced in the treatment of nuclear fuel materials. The viscosity of the solutions then sinks. Since a constant viscosity of the pouring solution is nevertheless to be desired for the process here in question for the manufacture of fuel and/or breeder materials, the processes above mentioned for the reprocessing of fuel and/or breeder materials is feasible only with great additional expense. In order to avoid the disadvantages mentioned above, processes have also been proposed that operate without the addition of a thickener. Thus, for example, a process is known from German Pat. No. 21 47 472 in which the drops are formed and prehardened in a liquid ketone phase that has a low ammonia concentration, before the hardening is completed in an aqueous ammonia solution located beneath the ketone phase. In this process, the nozzle for the generation of drops must be immersed in the ketone phase, which naturally leads to a lower drop frequency than in spraying into a space filled with gas. In order to prevent the stopping of the jet orifice, it is necessary to use a dual-flow nozzle, in which an outer surrounding stream of ammonia free ketone prevents the contact of the ammonia-containing ketone with the heavy metal solution or sol. When a ketone flowing out of the drop-pouring column is used as the surrounding stream, this material must be suitably prepared. SUMMARY OF THE INVENTION It is an object of the present invention to provide a process for manufacturing spherical fuel and/or breeder kernels or granules in which a hydrosol can be dripped into an aqueous ammonia solution at high drop frequency without the addition of a viscosity-increasing macromolecular substance. In addition, the economy of the manufacturing process should also be increased with this improvement. The starting point of the invention is an unpublished suggestion that a water solution or a hydrosol in which a heavy metal is dispersed should be injected into a gas-filled space above a liquid aqua ammonia phase at an angle of 90.degree. to the normal to the liquid surface of the precipitation bath, thereby providing a favorable relation between the prehardening of the drops in the gas phase and the impact velocity. Briefly, in a process of the general kind above described, a hydrosol containing thorium oxide in addition to thorium oxide, uranium (VI) oxide with the content of hexavalent uranium constituting up to 25% by weight of the aggregate heavy metal content, such that the heavy metal is contained in the hydrosol at a concentration between 1.5 and 3 moles per liter, is used and the precipitation bath is adjusted at a pH value of between 8 and 9. With horizontal injection of the drops and increase of the heavy metal concentration, the result is obtained that the hydrosol drops can be cast round without utilization of a thickener. As a result of the high heavy metal concentration of the hydrosols, the drops solidify in the gas phase substantially more strongly than in the case of hydrosols of lower concentration. Yet with the compositions of the precipitation bath conventional up to now, the drops falling into the bath would disintegrate upon increase of the heavy metal concentration. Surprisingly, it has been found that in combination with the previously described drop-casting conditions, the provision of a pH value in the precipitation bath in the range of from 8 to 9 prevents the disintegration of the drops, the pH value to be provided in the individual case being dependent substantially upon the parameters of heavy metal content, uranium content, and drop size. In accordance with a further development of the invention, it is useful to maintain the ammonium nitrate concentration at at least three moles liter in order to maintain the pH within the desired value range. The pH value is then determined by the buffering system NH.sub.4.sup.+ /NH.sub.3. Preferably, the precipitation bath has an aqua ammonia concentration between 0.5 and 3% by weight. The hydrosols and the gels produced therefrom, of course, have high ammonium nitrate concentrations as a result of the precipitation reaction, according to the equation EQU Th(NO.sub.3).sub.4 +4NH.sub.4 OH.fwdarw.Th(OH).sub.4 +4NH.sub.4 NO.sub.3, with the ammonium nitrate being in part washed out in the precipitation bath. It is, however, necessary to provide adjustment of the concentration value in the startup phase by the addition of ammonium nitrate or by substitution, for the ammonia-containing liquid bath, of a corresponding process step. The ammonium nitrate concentration can preferably be measured by measuring the electrical conductivity of the precipitation bath, and regulated in dependence thereon. In development of the invention, it is further contemplated that the hydrosol should be dripped at temperature of between 30.degree. C. and 60.degree. C. The dripping of a tempered hydrosol increases the yield of fuel and/or breeder cores having a quality satisfying strict requirements. In processes according to the invention, it is possible not only to utilize without difficulty highly concentrated hydrosols in a dripping process but also to dry and sinter immediately thereafter the fuel and/or breeder material grains without problems. Excess ammonium nitrate that adheres to the gel spheres taken from the precipitation bath are washed out in a water bath in which wash water containing about 1% by weight of ammonium hydroxide and a small quantity of surfactant is added. The drying of the gel spheres is then performed in air with a sufficient water vapor content, a process step already known from German published patent application (AS) 23 23 010. The process according to the present invention is particularly distinguished by uniformly high quality of the fuel and/or breeder kernels or granules produced. Yields greater than 99% can be obtained. An apparatus for carrying out the process of the invention is shown in the not previously known German patent application P 27 14 873.8-33 owned by the assignee of the present invention, which will be published in accordance with the German patent application procedure in due course.
	description	This application claims priority under 35 U.S.C. § 119 to U.S. Provisional Patent Application Ser. No. 62/515,050, filed on Jun. 5, 2017, and entitled “STORING HAZARDOUS MATERIAL IN A SUBTERRANEAN FORMATION,” the entire contents of which are incorporated by reference herein. This disclosure relates to storing hazardous material in a subterranean formation and, more particularly, storing spent nuclear fuel in a subterranean formation. Hazardous waste is often placed in long-term, permanent, or semi-permanent storage so as to prevent health issues among a population living near the stored waste. Such hazardous waste storage is often challenging, for example, in terms of storage location identification and surety of containment. For instance, the safe storage of nuclear waste (e.g., spent nuclear fuel, whether from commercial power reactors, test reactors, or even high-grade military waste) is considered to be one of the outstanding challenges of energy technology. Safe storage of the long-lived radioactive waste is a major impediment to the adoption of nuclear power in the United States and around the world. Conventional waste storage methods have emphasized the use of tunnels, and is exemplified by the design of the Yucca Mountain storage facility. Other techniques include boreholes, including vertical boreholes, drilled into crystalline basement rock. Other conventional techniques include forming a tunnel with boreholes emanating from the walls of the tunnel in shallow formations to allow human access. In a general implementation, a hazardous material storage repository includes a drillhole extending into the Earth and including an entry at least proximate a terranean surface, the drillhole including a substantially vertical drillhole portion, a transition drillhole portion coupled to the substantially vertical drillhole portion, and a hazardous material storage drillhole portion coupled to the transition drillhole portion, at least one of the transition drillhole portion or the hazardous material storage drillhole portion including an isolation drillhole portion that is directed vertically toward the terranean surface and away from an intersection between the substantially vertical drillhole portion and the transition drillhole portion; a storage canister positioned in the hazardous material storage drillhole portion, the storage canister sized to fit from the drillhole entry through the substantially vertical drillhole portion, the transition drillhole portion, and into the hazardous material storage drillhole portion of the drillhole, the storage canister including an inner cavity sized enclose hazardous material; and a seal positioned in the drillhole, the seal isolating the hazardous material storage drillhole portion of the drillhole from the entry of the drillhole. In an aspect combinable with the general implementation, the isolation drillhole portion includes a vertically inclined drillhole portion that includes a proximate end coupled to the transition drillhole portion at a first depth and a distal end opposite the proximate end at a second depth shallower than the first depth. In another aspect combinable with any of the previous aspects, the vertically inclined drillhole portion includes the hazardous material storage drillhole portion. In another aspect combinable with any of the previous aspects, an inclination angle of the vertically inclined drillhole portion is determined based at least in part on a distance associated with a disturbed zone of a geologic formation that surrounds the vertically inclined drillhole portion and a length of a distance tangent to a lowest portion of the storage canister and the substantially vertical drillhole portion. In another aspect combinable with any of the previous aspects, the distance associated with the disturbed zone of the geologic formation includes a distance between an outer circumference of the disturbed zone and a radial centerline of the vertically inclined drillhole portion. In another aspect combinable with any of the previous aspects, the inclination angle is about 3 degrees. In another aspect combinable with any of the previous aspects, the isolation drillhole portion includes a J-section drillhole portion coupled between the substantially vertical drillhole portion and the hazardous material storage drillhole portion. In another aspect combinable with any of the previous aspects, the J-section drillhole portion includes the transition drillhole portion. In another aspect combinable with any of the previous aspects, the hazardous material storage drillhole portion includes at least one of a substantially horizontal drillhole portion or a vertically inclined drillhole portion. In another aspect combinable with any of the previous aspects, the isolation drillhole portion includes a vertically undulating drillhole portion coupled to the transition drillhole portion. In another aspect combinable with any of the previous aspects, the transition drillhole portion includes a curved drillhole portion between the substantially vertical drillhole portion and the vertically undulating drillhole portion. In another aspect combinable with any of the previous aspects, the hazardous material storage drillhole portion is located within or below a barrier layer that includes at least one of a shale formation layer, a salt formation layer, or other impermeable formation layer. In another aspect combinable with any of the previous aspects, the hazardous material storage drillhole portion is vertically isolated, by the barrier layer, from a subterranean zone that includes mobile water. In another aspect combinable with any of the previous aspects, the hazardous material storage drillhole portion is formed below the barrier layer and is vertically isolated from the subterranean zone that includes mobile water by the barrier layer. In another aspect combinable with any of the previous aspects, the hazardous material storage drillhole portion is formed within the barrier layer, and is vertically isolated from the subterranean zone that includes mobile water by at least a portion of the barrier layer. In another aspect combinable with any of the previous aspects, the barrier layer includes a permeability of less than about 0.01 millidarcys. In another aspect combinable with any of the previous aspects, the barrier layer includes a brittleness of less than about 10 MPa, where brittleness includes a ratio of compressive stress of the barrier layer to tensile strength of the barrier layer. In another aspect combinable with any of the previous aspects, the barrier layer includes a thickness proximate the hazardous material storage drillhole portion of at least about 100 feet. In another aspect combinable with any of the previous aspects, the barrier layer includes a thickness proximate the hazardous material storage drillhole portion that inhibits diffusion of the hazardous material that escapes the storage canister through the barrier layer for an amount of time that is based on a half-life of the hazardous material. In another aspect combinable with any of the previous aspects, the barrier layer includes about 20 to 30% weight by volume of clay or organic matter. In another aspect combinable with any of the previous aspects, the barrier layer includes an impermeable layer. In another aspect combinable with any of the previous aspects, the barrier layer includes a leakage barrier defined by a time constant for leakage of the hazardous material of 10,000 years or more. In another aspect combinable with any of the previous aspects, the barrier layer includes a hydrocarbon or carbon dioxide bearing formation. In another aspect combinable with any of the previous aspects, the hazardous material includes spent nuclear fuel. Another aspect combinable with any of the previous aspects further includes at least one casing assembly that extends from at or proximate the terranean surface, through the drillhole, and into the hazardous material storage drillhole portion. In another aspect combinable with any of the previous aspects, the storage canister includes a connecting portion configured to couple to at least one of a downhole tool string or another storage canister. In another aspect combinable with any of the previous aspects, the isolation drillhole portion includes a spiral drillhole. In another aspect combinable with any of the previous aspects, the isolation drillhole portion has a specified geometry independent of a stress state of a rock formation into which the isolation drillhole portion is formed. In another general implementation, a method for storing hazardous material includes moving a storage canister through an entry of a drillhole that extends into a terranean surface, the entry at least proximate the terranean surface, the storage canister including an inner cavity sized enclose hazardous material; moving the storage canister through the drillhole that includes a substantially vertical drillhole portion, a transition drillhole portion coupled to the substantially vertical drillhole portion, and a hazardous material storage drillhole portion coupled to the transition drillhole portion, at least one of the transition drillhole portion or the hazardous material storage drillhole portion including an isolation drillhole portion that is directed vertically toward the terranean surface and away from an intersection between the substantially vertical drillhole portion and the transition drillhole portion; moving the storage canister into the hazardous material storage drillhole portion; and forming a seal in the drillhole that isolates the storage portion of the drillhole from the entry of the drillhole. In an aspect combinable with the general implementation, the isolation drillhole portion includes a vertically inclined drillhole portion that includes a proximate end coupled to the transition drillhole portion at a first depth and a distal end opposite the proximate end at a second depth shallower than the first depth. In another aspect combinable with any of the previous aspects, the vertically inclined drillhole portion includes the hazardous material storage drillhole portion. In another aspect combinable with any of the previous aspects, an inclination angle of the vertically inclined drillhole portion is determined based at least in part on a distance associated with a disturbed zone of a geologic formation that surrounds the vertically inclined drillhole portion and a length of a distance tangent to a lowest portion of the storage canister and the substantially vertical drillhole portion. In another aspect combinable with any of the previous aspects, the distance associated with the disturbed zone of the geologic formation includes a distance between an outer circumference of the disturbed zone and a radial centerline of the vertically inclined drillhole portion. In another aspect combinable with any of the previous aspects, the inclination angle is about 3 degrees. In another aspect combinable with any of the previous aspects, the isolation drillhole portion includes a J-section drillhole portion coupled between the substantially vertical drillhole portion and the hazardous material storage drillhole portion. In another aspect combinable with any of the previous aspects, the J-section drillhole portion includes the transition drillhole portion. In another aspect combinable with any of the previous aspects, the hazardous material storage drillhole portion includes at least one of a substantially horizontal drillhole portion or a vertically inclined drillhole portion. In another aspect combinable with any of the previous aspects, the isolation drillhole portion includes a vertically undulating drillhole portion coupled to the transition drillhole portion. In another aspect combinable with any of the previous aspects, the transition drillhole portion includes a curved drillhole portion between the substantially vertical drillhole portion and the vertically undulating drillhole portion. In another aspect combinable with any of the previous aspects, the hazardous material storage drillhole portion is located within or below a barrier layer that includes at least one of a shale formation layer, a salt formation layer, or other impermeable formation layer. In another aspect combinable with any of the previous aspects, the hazardous material storage drillhole portion is vertically isolated, by the barrier layer, from a subterranean zone that includes mobile water. In another aspect combinable with any of the previous aspects, the hazardous material storage drillhole portion is formed below the barrier layer and is vertically isolated from the subterranean zone that includes mobile water by the barrier layer. In another aspect combinable with any of the previous aspects, the hazardous material storage drillhole portion is formed within the barrier layer, and is vertically isolated from the subterranean zone that includes mobile water by at least a portion of the barrier layer. In another aspect combinable with any of the previous aspects, the barrier layer includes a permeability of less than about 0.01 millidarcys. In another aspect combinable with any of the previous aspects, the barrier layer includes a brittleness of less than about 10 MPa, where brittleness includes a ratio of compressive stress of the barrier layer to tensile strength of the barrier layer. In another aspect combinable with any of the previous aspects, the barrier layer includes a thickness proximate the hazardous material storage drillhole portion of at least about 100 feet. In another aspect combinable with any of the previous aspects, the barrier layer includes a thickness proximate the hazardous material storage drillhole portion that inhibits diffusion of the hazardous material that escapes the storage canister through the barrier layer for an amount of time that is based on a half-life of the hazardous material. In another aspect combinable with any of the previous aspects, the barrier layer includes about 20 to 30% weight by volume of clay or organic matter. In another aspect combinable with any of the previous aspects, the barrier layer includes an impermeable layer. In another aspect combinable with any of the previous aspects, the barrier layer includes a leakage barrier defined by a time constant for leakage of the hazardous material of 10,000 years or more. In another aspect combinable with any of the previous aspects, the barrier layer includes a hydrocarbon or carbon dioxide bearing formation. In another aspect combinable with any of the previous aspects, the hazardous material includes spent nuclear fuel. Another aspect combinable with any of the previous aspects further includes at least one casing assembly that extends from at or proximate the terranean surface, through the drillhole, and into the hazardous material storage drillhole portion. In another aspect combinable with any of the previous aspects, the storage canister includes a connecting portion configured to couple to at least one of a downhole tool string or another storage canister. Another aspect combinable with any of the previous aspects further includes prior to moving the storage canister through the entry of the drillhole that extends into the terranean surface, forming the drillhole from the terranean surface to a subterranean formation. Another aspect combinable with any of the previous aspects further includes installing a casing in the drillhole that extends from at or proximate the terranean surface, through the drillhole, and into the hazardous material storage drillhole portion. Another aspect combinable with any of the previous aspects further includes cementing the casing to the drillhole. Another aspect combinable with any of the previous aspects further includes, subsequent to forming the drillhole, producing hydrocarbon fluid from the subterranean formation, through the drillhole, and to the terranean surface. Another aspect combinable with any of the previous aspects further includes removing the seal from the drillhole; and retrieving the storage canister from the hazardous material storage drillhole portion to the terranean surface. Another aspect combinable with any of the previous aspects further includes monitoring at least one variable associated with the storage canister from a sensor positioned proximate the hazardous material storage drillhole portion; and recording the monitored variable at the terranean surface. In another aspect combinable with any of the previous aspects, the monitored variable includes at least one of radiation level, temperature, pressure, presence of oxygen, presence of water vapor, presence of liquid water, acidity, or seismic activity. Another aspect combinable with any of the previous aspects further includes based on the monitored variable exceeding a threshold value removing the seal from the drillhole; and retrieving the storage canister from the hazardous material storage drillhole portion to the terranean surface. In another general implementation, a method for storing hazardous material includes moving a storage canister through an entry of a drillhole that extends into a terranean surface, the entry at least proximate the terranean surface, the storage canister including an inner cavity sized enclose hazardous material; moving the storage canister through the drillhole that includes a substantially vertical drillhole portion, a transition drillhole portion coupled to the substantially vertical drillhole portion, and a hazardous material storage drillhole portion coupled to the transition drillhole portion, the hazardous material storage drillhole portion located below a self-healing geological formation, the hazardous material storage drillhole portion vertically isolated, by the self-healing geological formation, from a subterranean zone that includes mobile water; moving the storage canister into the hazardous material storage drillhole portion; and forming a seal in the drillhole that isolates the storage portion of the drillhole from the entry of the drillhole. In an aspect combinable with the general implementation, the self-healing geologic formation includes at least one of shale, salt, clay, or dolomite. In another general implementation, a hazardous material storage repository includes a drillhole extending into the Earth and including an entry at least proximate a terranean surface, the drillhole including a substantially vertical drillhole portion, a transition drillhole portion coupled to the substantially vertical drillhole portion, and a hazardous material storage drillhole portion coupled to the transition drillhole portion, the hazardous material storage drillhole portion located below a self-healing geological formation, the hazardous material storage drillhole portion vertically isolated, by the self-healing geological formation, from a subterranean zone that includes mobile water; a storage canister positioned in the hazardous material storage drillhole portion, the storage canister sized to fit from the drillhole entry through the substantially vertical drillhole portion, the transition drillhole portion, and into the hazardous material storage drillhole portion of the drillhole, the storage canister including an inner cavity sized enclose hazardous material; and a seal positioned in the drillhole, the seal isolating the hazardous material storage drillhole portion of the drillhole from the entry of the drillhole. In an aspect combinable with the general implementation, the self-healing geologic formation includes at least one of shale, salt, clay, or dolomite. Implementations of a hazardous material storage repository according to the present disclosure may include one or more of the following features. For example, a hazardous material storage repository according to the present disclosure may allow for multiple levels of containment of hazardous material within a storage repository located thousands of feet underground, decoupled from any nearby mobile water. A hazardous material storage repository according to the present disclosure may also use proven techniques (e.g., drilling) to create or form a storage area for the hazardous material, in a subterranean zone proven to have fluidly sealed hydrocarbons therein for millions of years. As another example, a hazardous material storage repository according to the present disclosure may provide long-term (e.g., thousands of years) storage for hazardous material (e.g., radioactive waste) in a shale formation that has geologic properties suitable for such storage, including low permeability, thickness, and ductility, among others. In addition, a greater volume of hazardous material may be stored at low cost—relative to conventional storage techniques—due in part to directional drilling techniques that facilitate long horizontal boreholes, often exceeding a mile in length. In addition, rock formations that have geologic properties suitable for such storage may be found in close proximity to sites at which hazardous material may be found or generated, thereby reducing dangers associated with transporting such hazardous material. Implementations of a hazardous material storage repository according to the present disclosure may also include one or more of the following features. Large storage volumes, in turn, allow for the storage of hazardous materials to be emplaced without a need for complex prior treatment, such as concentration or transfer to different forms or canisters. As a further example, in the case of nuclear waste material from a reactor for instance, the waste can be kept in its original pellets, unmodified, or in its original fuel rods, or in its original fuel assemblies, which contain dozens of fuel rods. In another aspect, the hazardous material may be kept in an original holder but a cement or other material is injected into the holder to fill the gaps between the hazardous materials and the structure. For example, if the hazardous material is stored in fuel rods which are, in turn, stored in fuel assemblies, then the spaces between the rods (typically filled with water when inside a nuclear reactor) could be filled with cement or other material to provide yet an additional layer of isolation from the outside world. As yet a further example, secure and low cost storage of hazardous material is facilitated while still permitting retrieval of such material if circumstances deem it advantageous to recover the stored materials. The details of one or more implementations of the subject matter described in this disclosure are set forth in the accompanying drawings and the description below. Other features, aspects, and advantages of the subject matter will become apparent from the description, the drawings, and the claims. FIG. 1A is a schematic illustration of example implementations of a hazardous material storage repository system, e.g., a subterranean location for the long-term (e.g., tens, hundreds, or thousands of years or more) but retrievable safe and secure storage of hazardous material, during a deposit or retrieval operation according to the present disclosure. For example, turning to FIG. 1A, this figure illustrates an example hazardous material storage repository system 100 during a deposit (or retrieval, as described below) process, e.g., during deployment of one or more canisters of hazardous material in a subterranean formation. As illustrated, the hazardous material storage repository system 100 includes a drillhole 104 formed (e.g., drilled or otherwise) from a terranean surface 102 and through multiple subterranean layers 112, 114, 116, and 132. Although the terranean surface 102 is illustrated as a land surface, terranean surface 102 may be a sub-sea or other underwater surface, such as a lake or an ocean floor or other surface under a body of water. Thus, the present disclosure contemplates that the drillhole 104 may be formed under a body of water from a drilling location on or proximate the body of water. The illustrated drillhole 104 is a directional drillhole in this example of hazardous material storage repository system 100. For instance, the drillhole 104 includes a substantially vertical portion 106 coupled to a radiussed or curved portion 108, which in turn is coupled to an inclined portion 110. As used in the present disclosure, “substantially” in the context of a drillhole orientation, refers to drillholes that may not be exactly vertical (e.g., exactly perpendicular to the terranean surface 102) or exactly horizontal (e.g., exactly parallel to the terranean surface 102), or exactly inclined at a particular incline angle relative to the terranean surface 102. In other words, vertical drillholes often undulate offset from a true vertical direction, that they might be drilled at an angle that deviates from true vertical, and inclined drillholes often undulate offset from a true incline angle. Further, in some aspects, an inclined drillhole may not have or exhibit an exactly uniform incline (e.g., in degrees) over a length of the drillhole. Instead, the incline of the drillhole may vary over its length (e.g., by 1-5 degrees). As illustrated in this example, the three portions of the drillhole 104—the vertical portion 106, the radiussed portion 108, and the inclined portion 110—form a continuous drillhole 104 that extends into the Earth. The illustrated drillhole 104, in this example, has a surface casing 120 positioned and set around the drillhole 104 from the terranean surface 102 into a particular depth in the Earth. For example, the surface casing 120 may be a relatively large-diameter tubular member (or string of members) set (e.g., cemented) around the drillhole 104 in a shallow formation. As used herein, “tubular” may refer to a member that has a circular cross-section, elliptical cross-section, or other shaped cross-section. For example, in this implementation of the hazardous material storage repository system 100, the surface casing 120 extends from the terranean surface through a surface layer 112. The surface layer 112, in this example, is a geologic layer comprised of one or more layered rock formations. In some aspects, the surface layer 112 in this example may or may not include freshwater aquifers, salt water or brine sources, or other sources of mobile water (e.g., water that moves through a geologic formation). In some aspects, the surface casing 120 may isolate the drillhole 104 from such mobile water, and may also provide a hanging location for other casing strings to be installed in the drillhole 104. Further, although not shown, a conductor casing may be set above the surface casing 120 (e.g., between the surface casing 120 and the surface 102 and within the surface layer 112) to prevent drilling fluids from escaping into the surface layer 112. As illustrated, a production casing 122 is positioned and set within the drillhole 104 downhole of the surface casing 120. Although termed a “production” casing, in this example, the casing 122 may or may not have been subject to hydrocarbon production operations. Thus, the casing 122 refers to and includes any form of tubular member that is set (e.g., cemented) in the drillhole 104 downhole of the surface casing 120. In some examples of the hazardous material storage repository system 100, the production casing 122 may begin at an end of the radiussed portion 108 and extend throughout the inclined portion 110. The casing 122 could also extend into the radiussed portion 108 and into the vertical portion 106. As shown, cement 130 is positioned (e.g., pumped) around the casings 120 and 122 in an annulus between the casings 120 and 122 and the drillhole 104. The cement 130, for example, may secure the casings 120 and 122 (and any other casings or liners of the drillhole 104) through the subterranean layers under the terranean surface 102. In some aspects, the cement 130 may be installed along the entire length of the casings (e.g., casings 120 and 122 and any other casings), or the cement 130 could be used along certain portions of the casings if adequate for a particular drillhole 102. The cement 130 can also provide an additional layer of confinement for the hazardous material in canisters 126. The drillhole 104 and associated casings 120 and 122 may be formed with various example dimensions and at various example depths (e.g., true vertical depth, or TVD). For instance, a conductor casing (not shown) may extend down to about 120 feet TVD, with a diameter of between about 28 in. and 60 in. The surface casing 120 may extend down to about 2500 feet TVD, with a diameter of between about 22 in. and 48 in. An intermediate casing (not shown) between the surface casing 120 and production casing 122 may extend down to about 8000 feet TVD, with a diameter of between about 16 in. and 36 in. The production casing 122 may extend inclinedly (e.g., to case the inclined portion 110) with a diameter of between about 11 in. and 22 in. The foregoing dimensions are merely provided as examples and other dimensions (e.g., diameters, TVDs, lengths) are contemplated by the present disclosure. For example, diameters and TVDs may depend on the particular geological composition of one or more of the multiple subterranean layers (112, 114, 116, and 132), particular drilling techniques, as well as a size, shape, or design of a hazardous material canister 126 that contains hazardous material to be deposited in the hazardous material storage repository system 100. In some alternative examples, the production casing 122 (or other casing in the drillhole 104) could be circular in cross-section, elliptical in cross-section, or some other shape. As illustrated, the vertical portion 106 of the drillhole 104 extends through subterranean layers 112, 114, 116, and 132, and, in this example, lands in a subterranean layer 119. As discussed above, the surface layer 112 may or may not include mobile water. Subterranean layer 114, which is below the surface layer 112, in this example, is a mobile water layer 114. For instance, mobile water layer 114 may include one or more sources of mobile water, such as freshwater aquifers, salt water or brine, or other source of mobile water. In this example of hazardous material storage repository system 100, mobile water may be water that moves through a subterranean layer based on a pressure differential across all or a part of the subterranean layer. For example, the mobile water layer 114 may be a permeable geologic formation in which water freely moves (e.g., due to pressure differences or otherwise) within the layer 114. In some aspects, the mobile water layer 114 may be a primary source of human-consumable water in a particular geographic area. Examples of rock formations of which the mobile water layer 114 may be composed include porous sandstones and limestones, among other formations. Other illustrated layers, such as the impermeable layer 116 and the storage layer 119, may include immobile water. Immobile water, in some aspects, is water (e.g., fresh, salt, brine), that is not fit for human or animal consumption, or both. Immobile water, in some aspects, may be water that, by its motion through the layers 116 or 119 (or both), cannot reach the mobile water layer 114, terranean surface 102, or both, within 10,000 years or more (such as to 1,000,000 years). Below the mobile water layer 114, in this example implementation of hazardous material storage repository system 100, is an impermeable layer 116. The impermeable layer 116, in this example, may not allow mobile water to pass through. Thus, relative to the mobile water layer 114, the impermeable layer 116 may have low permeability, e.g., on the order of nanodarcy permeability. Additionally, in this example, the impermeable layer 116 may be a relatively non-ductile (i.e., brittle) geologic formation. One measure of non-ductility is brittleness, which is the ratio of compressive stress to tensile strength. In some examples, the brittleness of the impermeable layer 116 may be between about 20 MPa and 40 MPa. As shown in this example, the impermeable layer 116 is shallower (e.g., closer to the terranean surface 102) than the storage layer 119. In this example rock formations of which the impermeable layer 116 may be composed include, for example, certain kinds of sandstone, mudstone, clay, and slate that exhibit permeability and brittleness properties as described above. In alternative examples, the impermeable layer 116 may be deeper (e.g., further from the terranean surface 102) than the storage layer 119. In such alternative examples, the impermeable layer 116 may be composed of an igneous rock, such as granite. Below the impermeable layer 116 is the storage layer 119. The storage layer 119, in this example, may be chosen as the landing for the inclined portion 110, which stores the hazardous material, for several reasons. Relative to the impermeable layer 116 or other layers, the storage layer 119 may be thick, e.g., between about 100 and 200 feet of total vertical thickness. Thickness of the storage layer 119 may allow for easier landing and directional drilling, thereby allowing the inclined portion 110 to be readily emplaced within the storage layer 119 during constructions (e.g., drilling). If formed through an approximate horizontal center of the storage layer 119, the inclined portion 110 may be surrounded by about 50 to 100 feet of the geologic formation that comprises the storage layer 119. Further, the storage layer 119 may also have only immobile water, e.g., due to a very low permeability of the layer 119 (e.g., on the order of milli- or nanodarcys). In addition, the storage layer 119 may have sufficient ductility, such that a brittleness of the rock formation that comprises the layer 119 is between about 3 MPa and 10 MPa. Examples of rock formations of which the storage layer 119 may be composed include: shale and anhydrite. Further, in some aspects, hazardous material may be stored below the storage layer, even in a permeable formation such as sandstone or limestone, if the storage layer is of sufficient geologic properties to isolate the permeable layer from the mobile water layer 114. In some examples implementations of the hazardous material storage repository system 100, the storage layer 119 (and/or the impermeable layer 116) is composed of shale. Shale, in some examples, may have properties that fit within those described above for the storage layer 119. For example, shale formations may be suitable for a long-term confinement of hazardous material (e.g., in the hazardous material canisters 126), and for their isolation from mobile water layer 114 (e.g., aquifers) and the terranean surface 102. Shale formations may be found relatively deep in the Earth, typically 3000 feet or greater, and placed in isolation below any fresh water aquifers. Other formations may include salt or other impermeable formation layer. Shale formations (or salt or other impermeable formation layers), for instance, may include geologic properties that enhance the long-term (e.g., thousands of years) isolation of material. Such properties, for instance, have been illustrated through the long term storage (e.g., tens of millions of years) of hydrocarbon fluids (e.g., gas, liquid, mixed phase fluid) without escape of substantial fractions of such fluids into surrounding layers (e.g., mobile water layer 114). Indeed, shale has been shown to hold natural gas for millions of years or more, giving it a proven capability for long-term storage of hazardous material. Example shale formations (e.g., Marcellus, Eagle Ford, Barnett, and otherwise) has stratification that contains many redundant sealing layers that have been effective in preventing movement of water, oil, and gas for millions of years, lacks mobile water, and can be expected (e.g., based on geological considerations) to seal hazardous material (e.g., fluids or solids) for thousands of years after deposit. In some aspects, the formation of the storage layer 119 and/or the impermeable layer 116 may form a leakage barrier, or barrier layer to fluid leakage that may be determined, at least in part, by the evidence of the storage capacity of the layer for hydrocarbons or other fluids (e.g., carbon dioxide) for hundreds of years, thousands of years, tens of thousands of years, hundreds of thousands of years, or even millions of years. For example, the barrier layer of the storage layer 119 and/or impermeable layer 116 may be defined by a time constant for leakage of the hazardous material more than 10,000 years (such as between about 10,000 years and 1,000,000 years) based on such evidence of hydrocarbon or other fluid storage. Shale (or salt or other impermeable layer) formations may also be at a suitable depth, e.g., between 3000 and 12,000 feet TVD. Such depths are typically below ground water aquifer (e.g., surface layer 112 and/or mobile water layer 114). Further, the presence of soluble elements in shale, including salt, and the absence of these same elements in aquifer layers, demonstrates a fluid isolation between shale and the aquifer layers. Another particular quality of shale that may advantageously lend itself to hazardous material storage is its clay content, which, in some aspects, provides a measure of ductility greater than that found in other, impermeable rock formations (e.g., impermeable layer 116). For example, shale may be stratified, made up of thinly alternating layers of clays (e.g., between about 20-30% clay by volume) and other minerals. Such a composition may make shale less brittle and, thus less susceptible to fracturing (e.g., naturally or otherwise) as compared to rock formations in the impermeable layer (e.g., dolomite or otherwise). For example, rock formations in the impermeable layer 116 may have suitable permeability for the long term storage of hazardous material, but are too brittle and commonly are fractured. Thus, such formations may not have sufficient sealing qualities (as evidenced through their geologic properties) for the long term storage of hazardous material. The present disclosure contemplates that there may be many other layers between or among the illustrated subterranean layers 112, 114, 116, and 119. For example, there may be repeating patterns (e.g., vertically), of one or more of the mobile water layer 114, impermeable layer 116, and storage layer 119. Further, in some instances, the storage layer 119 may be directly adjacent (e.g., vertically) the mobile water layer 114, i.e., without an intervening impermeable layer 116. In some examples, all or portions of the radiussed drillhole 108 and the inclined drillhole 110 may be formed below the storage layer 119, such that the storage layer 119 (e.g., shale or other geologic formation with characteristics as described herein) is vertically positioned between the inclined drillhole 110 and the mobile water layer 114. Further, in this example implementation, a self-healing layer 132 may be found below the terranean surface 102 and between, for example, the surface 102 and one or both of the impermeable layer 116 and the storage layer 119. In some aspects, the self-healing layer 132 may comprise a geologic formation that can stop or impede a flow of hazardous material (whether in liquid, solid, or gaseous form) from a storage portion of the drillhole 104 to or toward the terranean surface 102. For example, during formation of the drillhole 104 (e.g., drilling), all are portions of the geologic formations of the layers 112, 114, 116, and 119, may be damaged (as illustrated by a damaged zone 140), thereby affecting or changing their geologic characteristics (e.g., permeability). Indeed, although damaged zone 140 is illustrated between layers 114 and 132 for simplicity sake, the damaged zone 140 may surround an entire length (vertical, curved, and inclined portions) of the drillhole 104 a particular distance into the layers 112, 114, 116, 119, 132, and otherwise. In certain aspects, the location of the drillhole 104 may be selected so as to be formed through the self-healing layer 132. For example, as shown, the drillhole 104 may be formed such that at least a portion of the vertical portion 106 of the drillhole 104 is formed to pass through the self-healing layer 132. In some aspects, the self-healing layer 132 comprises a geologic formation that that does not sustain cracks for extended time durations even after being drilled therethrough. Examples of the geologic formation in the self-healing layer 132 include clay or dolomite. Cracks in such rock formations tend to heal, that is, they disappear rapidly with time due to the relative ductility of the material, and the enormous pressures that occur underground from the weight of the overlying rock on the formation in the self-healing layer. In addition to providing a “healing mechanism” for cracks that occur due to the formation of the drillhole 104 (e.g., drilling or otherwise), the self-healing layer 132 may also provide a barrier to natural faults and other cracks that otherwise could provide a pathway for hazardous material leakage (e.g., fluid or solid) from the storage region (e.g., in the inclined portion 110) to the terranean surface 102, the mobile water layer 114, or both. As shown in this example, the inclined portion 110 of the drillhole 104 includes a storage area 117 in a distal part of the portion 110 into which hazardous material may be retrievably placed for long-term storage. For example, as shown, a work string 124 (e.g., tubing, coiled tubing, wireline, or otherwise) may be extended into the cased drillhole 104 to place one or more (three shown but there may be more or less) hazardous material canisters 126 into long term, but in some aspects, retrievable, storage in the portion 110. For example, in the implementation shown in FIG. 1A, the work string 124 may include a downhole tool 128 that couples to the canister 126, and with each trip into the drillhole 104, the downhole tool 128 may deposit a particular hazardous material canister 126 in the inclined portion 110. The downhole tool 128 may couple to the canister 126 by, in some aspects, a threaded connection or other type of connection, such as a latched connection. In alternative aspects, the downhole tool 128 may couple to the canister 126 with an interlocking latch, such that rotation (or linear movement or electric or hydraulic switches) of the downhole tool 128 may latch to (or unlatch from) the canister 126. In alternative aspects, the downhole tool 124 may include one or more magnets (e.g., rare Earth magnets, electromagnets, a combination thereof, or otherwise) which attractingly couple to the canister 126. In some examples, the canister 126 may also include one or more magnets (e.g., rare Earth magnets, electromagnets, a combination thereof, or otherwise) of an opposite polarity as the magnets on the downhole tool 124. In some examples, the canister 126 may be made from or include a ferrous or other material attractable to the magnets of the downhole tool 124. As another example, each canister 126 may be positioned within the drillhole 104 by a drillhole tractor (e.g., on a wireline or otherwise), which may push or pull the canister into the inclined portion 110 through motorized (e.g., electric) motion. As yet another example, each canister 126 may include or be mounted to rollers (e.g., wheels), so that the downhole tool 124 may push the canister 126 into the cased drillhole 104. In some example implementations, the canister 126, one or more of the drillhole casings 120 and 122, or both, may be coated with a friction-reducing coating prior to the deposit operation. For example, by applying a coating (e.g., petroleum-based product, resin, ceramic, or otherwise) to the canister 126 and/or drillhole casings, the canister 126 may be more easily moved through the cased drillhole 104 into the inclined portion 110. In some aspects, only a portion of the drillhole casings may be coated. For example, in some aspects, the substantially vertical portion 106 may not be coated, but the radiussed portion 108 or the inclined portion 110, or both, may be coated to facilitate easier deposit and retrieval of the canister 126. FIG. 1A also illustrates an example of a retrieval operation of hazardous material in the inclined portion 110 of the drillhole 104. A retrieval operation may be the opposite of a deposit operation, such that the downhole tool 124 (e.g., a fishing tool) may be run into the drillhole 104, coupled to the last-deposited canister 126 (e.g., threadingly, latched, by magnet, or otherwise), and pull the canister 126 to the terranean surface 102. Multiple retrieval trips may be made by the downhole tool 124 in order to retrieve multiple canisters from the inclined portion 110 of the drillhole 104. Each canister 126 may enclose hazardous material. Such hazardous material, in some examples, may be biological or chemical waste or other biological or chemical hazardous material. In some examples, the hazardous material may include nuclear material, such as spent nuclear fuel recovered from a nuclear reactor (e.g., commercial power or test reactor) or military nuclear material. For example, a gigawatt nuclear plant may produce 30 tons of spent nuclear fuel per year. The density of that fuel is typically close to 10 (10 gm/cm3=10 kg/liter), so that the volume for a year of nuclear waste is about 3 m3. Spent nuclear fuel, in the form of nuclear fuel pellets, may be taken from the reactor and not modified. Nuclear fuel pellet are solid, although they can contain and emit a variety of radioactive gases including tritium (13 year half-life), krypton-85 (10.8 year half-life), and carbon dioxide containing C-14 (5730 year half-life). In some aspects, the storage layer 119 should be able to contain any radioactive output (e.g., gases) within the layer 119, even if such output escapes the canisters 126. For example, the storage layer 119 may be selected based on diffusion times of radioactive output through the layer 119. For example, a minimum diffusion time of radioactive output escaping the storage layer 119 may be set at, for example, fifty times a half-life for any particular component of the nuclear fuel pellets. Fifty half-lives as a minimum diffusion time would reduce an amount of radioactive output by a factor of 1×10−15. As another example, setting a minimum diffusion time to thirty half-lives would reduce an amount of radioactive output by a factor of one billion. For example, plutonium-239 is often considered a dangerous waste product in spent nuclear fuel because of its long half-life of 24,100 years. For this isotope, 50 half-lives would be 1.2 million years. Plutonium-239 has low solubility in water, is not volatile, and as a solid. its diffusion time is exceedingly small (e.g., many millions of years) through a matrix of the rock formation that comprises the illustrated storage layer 119 (e.g., shale or other formation). The storage layer 119, for example comprised of shale, may offer the capability to have such isolation times (e.g., millions of years) as shown by the geological history of containing gaseous hydrocarbons (e.g., methane and otherwise) for several million years. In contrast, in conventional nuclear material storage methods, there was a danger that some plutonium might dissolve in a layer that comprised mobile ground water upon confinement escape. As further shown in FIG. 1A, the storage canisters 126 may be positioned for long term storage in the inclined portion 110, which, as shown, is tilted upward at a small angle (e.g., 2-5 degrees) as it gets further away from the vertical portion 106 of the drillhole 104. As illustrated, the inclined portion 110 tilts upward toward the terranean surface 102. In some aspects, for example when there is radioactive hazardous material stored in the canisters 126, the inclination of the drillhole portion 110 may provide a further degree of safety and containment to prevent or impede the material, even if leaked from the canister 126, from reaching, e.g., the mobile water layer 114, the vertical portion 106 of the drillhole 104, the terranean surface 102, or a combination thereof. For example, radionuclides of concern in the hazardous material tend to be relatively buoyant or heavy (as compared to brine or other fluids that might fill the drillhole). Buoyant radionuclides may be the greatest concern for leakage, since heavy elements and molecules tend to sink, and would not diffuse upward towards the terranean surface 102. Krypton gas, and particularly 14CO2 (where 14C refers to radiocarbon, also called C-14, which is an isotope of carbon with a half-life of 5730 years), is a buoyant radioactive element that is heavier than air (as are most gases) but much lighter than water. Thus, should 14CO2 be introduced into a water bath, such gas would tend to float upward towards the terranean surface 102. Iodine, on the other hand, is denser than water, and would tend to diffuse downward if introduced into a water bath. By including the inclined portion 110 of the drillhole 104, any such diffusion of radioactive material (e.g., even if leaked from a canister 126 and in the presence of water or other liquid in the drillhole 104 or otherwise) would be directed angularly upward toward a distal end 121 of the inclined portion 110 and away from the radiussed portion 108 (and the vertical portion 106) of the drillhole 104. Thus, leaked hazardous material, even in a diffusible gas form, would not be offered a path (e.g., directly) to the terranean surface 102 (or the mobile water layer 114) through the vertical portion 106 of the drillhole 110. For instance, the leaked hazardous material (especially in gaseous form) would be directed and gathered at the distal end 121 of the drillhole portion 110. Alternative methods of depositing the canisters 126 into the inclined drillhole portion 110 may also be implemented. For instance, a fluid (e.g., liquid or gas) may be circulated through the drillhole 104 to fluidly push the canisters 126 into the inclined drillhole portion 110. In some example, each canister 126 may be fluidly pushed separately. In alternative aspects, two or more canisters 126 may be fluidly pushed, simultaneously, through the drillhole 104 for deposit into the inclined portion 110. The fluid can be, in some cases, water. Other examples include a drilling mud or drilling foam. In some examples, a gas may be used to push the canisters 126 into the drillhole, such as air, argon, or nitrogen. In some aspects, the choice of fluid may depend at least in part on a viscosity of the fluid. For example, a fluid may be chosen with enough viscosity to impede the drop of the canister 126 into the substantially vertical portion 106. This resistance or impedance may provide a safety factor against a sudden drop of the canister 126. The fluid may also provide lubrication to reduce a sliding friction between the canister 126 and the casings 120 and 122. The canister 126 can be conveyed within a casing filled with a liquid of controlled viscosity, density, and lubricant qualities. The fluid-filled annulus between the inner diameter of the casings 120 and 122 and the outer diameter of the conveyed canister 126 represents an opening designed to dampen any high rate of canister motion, providing automatic passive protection in an unlikely decoupling of the conveyed canister 126. In some aspects, other techniques may be employed to facilitate deposit of the canister 126 into the inclined portion 110. For example, one or more of the installed casings (e.g., casings 120 and 122) may have rails to guide the storage canister 126 into the drillhole 102 while reducing friction between the casings and the canister 126. The storage canister 126 and the casings (or the rails) may be made of materials that slide easily against one another. The casings may have a surface that is easily lubricated, or one that is self-lubricating when subjected to the weight of the storage canister 126. The fluid may also be used for retrieval of the canister 126. For example, in an example retrieval operation, a volume within the casings 120 and 122 may be filled with a compressed gas (e.g., air, nitrogen, argon, or otherwise). As the pressure increases at an end of the inclined portion 110, the canisters 126 may be pushed toward the radiussed portion 108, and subsequently through the substantially vertical portion 106 to the terranean surface. In some aspects, the drillhole 104 may be formed for the primary purpose of long-term storage of hazardous materials. In alternative aspects, the drillhole 104 may have been previously formed for the primary purpose of hydrocarbon production (e.g., oil, gas). For example, storage layer 119 may be a hydrocarbon bearing formation from which hydrocarbons were produced into the drillhole 104 and to the terranean surface 102. In some aspects, the storage layer 119 may have been hydraulically fractured prior to hydrocarbon production. Further in some aspects, the production casing 122 may have been perforated prior to hydraulic fracturing. In such aspects, the production casing 122 may be patched (e.g., cemented) to repair any holes made from the perforating process prior to a deposit operation of hazardous material. In addition, any cracks or openings in the cement between the casing and the drillhole can also be filled at that time. For example, in the case of spent nuclear fuel as a hazardous material, the drillhole may be formed at a particular location, e.g., near a nuclear power plant, as a new drillhole provided that the location also includes an appropriate storage layer 119, such as a shale formation. Alternatively, an existing well that has already produced shale gas, or one that was abandoned as “dry,” (e.g., with sufficiently low organics that the gas in place is too low for commercial development), may be selected as the drillhole 104. In some aspects, prior hydraulic fracturing of the storage layer 119 through the drillhole 104 may make little difference in the hazardous material storage capability of the drillhole 104. But such a prior activity may also confirm the ability of the storage layer 119 to store gases and other fluids for millions of years. If, therefore, the hazardous material or output of the hazardous material (e.g., radioactive gasses or otherwise) were to escape from the canister 126 and enter the fractured formation of the storage layer 119, such fractures may allow that material to spread relatively rapidly over a distance comparable in size to that of the fractures. In some aspects, the drillhole 102 may have been drilled for a production of hydrocarbons, but production of such hydrocarbons had failed, e.g., because the storage layer 119 comprised a rock formation (e.g., shale or otherwise) that was too ductile and difficult to fracture for production, but was advantageously ductile for the long-term storage of hazardous material. FIG. 1B is a schematic illustration of a portion of the example implementation of the hazardous material storage repository system 100 that shows an example determination of a minimum angle of the inclined portion 110 of the hazardous material storage repository system 100. For example, as shown in system 100, the inclined portion 110 provides that any path that leaking hazardous material (e.g., from one or more of the canister 126) takes to the terranean surface 102 through the drillhole 104 includes at least one downward component. In this case, the inclined portion 110 is the downward component. In other example implementations described later, such as systems 200 and 300, other portions (e.g., a J-section portion or undulating portion) may include at least one downward component. Such paths, as shown in this example, dip below a horizontal escape limit line 175 that intersects a canister 126 that is closest (when positioned in the storage area 117) to the vertical portion 106 of the drillhole 104. and therefore must include a downward component. In some aspects, an angle, a, of the inclined portion 110 of the drillhole 104 may be determined (and thereby guide the formation of the drillhole 104) according to a radius, R, of the damaged zone 140 of the drillhole 104 and a distance, D, from the canister 126 that is closest to the vertical portion 106 of the drillhole 104. As shown in the callout bubble in FIG. 1B, with knowledge of the distances R and D (or at least estimates), then the angle, a, can be computed according to the arctangent of R/D. In an example implementation, R may be about 1 meter while D may be about 20 meters. The angle, a, therefore, as the arctangent of R/D is about 3°. This is just one example of the determination of the angle, a, of a downward component (e.g., the inclined portion 110) of the drillhole 104 to ensure that such a downward component dips below the horizontal escape limit line 175. FIG. 2 is a schematic illustration of example implementations of another hazardous material storage repository system, e.g., a subterranean location for the long-term (e.g., tens, hundreds, or thousands of years or more) but retrievable safe and secure storage of hazardous material, during a deposit or retrieval operation according to the present disclosure. For example, turning to FIG. 2, this figure illustrates an example hazardous material storage repository system 200 during a deposit (or retrieval, as described below) process, e.g., during deployment of one or more canisters of hazardous material in a subterranean formation. As illustrated, the hazardous material storage repository system 200 includes a drillhole 204 formed (e.g., drilled or otherwise) from a terranean surface 202 and through multiple subterranean layers 212, 214, and 216. Although the terranean surface 202 is illustrated as a land surface, terranean surface 202 may be a sub-sea or other underwater surface, such as a lake or an ocean floor or other surface under a body of water. Thus, the present disclosure contemplates that the drillhole 204 may be formed under a body of water from a drilling location on or proximate the body of water. The illustrated drillhole 204 is a directional drillhole in this example of hazardous material storage repository system 200. For instance, the drillhole 204 includes a substantially vertical portion 206 coupled to a J-section portion 208, which in turn is coupled to a substantially horizontal portion 210. The J-section portion 208 as shown, has a shape that resembles the bottom portion of the letter “J” and may be shaped similar to a p-trap device used in a plumbing system that is used to prevent gasses from migrating from one side of the bend to the other side of the bend. As used in the present disclosure, “substantially” in the context of a drillhole orientation, refers to drillholes that may not be exactly vertical (e.g., exactly perpendicular to the terranean surface 202) or exactly horizontal (e.g., exactly parallel to the terranean surface 202), or exactly inclined at a particular incline angle relative to the terranean surface 202. In other words, vertical drillholes often undulate offset from a true vertical direction, that they might be drilled at an angle that deviates from true vertical, and horizontal drillholes often undulate offset from exactly horizontal. As illustrated in this example, the three portions of the drillhole 204—the vertical portion 206, the J-section portion 208, and the substantially horizontal portion 210—form a continuous drillhole 204 that extends into the Earth. As also shown in dashed line in FIG. 2, the J-section portion 208 may be coupled to an inclined portion 240 rather than (or in addition to) the substantially horizontal portion 210 of the drillhole 204. The illustrated drillhole 204, in this example, has a surface casing 220 positioned and set around the drillhole 204 from the terranean surface 202 into a particular depth in the Earth. For example, the surface casing 220 may be a relatively large-diameter tubular member (or string of members) set (e.g., cemented) around the drillhole 204 in a shallow formation. As used herein, “tubular” may refer to a member that has a circular cross-section, elliptical cross-section, or other shaped cross-section. For example, in this implementation of the hazardous material storage repository system 200, the surface casing 220 extends from the terranean surface through a surface layer 212. The surface layer 212, in this example, is a geologic layer comprised of one or more layered rock formations. In some aspects, the surface layer 212 in this example may or may not include freshwater aquifers, salt water or brine sources, or other sources of mobile water (e.g., water that moves through a geologic formation). In some aspects, the surface casing 220 may isolate the drillhole 204 from such mobile water, and may also provide a hanging location for other casing strings to be installed in the drillhole 204. Further, although not shown, a conductor casing may be set above the surface casing 220 (e.g., between the surface casing 220 and the surface 202 and within the surface layer 212) to prevent drilling fluids from escaping into the surface layer 212. As illustrated, a production casing 222 is positioned and set within the drillhole 204 downhole of the surface casing 220. Although termed a “production” casing, in this example, the casing 222 may or may not have been subject to hydrocarbon production operations. Thus, the casing 222 refers to and includes any form of tubular member that is set (e.g., cemented) in the drillhole 204 downhole of the surface casing 220. In some examples of the hazardous material storage repository system 200, the production casing 222 may begin at an end of the J-section portion 208 and extend throughout the substantially horizontal portion 210. The casing 222 could also extend into the J-section portion 208 and into the vertical portion 206. As shown, cement 230 is positioned (e.g., pumped) around the casings 220 and 222 in an annulus between the casings 220 and 222 and the drillhole 204. The cement 230, for example, may secure the casings 220 and 222 (and any other casings or liners of the drillhole 204) through the subterranean layers under the terranean surface 202. In some aspects, the cement 230 may be installed along the entire length of the casings (e.g., casings 220 and 222 and any other casings), or the cement 230 could be used along certain portions of the casings if adequate for a particular drillhole 202. The cement 230 can also provide an additional layer of confinement for the hazardous material in canisters 226. The drillhole 204 and associated casings 220 and 222 may be formed with various example dimensions and at various example depths (e.g., true vertical depth, or TVD). For instance, a conductor casing (not shown) may extend down to about 120 feet TVD, with a diameter of between about 28 in. and 60 in. The surface casing 220 may extend down to about 2500 feet TVD, with a diameter of between about 22 in. and 48 in. An intermediate casing (not shown) between the surface casing 220 and production casing 222 may extend down to about 8000 feet TVD, with a diameter of between about 16 in. and 36 in. The production casing 222 may extend inclinedly (e.g., to case the substantially horizontal portion 210 and/or the inclined portion 240) with a diameter of between about 11 in. and 22 in. The foregoing dimensions are merely provided as examples and other dimensions (e.g., diameters, TVDs, lengths) are contemplated by the present disclosure. For example, diameters and TVDs may depend on the particular geological composition of one or more of the multiple subterranean layers (212, 214, and 216), particular drilling techniques, as well as a size, shape, or design of a hazardous material canister 226 that contains hazardous material to be deposited in the hazardous material storage repository system 200. In some alternative examples, the production casing 222 (or other casing in the drillhole 204) could be circular in cross-section, elliptical in cross-section, or some other shape. As illustrated, the vertical portion 206 of the drillhole 204 extends through subterranean layers 212, 214, and 216, and, in this example, lands in a subterranean layer 219. As discussed above, the surface layer 212 may or may not include mobile water. Subterranean layer 214, which is below the surface layer 212, in this example, is a mobile water layer 214. For instance, mobile water layer 214 may include one or more sources of mobile water, such as freshwater aquifers, salt water or brine, or other source of mobile water. In this example of hazardous material storage repository system 200, mobile water may be water that moves through a subterranean layer based on a pressure differential across all or a part of the subterranean layer. For example, the mobile water layer 214 may be a permeable geologic formation in which water freely moves (e.g., due to pressure differences or otherwise) within the layer 214. In some aspects, the mobile water layer 214 may be a primary source of human-consumable water in a particular geographic area. Examples of rock formations of which the mobile water layer 214 may be composed include porous sandstones and limestones, among other formations. Other illustrated layers, such as the impermeable layer 216 and the storage layer 219, may include immobile water. Immobile water, in some aspects, is water (e.g., fresh, salt, brine), that is not fit for human or animal consumption, or both. Immobile water, in some aspects, may be water that, by its motion through the layers 216 or 219 (or both), cannot reach the mobile water layer 214, terranean surface 202, or both, within 10,000 years or more (such as to 1,000,000 years). Below the mobile water layer 214, in this example implementation of hazardous material storage repository system 200, is an impermeable layer 216. The impermeable layer 216, in this example, may not allow mobile water to pass through. Thus, relative to the mobile water layer 214, the impermeable layer 216 may have low permeability, e.g., on the order of 0.01 millidarcy permeability. Additionally, in this example, the impermeable layer 216 may be a relatively non-ductile (i.e., brittle) geologic formation. One measure of non-ductility is brittleness, which is the ratio of compressive stress to tensile strength. In some examples, the brittleness of the impermeable layer 216 may be between about 20 MPa and 40 MPa. As shown in this example, the impermeable layer 216 is shallower (e.g., closer to the terranean surface 202) than the storage layer 219. In this example rock formations of which the impermeable layer 216 may be composed include, for example, certain kinds of sandstone, mudstone, clay, and slate that exhibit permeability and brittleness properties as described above. In alternative examples, the impermeable layer 216 may be deeper (e.g., further from the terranean surface 202) than the storage layer 219. In such alternative examples, the impermeable layer 216 may be composed of an igneous rock, such as granite. Below the impermeable layer 216 is the storage layer 219. The storage layer 219, in this example, may be chosen as the landing for the substantially horizontal portion 210, which stores the hazardous material, for several reasons. Relative to the impermeable layer 216 or other layers, the storage layer 219 may be thick, e.g., between about 100 and 200 feet of total vertical thickness. Thickness of the storage layer 219 may allow for easier landing and directional drilling, thereby allowing the substantially horizontal portion 210 to be readily emplaced within the storage layer 219 during constructions (e.g., drilling). If formed through an approximate horizontal center of the storage layer 219, the substantially horizontal portion 210 may be surrounded by about 50 to 100 feet of the geologic formation that comprises the storage layer 219. Further, the storage layer 219 may also have only immobile water, e.g., due to a very low permeability of the layer 219 (e.g., on the order of milli- or nanodarcys). In addition, the storage layer 219 may have sufficient ductility, such that a brittleness of the rock formation that comprises the layer 219 is between about 3 MPa and 10 MPa. Examples of rock formations of which the storage layer 219 may be composed include: shale and anhydrite. Further, in some aspects, hazardous material may be stored below the storage layer, even in a permeable formation such as sandstone or limestone, if the storage layer is of sufficient geologic properties to isolate the permeable layer from the mobile water layer 214. In some examples implementations of the hazardous material storage repository system 200, the storage layer 219 (and/or the impermeable layer 216) is composed of shale. Shale, in some examples, may have properties that fit within those described above for the storage layer 219. For example, shale formations may be suitable for a long-term confinement of hazardous material (e.g., in the hazardous material canisters 226), and for their isolation from mobile water layer 214 (e.g., aquifers) and the terranean surface 202. Shale formations may be found relatively deep in the Earth, typically 3000 feet or greater, and placed in isolation below any fresh water aquifers. Other formations may include salt or other impermeable formation layer. Shale formations (or salt or other impermeable formation layers), for instance, may include geologic properties that enhance the long-term (e.g., thousands of years) isolation of material. Such properties, for instance, have been illustrated through the long term storage (e.g., tens of millions of years) of hydrocarbon fluids (e.g., gas, liquid, mixed phase fluid) without escape of such fluids into surrounding layers (e.g., mobile water layer 214). Indeed, shale has been shown to hold natural gas for millions of years or more, giving it a proven capability for long-term storage of hazardous material. Example shale formations (e.g., Marcellus, Eagle Ford, Barnett, and otherwise) has stratification that contains many redundant sealing layers that have been effective in preventing movement of water, oil, and gas for millions of years, lacks mobile water, and can be expected (e.g., based on geological considerations) to seal hazardous material (e.g., fluids or solids) for thousands of years after deposit. In some aspects, the formation of the storage layer 219 and/or the impermeable layer 216 may form a leakage barrier, or barrier layer to fluid leakage that may be determined, at least in part, by the evidence of the storage capacity of the layer for hydrocarbons or other fluids (e.g., carbon dioxide) for hundreds of years, thousands of years, tens of thousands of years, hundreds of thousands of years, or even millions of years. For example, the barrier layer of the storage layer 219 and/or impermeable layer 216 may be defined by a time constant for leakage of the hazardous material of more than 10,000 years (such as between 10,000 years and 1,000,000 years) based on such evidence of hydrocarbon or other fluid storage. Shale (or salt or other impermeable layer) formations may also be at a suitable depth, e.g., between 3000 and 12,000 feet TVD. Such depths are typically below ground water aquifer (e.g., surface layer 212 and/or mobile water layer 214). Further, the presence of soluble elements in shale, including salt, and the absence of these same elements in aquifer layers, demonstrates a fluid isolation between shale and the aquifer layers. Another particular quality of shale that may advantageously lend itself to hazardous material storage is its clay content, which, in some aspects, provides a measure of ductility greater than that found in other, impermeable rock formations (e.g., impermeable layer 216). For example, shale may be stratified, made up of thinly alternating layers of clays (e.g., between about 20-30% clay by volume) and other minerals. Such a composition may make shale less brittle and, thus less susceptible to fracturing (e.g., naturally or otherwise) as compared to rock formations in the impermeable layer (e.g., dolomite or otherwise). For example, rock formations in the impermeable layer 216 may have suitable permeability for the long term storage of hazardous material, but are too brittle and commonly are fractured. Thus, such formations may not have sufficient sealing qualities (as evidenced through their geologic properties) for the long term storage of hazardous material. The present disclosure contemplates that there may be many other layers between or among the illustrated subterranean layers 212, 214, 216, and 219. For example, there may be repeating patterns (e.g., vertically), of one or more of the mobile water layer 214, impermeable layer 216, and storage layer 219. Further, in some instances, the storage layer 219 may be directly adjacent (e.g., vertically) the mobile water layer 214, i.e., without an intervening impermeable layer 216. In some examples, all or portions of the J-section drillhole 208 and the substantially horizontal portion 210 (and/or the inclined portion 240) may be formed below the storage layer 219, such that the storage layer 219 (e.g., shale or other geologic formation with characteristics as described herein) is vertically positioned between the substantially horizontal portion 210 (and/or the inclined portion 240) and the mobile water layer 214. Although not illustrated in this particular example shown in FIG. 2, a self-healing layer (e.g., such as the self-healing layer 132) may be found below the terranean surface 202 and between, for example, the surface 202 and one or both of the impermeable layer 216 and the storage layer 219. In some aspects, the self-healing layer may comprise a geologic formation that can stop or impede a flow of hazardous material (whether in liquid, solid, or gaseous form) from a storage portion of the drillhole 204 to or toward the terranean surface 202. For example, during formation of the drillhole 204 (e.g., drilling), all are portions of the geologic formations of the layers 212, 214, 216, and 219, may be damaged, thereby affecting or changing their geologic characteristics (e.g., permeability). In certain aspects, the location of the drillhole 204 may be selected so as to be formed through the self-healing layer. For example, as shown, the drillhole 204 may be formed such that at least a portion of the vertical portion 206 of the drillhole 204 is formed to pass through the self-healing layer. In some aspects, the self-healing layer comprises a geologic formation that that does not sustain cracks for extended time durations even after being drilled therethrough. Examples of the geologic formation in the self-healing layer include clay or dolomite. Cracks in such rock formations tend to heal, that is, they disappear rapidly with time due to the relative ductility of the material, and the enormous pressures that occur underground from the weight of the overlying rock on the formation in the self-healing layer. In addition to providing a “healing mechanism” for cracks that occur due to the formation of the drillhole 204 (e.g., drilling or otherwise), the self-healing layer may also provide a barrier to natural faults and other cracks that otherwise could provide a pathway for hazardous material leakage (e.g., fluid or solid) from the storage region (e.g., in the substantially horizontal portion 210) to the terranean surface 202, the mobile water layer 214, or both. As shown in this example, the substantially horizontal portion 210 of the drillhole 204 includes a storage area 217 in a distal part of the portion 210 into which hazardous material may be retrievably placed for long-term storage. For example, as shown, a work string 224 (e.g., tubing, coiled tubing, wireline, or otherwise) may be extended into the cased drillhole 204 to place one or more (three shown but there may be more or less) hazardous material canisters 226 into long term, but in some aspects, retrievable, storage in the portion 210. For example, in the implementation shown in FIG. 2, the work string 224 may include a downhole tool 228 that couples to the canister 226, and with each trip into the drillhole 204, the downhole tool 228 may deposit a particular hazardous material canister 226 in the substantially horizontal portion 210. The downhole tool 228 may couple to the canister 226 by, in some aspects, a threaded connection or other type of connection, such as a latched connection. In alternative aspects, the downhole tool 228 may couple to the canister 226 with an interlocking latch, such that rotation (or linear movement or electric or hydraulic switches) of the downhole tool 228 may latch to (or unlatch from) the canister 226. In alternative aspects, the downhole tool 224 may include one or more magnets (e.g., rare Earth magnets, electromagnets, a combination thereof, or otherwise) which attractingly couple to the canister 226. In some examples, the canister 226 may also include one or more magnets (e.g., rare Earth magnets, electromagnets, a combination thereof, or otherwise) of an opposite polarity as the magnets on the downhole tool 224. In some examples, the canister 226 may be made from or include a ferrous or other material attractable to the magnets of the downhole tool 224. As another example, each canister 226 may be positioned within the drillhole 204 by a drillhole tractor (e.g., on a wireline or otherwise), which may push or pull the canister into the substantially horizontal portion 210 through motorized (e.g., electric) motion. As yet another example, each canister 226 may include or be mounted to rollers (e.g., wheels), so that the downhole tool 224 may push the canister 226 into the cased drillhole 204. In some example implementations, the canister 226, one or more of the drillhole casings 220 and 222, or both, may be coated with a friction-reducing coating prior to the deposit operation. For example, by applying a coating (e.g., petroleum-based product, resin, ceramic, or otherwise) to the canister 226 and/or drillhole casings, the canister 226 may be more easily moved through the cased drillhole 204 into the substantially horizontal portion 210. In some aspects, only a portion of the drillhole casings may be coated. For example, in some aspects, the substantially vertical portion 206 may not be coated, but the J-section portion 208 or the substantially horizontal portion 210, or both, may be coated to facilitate easier deposit and retrieval of the canister 226. FIG. 2 also illustrates an example of a retrieval operation of hazardous material in the substantially horizontal portion 210 of the drillhole 204. A retrieval operation may be the opposite of a deposit operation, such that the downhole tool 224 (e.g., a fishing tool) may be run into the drillhole 204, coupled to the last-deposited canister 226 (e.g., threadingly, latched, by magnet, or otherwise), and pull the canister 226 to the terranean surface 202. Multiple retrieval trips may be made by the downhole tool 224 in order to retrieve multiple canisters from the substantially horizontal portion 210 of the drillhole 204. Each canister 226 may enclose hazardous material. Such hazardous material, in some examples, may be biological or chemical waste or other biological or chemical hazardous material. In some examples, the hazardous material may include nuclear material, such as spent nuclear fuel recovered from a nuclear reactor (e.g., commercial power or test reactor) or military nuclear material. For example, a gigawatt nuclear plant may produce 30 tons of spent nuclear fuel per year. The density of that fuel is typically close to 10 (10 gm/cm3=10 kg/liter), so that the volume for a year of nuclear waste is about 3 m3. Spent nuclear fuel, in the form of nuclear fuel pellets, may be taken from the reactor and not modified. Nuclear fuel pellet are solid, although they can contain and emit a variety of radioactive gases including tritium (13 year half-life), krypton-85 (10.8 year half-life), and carbon dioxide containing C-14 (5730 year half-life). In some aspects, the storage layer 219 should be able to contain any radioactive output (e.g., gases) within the layer 219, even if such output escapes the canisters 226. For example, the storage layer 219 may be selected based on diffusion times of radioactive output through the layer 219. For example, a minimum diffusion time of radioactive output escaping the storage layer 219 may be set at, for example, fifty times a half-life for any particular component of the nuclear fuel pellets. Fifty half-lives as a minimum diffusion time would reduce an amount of radioactive output by a factor of 1×10−15. As another example, setting a minimum diffusion time to thirty half-lives would reduce an amount of radioactive output by a factor of one billion. For example, plutonium-239 is often considered a dangerous waste product in spent nuclear fuel because of its long half-life of 24,100 years. For this isotope, 50 half-lives would be 1.2 million years. Plutonium-239 has low solubility in water, is not volatile, and as a solid is not capable of diffusion through a matrix of the rock formation that comprises the illustrated storage layer 219 (e.g., shale or other formation). The storage layer 219, for example comprised of shale, may offer the capability to have such isolation times (e.g., millions of years) as shown by the geological history of containing gaseous hydrocarbons (e.g., methane and otherwise) for several million years. In contrast, in conventional nuclear material storage methods, there was a danger that some plutonium might dissolve in a layer that comprised mobile ground water upon confinement escape. As further shown in FIG. 2, the storage canisters 226 may be positioned for long term storage in the substantially horizontal portion 210, which, as shown, is coupled to the vertical portion 106 of the drillhole 104 through the J-section portion 208. As illustrated, the J-section portion 208 includes an upwardly directed portion angled toward the terranean surface 202. In some aspects, for example when there is radioactive hazardous material stored in the canisters 226, this inclination of the J-section portion 208 (and inclination of the inclined portion 240, if formed) may provide a further degree of safety and containment to prevent or impede the material, even if leaked from the canister 226, from reaching, e.g., the mobile water layer 214, the vertical portion 206 of the drillhole 204, the terranean surface 202, or a combination thereof. For example, radionuclides of concern in the hazardous material tend to be relatively buoyant or heavy (as compared to other components of the material). Buoyant radionuclides may be the greatest concern for leakage, since heavy elements and molecules tend to sink, and would not diffuse upward towards the terranean surface 202. Krypton gas, and particularly krypton-85, is a buoyant radioactive element that is heavier than air (as are most gases) but much lighter than water. Thus, should krypton-85 be introduced into a water bath, such gas would tend to float upward towards the terranean surface 202. Iodine, on the other hand, is denser than water, and would tend to diffuse downward if introduced into a water bath. By including the J-section portion 208 of the drillhole 204, any such diffusion of radioactive material (e.g., even if leaked from a canister 226 and in the presence of water or other liquid in the drillhole 204 or otherwise) would be directed angularly upward toward the substantially horizontal portion 210, and more specifically, toward a distal end 221 of the substantially horizontal portion 210 and away from the J-section portion 208 (and the vertical portion 206) of the drillhole 204. Thus, leaked hazardous material, even in a diffusible gas form, would not be offered a path (e.g., directly) to the terranean surface 202 (or the mobile water layer 214) through the vertical portion 206 of the drillhole 210. For instance, the leaked hazardous material (especially in gaseous form) would be directed and gathered at the distal end 221 of the drillhole portion 210, or, generally, within the substantially horizontal portion 210 of the drillhole 204. Alternative methods of depositing the canisters 226 into the inclined drillhole portion 210 may also be implemented. For instance, a fluid (e.g., liquid or gas) may be circulated through the drillhole 204 to fluidly push the canisters 226 into the inclined drillhole portion 210. In some example, each canister 226 may be fluidly pushed separately. In alternative aspects, two or more canisters 226 may be fluidly pushed, simultaneously, through the drillhole 204 for deposit into the substantially horizontal portion 210. The fluid can be, in some cases, water. Other examples include a drilling mud or drilling foam. In some examples, a gas may be used to push the canisters 226 into the drillhole, such as air, argon, or nitrogen. In some aspects, the choice of fluid may depend at least in part on a viscosity of the fluid. For example, a fluid may be chosen with enough viscosity to impede the drop of the canister 226 into the substantially vertical portion 206. This resistance or impedance may provide a safety factor against a sudden drop of the canister 226. The fluid may also provide lubrication to reduce a sliding friction between the canister 226 and the casings 220 and 222. The canister 226 can be conveyed within a casing filled with a liquid of controlled viscosity, density, and lubricant qualities. The fluid-filled annulus between the inner diameter of the casings 220 and 222 and the outer diameter of the conveyed canister 226 represents an opening designed to dampen any high rate of canister motion, providing automatic passive protection in an unlikely decoupling of the conveyed canister 226. In some aspects, other techniques may be employed to facilitate deposit of the canister 226 into the substantially horizontal portion 210. For example, one or more of the installed casings (e.g., casings 220 and 222) may have rails to guide the storage canister 226 into the drillhole 202 while reducing friction between the casings and the canister 226. The storage canister 226 and the casings (or the rails) may be made of materials that slide easily against one another. The casings may have a surface that is easily lubricated, or one that is self-lubricating when subjected to the weight of the storage canister 226. The fluid may also be used for retrieval of the canister 226. For example, in an example retrieval operation, a volume within the casings 220 and 222 may be filled with a compressed gas (e.g., air, nitrogen, argon, or otherwise). As the pressure increases at an end of the substantially horizontal portion 210, the canisters 226 may be pushed toward the J-section portion 208, and subsequently through the substantially vertical portion 206 to the terranean surface. In some aspects, the drillhole 204 may be formed for the primary purpose of long-term storage of hazardous materials. In alternative aspects, the drillhole 204 may have been previously formed for the primary purpose of hydrocarbon production (e.g., oil, gas). For example, storage layer 219 may be a hydrocarbon bearing formation from which hydrocarbons were produced into the drillhole 204 and to the terranean surface 202. In some aspects, the storage layer 219 may have been hydraulically fractured prior to hydrocarbon production. Further in some aspects, the production casing 222 may have been perforated prior to hydraulic fracturing. In such aspects, the production casing 222 may be patched (e.g., cemented) to repair any holes made from the perforating process prior to a deposit operation of hazardous material. In addition, any cracks or openings in the cement between the casing and the drillhole can also be filled at that time. For example, in the case of spent nuclear fuel as a hazardous material, the drillhole may be formed at a particular location, e.g., near a nuclear power plant, as a new drillhole provided that the location also includes an appropriate storage layer 219, such as a shale formation. Alternatively, an existing well that has already produced shale gas, or one that was abandoned as “dry,” (e.g., with sufficiently low organics that the gas in place is too low for commercial development), may be selected as the drillhole 204. In some aspects, prior hydraulic fracturing of the storage layer 219 through the drillhole 204 may make little difference in the hazardous material storage capability of the drillhole 204. But such a prior activity may also confirm the ability of the storage layer 219 to store gases and other fluids for millions of years. If, therefore, the hazardous material or output of the hazardous material (e.g., radioactive gasses or otherwise) were to escape from the canister 226 and enter the fractured formation of the storage layer 219, such fractures may allow that material to spread relatively rapidly over a distance comparable in size to that of the fractures. In some aspects, the drillhole 202 may have been drilled for a production of hydrocarbons, but production of such hydrocarbons had failed, e.g., because the storage layer 219 comprised a rock formation (e.g., shale or otherwise) that was too ductile and difficult to fracture for production, but was advantageously ductile for the long-term storage of hazardous material. FIG. 3 is a schematic illustration of example implementations of another hazardous material storage repository system, e.g., a subterranean location for the long-term (e.g., tens, hundreds, or thousands of years or more) but retrievable safe and secure storage of hazardous material, during a deposit or retrieval operation according to the present disclosure. For example, turning to FIG. 3, this figure illustrates an example hazardous material storage repository system 300 during a deposit (or retrieval, as described below) process, e.g., during deployment of one or more canisters of hazardous material in a subterranean formation. As illustrated, the hazardous material storage repository system 300 includes a drillhole 304 formed (e.g., drilled or otherwise) from a terranean surface 302 and through multiple subterranean layers 312, 314, and 316. Although the terranean surface 302 is illustrated as a land surface, terranean surface 302 may be a sub-sea or other underwater surface, such as a lake or an ocean floor or other surface under a body of water. Thus, the present disclosure contemplates that the drillhole 304 may be formed under a body of water from a drilling location on or proximate the body of water. The illustrated drillhole 304 is a directional drillhole in this example of hazardous material storage repository system 300. For instance, the drillhole 304 includes a substantially vertical portion 306 coupled to a curved portion 308, which in turn is coupled to a vertically undulating portion 310. As used in the present disclosure, “substantially” in the context of a drillhole orientation, refers to drillholes that may not be exactly vertical (e.g., exactly perpendicular to the terranean surface 302) or exactly horizontal (e.g., exactly parallel to the terranean surface 302), or exactly inclined at a particular incline angle relative to the terranean surface 302. In other words, vertical drillholes often undulate offset from a true vertical direction, that they might be drilled at an angle that deviates from true vertical, and horizontal drillholes often undulate offset from exactly horizontal. Further, in some aspects, an undulating portion may not undulate with regularity, i.e., have peaks that are uniformly spaced apart or valleys that are uniformly spaced apart. Instead, an undulating drillhole may undulate irregularly, e.g., with peaks that are non-uniformly spaced and/or valleys that are non-uniformly spaced. Further, an undulated drillhole may have a peak-to-valley distance that varies along a length of the drillhole. As illustrated in this example, the three portions of the drillhole 304—the vertical portion 306, the curved portion 308, and the vertically undulating portion 310—form a continuous drillhole 304 that extends into the Earth. The illustrated drillhole 304, in this example, has a surface casing 320 positioned and set around the drillhole 304 from the terranean surface 302 into a particular depth in the Earth. For example, the surface casing 320 may be a relatively large-diameter tubular member (or string of members) set (e.g., cemented) around the drillhole 304 in a shallow formation. As used herein, “tubular” may refer to a member that has a circular cross-section, elliptical cross-section, or other shaped cross-section. For example, in this implementation of the hazardous material storage repository system 300, the surface casing 320 extends from the terranean surface through a surface layer 312. The surface layer 312, in this example, is a geologic layer comprised of one or more layered rock formations. In some aspects, the surface layer 312 in this example may or may not include freshwater aquifers, salt water or brine sources, or other sources of mobile water (e.g., water that moves through a geologic formation). In some aspects, the surface casing 320 may isolate the drillhole 304 from such mobile water, and may also provide a hanging location for other casing strings to be installed in the drillhole 304. Further, although not shown, a conductor casing may be set above the surface casing 320 (e.g., between the surface casing 320 and the surface 302 and within the surface layer 312) to prevent drilling fluids from escaping into the surface layer 312. As illustrated, a production casing 322 is positioned and set within the drillhole 304 downhole of the surface casing 320. Although termed a “production” casing, in this example, the casing 322 may or may not have been subject to hydrocarbon production operations. Thus, the casing 322 refers to and includes any form of tubular member that is set (e.g., cemented) in the drillhole 304 downhole of the surface casing 320. In some examples of the hazardous material storage repository system 300, the production casing 322 may begin at an end of the curved portion 308 and extend throughout the vertically undulating portion 310. The casing 322 could also extend into the curved portion 308 and into the vertical portion 306. As shown, cement 330 is positioned (e.g., pumped) around the casings 320 and 322 in an annulus between the casings 320 and 322 and the drillhole 304. The cement 330, for example, may secure the casings 320 and 322 (and any other casings or liners of the drillhole 304) through the subterranean layers under the terranean surface 302. In some aspects, the cement 330 may be installed along the entire length of the casings (e.g., casings 320 and 322 and any other casings), or the cement 330 could be used along certain portions of the casings if adequate for a particular drillhole 302. The cement 330 can also provide an additional layer of confinement for the hazardous material in canisters 326. The drillhole 304 and associated casings 320 and 322 may be formed with various example dimensions and at various example depths (e.g., true vertical depth, or TVD). For instance, a conductor casing (not shown) may extend down to about 120 feet TVD, with a diameter of between about 28 in. and 60 in. The surface casing 320 may extend down to about 2500 feet TVD, with a diameter of between about 22 in. and 48 in. An intermediate casing (not shown) between the surface casing 320 and production casing 322 may extend down to about 8000 feet TVD, with a diameter of between about 16 in. and 36 in. The production casing 322 may extend inclinedly (e.g., to case the vertically undulating portion 310) with a diameter of between about 11 in. and 22 in. The foregoing dimensions are merely provided as examples and other dimensions (e.g., diameters, TVDs, lengths) are contemplated by the present disclosure. For example, diameters and TVDs may depend on the particular geological composition of one or more of the multiple subterranean layers (312, 314, and 316), particular drilling techniques, as well as a size, shape, or design of a hazardous material canister 326 that contains hazardous material to be deposited in the hazardous material storage repository system 300. In some alternative examples, the production casing 322 (or other casing in the drillhole 304) could be circular in cross-section, elliptical in cross-section, or some other shape. As illustrated, the vertical portion 306 of the drillhole 304 extends through subterranean layers 312, 314, and 316, and, in this example, lands in a subterranean layer 319. As discussed above, the surface layer 312 may or may not include mobile water. Subterranean layer 314, which is below the surface layer 312, in this example, is a mobile water layer 314. For instance, mobile water layer 314 may include one or more sources of mobile water, such as freshwater aquifers, salt water or brine, or other source of mobile water. In this example of hazardous material storage repository system 300, mobile water may be water that moves through a subterranean layer based on a pressure differential across all or a part of the subterranean layer. For example, the mobile water layer 314 may be a permeable geologic formation in which water freely moves (e.g., due to pressure differences or otherwise) within the layer 314. In some aspects, the mobile water layer 314 may be a primary source of human-consumable water in a particular geographic area. Examples of rock formations of which the mobile water layer 314 may be composed include porous sandstones and limestones, among other formations. Other illustrated layers, such as the impermeable layer 316 and the storage layer 319, may include immobile water. Immobile water, in some aspects, is water (e.g., fresh, salt, brine), that is not fit for human or animal consumption, or both. Immobile water, in some aspects, may be water that, by its motion through the layers 316 or 319 (or both), cannot reach the mobile water layer 314, terranean surface 302, or both, within 10,000 years or more (such as to 1,000,000 years). Below the mobile water layer 314, in this example implementation of hazardous material storage repository system 300, is an impermeable layer 316. The impermeable layer 316, in this example, may not allow mobile water to pass through. Thus, relative to the mobile water layer 314, the impermeable layer 316 may have low permeability, e.g., on the order of nanodarcy permeability. Additionally, in this example, the impermeable layer 316 may be a relatively non-ductile (i.e., brittle) geologic formation. One measure of non-ductility is brittleness, which is the ratio of compressive stress to tensile strength. In some examples, the brittleness of the impermeable layer 316 may be between about 20 MPa and 40 MPa. As shown in this example, the impermeable layer 316 is shallower (e.g., closer to the terranean surface 302) than the storage layer 319. In this example rock formations of which the impermeable layer 316 may be composed include, for example, certain kinds of sandstone, mudstone, clay, and slate that exhibit permeability and brittleness properties as described above. In alternative examples, the impermeable layer 316 may be deeper (e.g., further from the terranean surface 302) than the storage layer 319. In such alternative examples, the impermeable layer 316 may be composed of an igneous rock, such as granite. Below the impermeable layer 316 is the storage layer 319. The storage layer 319, in this example, may be chosen as the landing for the vertically undulating portion 310, which stores the hazardous material, for several reasons. Relative to the impermeable layer 316 or other layers, the storage layer 319 may be thick, e.g., between about 100 and 200 feet of total vertical thickness. Thickness of the storage layer 319 may allow for easier landing and directional drilling, thereby allowing the vertically undulating portion 310 to be readily emplaced within the storage layer 319 during constructions (e.g., drilling). If formed through an approximate horizontal center of the storage layer 319, the vertically undulating portion 310 may be surrounded by about 50 to 100 feet of the geologic formation that comprises the storage layer 319. Further, the storage layer 319 may also have only immobile water, e.g., due to a very low permeability of the layer 319 (e.g., on the order of milli- or nanodarcys). In addition, the storage layer 319 may have sufficient ductility, such that a brittleness of the rock formation that comprises the layer 319 is between about 3 MPa and 10 MPa. Examples of rock formations of which the storage layer 319 may be composed include: shale and anhydrite. Further, in some aspects, hazardous material may be stored below the storage layer, even in a permeable formation such as sandstone or limestone, if the storage layer is of sufficient geologic properties to isolate the permeable layer from the mobile water layer 314. In some examples implementations of the hazardous material storage repository system 300, the storage layer 319 (and/or the impermeable layer 316) is composed of shale. Shale, in some examples, may have properties that fit within those described above for the storage layer 319. For example, shale formations may be suitable for a long-term confinement of hazardous material (e.g., in the hazardous material canisters 326), and for their isolation from mobile water layer 314 (e.g., aquifers) and the terranean surface 302. Shale formations may be found relatively deep in the Earth, typically 3000 feet or greater, and placed in isolation below any fresh water aquifers. Other formations may include salt or other impermeable formation layer. Shale formations (or salt or other impermeable formation layers), for instance, may include geologic properties that enhance the long-term (e.g., thousands of years) isolation of material. Such properties, for instance, have been illustrated through the long term storage (e.g., tens of millions of years) of hydrocarbon fluids (e.g., gas, liquid, mixed phase fluid) without escape of such fluids into surrounding layers (e.g., mobile water layer 314). Indeed, shale has been shown to hold natural gas for millions of years or more, giving it a proven capability for long-term storage of hazardous material. Example shale formations (e.g., Marcellus, Eagle Ford, Barnett, and otherwise) has stratification that contains many redundant sealing layers that have been effective in preventing movement of water, oil, and gas for millions of years, lacks mobile water, and can be expected (e.g., based on geological considerations) to seal hazardous material (e.g., fluids or solids) for thousands of years after deposit. In some aspects, the formation of the storage layer 319 and/or the impermeable layer 316 may form a leakage barrier, or barrier layer to fluid leakage that may be determined, at least in part, by the evidence of the storage capacity of the layer for hydrocarbons or other fluids (e.g., carbon dioxide) for hundreds of years, thousands of years, tens of thousands of years, hundreds of thousands of years, or even millions of years. For example, the barrier layer of the storage layer 319 and/or impermeable layer 316 may be defined by a time constant for leakage of the hazardous material more than 10,000 years (such as between 10,000 years and 1,000,000 years) based on such evidence of hydrocarbon or other fluid storage. Shale (or salt or other impermeable layer) formations may also be at a suitable depth, e.g., between 3000 and 12,000 feet TVD. Such depths are typically below ground water aquifer (e.g., surface layer 312 and/or mobile water layer 314). Further, the presence of soluble elements in shale, including salt, and the absence of these same elements in aquifer layers, demonstrates a fluid isolation between shale and the aquifer layers. Another particular quality of shale that may advantageously lend itself to hazardous material storage is its clay content, which, in some aspects, provides a measure of ductility greater than that found in other, impermeable rock formations (e.g., impermeable layer 316). For example, shale may be stratified, made up of thinly alternating layers of clays (e.g., between about 20-30% clay by volume) and other minerals. Such a composition may make shale less brittle and, thus less susceptible to fracturing (e.g., naturally or otherwise) as compared to rock formations in the impermeable layer (e.g., dolomite or otherwise). For example, rock formations in the impermeable layer 316 may have suitable permeability for the long term storage of hazardous material, but are too brittle and commonly are fractured. Thus, such formations may not have sufficient sealing qualities (as evidenced through their geologic properties) for the long term storage of hazardous material. The present disclosure contemplates that there may be many other layers between or among the illustrated subterranean layers 312, 314, 316, and 319. For example, there may be repeating patterns (e.g., vertically), of one or more of the mobile water layer 314, impermeable layer 316, and storage layer 319. Further, in some instances, the storage layer 319 may be directly adjacent (e.g., vertically) the mobile water layer 314, i.e., without an intervening impermeable layer 316. In some examples, all or portions of the curved portion 308 and the vertically undulating portion 310 may be formed below the storage layer 319, such that the storage layer 319 (e.g., shale or other geologic formation with characteristics as described herein) is vertically positioned between the vertically undulating portion 310 and the mobile water layer 314. Although not illustrated in this particular example shown in FIG. 3, a self-healing layer (e.g., such as the self-healing layer 132) may be found below the terranean surface 302 and between, for example, the surface 302 and one or both of the impermeable layer 316 and the storage layer 319. In some aspects, the self-healing layer may comprise a geologic formation that can stop or impede a flow of hazardous material (whether in liquid, solid, or gaseous form) from a storage portion of the drillhole 304 to or toward the terranean surface 302. For example, during formation of the drillhole 304 (e.g., drilling), all are portions of the geologic formations of the layers 312, 314, 316, and 319, may be damaged, thereby affecting or changing their geologic characteristics (e.g., permeability). In certain aspects, the location of the drillhole 304 may be selected so as to be formed through the self-healing layer. For example, as shown, the drillhole 304 may be formed such that at least a portion of the vertical portion 306 of the drillhole 304 is formed to pass through the self-healing layer. In some aspects, the self-healing layer comprises a geologic formation that that does not sustain cracks for extended time durations even after being drilled therethrough. Examples of the geologic formation in the self-healing layer include clay or dolomite. Cracks in such rock formations tend to heal, that is, they disappear rapidly with time due to the relative ductility of the material, and the enormous pressures that occur underground from the weight of the overlying rock on the formation in the self-healing layer. In addition to providing a “healing mechanism” for cracks that occur due to the formation of the drillhole 304 (e.g., drilling or otherwise), the self-healing layer may also provide a barrier to natural faults and other cracks that otherwise could provide a pathway for hazardous material leakage (e.g., fluid or solid) from the storage region (e.g., in the vertically undulating portion 310) to the terranean surface 302, the mobile water layer 314, or both. As shown in this example, the vertically undulating portion 310 of the drillhole 304 includes a storage area 317 in a distal part of the portion 310 into which hazardous material may be retrievably placed for long-term storage. For example, as shown, a work string 324 (e.g., tubing, coiled tubing, wireline, or otherwise) may be extended into the cased drillhole 304 to place one or more (three shown but there may be more or less) hazardous material canisters 326 into long term, but in some aspects, retrievable, storage in the portion 310. For example, in the implementation shown in FIG. 3, the work string 324 may include a downhole tool 328 that couples to the canister 326, and with each trip into the drillhole 304, the downhole tool 328 may deposit a particular hazardous material canister 326 in the vertically undulating portion 310. The downhole tool 328 may couple to the canister 326 by, in some aspects, a threaded connection or other type of connection, such as a latched connection. In alternative aspects, the downhole tool 328 may couple to the canister 326 with an interlocking latch, such that rotation (or linear movement or electric or hydraulic switches) of the downhole tool 328 may latch to (or unlatch from) the canister 326. In alternative aspects, the downhole tool 324 may include one or more magnets (e.g., rare Earth magnets, electromagnets, a combination thereof, or otherwise) which attractingly couple to the canister 326. In some examples, the canister 326 may also include one or more magnets (e.g., rare Earth magnets, electromagnets, a combination thereof, or otherwise) of an opposite polarity as the magnets on the downhole tool 324. In some examples, the canister 326 may be made from or include a ferrous or other material attractable to the magnets of the downhole tool 324. As another example, each canister 326 may be positioned within the drillhole 304 by a drillhole tractor (e.g., on a wireline or otherwise), which may push or pull the canister into the vertically undulating portion 310 through motorized (e.g., electric) motion. As yet another example, each canister 326 may include or be mounted to rollers (e.g., wheels), so that the downhole tool 324 may push the canister 326 into the cased drillhole 304. In some example implementations, the canister 326, one or more of the drillhole casings 320 and 322, or both, may be coated with a friction-reducing coating prior to the deposit operation. For example, by applying a coating (e.g., petroleum-based product, resin, ceramic, or otherwise) to the canister 326 and/or drillhole casings, the canister 326 may be more easily moved through the cased drillhole 304 into the vertically undulating portion 310. In some aspects, only a portion of the drillhole casings may be coated. For example, in some aspects, the substantially vertical portion 306 may not be coated, but the curved portion 308 or the vertically undulating portion 310, or both, may be coated to facilitate easier deposit and retrieval of the canister 326. FIG. 3 also illustrates an example of a retrieval operation of hazardous material in the vertically undulating portion 310 of the drillhole 304. A retrieval operation may be the opposite of a deposit operation, such that the downhole tool 324 (e.g., a fishing tool) may be run into the drillhole 304, coupled to the last-deposited canister 326 (e.g., threadingly, latched, by magnet, or otherwise), and pull the canister 326 to the terranean surface 302. Multiple retrieval trips may be made by the downhole tool 324 in order to retrieve multiple canisters from the vertically undulating portion 310 of the drillhole 304. Each canister 326 may enclose hazardous material. Such hazardous material, in some examples, may be biological or chemical waste or other biological or chemical hazardous material. In some examples, the hazardous material may include nuclear material, such as spent nuclear fuel recovered from a nuclear reactor (e.g., commercial power or test reactor) or military nuclear material. For example, a gigawatt nuclear plant may produce 30 tons of spent nuclear fuel per year. The density of that fuel is typically close to 10 (10 gm/cm3=10 kg/liter), so that the volume for a year of nuclear waste is about 3 m3. Spent nuclear fuel, in the form of nuclear fuel pellets, may be taken from the reactor and not modified. Nuclear fuel pellet are solid, although they can contain and emit a variety of radioactive gases including tritium (13 year half-life), krypton-85 (10.8 year half-life), and carbon dioxide containing C-14 (5730 year half-life). In some aspects, the storage layer 319 should be able to contain any radioactive output (e.g., gases) within the layer 319, even if such output escapes the canisters 326. For example, the storage layer 319 may be selected based on diffusion times of radioactive output through the layer 319. For example, a minimum diffusion time of radioactive output escaping the storage layer 319 may be set at, for example, fifty times a half-life for any particular component of the nuclear fuel pellets. Fifty half-lives as a minimum diffusion time would reduce an amount of radioactive output by a factor of 1×10−15. As another example, setting a minimum diffusion time to thirty half-lives would reduce an amount of radioactive output by a factor of one billion. For example, plutonium-239 is often considered a dangerous waste product in spent nuclear fuel because of its long half-life of 24,100 years. For this isotope, 50 half-lives would be 1.2 million years. Plutonium-239 has low solubility in water, is not volatile, and as a solid is not capable of diffusion through a matrix of the rock formation that comprises the illustrated storage layer 319 (e.g., shale or other formation). The storage layer 319, for example comprised of shale, may offer the capability to have such isolation times (e.g., millions of years) as shown by the geological history of containing gaseous hydrocarbons (e.g., methane and otherwise) for several million years. In contrast, in conventional nuclear material storage methods, there was a danger that some plutonium might dissolve in a layer that comprised mobile ground water upon confinement escape. As further shown in FIG. 3, the storage canisters 326 may be positioned for long term storage in the vertically undulating portion 310, which, as shown, is coupled to the vertical portion 106 of the drillhole 104 through the curved portion 308. As illustrated, the curved portion 308 includes an upwardly directed portion angled toward the terranean surface 302. Further, as shown, the undulating portion 310 of the drillhole 304 includes several upwardly and downwardly (relative to the surface 302) inclined portions, thereby forming several peaks and valleys in the undulating portion 310. In some aspects, for example when there is radioactive hazardous material stored in the canisters 326, these inclinations of the curved portion 308 and undulating portion 310 may provide a further degree of safety and containment to prevent or impede the material, even if leaked from the canister 326, from reaching, e.g., the mobile water layer 314, the vertical portion 306 of the drillhole 304, the terranean surface 302, or a combination thereof. For example, radionuclides of concern in the hazardous material tend to be relatively buoyant or heavy (as compared to other components of the material). Buoyant radionuclides may be the greatest concern for leakage, since heavy elements and molecules tend to sink, and would not diffuse upward towards the terranean surface 302. Krypton gas, and particularly krypton-85, is a buoyant radioactive element that is heavier than air (as are most gases) but much lighter than water. Thus, should krypton-85 be introduced into a water bath, such gas would tend to float upward towards the terranean surface 302. Iodine, on the other hand, is denser than water, and would tend to diffuse downward if introduced into a water bath. By including the curved portion 308 of the drillhole 304 and the undulating portion 310, any such diffusion of radioactive material (e.g., even if leaked from a canister 326 and in the presence of water or other liquid in the drillhole 304 or otherwise) would be directed toward the vertically undulating portion 310, and more specifically, to peaks within the vertically undulating portion 310 and away from the curved portion 308 (and the vertical portion 306) of the drillhole 304. Thus, leaked hazardous material, even in a diffusible gas form, would not be offered a path (e.g., directly) to the terranean surface 302 (or the mobile water layer 314) through the vertical portion 306 of the drillhole 310. For instance, the leaked hazardous material (especially in gaseous form) would be directed and gathered at the peaks of the drillhole portion 310, or, generally, within the vertically undulating portion 310 of the drillhole 304. Alternative methods of depositing the canisters 326 into the inclined drillhole portion 310 may also be implemented. For instance, a fluid (e.g., liquid or gas) may be circulated through the drillhole 304 to fluidly push the canisters 326 into the inclined drillhole portion 310. In some example, each canister 326 may be fluidly pushed separately. In alternative aspects, two or more canisters 326 may be fluidly pushed, simultaneously, through the drillhole 304 for deposit into the vertically undulating portion 310. The fluid can be, in some cases, water. Other examples include a drilling mud or drilling foam. In some examples, a gas may be used to push the canisters 326 into the drillhole, such as air, argon, or nitrogen. In some aspects, the choice of fluid may depend at least in part on a viscosity of the fluid. For example, a fluid may be chosen with enough viscosity to impede the drop of the canister 326 into the substantially vertical portion 306. This resistance or impedance may provide a safety factor against a sudden drop of the canister 326. The fluid may also provide lubrication to reduce a sliding friction between the canister 326 and the casings 320 and 322. The canister 326 can be conveyed within a casing filled with a liquid of controlled viscosity, density, and lubricant qualities. The fluid-filled annulus between the inner diameter of the casings 320 and 322 and the outer diameter of the conveyed canister 326 represents an opening designed to dampen any high rate of canister motion, providing automatic passive protection in an unlikely decoupling of the conveyed canister 326. In some aspects, other techniques may be employed to facilitate deposit of the canister 326 into the vertically undulating portion 310. For example, one or more of the installed casings (e.g., casings 320 and 322) may have rails to guide the storage canister 326 into the drillhole 302 while reducing friction between the casings and the canister 326. The storage canister 326 and the casings (or the rails) may be made of materials that slide easily against one another. The casings may have a surface that is easily lubricated, or one that is self-lubricating when subjected to the weight of the storage canister 326. The fluid may also be used for retrieval of the canister 326. For example, in an example retrieval operation, a volume within the casings 320 and 322 may be filled with a compressed gas (e.g., air, nitrogen, argon, or otherwise). As the pressure increases at an end of the vertically undulating portion 310, the canisters 326 may be pushed toward the curved portion 308, and subsequently through the substantially vertical portion 306 to the terranean surface. In some aspects, the drillhole 304 may be formed for the primary purpose of long-term storage of hazardous materials. In alternative aspects, the drillhole 304 may have been previously formed for the primary purpose of hydrocarbon production (e.g., oil, gas). For example, storage layer 319 may be a hydrocarbon bearing formation from which hydrocarbons were produced into the drillhole 304 and to the terranean surface 302. In some aspects, the storage layer 319 may have been hydraulically fractured prior to hydrocarbon production. Further in some aspects, the production casing 322 may have been perforated prior to hydraulic fracturing. In such aspects, the production casing 322 may be patched (e.g., cemented) to repair any holes made from the perforating process prior to a deposit operation of hazardous material. In addition, any cracks or openings in the cement between the casing and the drillhole can also be filled at that time. For example, in the case of spent nuclear fuel as a hazardous material, the drillhole may be formed at a particular location, e.g., near a nuclear power plant, as a new drillhole provided that the location also includes an appropriate storage layer 319, such as a shale formation. Alternatively, an existing well that has already produced shale gas, or one that was abandoned as “dry,” (e.g., with sufficiently low organics that the gas in place is too low for commercial development), may be selected as the drillhole 304. In some aspects, prior hydraulic fracturing of the storage layer 319 through the drillhole 304 may make little difference in the hazardous material storage capability of the drillhole 304. But such a prior activity may also confirm the ability of the storage layer 319 to store gases and other fluids for millions of years. If, therefore, the hazardous material or output of the hazardous material (e.g., radioactive gasses or otherwise) were to escape from the canister 326 and enter the fractured formation of the storage layer 319, such fractures may allow that material to spread relatively rapidly over a distance comparable in size to that of the fractures. In some aspects, the drillhole 302 may have been drilled for a production of hydrocarbons, but production of such hydrocarbons had failed, e.g., because the storage layer 319 comprised a rock formation (e.g., shale or otherwise) that was too ductile and difficult to fracture for production, but was advantageously ductile for the long-term storage of hazardous material. FIG. 4A-4C are schematic illustrations of other example implementations of a hazardous material storage repository system according to the present disclosure. FIG. 4A shows hazardous material storage repository system 400, FIG. 4B shows hazardous material storage repository system 450, and FIG. 4C shows hazardous material storage repository system 480. Each of the systems 400, 450, and 480 include a substantially vertical drillhole (404, 454, and 484, respectively) drilled from a terranean surface (402, 452, and 482, respectively). Each substantially vertical drillhole (404, 454, 484) couples to (or continues into) a transition drillhole (406, 456, and 486, respectively) that is a curved or radiussed drillhole. Each transition drillhole (406, 456, and 486) then couples to (or continues into) an isolation drillhole (408, 458, and 488, respectively) that includes or comprises a hazardous material storage repository into which one or more hazardous material storage canisters (e.g., canisters 126) may be placed for long-term storage and, if necessary retrieved according to the present disclosure. As shown in FIG. 4A, the isolation drillhole 408 is a spiral drillhole that, at the point where it connects to the transition drillhole 406, starts to curve to the horizontal and simultaneously begins to curve to the side, i.e. in a horizontal direction. Once the spiral drillhole reaches its lowest point, it continues to curve in both directions, giving it a slight upward spiral. At that point the horizontal curve may be made a little bigger so that the curve does not intersect the vertical drillhole 404. Once the spiral drillhole begins to rise, a curved hazardous material storage repository section may commence. The storage section may continue until a highest point (e.g., point closest to the terranean surface 402), which is a dead-end trap (e.g., for escaped hazardous material solid, liquid, or gas). The rise of the spiral drillhole can be typically 3 degrees. In some aspects, the path of the spiral drillhole 408 can be down the axis of the spiral (that is, in the center of the spiraling circles) or displaced. Also, as shown in FIG. 4A, the vertical drillhole 404 is formed within the spiral drillhole 408. In other words, the spiral drillhole 408 may be formed symmetrically around the vertical drillhole 404. Turning briefly to FIG. 4C, the system 480 shows a spiral drillhole 488 similar to that of the spiral drillhole 408. However, spiral drillhole 488 is formed offset and to a side of the vertical drillhole 484. In some aspects, the spiral drillhole 488 can be formed offset of any side of the vertical drillhole 484. Turning to FIG. 4B, the system 450 includes a spiral drillhole 458 that is coupled to the transition drillhole 456 that turns from the vertical drillhole 454. Here, the spiral drillhole 458, rather than being oriented vertically (e.g., with an axis of rotation parallel of the vertical drillhole), is oriented horizontally (e.g., with an axis of rotation perpendicular to the vertical drillhole 454). At an end of or within the spiral drillhole 458 (or both) is a hazardous material storage section. In the implementations of systems 400, 450, and 480, a radius of curvature of the transition drillholes may be about 1000 feet. The circumference of each spiral in the spiral drillholes may be about 2π times the radius of curvature, or about 6,000 feet. Thus, each spiral in the spiral drillholes may contain a bit over one mile of storage area of hazardous material canisters. In some alternative aspects, the radius of curvature may be about 500 feet. Then, each spiral of the spiral drillholes may include about 0.5 miles of storage area of hazardous material canisters. If two miles of storage is desired then there may be four spirals for each spiral drillholes of this size. As shown in FIGS. 4A-4C, each of the systems 400, 450, and 480 include drillhole portions that serve as hazardous material storage areas and are directed vertically toward the terranean surface and away from an intersection between the transition drillhole of each system and the vertical drillhole of each section. Thus, any leaked hazardous material (e.g., such as radioactive waste gas) may be directed to such vertically-directed storage areas and away from the vertical drillholes. Each of the drillholes shown in FIGS. 4A-4C may be cased or uncased; the casing may serve as an additional layer of protection to prevent hazardous material from reaching mobile water. If casing is omitted, then angular changes to any drillhole can be more rapid with a constraint being the accommodation of movement of any canister therethrough. If there is casing, angular changes in direction for the drillholes may be done sufficiently slowly (as they are in standard directional drilling) that the casing can be pushed into the drillhole. Further, in some aspects, all or a portion of each of the illustrated isolation drillholes (408, 458, and 488) may be formed in or under an impermeable layer (as described in the present disclosure). In some aspects, implementations of a spiral drillhole may have a constant curvature around an axis of rotation. Alternative implementations of a spiral drillhole may have a gradually changing curvature, making the spirals in the spiral drillhole either tighter or less confined. Still additional implementations of a spiral drillhole may have the spirals changing in radius (making it tighter or less tight) but have little or no vertical rise (e.g., for situations in which it might be useful if the geologic layer in which the hazardous material storage section of the isolation drillholes is not very thick in the vertical dimension). FIG. 5A is a top view, and FIGS. 5B-5C are side views, of schematic illustrations of another example implementation of a hazardous material storage repository system 500. As shown, the system includes a vertical drillhole 504 formed from a terranean surface 502. The vertical drillhole 504 is coupled to or continues into a transition drillhole 506. The transition drillhole 506 is coupled to or turns into an isolation drillhole 508. In this example, the isolation drillhole 508 includes or comprise an undulating drillhole in which the undulations are substantially side-to-side. As shown in FIG. 5B, the isolation drillhole 508 rises toward the terranean surface 502 and vertically away from the transition drillhole 506 as it undulates side-to-side. As shown in FIG. 5C, alternatively, the isolation drillhole 508 stays in a plane substantially parallel to the terranean surface 502 as it undulates side-to-side. In some aspects, the spiral or undulating drillholes may be oriented without regard to the stress pattern of any gas or oil bearing layer in which they are formed. This is because the orientation need not take into account any fracturing of the drillhole as is the case for hydrocarbon production. Thus, drillhole geometers that are not oriented in the direction of the rock stress pattern, and are more compact, can be utilized. These drillholes may also have significant value in reducing the amount of terranean land under which the drillholes are formed. This may also reduce a cost of the land and of any mineral rights that must be bought to allow the hazardous material storage repository systems to be built. The drillholes are therefore determined not by the pattern of stresses in the rock, but primarily by the efficient and practical use of the available land. Each of the drillholes shown in FIGS. 5A-5C may be cased or uncased; the casing may serve as an additional layer of protection to prevent hazardous material from reaching mobile water. If casing is omitted, then angular changes to any drillhole can be more rapid with a constraint being the accommodation of movement of any canister therethrough. If there is casing, angular changes in direction for the drillholes may be done sufficiently slowly (as they are in standard directional drilling) that the casing can be pushed into the drillhole. Further, in some aspects, all or a portion of the isolation drillhole 508 may be formed in or under an impermeable layer (as described in the present disclosure). Referring generally to FIGS. 1A, 2, 3, 4A-4C, and 5A-5C, the example hazardous material storage repository systems (e.g., 100, 200, 300, 400, 450, 480, and 500) may provide for multiple layers of containment to ensure that a hazardous material (e.g., biological, chemical, nuclear) is sealingly stored in an appropriate subterranean layer. In some example implementations, there may be at least twelve layers of containment. In alternative implementations, a fewer or a greater number of containment layers may be employed. First, using spent nuclear fuel as an example hazardous material, the fuel pellets are taken from the reactor and not modified. They may be made from sintered uranium dioxide (UO2), a ceramic, and may remain solid and emit a variety of radioactive gases including tritium (13 year half-life), krypton-85 (10.8 year half-life), and carbon dioxide containing C-14 (5730 year half-life). Unless the pellets are exposed to extremely corrosive conditions or other effects that damage the multiple layers of containment, most of the radioisotopes (including the C-14, tritium or krypton-85) will be contained in the pellets. Second, the fuel pellets are surrounded by the zircaloy tubes of the fuel rods, just as in the reactor. As described, the tubes could be mounted in the original fuel assemblies, or removed from those assemblies for tighter packing. Third, the tubes are placed in the sealed housings of the hazardous material canister. The housing may be a unified structure or multi-panel structure, with the multiple panels (e.g., sides, top, bottom) mechanically fastened (e.g., screws, rivets, welds, and otherwise). Fourth, a material (e.g., solid or fluid) may fill the hazardous material canister to provide a further buffer between the material and the exterior of the canister. Fifth, the hazardous material canister(s) are positioned (as described above), in a drillhole that is lined with a steel or other sealing casing that extends, in some examples, throughout the entire drillhole (e.g., a substantially vertical portion, a radiussed portion, and a inclined portion). The casing is cemented in place, providing a relatively smooth surface (e.g., as compared to the drillhole wall) for the hazardous material canister to be moved through, thereby reducing the possibility of a leak or break during deposit or retrieval. Sixth, the cement that holds or helps hold the casing in place, may also provide a sealing layer to contain the hazardous material should it escape the canister. Seventh, the hazardous material canister is stored in a portion of the drillhole (e.g., the inclined portion) that is positioned within a thick (e.g., 100-200 feet) seam of a rock formation that comprises a storage layer. The storage layer may be chosen due at least in part to the geologic properties of the rock formation (e.g., only immobile water, low permeability, thick, appropriate ductility or non-brittleness). For example, in the case of shale as the rock formation of the storage layer, this type of rock may offers a level of containment since it is known that shale has been a seal for hydrocarbon gas for millions of years. The shale may contain brine, but that brine is demonstrably immobile, and not in communication with surface fresh water. Eighth, in some aspects, the rock formation of the storage layer may have other unique geological properties that offer another level of containment. For example, shale rock often contains reactive components, such as iron sulfide, that reduce the likelihood that hazardous materials (e.g., spent nuclear fuel and its radioactive output) can migrate through the storage layer without reacting in ways that reduce the diffusion rate of such output even further. Further, the storage layer may include components, such as clay and organic matter, that typically have extremely low diffusivity. For example, shale may be stratified and composed of thinly alternating layers of clays and other minerals. Such a stratification of a rock formation in the storage layer, such as shale, may offer this additional layer of containment. Ninth, the storage layer may be located deeper than, and under, an impermeable layer, which separates the storage layer (e.g., vertically) from a mobile water layer. Tenth, the storage layer may be selected based on a depth (e.g., 3000 to 12,000 ft.) of such a layer within the subterranean layers. Such depths are typically far below any layers that contain mobile water, and thus, the sheer depth of the storage layer provides an additional layer of containment. Eleventh, example implementations of the hazardous material storage repository system of the present disclosure facilitate monitoring of the stored hazardous material. For example, if monitored data indicates a leak or otherwise of the hazardous material (e.g., change in temperature, radioactivity, or otherwise), or even tampering or intrusion of the canister, the hazardous material canister may be retrieved for repair or inspection. Twelfth, the one or more hazardous material canisters may be retrievable for periodic inspection, conditioning, or repair, as necessary (e.g., with or without monitoring). Thus, any problem with the canisters may be addressed without allowing hazardous material to leak or escape from the canisters unabated. Thirteenth, even if hazardous material escaped from the canisters and no impermeable layer was located between the leaked hazardous material and the terranean surface, the leaked hazardous material may be contained within the drillhole at a location that has no upward path to the surface or to aquifers (e.g., mobile water layers) or to other zones that would be considered hazardous to humans. For example, the location, which may be a dead end of an inclined drillhole, a J-section drillhole, or peaks of a vertically undulating drillhole, may have no direct upward (e.g., toward the surface) path to a vertical portion of the drillhole. A number of implementations have been described. Nevertheless, it will be understood that various modifications may be made without departing from the spirit and scope of the disclosure. For example, example operations, methods, or processes described herein may include more steps or fewer steps than those described. Further, the steps in such example operations, methods, or processes may be performed in different successions than that described or illustrated in the figures. Accordingly, other implementations are within the scope of the following claims.
	description	Referring now to FIG. 1, the essential principle of operation for the devices of the present invention is illustrated. FIG. 1 is a conceptual cross section view of a single neutron detector comprising a means for detecting neutrons 10 stacked on an absorbing layer 11. The absorbing layer 11, being composed of a first material that absorbs protons, such as titanium, is stacked on a hydrogenous substrate 12. Hydrogenous substrate 12 is composed of a second material having hydrogen atoms interacting with an unknown source of neutrons, indicated by box 13. When a single neutron detector is placed in a field of a neutron spectrum, the incident neutrons, indicated by arrow 14, from suspected neutron source 13 interact with hydrogen atoms within hydrogenous substrate 12. This interaction produces proton recoils that travel in fairly straight lines, one of which is indicated by arrow 15, through the absorber layer 11 and the detector means 10. Scattered neutrons, indicated by arrow 16, are deflected away from the hydrogenous substrate 12. Detector means 10 is connected to a data processing means, indicated by box 17, and a ground 18. The data processing means 17 includes a means for proton distribution. Using several detector means 10 with each absorbing layer 11 having a different thickness allows protons with energies and corresponding ranges greater than the thickness of a particular absorbing layer 11 to reach detector means 10 and produce proton counts. The amount of absorber layers 11 and their thickness can be selected to correspond to ranges of protons from a low value for 1 MeV and larger thicknesses of 250 MeV. Hydrogenous substrate 12 converts part of the kinetic neutron energy to energy of the recoil protons 15 and the detector means 10 detects protons passing through the absorbing layer 12. This approach is demonstrated by considering the energy transfer behavior of neutrons and protons. The maximum energy a neutron of energy En can transfer to a proton Ep (max) equals En (1,2). For this example, assume an absorbing layer 11 thickness of d. For monoenergetic neutrons (En), the number of recoil protons reaching detecting means 10 and producing proton counts decreases as energy En decreases. The number of protons will eventually equal zero when the range of maximum energy recoil protons becomes smaller than d. Recoil particles due to elastic scattering do occur in the higher atomic number non-hydrogenous absorber but, except for very high En, they do not contribute to the counts due to their small range and the unfavorable quantum energy transfer in elastic scattering. Having a system with K units, each with a different d and exposing them to a neutron spectrum, one obtains data which consist of K counts or count rate values Ci(di) i=1, 2, . . . K where for dixe2x88x921 less than di less than di+1, Cixe2x88x921 (di greater than Ci greater than Ci+1. From these numbers one can unfold the incident spectrum of neutrons. The detector means 10 can be of any shape or configuration and can be any type of solid state device. The inventors herein have employed a depleted n/p diode used to measure alpha particles, which was relatively insensitive to beta particles because of their low LET (Linear Energy Transfer) values as a detector means 10. Spectroscopic grade detectors are not required for this device since only event counting is required and data describing the energy spectrum are not needed. In considering the thicknesses of absorbing layers 11 and the ranges of protons to be measured, an energy range of 1 to 250 MeV was selected to match the expected neutron spectrum distribution. One solution to achieve this objective is to fabricate an instrument that converts a distribution of neutrons to one of recoil protons, which are charged particles that can be easily counted. By employing 12 detector means 10 within a given chamber, the recoil protons are essentially sorted into 12 bins where they can be readily counted. Said absorber layers 11 can be constructed of aluminum for detecting the lower energy levels or tantalum for the higher values. The hydrogenous substrate 12 for each detector means 10 could be constructed of polyethylene. The data processing means 17 and its means for proton distribution provides a hitherto unavailable capability to determine a proton distribution pattern to construct a neutron spectrum indicating the spectrum of neutrons from an unknown source of neutrons 13. In operation, results of a spectral measurement are a set of pairs from the detector means 10 and the absorbing layer 11 that allows protons with energies and corresponding ranges greater than the absorbing layer 11""s thickness to reach the detector means 10 and produce proton recoil counts. One data processing means 17 successfully employed by the present inventors is a 3-dimensional Monte Carlo Adjoint Transport code, NOVICE, which is described in Jordan, T., xe2x80x9cNovice, A Radiation Transport and Shielding Codexe2x80x9d, Experimental and Mathematical Physics Consultant, Report EMP. L 82.001, January 1982. FIG. 2 is a chart showing plots of counts in the detector versus proton energy with different thicknesses indicated as a parameter on the curves, and these results were obtained using the NOVICE program and a flat spectrometer 20 depicted in FIG. 6, which will be described below. The FIG. 2 plots are counts in the detector versus proton energy with the aluminum and tantalum thicknesses indicated as a parameter on the curves. In this preliminary assessment of the feasibility of neutron monitor with multiple neutron detectors, an incident neutron spectrum and the subsequent unfolding software were not included in the code""s run. The proton recoil spectrum was assumed to exist in the converter material of hydrogenous substrate 12. The separation or resolution of proton energy shown in FIG. 2 provides useful information about detecting 12 ranges of neutron energy. The flat configuration of monitor 20, depicted in FIG. 6, along with the use of tantalum for the absorber layers 11 and for the chamber 21 make it too heavy for spacecraft or other airborne applications. Using a data processing device with the NOVICE computer software to analyze the monitor revealed other more useful potential configurations for neutron spectrometers, which were modeled and analyzed by the computer. One configuration suggested by the FIG. 2 NOVICE results is a pentagon dodecahedron, which allows for a full measurement range because of its 12 surfaces, each supporting a detector-absorber pair with different absorber layer thicknesses. FIGS. 3A and 3B, are perspective drawings depicting a detector means 41 stacked on a pentagonal absorbing layer 42 and a dodecahedron neutron spectrometer monitor 40, respectively. Referring now to FIG. 3A, which depicts a perspective view of a neutron detector comprising a detector means 41 stacked on an absorbing layer 42. Absorbing layer 42 is composed of a first material that absorbs protons, such as aluminum in this embodiment, or tantalum or titanium in other embodiments. By placing this assembly on an appropriate hydrogenous substrate, a neutron detector is provided. Referring now to FIG. 3B, decahedron neutron spectrometer monitor 40 is depicted with 11 of 12 of the absorbing layers 42 with varying thicknesses stacked on a surface facet of a solid dodecahedron substrate 43, which provides the hydrogenous substrate. Dodecahedron substrate 43 is shown partially exposed without one absorbing layer for illustrative purposes. FIG. 4 is a front view drawing of the dodecahedron neutron spectrometer monitor 40 with all absorbing layers 51-62, respectively, covering each of the 12 facets of substrate 43 and representative dimensions. For the sake of clarity, only one detector means 42 is shown stacked on absorbing layer 54, with 11 other detector means 42 for the other 11 absorbing layers 51-53 and 55-62, respectively, not shown. Each of the 12 absorbing layers 51-62 are constructed with a varying thickness and are stacked on a surface facet of the solid dodecahedron substrate 43. Substrate 43 is composed of a hydrogenous material, such as polyethylene, having hydrogen atoms and functions as a neutron converter when interacting with said absorbing layers 51-62 in the presence of an unknown energy distribution, indicated by box 44, which emits incident neutrons, indicated by arrow 63. In operation, said hydrogenous substrate 43 converts said neutrons to recoil protons and each of said detector means 42 detects recoil protons passing through each absorbing layer 51-62, respectively. Each absorbing layer 51-62, respectively has a different thickness, as depicted in FIG. 5, to absorb neutron energies from 1 to 250 MeV. Returning now to FIG. 4, the hydrogenous substrate 43 is housed in a concentrically hollow spherical chamber, indicated by broken line 45. Each detector means 42 is coupled to a means for data processing, indicated by box 46, outside the spherical chamber 45, which provides a count of recoil protons to a means for proton distribution, not shown, residing within said data processing means 46. The means for proton distribution determines a proton distribution pattern to construct a neutron spectrum pattern indicating the spectrum of neutrons from said suspected source of neutron radiation 44. FIG. 4 also includes representative dimensions. Each absorbing layer 51-62 is pentagonally shaped in this embodiment, with each side 2.03 cm in length. Each of said detector means 42 are circular and 0.5xe2x80x3 wide and 0.015xe2x80x3 thick. Covered hydrogenous substrate 43 is 4.47 cm in height and housed concentrically within hollow spherical chamber 45. Hydrogenous substrate 43 was fabricated from a solid block of Lucite(trademark) The hollow spherical chamber 45 is composed of titanium in this embodiment with an inner diameter of 10.8 cm and a wall thickness of 2.5 cm. Each of said 12 absorbing layers 51-62 is composed of aluminum in this embodiment with a varying thickness. In another embodiment with titanium absorbing layers, the thickness of the absorbing layers varied from 0.00105 cm to 2.4217 cm, as described in Table I below. The absorbing layers 51-62 may also be composed of tantalum or titanium. Detector means 42 can be constructed from a depleted n/p diode. It should be understood to those skilled in the art that these dimensions are merely representative and numerous other choices of dimensions are possible. FIG. 5 is a perspective drawing of hydrogenous substrate 43, sing like numerals for similar structural elements, illustrating number of absorbing layers with a varying thickness. In this drawing, covered hydrogenous substrate 43 is shown removed from the hollow spherical shell 45 to better illustrate each absorbing layer having a different thickness. Referring back to FIG. 2, which is the chart showing plots of counts in the detector versus proton energy with different thicknesses indicated as a parameter on the curves from the NOVICE program. Those plots from the FIG. 6 flat spectrometer 20, which will be described shortly, are based on using aluminum and tantalum as absorber material. These results suggested using titanium as the preferred absorber material for the FIG. 4 absorbing layers 51-62 for all energy levels, because titanium is lighter than tantalum and its neutrons do not generate nuclear interactions. Only elastic scattering takes place. The proton energy resolution from this embodiment is also relatively good. The FIG. 2 results also indicate that aluminum absorbers produced a slightly better energy resolution for the lower range of energies, 1 to 10 MeV. The size of this dodecahedron configuration is small and light in weight and very practical for a spacecraft application. In order to insure that an unknown neutron spectrum has an isotropic distribution, the spectrometer 40 can also be located at the center of a titanium sphere with a diameter of 3 inches. FIG. 6 is a perspective conceptual drawing of the flat embodiment of the present invention""s neutron spectrometer monitor 70. Monitor 70 comprises a group of the FIG. 1 neutron detector means 10 arranged in a chamber 71. As described above, having several detector means 10 stacked onto absorbing layers, not shown, each having a different thickness, allows protons with energies and corresponding ranges greater than the thickness of each absorbing layer to reach the detector means 10 and produce proton counts. FIG. 6 depicts 12 detector means 10 which correspond to 12 energy bins and thus detect protons with ranges corresponding to energies from 1 MeV up to 250 MeV. The floor of chamber 71 serves as the hydrogenous substrate. Monitor 70 is placed in proximity to an unknown source of neutrons, shown as box 76. Detecting means 10 is coupled to a means for data processing, indicated by box 77, and provides a separate count of recoil protons for each different thickness employed in the absorbing layers. The data processing means 77 transmits the count of recoil protons to a means for proton distribution, not shown, residing within the data processing means 77. The means for proton distribution determines a proton distribution pattern to construct a neutron spectrum pattern indicating the spectrum of neutrons from the suspected concentration of neutrons 76. Bulkhead output connector 72 on the chamber 71 allows correction of voltage to the detector as well as correction of output counts to counting instruments. In the flat configuration, said chamber 71 is shown in a rectangular shape, and its walls 78, lid, not shown, and unit compartments 79 can be composed of-tantalum. Each detector means 10 in the egg-crate-like structure is numbered 1xe2x80x2-12xe2x80x2, respectively, to correspond with readings shown in the FIG. 2 chart. Detector means 7xe2x80x2 is depicted with representative dimensions of 2 cm in width and 2 cm in length. A gap 80 between detector means 11xe2x80x2 and 12xe2x80x2 is 0.471 cm. The thickness of each wall 78 is 1 cm and its height is about 3 cm. The chamber 71 is depicted as 15 cm in length and 5.41 cm in width. These dimensions are merely representative and numerous other choices of dimensions are possible, however, it is critical that each absorber layer is constructed with a different thickness according to the minimum and maximum energies of neutrons in the spectrum. Similarly, the materials used for constructing the absorber layers, detector means 10 and chamber 71 can also be varied according to the minimum and maximum energies of neutrons in the spectrum. It is to be understood that details concerning materials, shapes and dimensions are merely illustrative, and that other combinations of materials, shapes and dimensions can also be advantageously employed and are considered to be within the contemplation of the present invention. We also wish it to be understood that we do not desire to be limited to the exact details of construction shown and described. It will be apparent that various structural modifications may be made without departing from the spirit of the invention and the scope of the appended claims.
	summary
	claims	1. A laser-driven particle beam irradiation apparatus comprising:a particle beam generator irradiating a target with pulsed laser light to emit a laser-driven particle ray;a beam converging unit forming a transportation path which guides the emitted laser-driven particle ray to an object to be irradiated and spatially converging the laser-driven particle ray;an energy selector selecting an energy and an energy width of the laser-driven particle ray;an irradiation port causing the laser-driven particle ray to scan the object to be irradiated to adjust an irradiation position in the object; andan irradiation controller controlling operation of the particle beam generator, the beam converging unit, the energy selector, and the irradiation port, whereinthe beam converging unit generates a magnetic field on a trajectory of the laser-driven particle ray and converging the laser-driven particle ray by the magnetic field, the magnetic field forcing divergence components of the laser-driven particle ray that go away from a center of the trajectory back to the center of the trajectory, andthe beam converging unit is provided between the particle beam generator and the energy selector. 2. A laser-driven particle beam irradiation apparatus comprising:a particle beam generator irradiating a target with pulsed laser light to emit a laser-driven particle ray;a beam converging unit forming a transportation path which guides the emitted laser-driven particle ray to an object to be irradiated and spatially converging the laser-driven particle ray;an energy selector selecting an energy and an energy width of the laser-driven particle ray;an irradiation port causing the laser-driven particle ray to scan the object to be irradiated to adjust an irradiation position in the object;an irradiation controller controlling operation of the particle beam generator, the beam converging unit, the energy selector, and the irradiation port; andan energy distribution converging unit forming the transportation path of the laser-driven particle ray and converging an energy distribution of the laser-driven particle ray through the transportation path to provide a peak at a particular energy, whereinthe beam converging unit generates a magnetic field on a trajectory of the laser-driven particle ray and converging the laser-driven particle ray by the magnetic field, the magnetic field forcing divergence components of the laser-driven particle ray that go away from a center of the trajectory back to the center of the trajectory, andthe energy distribution converging unit includes a phase rotation cavity unit forming a transportation path of the laser-driven particle ray and, under application of a high-frequency voltage, generating in the transportation path a high-frequency electric field in which a state in which protons in a bunch are accelerated and a state in which protons in a bunch are decelerated appear to converge the energy distribution of the laser-driven proton ray to a particular energy, and wherein the irradiation controller adjusts the phase of the high-frequency voltage to be applied to the phase rotation cavity unit to adjust the position of the energy peak of the energy distribution of the laser-driven particle ray. 3. The laser-driven particle beam irradiation apparatus according to claim 2, wherein the phase rotation cavity unit of the energy distribution converging unit includes an outer cavity forming the transportation path of the laser-driven particle ray and a plurality of inner cavities which are spaced in a row in the outer cavity and to which a high-frequency voltage is applied, wherein a high-frequency electric field is formed in a gap between adjacent inner cavities to converge the energy distribution of a proton beam around the energy of protons that enter the gap at a timing of being synchronized with the phase of the high-frequency voltage applied to the inner cavities among the protons in a bunch in the outer cavity. 4. The laser-driven particle beam irradiation apparatus according to claim 3, wherein the irradiation controller applies a pulse width compressing voltage to the inner cavities of the energy distribution converging unit to generate a high-frequency electric field in the gap between adjacent inner cavities, the pulse width compressing voltage being defined as V > E 0 ⁢ β 0 2 ⁢ γ 0 2 q · 1 - m 2 ⁢ c 4 / E 0 2 1 - m 2 ⁢ c 4 / E 0 2 + fL / c wherein f is the frequency of the high-frequency voltage to be applied to the inner cavities, L is the distance from a laser-driven particle ray emission point in the target to the gap between adjacent inner cavities, β0 and γ0 are Lorentz factors, E0 is the total energy of the laser-driven particle ray, c is the speed of light, m is the mass of the laser-driven-particle ray, and q is the charge of the laser-driven particle ray. 5. A laser-driven particle beam irradiation method, comprising:a particle beam generating step of irradiating a target with pulsed laser light to extract a laser-driven particle ray;a beam converging step of spatially converging the laser-driven particle ray;an energy selecting step of selecting an energy and an energy width of the laser-driven particle ray according to a depth of an irradiation position set in an object to be irradiated;an irradiation step of adjusting the irradiation position of the laser-driven particle ray in the object to be irradiated; anda pulse width compressing step of reducing the pulse width of the laser-driven particle ray,wherein, in the beam converging step, a magnetic field forcing divergence components of the laser-driven particle ray that go away from a center of the trajectory of the laser-driven particle ray back to the center of the trajectory is generated on the trajectory and the laser-driven particle ray is converged by the magnetic field. 6. The laser-driven particle beam irradiation method according to claim 5, wherein, in the pulse width compressing step, a high-frequency electric field induced by a pulse width compressing voltage is generated and the laser-driven particle ray is guided to and passed through the high-frequency electric field to reduce the pulse width of the laser-driven particle ray, the pulse width compressing voltage being defined as V > E 0 ⁢ β 0 2 ⁢ γ 0 2 q · 1 - m 2 ⁢ c 4 / E 0 2 1 - m 2 ⁢ c 4 / E 0 2 + fL / c wherein f is the frequency of the high-frequency voltage, L is the distance from a laser-driven particle ray emission point, β0 and γ0 are Lorentz factors, E0 is the total energy of the laser-driven particle ray, c is the speed of light, m is the mass of the laser-driven-particle ray, and q is the charge of the laser-driven particle ray.
	abstract	An electron beam exposure apparatus comprising: column 1 for irradiating an electron beam to wafer 10 serving as a sample; sample chamber 3 having vacuum pump 40 as a vacuum exhaustion unit for controlling the internal unit to a vacuum atmosphere; stage 4A arranged in the sample chamber 3 for holding and moving the wafer 10; and first mounting 5A for elastically supporting the column 1 with respect to the sample chamber 3.
	abstract	A transmission electron microscope (TEM) specimen and a method of manufacturing the specimen are provided. The specimen comprises an analysis point. The specimen is formed by forming a dimple at a surface portion of the preliminary specimen, and ion milling the preliminary specimen having the dimple.
	abstract	A method to assess and predict pressurized water stress corrosion cracking in operational nuclear power plants and the effect of adding zinc compounds into a reactor coolant system of the nuclear power plant.
	claims	1. A shroud segment for a nuclear reactor vessel, the shroud segment comprising:an elongated inner shell;an elongated outer shell;a plurality of elongated intermediate shells disposed between the inner and outer shells;the inner shell, outer shell, and intermediate shells being radially spaced apart forming a plurality of annular cavities for holding water;a top closure plate attached to the top of the shroud segment; anda bottom closure plate attached to the bottom of the shroud segment;wherein the top and bottom closure plates are configured for detachable coupling to adjoining shroud segments to form a stacked array of shroud segments;a mounting clamp pivotably attached to the bottom closure plate, the mounting clamp movable between an open unlocked position and a closed locked position; andan elastically deformable and arcuately shaped seismic restraint fixedly attached to the shroud segment;wherein the seismic restraint is a leaf spring fixedly attached to the clamp and pivotable with the clamp. 2. The shroud segment of claim 1, wherein the annular cavities extend from the top closure plate to the bottom closure plate. 3. The shroud segment of claim 1, wherein the annular cavities are fluidly interconnected by a plurality of drain holes extending through the intermediate shells. 4. The shroud segment of claim 3, wherein the drain holes are radially staggered between adjacent intermediate shells. 5. The shroud segment of claim 3, wherein the outer shell includes drains holes to fluidly connect the annular cavities to an exterior of the shell segment. 6. The shroud segment of claim 1, wherein the intermediate shells have a length that is coextensive with the length of the shroud. 7. The shroud segment of claim 1, wherein the mounting clamp is attached to a radially extending mounting lug on the bottom closure plate. 8. The shroud segment of claim 1, wherein the mounting clamp includes a recess configured to receive a mounting lug on a top closure plate of a second shroud segment. 9. The shroud segment of claim 1, wherein the mounting clamp includes a locking fastener for holding the mounting clamp in the closed locked position. 10. A shroud segment for a nuclear reactor vessel, the shroud segment comprising:an elongated inner shell;an elongated outer shell;a plurality of elongated intermediate shells disposed between the inner and outer shells;the inner shell, outer shell, and intermediate shells being radially spaced apart forming a plurality of annular cavities for holding water;a top closure plate attached to the top of the shroud segment;a bottom closure plate attached to the bottom of the shroud segment;wherein the top and bottom closure plates are configured for detachable coupling to adjoining shroud segments to form a stacked array of shroud segments;a mounting clamp pivotably attached to the bottom closure plate, the mounting clamp movable between an open unlocked position and a closed locked position; andan elastically deformable and arcuately shaped seismic restraint fixedly attached to the shroud segment;wherein the seismic restraint defines an outward facing concave recess, the seismic restraint configured to engage an interior of the reactor vessel when the shroud segment is mounted therein. 11. The shroud segment according to claim 1, wherein the intermediate shells each have a length that is coextensive with respective lengths of the inner and outer shells, and wherein the shroud segment forms a riser region inside the shroud segment. 12. The shroud segment of claim 11, wherein the inner, outer, and intermediate shells of the shroud segment are seal welded to the top and bottom closure plates collectively forming a self-supporting shroud segment structure. 13. The shroud segment of claim 1, wherein the top closure plate includes radially extending lifting lugs configured for rigging. 14. A shroud segment for a nuclear reactor vessel, the shroud segment comprising:an elongated inner shell;an elongated outer shell;a plurality of elongated intermediate shells disposed between the inner and outer shells;the inner shell, outer shell, and intermediate shells being radially spaced apart forming a plurality of annular cavities between the inner shell, outer shell, and intermediate shells configured for holding water;a top closure plate attached to the top of the shroud segment; anda bottom closure plate attached to the bottom of the shroud segment;wherein the annular cavities are fluidly interconnected by a plurality of drain holes formed through the outer shell and intermediate shells, the drain holes in the outer shell spaced longitudinally apart along a length of the outer shell, and the drain holes in the intermediate shells being spaced longitudinally apart along a respective length of each intermediate shell;the top and bottom closure plates configured for detachable coupling to adjoining shroud segments to form a stacked array of shroud segments;a mounting clamp configured to detachably couple to a top closure plate of an adjoining shroud segment;the mounting clamp pivotably coupled to the bottom closure plate via a pivot pin extending through the bottom closure plate and defining a stationary pivot axis, the mounting clamp movable about the pivot axis between an open unlocked position and a closed locked position; anda locking fastener rotatably mounted to the bottom closure plate and operable to retain the mounting claim in the closed locked position. 15. The shroud segment of claim 14, wherein the inner shell does not include drain holes. 16. A shroud segment for a nuclear reactor vessel, the shroud segment comprising:an elongated inner shell;an elongated outer shell;a plurality of radially spaced apart elongated intermediate shells disposed between the inner and outer shells;a top closure plate attached to the top of the shroud segment and having an outer circumferentially-extending peripheral edge which is substantially flush in radial direction with the outer shell;a bottom closure plate attached to the bottom of the shroud segment and having an outer circumferentially-extending peripheral edge which is substantially flush in radial direction with the outer shell;wherein the top and bottom closure plates each comprise a plurality of mounting tabs protruding radially outwards beyond their respective peripheral edges, the mounting tabs being configured for coupling to mounting tabs of adjoining shroud segments to form a stacked array of shroud segments;an outer annular cavity formed between the intermediate shells and the outer shell;an inner annular cavity formed between the intermediate shells and the inner shell;a plurality of intermediate annular cavities formed between the intermediate shells; anda mounting clamp pivotably attached to each of the mounting tabs of the bottom closure plate, the mounting clamps movable between an open unlocked position and a closed locked position;wherein the inner, outer, and intermediate shells of each shroud segment is seal welded to their respective top and bottom closure plates collectively forming a self-supporting shroud segment structure, each shroud segment detachably coupled to an adjoining shroud segment.
	summary
	summary
	summary
	claims	1. A method for securing and confining a gasket plane of a vessel of a nuclear reactor during an operation for closing the vessel in order to prevent a migration of solid material particles towards an inside of the vessel, the gasket plane being formed by a ring-shaped recess on a flange of the vessel and by a ring shaped shoulder on a lid of the vessel, the ring shaped recess on the flange and the shoulder facing each other and each including a vertical wall delimiting between them a gap after laying the lid on the vessel flange, the shoulder including a horizontal wall provided with two O-ring seal gaskets wherein after having disassembled the lid and laying the lid on a support, the method comprises the following steps:placing around the shoulder of the lid a ring-shaped gasket formed by a metal strip maintained on the vertical wall of the shoulder by self-maintaining members firmly attached to the metal strip;placing the lid bearing the ring-shaped gasket above the vessel by positioning it at a determined height and performing a cleanliness inspection of the gasket plane;lowering and laying the lid on the vessel so that end side edges of the metal strip come into contact with respective horizontal walls of the lid and of the vessel flange;attaching the lid on the vessel flange; andleaving the ring-shaped gasket in place during a whole operating cycle of the reactor. 2. A ring-shaped gasket for securing and confining a gasket plane of a vessel of a nuclear reactor during an operation for closing the vessel in order to prevent a migration of solid material particles towards the inside of the vessel, the gasket plane being formed by a ring-shaped recess made on a flange of the vessel and by a ring-shaped shoulder made on a vessel lid, the recess and the shoulder facing each other and each including a vertical wall delimiting between the recess and the shoulder a gap, the shoulder including a horizontal wall provided with two O-ring seal gaskets, the gasket comprising:a metal strip including two end side edges, each edge capable of bearing upon respective horizontal walls of the lid and of the vessel flange, and self-maintaining members of the ring-shaped gasket bearing upon the vertical wall of the shoulder of the lid. 3. The ring-shaped gasket according to claim 2 wherein the self-maintaining members include at least two opposite metal tabs attached on an outer face of the metal strip, the metal tabs facing the vertical wall of the shoulder of the lid. 4. The ring-shaped gasket according to claim 3 wherein each metal tab is attached on the outer face of the strip by welding and has a thickness between 0.2 mm and 0.5mm. 5. The ring shaped gasket according to claim 4 wherein the metal tab thickness is 0.25 mm. 6. The ring shaped gasket according to claim 4 wherein the metal tab thickness is 0.5 mm. 7. The ring-shaped gasket according to claim 2 wherein the self-maintaining members include several metal tabs uniformly distributed over a perimeter of the metal strip and attached on an outer face of the metal strip facing the vertical wall of the shoulder of the lid. 8. The ring-shaped gasket according to claim 2 wherein the metal strip of the ring-shaped gasket has the shape of a flexible ring with a thickness between 0.3 mm and 1 mm. 9. The ring-shaped gasket according to claim 2 wherein the metal strip of the ring-shaped gasket has an L-shaped cross-section including two walls, an upper wall and a lower wall, the upper wall and the lower wall forming an angle between them, the upper wall having an end side edge bearing upon the horizontal wall of the lid and the lower wall having a smaller width than the upper wall, the lower wall having an end side edge bearing upon the horizontal wall of the vessel flange. 10. The ring-shaped gasket according to claim 9 wherein the self-maintaining members include metal tabs attached on an outer face of the upper wall of the metal strip. 11. The ring-shaped gasket according to claim 9 wherein the self-maintaining members include metal tabs attached on an outer face of the lower wall of the strip.
	abstract	A method of removing, from a fusion power reactor, a tile that comprises a tile-support tube, which is attached to a back portion of the tile and which comprises a coolant channel that is configured in a horizontal orientation, comprises rotating the tile, which is installed in a locked orientation in a manifold channel of a first wall of the fusion power reactor, until the tile is in an install/remove orientation. The method further comprises grasping, with a removal tool, the tile-support tube. The method additionally comprises lifting the tile away from the first wall of the fusion power reactor with the removal tool such that the tile is completely removed from the manifold channel of the first wall of the fusion power reactor.
058898344	claims	1. Blade collimator for radiation therapy, comprising a support device, two opposing sets of blades mounted in one layer in said support device, each blade capable of reciprocable movement in said support device and having a dimension perpendicular to the movement, each set of blades comprising a central blade adjacent to the blades of the other set, an exterior blade, and a blade interposed between the central blade and the exterior blade, the dimension of the exterior blade being greater than the dimension of the interposed blade and the dimension of the interposed blade being greater than the dimension of the central blade. 3 blades 4 mm in width, 3blades 3 mm in width and 7 blades 2 mm in width. the blades are moved by electric motors, and the electric motors each have a drive shaft and a drive threaded rod, whereby the blades are moved by electric motors, and the electric motors each have a drive shaft and a drive threaded rod, whereby the blades are moved by electric motors, and the electric motors each have a drive shaft and a drive threaded rod, whereby 2. Blade collimator according to claim 1 in which the dimensions of several adjacent blades of each set are equal. 3. Blade collimator according to claim 1 in which the blades are leak-proof against radiation by means of reciprocal interlocking of teeth. 4. Blade collimator according to claim 1 in which the dimensions of the respective blades are symmetrical to an axis of symmetry in a direction of the movement. 5. Blade collimator according to claim 1 in which mechanisms for moving the blades separately or in sets are mounted on the blade support, and the mechanisms are activated by at least one of electrical and mechanical devices. 6. Blade collimator according to claim 5 in which the devices are coupled to a control unit that controls the activation and path adjustment of the blades by means of stored patient data. 7. Blade collimator according to claim 6 in which the sets of blades are arrayed with their support or such that they rotate relative to it. 8. Blade collimator according to claim 1 in which sets of blades in modular form are interchangeable. 9. Blade collimator according to claim 1 in which the blades are leak-proof against radiation by means of overfocusing. 10. Blade collimator according to claim 1 in which an additional blade is inserted between two blades and the additional blade is positioned by mechanical devices such that its front face in the direction of movement of the blades always occupies an essentially intermediate position between the front faces of the two blades adjacent to it. 11. Blade collimator according to claim 1 in which the sets of blades consist of from 20 to 32 blades. 12. Blade collimator according to claim 11 with 26 blades in which each set has: 13. Blade collimator according to claim 1, in which an end of oblong connecting cords are mounted on upper edges, toward the rear, of the individual blades, whereby other ends of the connecting cords engage rods of a secondary position measurement device and the connecting cords, seen from the direction of movement of the blades, spread out upwards in roughly a fan shape to meet contact points on the rods, which are more widely separated than the blades. 14. Blade collimator according to claim 13 in which the connecting cords consist of flat strips that bend in their course from the blades to the contact points on the rods of the position measurement device. 15. Blade collimator according to claim 1, in which the blades, as seen from the direction of movement, exhibit from top to bottom a cross-sectional shape with widened sections on both sides of a bisecting line of the individual blades, as well as matching narrowed sections, whereby the adjacent, identically shaped blades exhibit their widened sections and narrowed sections at corresponding longitudinally displaced positions, so that the side faces of the blades nestle against each adjacent blade in essentially flat contact. 16. Blade collimator according to claim 15 in which tapped holes as counterparts to drive threaded rods are created in the widened cross-sectional areas of each blade. 17. Blade collimator according to claim 5, in which 18. Blade collimator according to claim 13, in which 19. Blade collimator according to claim 15, in which 20. Blade collimator according to claim 5 in which the mechanisms are activated by springs. 21. Blade collimator according to claim 5 in which the mechanisms are activated by connecting rods. 22. Blade collimator according to claim 5 in which the mechanisms are activated by electric motors. 23. Blade collimator according to claim 5 in which the stored patient data includes at least one of x-rays, computer tomograms or nuclear spin resonance tomograms. 24. Blade collimator according to claim 14 in which the connecting cords consist of flat metal strips, and whose end segments are straight.
	description	This application is a continuation of U.S. patent application Ser. No. 15/428,947, filed Feb. 9, 2017, now U.S. Pat. No. 10,157,687, which is a continuation-in-part of U.S. patent application Ser. No. 15/076,475, filed Mar. 21, 2016, now abandoned. U.S. patent application Ser. No. 15/076,475 is a continuation of U.S. patent application Ser. No. 13/794,589, filed Mar. 11, 2013, now U.S. Pat. No. 9,303,295, which application claims the benefit of U.S. Provisional Application No. 61/747,054, filed Dec. 28, 2012, which applications are incorporated herein by reference in their entirety. The present patent application relates to a fuel element including a cladding material and methods related to same. Disclosed embodiments include fuel elements, fuel assemblies, cladding materials, and methods of making and using same. The foregoing is a summary and thus may contain simplifications, generalizations, inclusions, and/or omissions of detail; consequently, those skilled in the art will appreciate that the summary is illustrative only and is NOT intended to be in any way limiting. In addition to any illustrative aspects, embodiments, and features described above, further aspects, embodiments, and features will become apparent by reference to the drawings and the following detailed description. Other aspects, features, and advantages of the devices and/or processes and/or other subject matter described herein will become apparent in the teachings set forth herein. Introduction In the following detailed description, reference is made to the accompanying drawings, which form a part hereof. In the drawings, the use of similar or the same symbols in different drawings typically indicates similar or identical items, unless context dictates otherwise. The illustrative embodiments described in the detailed description, drawings, and claims are not meant to be limiting. Other embodiments may be utilized, and other changes may be made, without departing from the spirit or scope of the subject matter presented here. One skilled in the art will recognize that the herein described components (e.g., operations), devices, objects, and the discussion accompanying them are used as examples for the sake of conceptual clarity and that various configuration modifications are contemplated. Consequently, as used herein, the specific exemplars set forth and the accompanying discussion are intended to be representative of their more general classes. In general, use of any specific exemplar is intended to be representative of its class, and the non-inclusion of specific components (e.g., operations), devices, and objects should not be taken as limiting. The present application uses formal outline headings for clarity of presentation. However, it is to be understood that the outline headings are for presentation purposes, and that different types of subject matter may be discussed throughout the application (e.g., device(s)/structure(s) may be described under process(es)/operations heading(s) and/or process(es)/operations may be discussed under structure(s)/process(es) headings; and/or descriptions of single topics may span two or more topic headings). Hence, the use of the formal outline headings is not intended to be in any way limiting. Overview By way of overview, provided in one embodiment is a method of making a composition, the method comprising: heat treating a material including an iron-based composition at a first temperature under a first condition in which at least some of the iron-based composition is transformed into an austenite phase; cooling the material to a second temperature at a cooling rate under a second condition in which at least some of the iron-based composition is transformed into a martensite phase; and heat treating the material at a third temperature under a third condition in which carbides are precipitated. Provided in another embodiment is a method of making a composition, the method comprising: subjecting a material to at least one of cold drawing, cold rolling, and pilgering; heat treating the material including an iron-based composition at a first temperature under a first condition in which at least some of the iron-based composition is transformed into an austenite phase; cooling the material to a second temperature at a cooling rate under a second condition in which at least some of the iron-based composition is transformed into a martensite phase; and heat treating the material at a third temperature under a third condition, in which carbides are precipitated. Provided in another embodiment is a composition comprising: (Fe)a(Cr)b(M)c; wherein a, b, and c are each a number greater than zero representing a weight percentage; M is at least one transition metal element; b is between 11 and 12; c is between about 0.25 and about 0.9; and balanced by a; and the composition further includes at least N at between about 0.01 wt % and about 0.04 wt %. Provided in another embodiment is a composition, comprising: (Fe)a(Cr)b(Mo, Ni, Mn, W, V)c; wherein a, b, and c are each a number greater than zero representing a weight percentage; b is between 11 and 12; c is between about 0.25 and about 0.9; and balanced by a; at least substantially all of the composition has a martensite phase; and the composition includes N at between about 0.01 wt % and about 0.04 wt %. Provided in another embodiment is a method of using a fuel assembly, comprising: generating power using a fuel assembly, a fuel element of which includes a composition, which is represented by a chemical formula: (Fe)a(Cr)b(M)c; wherein a, b, and c are each a number greater than zero representing a weight percentage; M is at least one transition metal element; b is between 11 and 12; c is between about 0.25 and about 0.9; and balanced by a; and the composition further includes at least N at between about 0.01 wt % and about 0.04 wt %. Provided in another embodiment is a fuel element comprising a tubular composition made by a method comprising: heat treating a material including an iron-based composition at a first temperature under a first condition in which at least some of the iron-based composition is transformed into an austenite phase; cooling the material to a second temperature at a cooling rate under a second condition in which at least some of the iron-based composition is transformed into a martensite phase; and heat treating the material at a third temperature under a third condition, in which carbides are precipitated. In one embodiment, in compositions where nitrogen is present, the precipitation of carbides may be accompanied by precipitation of nitrides and carbonitrides. Fuel Assembly FIG. 1a provides a partial illustration of a nuclear fuel assembly 10 in accordance with one embodiment. The fuel assembly may be a fissile nuclear fuel assembly or a fertile nuclear fuel assembly. The assembly may include fuel elements (or “fuel rods” or “fuel pins”) 11. FIG. 1b provides a partial illustration of a fuel element 11 in accordance with one embodiment. As shown in this embodiment, the fuel element 11 may include a cladding material 13, a fuel 14, and, in some instances, at least one gap 15. Fuel may be sealed within a cavity by the exterior cladding material 13. In some instances, the multiple fuel materials may be stacked axially as shown in FIG. 1b, but this need not be the case. For example, a fuel element may contain only one fuel material. In one embodiment, gap(s) 15 may be present between the fuel material and the cladding material, though gap(s) need not be present. In one embodiment, the gap is filled with a pressurized atmosphere, such as a pressured helium atmosphere. A fuel may contain any fissionable material. A fissionable material may contain a metal and/or metal alloy. In one embodiment, the fuel may be a metal fuel. It can be appreciated that metal fuel may offer relatively high heavy metal loadings and excellent neutron economy, which is desirable for breed-and-burn process of a nuclear fission reactor. Depending on the application, fuel may include at least one element chosen from U, Th, Am, Np, and Pu. The term “element” as represented by a chemical symbol herein may refer to one that is found in the Periodic Table—this is not to be confused with the “element” of a “fuel element”. In one embodiment, the fuel may include at least about 90 wt % U—e.g., at least 95 wt %, 98 wt %, 99 wt %, 99.5 wt %, 99.9 wt %, 99.99 wt %, or higher of U. The fuel may further include a refractory material, which may include at least one element chosen from Nb, Mo, Ta, W, Re, Zr, V, Ti, Cr, Ru, Rh, Os, Ir, and Hf. In one embodiment, the fuel may include additional burnable poisons, such as boron, gadolinium, or indium. The metal fuel may be alloyed with about 3 wt % to about 10 wt % zirconium to stabilize dimensionally the alloy during irradiation and to inhibit low-temperature eutectic and corrosion damage of the cladding. A sodium thermal bond fills the gap that exists between the alloy fuel and the inner wall of the cladding tube to allow for fuel swelling and to provide efficient heat transfer, which may keep the fuel temperatures low. In one embodiment, individual fuel elements 11 may have a thin wire 12 from about 0.8 mm diameter to about 1.6 mm diameter helically wrapped around the circumference of the clad tubing to provide coolant space and mechanical separation of individual fuel elements 56 within the housing of the fuel assemblies 18 and 20 (that also serve as the coolant duct). In one embodiment, the cladding 13, and/or wire wrap 12 may be fabricated from ferritic-martensitic steel because of its irradiation performance as indicated by a body of empirical data. Fuel Element A “fuel element”, such as element 11 shown in FIGS. 1a-1b, in a fuel assembly of a power generating reactor may generally take the form of a cylindrical rod. The fuel element may be a part of a power generating reactor, which is a part of a nuclear power plant. Depending on the application, the fuel element may have any suitable dimensions with respect to its length and diameter. The fuel element may include a cladding layer 13 and a fuel 14 disposed interior to the cladding layer 13. In the case of a nuclear reactor, the fuel may contain (or be) a nuclear fuel. In one embodiment, the nuclear fuel may be an annular nuclear fuel. The fuel element may additionally include a liner disposed between the nuclear fuel 14 and the cladding layer 13. The liner may contain multiple layers. The fuel may have any geometry. In one embodiment, the fuel has an annular geometry. In such an embodiment, a fuel in an annular form may allow a desirable level of fuel density to be achieved after a certain level of burn-up. Also, such an annular configuration may maintain compressive forces between the fuel and the cladding to promote thermal transport. The fuel may be tailored to have various properties, depending on the application. For example, the fuel may have any level of density. In one embodiment, it is desirable to have a high density of fuel, such as one as close to theoretical density uranium (in the case of a fuel containing uranium) as possible. In another embodiment, having a high porosity (low density) may prevent formation of additional internal voids during irradiation, decreasing fuel pressure on structural material, such as cladding, during operation of the nuclear fuel. The cladding material for the cladding layer 13 may include any suitable material, depending on the application. In one embodiment, the cladding layer 13 may include at least one material chosen from a metal, a metal alloy, and a ceramic. In one embodiment, the cladding layer 13 may contain a refractory material, such as a refractory metal including at least one element chosen from Nb, Mo, Ta, W, Re, Zr, V, Ti, Cr, Ru, Rh, Os, Ir, Nd, and Hf. In another embodiment, the cladding material may be chosen from a ceramic material, such as silicon carbide or aluminum oxide (alumina). A metal alloy in cladding layer 13 may be, in one exemplary embodiment, steel. The steel may be chosen from an austenitic steel, a ferritic-martensitic steel, an oxide-dispersed steel, T91 steel, T92 steel, HT9 steel, 316 steel, and 304 steel. The steel may have any type of microstructure. For example, the steel may include at least one of a martensite phase, a ferrite phase, and an austenite phase. In one embodiment, substantially all of the steel has at least one phase chosen from a martensite phase, a ferrite phase, and an austenite phase. Depending on the application, the microstructure may be tailored to have a particular phase (or phases). The cladding layer 13 may include an iron-based composition as described below. At least some of the components of the fuel elements may be bonded. The bonding may be physical (e.g., mechanical) or chemical. In one embodiment, the nuclear fuel and the cladding are mechanically bonded. In one embodiment, the first layer and the second layer are mechanically bonded. Iron-Based Composition Provided in one embodiment herein is a composition including a metal. The metal may include at least one of a metal, metal alloy, and intermetallic composition. In one embodiment, the metal includes iron. In one embodiment, the composition includes an iron-based composition. The term “X-based” composition in one embodiment may refer to a composition including a significant amount an element X (e.g., metal element). The amount may be, for example, at least 30%—e.g., at least 40%, at least 50%, at least 60%, at least 70%, at least 80%, at least 90%, at least 95%, at least 99%, or more. The percentage herein may refer to weight percent or a volume (or atomic) percent, depending on the context. In one embodiment, the iron-based composition may include steel. The compositions described herein may be employed as a component of a nuclear fuel element, such as the cladding material thereof. However, the metal-containing composition need not be limited to cladding material and may be employed wherever such a composition is employed. For example, provided in one embodiment is a composition that is represented by the chemical formula (Fe)a(Cr)b(M)c, wherein a, b, and c are each a number greater than zero representing a weight percentage; depending on the context, these numbers may alternatively represent a volume percentage. In one embodiment, b is a number between 11 and 12, c is between about 0.25 and about 0.9; balanced by a. In one embodiment, the composition includes at least nitrogen (“N”) at between about 0.005 wt % and about 0.05 wt %—e.g., about 0.01 wt % and about 0.04 wt %, between about 0.01 wt % and about 0.03 wt %, between about 0.02 wt % and about 0.03 wt %, etc. The element M may represent at least one transition metal element. The element M in this iron-based composition may be any transition metal element found in the Periodic Table—e.g., the elements in Groups 3-12 of the Periodic Table. In one embodiment, M represents at least one of Mo, Ni, Mn, W, and V. In another embodiment, the composition may include (or be) an iron-based composition including a steel composition. The composition may be represented by the chemical formula: (Fe)a(Cr)b(Mo, Ni, Mn, W, V)c, wherein a, b, and c are each a number greater than zero representing a weight percentage; depending on the context, the numbers may alternatively represent a volume percentage. In one embodiment, the number b is between 11 and 12; c is between about 0.25 and about 0.9; balanced by a. In one embodiment, the composition includes N at between about 0.01 wt % and about 0.04 wt %. The composition may contain at least one additional element. The additional element may be a non-metal element. In one embodiment, the non-metal element may be at least one element chosen from Si, S, C, and P. The additional element may be a metal element, including Cu, Cr, Mo, Mn, V, W, Ni, etc. In one embodiment, the composition further includes Cr at between about 10 wt % and about 12.5 wt %; C at between about 0.17 wt % and about 0.22 wt %; Mo at between about 0.80 wt % and about 1.2 wt %; Si less than or equal to about 0.5 wt %; Mn less than or equal to about 1.0 wt %; V at between about 0.25 wt % and about 0.35 wt %; W at between about 0.40 wt % and about 0.60 wt %; P less than or equal to about 0.03 wt %; and S less than or equal to about 0.3 wt %. In another embodiment, the composition further includes Ni at between about 0.3 wt % and 0.7 wt %. In another embodiment, the composition further includes Cr at about 11.5 wt %; C at about 0.20 wt %; Mo at about 0.90 wt %; Ni at about 0.55 wt %; Mn at about 0.65 wt %; V at about 0.30 wt %; W at about 0.50 wt %; Si at about 0.20 wt % and N at about 0.02 wt %. Other elements may also be present in any suitable amount. In some cases, certain incidental impurities may be present. The composition may include an iron-based composition that includes a steel composition including a tailored microstructure. For example, the compositions provided herein may have a small amount of a delta-ferrite phase. In one embodiment, the composition is at least substantially free of delta-ferrite. In another embodiment, the composition is completely free of delta-ferrite. Instead of a ferrite phase, the composition may include a martensite phase (e.g., tempered martensite). In one embodiment, substantially all of the composition has a martensite phase. In another embodiment, completely all of the composition has a martensite phase. As described below, one technique of tailoring the microstructure (e.g., to mitigate formation of a ferrite phase) may be to control the content of nitrogen within the range provided herein. Mitigation herein may refer to reduction and/or prevention but need not refer to total elimination. The microstructure, including the phases, may be described in terms of a chromium equivalent. In one embodiment, chromium equivalent (“Creq”) is the sum of ferrite forming elements plotted in constitution diagrams for the estimation of phases in stainless steel, weld metal, and calculated from various equations. In some instances, chromium equivalent may be used in conjunction with nickel equivalent, which is the sum of austenite forming elements. The equation may be any suitable equation, depending at least on the material chemistry. In one embodiment, the equation may be represented by the net chromium equivalent, net Creq which is the difference between chromium equivalent and nickel equivalent. Net Creq (wt %)=(% Cr)+6(% Si)+4(% Mo)+11(% V)+5(% Nb)+1.5(% W)+8(% Ti)+12(% Al)−4(% Ni)−2(% Co)−2(% Mn)−(% Cu)−40(% C)−30(% N). In one embodiment, the compositions described herein may have Creq of less than or equal to about 10—e.g., less than or equal to about 9, 8, 7, 6, 5, 4, 3, 2, or less. In one embodiment, the Creq may be kept under 9 to mitigate formation of ferrites. Based on the equation above, N-content may play an important role in the value of Creq, and hence the ferrite formation (or lack thereof). Due at least in part to the microstructure, the compositions described herein may have tailored material properties. For example, the compositions may have a high thermal stability. Thermal stability of a composition in one embodiment may refer to the resistance of a particular phase of the composition to decomposition (or dissociation) into another phase at an elevated temperature. In one embodiment, the compositions described herein are substantially thermally stable at a temperature of greater or equal to about 500° C.—e.g., greater or equal to about 550° C., about 600° C., or higher. The compositions provided herein may include additional phase(s) or material(s). For example, in a case where the composition includes carbon, the carbon element may be present in the form of a carbide. In one embodiment, the composition may include carbides distributed substantially uniformly in the composition. The carbides may have any suitable sizes, depending on the application. In one embodiment, the carbides have a size of less than or equal to about 2 microns—e.g., less than or equal to about 1 micron, 0.5 microns, 0.2 microns, 0.1 microns, or smaller. Methods of Making/Using the Iron-Based Composition The iron-based composition and a fuel element including the composition described herein may be manufactured by a variety of techniques. The iron-based composition may be any of the compositions described herein. For example, the composition may include a steel. Provided in another embodiment is a fuel element having a tubular structure made by the methods described herein. For example, referring to FIG. 2a, provided in one embodiment is a method of making a composition; the method includes heat treating a material including an iron-based composition at a first temperature under a first condition in which at least some of the iron-based composition is transformed into an austenite phase (step 201); cooling the material to a second temperature at a cooling rate under a second condition in which at least some of the iron-based composition is transformed into a martensite phase (step 202); and heat treating the material at a third temperature under a third condition in which carbides are precipitated (step 203). In one embodiment, steps 201 and 202 together may be referred to as normalization, whereas step 203 may be referred to as tempering. The first temperature may be any temperature suitable for the first condition. In one example, the first temperature may be above the austenitization temperature of the composition-the temperature at which substantially all of the ferrite phase of the iron-based composition transforms to an austenite phase. The austenite temperature varies with the material chemistry. In one embodiment, the first temperature is between about 900° C. and about 1200° C.-e.g., about 1000° C. and about 1150° C., about 1025° C. and about 1100° C., etc. The first temperature may be higher than 1200° C. or lower than 900° C., depending on the material. For example, in an embodiment the processing of the steel composition includes transforming at least some of the steel composition into an austenite phase by heating the steel composition to a temperature from 1100° C. to 1300° C. for 40-60 hours. Referring to FIG. 2b, the process of heat treating at the first temperature may further comprise heating the material to the first temperature (step 204). Heat treating at the first temperature may be carried out for any suitable length of time, depending on the material involved. The time may be adjusted such that the length is sufficiently long to promote formation of a homogeneous austenite solid solution. In one embodiment, the heat treatment may be carried out for about at least 3 minutes—e.g., at least 4 minutes, 5 minutes, 15 minutes, 20 minutes, 30 minutes, 60 minutes, 90 minutes, 120 minutes, 150 minutes, 180 minutes, or more. A longer or shorter length of time is also possible. In one embodiment, heat treating at the first temperature may be carried out for between about 1 minute and about 200 minutes—e.g., about 2 minutes and about 150 minutes, about 3 minutes and about 120 minutes, about 5 minutes and about 60 minutes, etc. In one embodiment, during heat treating at the first temperature (e.g. at the end of the treatment), at least some of the iron-based composition is transformed into an austenite phase. In one embodiment, substantially all of the composition is transformed into an austenite phase. In another embodiment, completely all of the composition is transformed into an austenite phase. In one embodiment, the first condition mitigates formation of a delta-ferrite phase of the iron-based composition. In another embodiment, the first condition promotes transformation of substantially all of the iron-based composition into an austenite phase. Referring to FIG. 2c, the process of heat treating at the first temperature (step 201) may further comprise dissolving at least substantially all of the carbides, if any, present in the iron-based composition of the material (step 205). The second temperature in step 202 may be any temperature suitable for the second condition. In one embodiment, the second temperature is less than or equal to 60° C.—e.g., less than or equal to 50° C., 40° C., 30° C., 20° C., 10° C., or less. In one embodiment, the second temperature is about room temperature (e.g., 20° C.). Cooling may be carried out via any suitable techniques. In one embodiment, cooling includes cooling by at least one of air and liquid. In one embodiment, the second condition promotes transformation of substantially all of the iron-based composition into a martensite phase. For example, the cooling may be carried out at a sufficient rate such that during cooling (e.g. at the end of the treatment), at least some of the iron-based composition is transformed into a martensite phase. In one embodiment, the rate is high enough that substantially all of the composition is transformed into a martensite phase. In another embodiment, the rate is high enough that completely all of the composition is transformed into a martensite phase. In one embodiment, at the end of cooling the composition is substantially free of at least one phase chosen from a ferrite phase and an austenite phase. In one embodiment, at the end of cooling the composition is completely free of at least one phase chosen from a ferrite phase and an austenite phase. The third temperature in step 203 may be any temperature suitable for the third condition. The third temperatures may be lower than the temperature above which austenite begins to form. In one embodiment, the third temperature may be lower than the first temperature. In one embodiment, the third temperature is at least 500° C.—e.g., at least 550° C., 600° C., 650° C., 700° C., 750° C., 800° C., 850° C., 900° C., or more. In one embodiment, the third temperature is between about 500° C. and about 900° C.—e.g., about 550° C. and about 850° C., about 600° C. and about 800° C., about 650° C. and about 780° C., about 700° C. and about 750° C., etc. A higher or lower temperature is also possible. The third temperature may be high enough to precipitate carbides and impart high temperature stability of carbides but low enough that carbide density is high and the carbide size is small with a homogeneous distribution of carbides for void swelling resistance. Referring to FIG. 2d, heat treating at the third temperature may include heating the material to the third temperature (step 206). Heat treating at the third temperature may be carried out for any suitable length of time, depending on the material involved. In one embodiment, heat treating at the third temperature may be carried out for between about 0.1 hours and about 5 hours—e.g., between about 0.2 hours to about 4 hours, about 0.5 hours to about 3 hours, about 1 hours and about 2 hours, etc. A longer or shorter length of time is also possible. In one embodiment, the third condition may mitigate the formation of a ferrite phase and/or an austenite phase of the iron-based composition. In one embodiment, the composition is substantially free of a ferrite phase and/or an austenite phase. The heat treatment may be carried out by any suitable techniques. In one embodiment, heat treating at the third temperature is carried out in a vertical furnace. Additional process(es) may be involved. For example, referring to FIG. 2e, the method may further comprise cooling the composition from the third temperature to a fourth temperature (step 207). The fourth temperature may be lower than the third temperatures. For example, the fourth temperature may less than or equal to 60° C.—e.g., less than or equal to 50° C., 40° C., 30° C., 20° C., 10° C., or less. In one embodiment, the fourth temperature is about room temperature (e.g., 20° C.). Referring to FIG. 2f, the method may further comprise controlling the wt % of N in the iron-based composition of the material to mitigate growth of a carbide phase of the iron-based composition (step 208). Referring to FIGS. 3a-3c, the differences in microstructure of the iron-based composition are illustrated in the figures. FIG. 3a shows a microstructure with blocky delta ferrite grains, loss of tempered martensite microstructure, and many grains devoid of complex carbide microstructure, in conventional steel. FIG. 3b shows an improved microstructure, with more homogenous carbide microstructure in the tempered martensite grains—note there are still some small delta ferrite grains in the microstructure. FIG. 3c shows a result of subjecting a steel sample to the processes described herein. The figure shows improved microstructure substantially free of delta ferrite with most grain regions having a high density of finely distributed carbides. Another embodiment provides an alternative method of making of a composition. Referring to FIG. 4a, the method includes: subjecting a material to at least one of cold drawing, cold rolling, and pilgering (step 401); heat treating the material including an iron-based composition at a first temperature under a first condition in which at least some of the iron-based composition is transformed into an austenite phase (step 402); cooling the material to a second temperature at a cooling rate under a second condition in which at least some of the iron-based composition is transformed into a martensite phase (step 403); and heat treating the material at a third temperature under a third condition, in which carbides are precipitated (step 404). In step 401, the material is cold worked; cold drawing, cold rolling, and pilgering are only some examples of processes that the material may be subjected to. One result of cold working is that the material dimension may be changed to a desired value. For example, the thickness of the material may be reduced as a result of cold working. In one embodiment, the reduction in thickness may be, for example, by at least 5%—e.g., at least 10%, 15%, 20%, 25%, or more. In one embodiment, the reduction is between about 5% and about 20%—e.g., between about 8% and about 16%, between about 10% and about 15%, etc. Higher or lower values are also possible. The dimension(s) of the material may be controlled via additional processes. In one embodiment, the ingot may undergo a thermo-mechanical processing to form the material with the final desired dimension(s). Referring to FIG. 4b, the starting material to be processed may be a billet, ingot, forging, etc., that has a cylindrical shape (step 405). The starting material is then mechanically worked (e.g., cold worked) by suitable tube-manufacturing process(es) (step 406). When the tube-manufacturing process involves cold work, the work piece may be annealed (“intermediate annealing”) after the working process at a temperature below the temperature above which austenite begins to form—below a transformation temperature from ferrite phase to an austenite phase (step 407). In one embodiment, austenite needs to be avoided because it would transform on cooling to hard martensite, thus counteracting the softening process. Steps 406 and 407 are repeated until the final dimensions are achieved. In one embodiment, after the final cold working step (step 408) that provides the tube with its final dimensions, the tube is not annealed again. The tube then may undergo normalization and tempering, as described above. The method may include additional processes. Referring to FIG. 4c, the method may further comprise extruding an ingot including the composition (step 409). Referring to FIG. 4d, the method may further comprise forming an ingot including the iron-based composition before the subjecting step, wherein the forming includes at least one process chosen from cold cathode induction melting, vacuum induction melting, vacuum arc re-melting, and electro-slag remelting (step 410). Referring to FIG. 4e, the method may further comprise forming an ingot including the iron-based composition and purifying the ingot to remove impurities (e.g., P, S, etc.) before the subjecting step (step 411). The forming and the purifying processes may involve any suitable techniques. The aforedescribed temperatures may vary depending on the materials and/or applications thereof involved. A fuel element (and fuel assemblies) including the composition (e.g., as the cladding) may be used in a variety of applications. Provided in one embodiment is a method of using a fuel assembly. Referring to FIG. 5a, the method includes generating power using a fuel assembly, a fuel element of which includes any of the iron-based compositions described herein (step 501). Referring to FIG. 5b, the generation of power may include generating at least one of electrical power and thermal power (step 502). Power Generation As described above, the fuel assemblies described herein may be a part of a power or energy generator, which may be a part of a power generating plant. The fuel assembly may be a nuclear fuel assembly. In one embodiment, the fuel assembly may include a fuel, a plurality of fuel elements, and a plurality of fuel ducts, such as those described above. The fuel ducts may include the plurality of fuel elements disposed therein. The fuel assembly described herein may be adapted to produce a peak areal power density of at least about 50 MW/m2—e.g., at least about 60 MW/m2, about 70 MW/m2, about 80 MW/m2, about 90 MW/m2, about 100 MW/m2, or higher. In some embodiments, the fuel assembly may be subjected to radiation damage at a level of at least about 120 displacements per atom (“DPA”)—e.g., at least about 150 DPA, about 160 DPA, about 180 DPA, about 200 DPA, or higher. All of the above U.S. patents, U.S. patent application publications, U.S. patent applications, foreign patents, foreign patent applications and non-patent publications referred to in this specification and/or listed in any Application Data Sheet, are incorporated herein by reference in their entirety, to the extent not inconsistent herewith. In the event that one or more of the incorporated literature and similar materials differs from or contradicts this application, including but not limited to defined terms, term usage, described techniques, or the like, this application controls. Embodiments of the composition described above were made and tested for void swelling performance. Three heats, identified as Heats FD, CH, and DH, of the composition were prepared to meet the specification listed above. Heats CH and DH are the same composition that differ only in a slight variation in the final heat treatment. For relative comparison between historic HT9 and the embodiments of the composition described herein, a historical HT9 sample of heat 84425 from the ACO-3 duct used in the Fast Flux Test Facility (FFTF) was tested for swelling using the same protocol. The actual composition of the final plate product of each heat was determined by analysis and is shown in Table 1. The actual composition of the historical sample was also determined and is likewise presented in Table 1. TABLE 1Heats CHHistorical HT9Heat FDand DHACO-3ElementMax.Min.ActualActualActualFeBal.Bal.Bal.Bal.Bal.C0.190.170.1760.200.20Si0.230.170.210.220.27Mn0.530.470.500.690.58P0.01—<0.0050.0040.003S0.003—0.00170.0010.004Cr12.211.812.1211.5611.87Mo1.050.951.010.881.02Ni0.550.450.520.560.53V0.330.270.300.3150.30W0.650.550.600.490.37N0.0130.0070.0110.0230.0017Cu0.02—<0.01—0.013Al————0.002Nb———<0.004<0.010Co————0.011Heat FD Preparation A 50 kg VIM ingot of Heat FD steel was heated at 1,200° C. for 48 hours to homogenize the cast structure and then was forged to approximately 70t×100w×450L (mm). The temperature of furnace for homogenizing was controlled by PID temperature controller and by using calibrated thermocouple. The forged plate was soaked at 1,200° C. for 2 hours and hot-rolled from 70t×100w×450L (mm) to approximately 24t×110w×1,050L (mm). A portion of the hot-rolled steel plate was annealed at 800° C. for 1 hour in order to make easy surface machining and approximately 0.3 mm per side was machined off the surface plate to remove any oxide film. The plate was then cold-rolled to a thickness of 5.4 mm by multiple steps. At the intermediate passes during cold rolling, the plate was annealed at 800° C. for 1 hour for softening cold-worked structure. Again, the furnace temperature for intermediate heat treatment is controlled by the PID temperature controller of furnace and by using calibrated thermocouple. After cold rolling, the plate was annealed at 800° C. for 1 hour in order to make easy sawing and was cut to smaller pieces for final heat treatment. After cutting, final heat treatment was performed for on one of the smaller pieces. This piece, designated Heat FD, was heat-treated (as part of a batch of pieces of other steels) at 1,000° C. for 30 minutes and then air-cooled to room temperature in order to obtain martensite structure. The temperature of furnace for normalization heat treatment was controlled by PID temperature controller and by using calibrated thermocouple. Furthermore, new thermocouples were attached by spot welding on the surface of one of the pieces in the heat treatment batch. The batch including the piece attached to thermocouples was put in the furnace at the normalization temperature, and final normalization heat treatment time started to count after thermocouples attached on piece reached the normalization temperature. After holding for prescribed time, the batch was taken out of the furnace. The normalized piece of Heat FD was heat-treated at 750° C. for 0.5 hour in order to temper the martensite structure, and then was air-cooled to room temperature. The temperature of furnace for final temper heat treatment was controlled by PID temperature controller and by using calibrated thermocouple. Again, the Heat FD piece was part of a batch of other steel pieces that included a piece with attached thermocouples as described above. The batch was put in the furnace kept at the tempering temperature, and final tempering treatment time started to count after thermocouples attached on piece reached the tempering temperature or 750° C. After holding for 30 minutes, the tempered batch was taken out of the furnace. The Vickers hardness of the tempered Heat FD piece was tested three times and determined to be 238, 246, and 241 for an average of 242. Heat CH and DH Preparation FIGS. 6a and 6b show a process outline of the major process steps used to fabricate plate and tube products of Heats CH and DH. The early processing steps (vacuum induction melting (VIM), vacuum arc re-melting (VAR) and homogenization were applied for both Heats. One peculiarity of the fabrication process is the application of a second homogenization heat treatment at 1180° C. for 48 hours, either after hot rolling of the plate or after the 2nd or 3rd cold rolling step for the tube. Swelling Testing Heavy ion irradiation testing was conducted on plates of each of the three heats and the historic control sample to determine the swelling performance of the composition. Irradiations were conducted in an ion beam laboratory using a dual ion (Fe++ and He++) irradiation beam to simulate the production of He from (n,α) reactions and the subsequent formation of voids in a neutron environment. Energetic 5 MeV Fe++ and low current He++ ions were directed at the steel samples at temperatures of 440, 460, and 480° C. to an irradiation dose level of 188 dpa. ˜2 MeV He++ ions are transmitted through an Al foil with a thickness of ˜3 μm in order to degrade their energy and deposit the He+ at the appropriate depth in the steel. The precise He++ beam energy is dependent on the exact thickness of the Al foil. The Al foil is rotated relative to the He++ beam in order change the incidence angle of the beam and modify the depth of implantation in the steel to range from 300-1000 nm. The incidence angle varies from 0-60° at five different intervals, with different hold times for each incidence angle, producing five separate depth profiles that cumulatively provide a roughly uniform (±10%) He concentration from 300-1000 nm into the steel. The irradiations were conducted on the three heats and the historical control sample using a 3 MV Pelletron accelerator. Samples were irradiated using a combination of a defocused 5 MeV Fe++ ion beam with typical beam current of ˜100-400 nA on the samples and a 3 mm diameter focused ˜2 MeV He++ beam that was raster scanned at 0.255 kHz in x and 1.055 kHz in y. Before each irradiation, the stage was outgassed to a pressure below 1×10−7 torr. The beam current was recorded every 30-60 minutes using the Faraday cup immediately in front of the samples and the integrated charge (current x time) was converted to dose based on the damage rate output of Stopping Range of Ions in Matter (SRIM) calculation at a depth of 600 nm using the Quick Kinchin-Pease mode and a 40 eV displacement energy. The samples were mechanically polished using SiC paper up to a fine grit of #4000 followed by final polishing with diamond solutions up to 0.25 μm, with a final mechanical polishing of 0.02 colloidal silica solution prior to irradiation. After mechanical polishing, specimens were electropolished for 20 seconds in a 90% methanol and 10% perchloric acid solution, at temperatures between −40° C. and −50° C., with an applied potential of 35 V between the specimen and platinum mesh cathode. Temperature control was achieved by using a series of thermocouples affixed to irradiation samples that are heated and then used to calibrate a two-dimensional imaging pyrometer at the irradiation temperature. Temperature was controlled using the imaging pyrometer to ±10° C. throughout the irradiation. Irradiated sample preparation was accomplished using cross-section focused ion beam (FIB) liftouts from the irradiated surface of each sample. The liftout method allows the entire irradiation damage region to be imaged, and for void imaging analysis to be consistently performed only at the desired depth. FIG. 7 illustrates a representative transmission electron microscope (TEM) image illustrating the depth effect on voids created by irradiation. Void imaging was done on a JEOL 2100F TEM. Void measurements included only voids that were within a damage zone depth of 300-700 nm into the sample, as represented by FIG. 7. By performing the analysis in this way, all voids at the surface (0-300 nm), which would be influenced by surface effects and changes in surface composition, were not taken into account. So, too, all voids at the end of damage curve (>700 nm) that may be affected by self-interstitial implantation of the Fe++ ion were not considered. Self-interstitial ions at the end of the damage curve tend to suppress void nucleation by affecting the vacancy/interstitial bias that causes void nucleation. Sample thickness was measured using electron energy loss spectroscopy (EELS) to measure the zero energy loss fraction and determine sample thickness. Using sample thickness and image area, void density and swelling measurements can be made. As mentioned above, the irradiations included a sample from the archived ACO-3 duct HT9 material from FFTF for a relative swelling comparison to the composition embodiments described above. Heavy ion irradiations were conducted on the four heats (CH, DH, FD, and ACO-3) described above in order to generate a relative comparison in swelling behavior among the different heats. The swelling response could also be compared to the archive (heat 84425) of HT9 from ACO-3 duct wall from the FFTF program, irradiated at 443° C. to a dose of 155 dpa, which demonstrated swelling of ˜0.3% based on TEM imaging of the voids. Information regarding the historic heat of HT9 from FFTF program can be found in the article Phase Stability of an HT-9 Duct Irradiate in FFTF, by O. Anderoglu, et al., Journal of Nuclear Materials 430 (2012) pp. 194-204. To quantify the difference in swelling performance between the embodiments of the present compositions and the historic ACO-3 steel, the swelling % data in FIG. 8 were determined using process identified in Section 2.2 of the article Void Swelling And Microstructure Evolution At Very High Damage Level In Self-Ion Irradiated Ferritic-Martensitic Steels, by E. Getto, et al., Journal of Nuclear Materials 480 (2016) pp. 159-176, which section is incorporated herein by reference. Wherever swelling % is used in this disclosure, it is calculated by the process identified in the incorporated Section. FIG. 8 shows the swelling results for the heats. FIG. 8 clearly shows the difference in void swelling performance of the composition embodiments relative to the archived ACO-3. At the lower and higher temperatures, 440° C. and 500° C., little swelling was detectable in any of the heats. However, at temperatures of 460° C. and 480° C., each of the three heats of the present composition show significant improvements in swelling over the historic ACO-3 steel. FIG. 9 shows a TEM collage of void microstructure in the four heats after irradiation at 480° C. to 188 dpa with 0.2 appm He/dpa, in which the voids appear as the black features. The ACO-3 sample showed an inhomogeneous distribution of voids, but with a large cluster of many voids. The heats of the present composition each show a clear improvement over the ACO-3. The differences between ACO-3 and the heats of the present composition are striking and reflect a difference in void incubation between ACO-3 and the embodiments of the steel compositions described herein. FIG. 10 shows a TEM collage of void microstructure in the four heats after irradiation at 460° C. to 188 dpa with 0.015 appm He/dpa. Again, the heats of the present composition each show a clear improvement over the ACO-3. The examples provided above show that a steel of the composition: Cr at between about 10.0 wt % and about 13.0 wt %; C at between about 0.17 wt % and about 0.23 wt %; Mo at between about 0.80 wt % and about 1.2 wt %; Si less than or equal to about 0.5 wt %; Mn less than or equal to about 1.0 wt %; V at between about 0.25 wt % and about 0.35 wt %; W at between about 0.40 wt % and about 0.60 wt %; and Fe at least 80 wt %; can be manufactured that exhibits a swelling of less than 0.9% by volume, and in some cases less than 0.75%, less than 0.5%, and even less than 0.3% at a depth between 500-700 nm below the surface after dual-beam Fe++ and He++ irradiation to doses of 188 displacements per atom (dpa) with 0.2 appm He/dpa, as calculated using the Stopping Range in Matter simulation with the K-P option for damage cascades and a 40 eV displacement energy, created by irradiating the steel composition at 460° C. with a defocused beam of 5 MeV Fe++ ions and a raster-scanned beam of ˜2 MeV He++ ions transmitted through a thin Al foil for scattering and energy reduction to create a uniform He profile at the irradiation depth of the sample. With respect to the use of substantially any plural and/or singular terms herein, those having skill in the art can translate from the plural to the singular and/or from the singular to the plural as is appropriate to the context and/or application. The various singular/plural permutations are not expressly set forth herein for sake of clarity. The herein described subject matter sometimes illustrates different components contained within, or connected with, different other components. It is to be understood that such depicted architectures are merely exemplary, and that in fact many other architectures may be implemented which achieve the same functionality. In a conceptual sense, any arrangement of components to achieve the same functionality is effectively “associated” such that the desired functionality is achieved. Hence, any two components herein combined to achieve a particular functionality can be seen as “associated with” each other such that the desired functionality is achieved, irrespective of architectures or intermedial components. Likewise, any two components so associated can also be viewed as being “operably connected,” or “operably coupled,” to each other to achieve the desired functionality, and any two components capable of being so associated can also be viewed as being “operably couplable,” to each other to achieve the desired functionality. Specific examples of operably couplable include, but are not limited to, physically mateable and/or physically interacting components, and/or wirelessly interactable, and/or wirelessly interacting components, and/or logically interacting, and/or logically interactable components. In some instances, one or more components may be referred to herein as “configured to,” “configured by,” “configurable to,” “operable/operative to,” “adapted/adaptable,” “able to,” “conformable/conformed to,” etc. Those skilled in the art will recognize that such terms (e.g. “configured to”) can generally encompass active-state components and/or inactive-state components and/or standby-state components, unless context requires otherwise. While particular aspects of the present subject matter described herein have been shown and described, it will be apparent to those skilled in the art that, based upon the teachings herein, changes and modifications may be made without departing from the subject matter described herein and its broader aspects and, therefore, the appended claims are to encompass within their scope all such changes and modifications as are within the true spirit and scope of the subject matter described herein. It will be understood by those within the art that, in general, terms used herein, and especially in the appended claims (e.g., bodies of the appended claims) are generally intended as “open” terms (e.g., the term “including” should be interpreted as “including but not limited to,” the term “having” should be interpreted as “having at least,” the term “includes” should be interpreted as “includes but is not limited to,” etc.). It will be further understood by those within the art that if a specific number of an introduced claim recitation is intended, such an intent will be explicitly recited in the claim, and in the absence of such recitation no such intent is present. For example, as an aid to understanding, the following appended claims may contain usage of the introductory phrases “at least one” and “one or more” to introduce claim recitations. However, the use of such phrases should not be construed to imply that the introduction of a claim recitation by the indefinite articles “a” or “an” limits any particular claim containing such introduced claim recitation to claims containing only one such recitation, even when the same claim includes the introductory phrases “one or more” or “at least one” and indefinite articles such as “a” or “an” (e.g., “a” and/or “an” should typically be interpreted to mean “at least one” or “one or more”); the same holds true for the use of definite articles used to introduce claim recitations. In addition, even if a specific number of an introduced claim recitation is explicitly recited, those skilled in the art will recognize that such recitation should typically be interpreted to mean at least the recited number (e.g., the bare recitation of “two recitations,” without other modifiers, typically means at least two recitations, or two or more recitations). Furthermore, in those instances where a convention analogous to “at least one of A, B, and C, etc.” is used, in general such a construction is intended in the sense one having skill in the art would understand the convention (e.g., “a system having at least one of A, B, and C” would include but not be limited to systems that have A alone, B alone, C alone, A and B together, A and C together, B and C together, and/or A, B, and C together, etc.). In those instances where a convention analogous to “at least one of A, B, or C, etc.” is used, in general such a construction is intended in the sense one having skill in the art would understand the convention (e.g., “a system having at least one of A, B, or C” would include but not be limited to systems that have A alone, B alone, C alone, A and B together, A and C together, B and C together, and/or A, B, and C together, etc.). It will be further understood by those within the art that typically a disjunctive word and/or phrase presenting two or more alternative terms, whether in the description, claims, or drawings, should be understood to contemplate the possibilities of including one of the terms, either of the terms, or both terms unless context dictates otherwise. For example, the phrase “A or B” will be typically understood to include the possibilities of “A” or “B” or “A and B.” With respect to the appended claims, those skilled in the art will appreciate that recited operations therein may generally be performed in any order. Also, although various operational flows are presented in a sequence(s), it should be understood that the various operations may be performed in other orders than those which are illustrated, or may be performed concurrently. Examples of such alternate orderings may include overlapping, interleaved, interrupted, reordered, incremental, preparatory, supplemental, simultaneous, reverse, or other variant orderings, unless context dictates otherwise. Furthermore, terms like “responsive to,” “related to,” or other past-tense adjectives are generally not intended to exclude such variants, unless context dictates otherwise. Those skilled in the art will appreciate that the foregoing specific exemplary processes and/or devices and/or technologies are representative of more general processes and/or devices and/or technologies taught elsewhere herein, such as in the claims filed herewith and/or elsewhere in the present application. While various aspects and embodiments have been disclosed herein, other aspects and embodiments will be apparent to those skilled in the art. The various aspects and embodiments disclosed herein are for purposes of illustration and are not intended to be limiting, with the true scope and spirit being indicated by the following claims. Any portion of the processes described herein may be automated. The automation may be accomplished by involving at least one computer. The automation may be executed by program that is stored in at least one non-transitory computer readable medium. The medium may be, for example, a CD, DVD, USB, hard drive, etc. The selection and/or design of the fuel element structure, including the assembly, may also be optimized by using the computer and/or a software program. The above-described embodiments of the invention can be implemented in any of numerous ways. For example, some embodiments may be implemented using hardware, software or a combination thereof. When any aspect of an embodiment is implemented at least in part in software, the software code can be executed on any suitable processor or collection of processors, whether provided in a single computer or distributed among multiple computers. Also, the technology described herein may be embodied as a method, of which at least one example has been provided. The acts performed as part of the method may be ordered in any suitable way. Accordingly, embodiments may be constructed in which acts are performed in any order different from that illustrated, which may include performing some acts simultaneously, even though shown as sequential acts in illustrative embodiments. All definitions, as defined and used herein, should be understood to control over dictionary definitions, definitions in documents incorporated by reference, and/or ordinary meanings of the defined terms. The indefinite articles “a” and “an,” as used herein in the specification and in the claims, unless clearly indicated to the contrary, should be understood to mean “at least one.” The phrase “and/or,” as used herein in the specification and in the claims, should be understood to mean “either or both” of the elements so conjoined, i.e., elements that are conjunctively present in some cases and disjunctively present in other cases. Multiple elements listed with “and/or” should be construed in the same fashion, i.e., “one or more” of the elements so conjoined. Other elements may optionally be present other than the elements specifically identified by the “and/or” clause, whether related or unrelated to those elements specifically identified. Thus, as a non-limiting example, a reference to “A and/or B”, when used in conjunction with open-ended language such as “including” can refer, in one embodiment, to A only (optionally including elements other than B); in another embodiment, to B only (optionally including elements other than A); in yet another embodiment, to both A and B (optionally including other elements); etc. As used herein in the specification and in the claims, “or” should be understood to have the same meaning as “and/or” as defined above. For example, when separating items in a list, “or” or “and/or” shall be interpreted as being inclusive, i.e., the inclusion of at least one, but also including more than one, of a number or list of elements, and, optionally, additional unlisted items. Only terms clearly indicated to the contrary, such as “only one of” or “exactly one of,” or, when used in the claims, “consisting of,” will refer to the inclusion of exactly one element of a number or list of elements. In general, the term “or” as used herein shall only be interpreted as indicating exclusive alternatives (i.e. “one or the other but not both”) when preceded by terms of exclusivity, such as “either,” “one of,” “only one of,” or “exactly one of,” “Consisting essentially of,” when used in the claims, shall have its ordinary meaning as used in the field of patent law. As used herein in the specification and in the claims, the phrase “at least one,” in reference to a list of one or more elements, should be understood to mean at least one element selected from any one or more of the elements in the list of elements, but not necessarily including at least one of each and every element specifically listed within the list of elements and not excluding any combinations of elements in the list of elements. This definition also allows that elements may optionally be present other than the elements specifically identified within the list of elements to which the phrase “at least one” refers, whether related or unrelated to those elements specifically identified. Thus, as a non-limiting example, “at least one of A and B” (or, equivalently, “at least one of A or B,” or, equivalently “at least one of A and/or B”) can refer, in one embodiment, to at least one, optionally including more than one, A, with no B present (and optionally including elements other than B); in another embodiment, to at least one, optionally including more than one, B, with no A present (and optionally including elements other than A); in yet another embodiment, to at least one, optionally including more than one, A, and at least one, optionally including more than one, B (and optionally including other elements); etc. Any ranges cited herein are inclusive. The terms “substantially” and “about” used throughout this Specification are used to describe and account for small fluctuations. For example, they can refer to less than or equal to ±5%, such as less than or equal to ±2%, such as less than or equal to ±1%, such as less than or equal to ±0.5%, such as less than or equal to ±0.2%, such as less than or equal to ±0.1%, such as less than or equal to ±0.05%. In the claims, as well as in the specification above, all transitional phrases such as “including,” “carrying,” “having,” “containing,” “involving,” “holding,” “composed of,” and the like are to be understood to be open-ended, i.e., to mean including but not limited to. Only the transitional phrases “consisting of” and “consisting essentially of” shall be closed or semi-closed transitional phrases, respectively, as set forth in the United States Patent Office Manual of Patent Examining Procedures, Section 2111.03. The claims should not be read as limited to the described order or elements unless stated to that effect. It should be understood that various changes in form and detail may be made by one of ordinary skill in the art without departing from the spirit and scope of the appended claims. All embodiments that come within the spirit and scope of the following claims and equivalents thereto are claimed.
	summary
	summary
	claims	1. A process of treating radioactive waste, comprising:transferring a solidifying container to a first location of a radioactive waste treatment facility along a transferring line; preparing a solidifying agent paste by kneading a solidifying agent and an additive water and injecting the solidifying agent paste into the solidifying container at the first location; transferring the solidifying container to a second location downstream of the first location with respect to a process flow; and introducing radioactive waste into the solidifying container and kneading the radioactive waste and the solidifying agent paste in the solidifying container at the second location. 2. The process of treating radioactive waste, according to claim 1, further comprising the step of elevating the solidifying container from the transferring line. 3. The process of treating radioactive waste according to claim 1, wherein the kneading of the radioactive waste is performed by an in-drum type waste kneader. 4. The process of treating radioactive waste according to claim 2, wherein the kneading of the radioactive waste is performed by an in-drum type waste kneader. 5. The process of treating radioactive waste according to claim 1, therein the kneading to the radioactive waste is performed by kneading blades. 6. The process of treating radioactive waste according to claim 1, further comprising the step of providing a partition wall between said first location and said second location.
	summary
047708433	abstract	A method and apparatus for controlling the stability of a boiling water reactor using a digital computer to calculate on-line, from distributed steady state values of only power, flow, enthalpy and pressure; a stability index for selected fuel assemblies taking into account nuclear feedback as well as detailed hydrodynamic effects. Such calculations are only made for fuel assemblies selected as most susceptible to instability on the basis of the level and axial distribution of power generated. If the least stable fuel assembly is unstable, its stability index is iteratively recalculated using assumed incremental changes in either flow or control rod position until a stable condition is indicated. The cumulative adjustment in flow or rod position required for stability is reported to the operator and can be implemented manually or by an automatic control system.
	description	None. None. None. This invention relates to a process and apparatus for growing agricultural products with a reduced abundance of carbon-14 (14C) by employing centrifugal separation of atmospheric gases to remove carbon dioxide (CO2) with radioactive 14C. Agricultural products with reduced 14C content can be grown in controlled environments for the benefit of reducing harmful damage to human DNA that is unavoidable with our current food chain, due to the natural abundance of 14C in atmospheric gases. Radioactive 14C decay to nitrogen-14 with the release of 156 KeV has long been known to have biological effects (Purdom, C. E.). Sequencing of the human genome has identified 6.1 billion base pairs in human DNA, with 119 billion carbon atoms in the DNA of each nucleated cell (Lander, E. S., and Genome Reference Consortium (GRC) Human Genome Assembly build 38 (GRCh38)). Recent quantitative analysis of human tissues has estimated 3 trillion nucleated cells in the human body (Sender, R., Fuchs, S., & Milo, R.). Given the natural abundance and half-life of 14C and composition of our genome (i.e., a mean of roughly 6.0×109 base pairs with 19.5 carbon atoms each), in the average human this decay is occurring once per second in human DNA, resulting in potential bond ruptures, DNA strand breakage, and nitrogen substitution in canonical bases (Sassi, M., et. al.). This cumulative damage has been positively correlated to cancer diagnoses (Patrick, A. D., & Patrick, B. E.), and may have other yet-to-be-quantified effects on human tissues as we age. In fact, no mammal has yet lived without this cumulative damage, so the qualitative benefits of precluding this genetic alteration are yet-to-be-quantified. To preclude this cumulative damage and genetic alteration, it is necessary to perform isotope separation on large volumes of atmospheric gases to remove 14C from agricultural products and their derivatives in the food chain. This requires an economical means for the filtration of atmospheric gases and the growth of agricultural products in controlled environments. In commercial applications, isotope separation has most commonly been applied to uranium isotopes utilizing a centrifugal separation process. The helikon vortex has been applied to uranium isotope enrichment in South Africa utilizing a multi-stage cascade design (Feiverson, H. A., Glaser, A., Mian, Z., & Von Hippel, F. N., and Moore, J. D. L.), but has not been applied to the selective isotope separation of CO2 from atmospheric gases in prior art. Turner, et al., in U.S. Pat. No. 8,460,434, shows that a helikon vortex can be utilized as a centrifugal separator in a multi-stage cascade design as one part of a process to separate methane from landfill gas. Although the multi-stage cascade design of the helikon vortex can separate gases by molecular density, it was developed for the separation of uranium isotopes, which are very heavy and differ in mass by a small amount (i.e., 235U and 238U, which differ in mass by 1.3%), which is one of the most challenging applications for centrifugal separation. Due to this multi-stage cascade design, it is very energy intensive to operate, and although it can be applied to the separation other gases by molecular density, it is uneconomical for the filtration of atmospheric gases on a large-scale for agricultural production. Shacter, in U.S. Pat. No. 3,925,036, shows a method for cycling gases through a cascade of multiple stages to achieve the separation other gases by molecular density. This multi-stage cascade design was also intended for the separation of uranium isotopes, and due to the reasons noted above is very energy intensive to operate, and although it can be applied to the separation other gases by molecular density, it is uneconomical for the filtration of atmospheric gases on a large-scale for agricultural production. Steimel, in U.S. Pat. No. 3,004,158, shows that a gas centrifuge can separate molecules of different masses by applying extremely high velocities while utilizing ionization of the gas with electric currents and the control of magnetic fields around the gas chamber. Although this process is effective for the separation of isotopes of heavy elements, such as uranium (i.e., 235U and 238U, which differ in mass by 1.3%), it is very energy intensive to operate and the apparatus itself is complex to construct, including a large electromagnet, electrodes, and controlling mechanisms. While all of this may be essential for the difficult and energy intensive separation of heavy isotopes from each other (e.g., 235U and 238U), the separation of carbon isotopes (e.g., 12C and 14C, which differ in mass by 16.7%) is much less energy intensive, due to the relatively large mass difference between isotopes. Being more energy intensive than necessary for the desired application, this process is uneconomical for the filtration of atmospheric gases on a large-scale for agricultural production. Gerber, in U.S. Pat. No. 3,594,573, shows that heavy and light isotopes can be separated from a fluid by applying a rotating electric field and ionization of the liquid with electrodes or a radioactive source. Although this process may have economical applications for liquids at atmospheric pressures, utilization of this process for the separation of CO2 with 14C from atmospheric gases would first require the separation of CO2 from other atmospheric gases, the liquification of the removed CO2, and then the application of the described process. After this, the CO2 without 14C would need then to be re-combined with atmospheric gases without CO2. Together, with the added complexity of removing CO2 from atmospheric gases, liquification of this gas, application of the described process, and then recombination of gases, this approach is uneconomical for the filtration of atmospheric gases on a large-scale for agricultural production. Janes, in U.S. Pat. No. 3,939,354, shows that ions can be separated from a plasma source utilizing mass acceleration. Similarly, Drummond, et al., in U.S. Pat. No. 3,942,975, shows that matter can be converted by an arc heater into an ionized plasma in excess of 5,000° K and stabilized with magnetic fields. Although this process was developed for the separation of rare valuable elements, such as metals, these could be adapted to separate carbon isotopes from sources of carbon. Utilization of these methodologies for the separation of CO2 with 14C from atmospheric gases would first require the separation of CO2 from other atmospheric gases, then application of the described process to the removed CO2 (or conversion of some other carbon source to plasma) and then removal of 14C. After this, the carbon without 14C would need to be combined with oxygen to produce CO2, which would then need to be mixed with the atmospheric gases that had the CO2 removed earlier. Together, with the added complexity of removing CO2 from atmospheric gases, application of the described process, conversion of carbon to CO2, and then recombination of gases, this appears to be an uneconomical alternative for the filtration of atmospheric gases on a large-scale for agricultural production. McKinney, et al., in U.S. Pat. No. 3,421,334, shows that isotopes of helium can be separated while in liquid form by exploiting unique physical properties of different isotopes. Although the claim was limited for use with helium, a similar approach could exploit the physical properties of CO2 in a liquid state. This approach would be complicated by the fact CO2 is a compound rather than an element and that there are three stable isotopes of oxygen (i.e., 16O, 17O, and 18O) that are naturally found in combinations with three naturally occurring isotopes of carbon (i.e., 12C, 13C, and 14C). Even so, exploiting the unique molecular weight of 12CO6O2 in a liquid state would require the removal of all CO2 from atmospheric gases, application of this new process, and then recombination of the CO2 without 14C with the atmospheric gases without CO2. Altogether, even if this claim were modified for this application, it would also appear to be an uneconomical alternative for the filtration of atmospheric gases to remove 14C on a large-scale for agricultural production. Russ, Fischer, and Crawford, in U.S. Pat. No. 7,332,715 (2008), shows that gas at an atmospheric pressure can be passed through an ionization chamber with an electrode that generates ions, which pass through an ion filter apparatus with voltage differentials, thereby performing mass spectrometry, which demonstrates one form of isotope separation. Although this process is useful for the identification and measurement of the molecular and isotopic constituents of a gas, it is not readily extensible or adaptable to the removal of one isotopic component of atmospheric gases on a large scale, since each molecule of atmospheric gas needs to be ionized prior to filtration. Lashoda, et al, in U.S. Pat. No. 4,584,073, shows that isotopes of an element in a compound can be separated utilizing a laser when the compound is deposited in a monolayer on small glass beads. Although this process has useful applications, utilization of this process for separation of CO2 with 14C from atmospheric gases would first require the separation of CO2 from all other atmospheric gases, the liquification of the removed CO2, and then the application of the described process. After this, the CO2 without 14C would then need to be re-combined with atmospheric gases without CO2. Together, with the added complexity of removing CO2 from atmospheric gases, liquification of the removed CO2 gas, application of the described process, and then recombination of gases, this approach is uneconomical for the filtration of atmospheric gases on a large-scale for agricultural production. Several instances of prior art utilize condensation of gases or condensates as part of a system or method to remove isotopes. Redmann, in U.S. Pat. No. 4,638,674, shows that isotopes can be removed from a continuous stream of gas through condensation, although the claims are limited to gas streams from a nuclear plant rather than atmospheric gases. Similarly, Schweiger in U.S. Pat. No. 4,816,209, shows that radioactive tritium isotopes can be removed from gas from a nuclear reactor by utilizing condensation. These claims are also limited to gases from nuclear reactors. Janner, et al., in U.S. Pat. No. 4,311,674, shows that one isotope component of gases can be selectively excited from a condensate using radiation from a laser. Utilization of this process for separation of CO2 with 14C from atmospheric gases would first require the condensation of CO2 from all other atmospheric gases by increasing the pressure of the gases to exceed 5.1 bars, and then application of the described process. After this, the CO2 without 14C would then need to be re-combined with atmospheric gases without CO2. Together, with the added complexity of removing CO2 from atmospheric gases, liquification of the removed CO2 gas, application of the described process, and then recombination of gases, this approach is uneconomical for the filtration of atmospheric gases on a large-scale for agricultural production. Wikdahl, in U.S. Pat. No. 4,070,171, shows that gas mixtures can be separated by molecular or atomic weight by centrifugal force in a vortex. The described apparatus utilizes velocities exceeding the speed of sound and has been utilized for uranium isotope separation, which is among the most technically difficult isotope separation applications. This apparatus could be adapted for the less rigorous application of 14C separation, although the small diameter limits the utility for the filtration of atmospheric gases on a large-scale for agricultural production, and effective 14C separation can be achieved at lower velocities than those required for more demanding applications. Therefore, this apparatus would be less economical than an alternative that does not require such extremely high velocities, which limits efficiency, and such a small diameter, which limits the volume of throughput. Mangadoddy, et al., in U.S. Pat. No. 9,579,666 B2, shows that dense medium can be separated by centrifugal force in a vortex. Although this apparatus appears very similar to Wikdahl's apparatus, as noted above, it has a larger diameter, is intended for the separation of particles rather than molecules, and is functional at lower velocities. Although this apparatus was not intended for isotope separation, and that subject is outside the scope of the claims, it could be modified and adapted for the application of separating CO2 with 14C from atmospheric gases. In conclusion, no method or process has been formerly developed for maintaining a controlled environment from which CO2 with 14C has either been removed or reduced to a lower level than the natural abundance of 14C, as required for growing agricultural products with reduced 14C content. Similarly, no apparatus has been formerly developed with the specific intent to efficiently and economically remove CO2 with 14C from atmospheric gases with a single filtration pass, as required for large scale agricultural production. A process to grow agricultural products with a reduced abundance of radioactive 14C will have health benefits by reducing harmful damage to human DNA, which has been correlated to cancer. Other benefits of reduced cumulative genetic damage over long periods of time have yet to be quantified. To-date, removal of 14C from agricultural products on a large scale has not been possible due to a lack of an economical means to remove 14C from CO2 on a scale sufficient for agricultural production. Such agricultural products can be grown in a large variety of controlled environments so long as they are airtight, such as a sealed container, greenhouse, or building, and provided the other requirements for agricultural growth are also satisfied, such as light, water, and micronutrients. The controlled environment must be airtight so that the gases therein can be controlled and constitute filtered atmospheric gases from which CO2 with 14C has been removed. With the proper sensors, control valves, and control systems, 1) the abundance of CO2 in the controlled environment can be automatically maintained by circulating atmospheric gases through the filtration system, operating control valves, and circulation of fresh filtered air through the controlled environment, 2) to ensure the quality of the agricultural products, the control system can also ensure the filtration system is effective prior to routing filtered atmospheric gases into the controlled environment, and 3) the air pressure inside the controlled environment can be maintained at a positive pressure with respect to the external atmospheric air pressure, to prevent any leakage that could contaminate the controlled environment. Together with hydroponic growing methodologies, this process enables the complete automation of large scale agricultural production with reduced 14C. The bilateral and unilateral compression helikon vortex designs provide efficient, single-pass systems for the effective filtration of 14C from CO2 that is suitable for the filtration of large quantities of atmospheric gases as required for agricultural production (Patrick, A. D., & Patrick, B. E.). These designs are effective due to the relatively large mass difference between stable carbon and unstable carbon isotopes (i.e., 12C and 14C, which differ in mass by 16.7%), which is much less energy intensive to separate than the typical subjects of nuclear isotope separation, i.e., the heavy element isotopes of uranium, such as 235U and 238U, which differ in mass by 1.3% and require much more energy to separate. The designs also benefit from the fact unlike uranium, which is a scarce resource and cannot be wasted, atmospheric gases are relatively abundant and available for filtration at no material cost. Therefore, if a portion of perfectly usable air is lost as “waste” from the filtration process, there is no material cost for the separation process, and consequently, the filtration process does not require a high level of material efficiency to be successful or effective at removing 14C. The designs are simple without requiring electromagnets or electrodes for the ionization of gas, like some isotope separation methodologies. Also, many of the designs that utilize or require the ionization of gas are more complex and resource intensive to construct and operate. The single-pass system designs are also efficient without requiring a multi-stage cascade design, which requires many more resources to build than a single-pass filtration system, as well as much more energy to operate. The designs are more efficient in both design and operation than any of the designs that require liquification of the gases, or ionization of liquified gases, which introduce the process complexities of liquifying atmospheric gases, the maintenance hazards of operating with highly pressurized systems, and the recombination of filtered gases after liquification. The designs are also more efficient and economical than processes that would require converting CO2 to plasma and stabilizing ionized plasma with magnetic fields. Since the designs only require the acceleration of atmospheric gases, they are also more efficient than processes that require ionization and processing of each molecule of gas in mixtures of gases being separated. Since the designs utilize atmospheric gases directly, they do not require condensation of gases from nuclear power plants or require the excitation of condensates by lasers, which would only add inefficiencies. The designs do not require the acceleration of gases to velocities exceeding the speed of sound, which is required for centrifugal gas separation methodologies applied to more technically difficult isotope separation applications. The designs also do not require the very small diameter of apparatus required by centrifugal gas separation systems intended for more technically challenging isotope separation applications. Since the designs are effective at lower velocities and larger diameters, they are more efficient and well suited for the high throughput of atmospheric gases volumes required for large scale agricultural production applications. The designs are not constrained by particulate separation, only the densities of atmospheric gases, and any particulates that enter the designs would generally be discarded with the high-density atmospheric gases, including the CO2 with 14C. The designs are intended to efficiently and economically remove CO2 with 14C from atmospheric gases with a single-pass filtration, as required for large scale agricultural production. FIG. 1. is a flow diagram for the separation of atmospheric gases to remove CO2 with 14C in accordance with the process, control system, and Helikon Vortex Bilateral and Unilateral Compression designs within the invention. The Helikon Vortex 1 (see FIG. 2 or FIG. 3 for details) constitutes a means to remove CO2 with 14C from the atmospheric gases 2. Several alternative processes or apparatus could substitute 1 in this flow diagram, with respective losses of efficiency as described in the background section, and constitute an alternative means to remove CO2 with 14C from 2. The atmospheric pressure p1 of the atmospheric gases 2 is measured by pressure sensor 3 and CO2 abundance c1 in the atmospheric gases 2 is measured by CO2 sensor 4, both of which are monitored by a control system 13. A commercial high-speed air blower 5, which can be activated by the control system 13, accelerates the atmospheric gases to velocity v and volume V0 per second which is output directly into an airflow adapter 6 which is connected to the vortex chamber 7, into which the air is injected tangentially to maximize centrifugal acceleration. A cone 8 which is aligned with the vortex chamber 7 by the vortex exhaust/cone alignment base 9. The position of the cone 8 can be raised or lowered relative to the vortex chamber 7 to reduce or widen the gap between the vortex chamber 7 and the cone 8. The positioning of the cone 8 to achieve the desired separation is hereafter referred to as calibration. Dense molecular gas 10 is forced to the outside of the vortex chamber 7 by centrifugal acceleration a and exits the vortex chamber 7 through the gap near the cone 8, where it is exhausted to the atmosphere, reentering the atmospheric gases 2. Low density molecular gas 11 with reduced 14C content is slowed by the cone 8 and exits the vortex chamber opposite the cone at the top. The calibration (or cone position) can be adjusted by an electrical motor 12 which can raise or lower the cone 8 position relative to the vortex chamber 7 through axial rotation. Low density molecular gas 11 can exit through either manual or solenoid operated electrical control valves 14 and 17, which can be controlled by the control system 13. Control valve 14 is a relief valve which opens and releases gases while the high-speed blower 5 is starting, while the vortex chamber is pressurizing, or while the cone position is changing during calibration. CO2 abundance c2 of the relief valve gas output 15 is measured at CO2 sensor 16 and monitored by the control system 13. Once the vortex chamber 7 is pressurized and CO2 separation is adequate per the helikon vortex calibration, relief control valve 14 is closed and the vortex chamber control valve 17 is simultaneously opened by the control system 13. CO2 separation is adequate when CO2 sensor calibration adjusted measurements c2/c1<S, where the required separation S<1, and S is dependent on the efficiency of the helikon vortex. While the vortex chamber control valve 17 (i.e., the control valve for gaseous input to the controlled environment) is open, the CO2 abundance c3 of the vortex chamber control valve output 18 is monitored by CO2 sensor 19 to ensure CO2 separation is adequate, per the helikon vortex calibration, and proper operation of the vortex. CO2 separation is adequate when CO2 sensor calibration adjusted measurements c3/c1<S. The vortex chamber control valve output 18 passes directly into a controlled environment 20 which can be used for applications requiring CO2 with reduced 14C content (e.g., agricultural production applications). The pressure p2 of gases inside the controlled environment 20 is measured by a pressure sensor 21 and monitored by the control system 13 with to ensure a positive pressure (i.e., p2>p1) is maintained inside the controlled environment 20 to preclude contamination with CO2 containing 14C in the event of a leak or rupture. Control valve 22 remains closed while p2<p1 when 17 is open until 20 has a positive pressure differential over the atmospheric pressure (as determined by comparing pressure sensors 3 and 21), or p2>p1+p0, where p0 is the minimum additional pressure required by 20, to ensure atmospheric gases 2 do not enter 20 through 22. When control valve 17 is open and a sufficient positive pressure exists in the controlled environment 20, or p2>p1+p0, control valve 22 will be opened by the control system 13, allowing controlled environment gases 23 to exit through 22, where it is exhausted to the atmosphere, reentering atmospheric gases 2. Control valve 22 may also be opened by 13 when atmospheric pressure p1 decreases so that p2>p1+2*p0, as an emergency relief, to ensure the pressure in 20 is not so high that controlled environment gases 23 do not enter 7 through 17 when 17 is opened. When p1 is rising, 13 can also turn on 5 to increase p2 to maintain a positive pressure in 20; as described above, 5 pressurizes 7, whereby 17 is opened, increasing p2. When CO2 abundance decreases in 20 due to utilization or consumption by applications, as measured by c3, and c3<c0, where c0 is the minimum CO2 abundance required by 20, 13 will turn on 5 to replace the controlled environment gases in 20. In this manner, 13 can regulate both the pressure and CO2 abundance in the controlled environment 20 as the natural atmospheric pressure p1 of 2 fluctuates and CO2 with reduced 14C content is utilized in 20. The control system 13 can either be programmed or configured to operate 5, 14, 17, and 22 utilizing electronic controls or switches with digital or analog signals, constituting a means to operate the blower and control valves. Similarly, 13 can either be programmed or configured to monitor digital or analog signals from 3, 4, 16, 19, and 21, constituting a means to monitor the sensors. FIG. 2 is a Bilateral Compression Helikon Vortex Overview, with a front view (FIG. 2a), top view (FIG. 2b), and right-side view (FIG. 2c), and cross-section of the tangential airflow stabilizer (FIG. 2d). This assembly is one instantiation of the helikon vortex 1 in FIG. 1, and several components from FIG. 1 are recognizable here, including the airflow adapter 6, helikon vortex chamber 7, cone 8, and helikon vortex exhaust/cone alignment base 9. The vortex output adapter 24 is where CO2 with reduced 14C content is output, and this is attached to the narrow vortex chamber cap/outlet 25, which is on top of 7. The vortex chamber consists of the upper narrow vortex chamber 26, extends through the center of the upper lateral vortex chamber adapter 27, the center of the airflow adapter 6, the center of the lower lateral vortex chamber adapter 32, and the lower narrow vortex chamber 33. The upper and lower narrow vortex chambers have an interior radius of r1 and combined height of h1, where the height of 26 is less than or equal to half the height of 33. The airflow adapter 6 consists of several components identifiable here, including the blower input connector 28, radial to tangential airflow adapter 29, tangential airflow stabilizer 30, and the wide vortex chamber with tangential input 31. The wide vortex chamber has an interior radius of r2 and height of h2, and is connected to the narrow vortex chambers 26 and 33 of interior radius r1 by 27 and 32, each with a height h3. The blower input connector 28 is a circular adapter with an interior radius of r0 and thickness of to for an exterior radius of r0+t0, providing a cross-section area of πr02 for V0 per second of input from the high-speed blower 5. The radial to tangential airflow adapter 29 changes the radial airflow at 28 to a vertical stream at the tangential airflow stabilizer 30 with an interior stream height of h0, a maximum width of w0 where πr02≥h0w0. The stream cross-section 34 can be compressed to increase pressure in the vortex chamber or to achieve a higher input velocity based on the performance of 5. The stream can also be tapered or shaped at the top and bottom excluding wedges from the tangential airflow 35 of height h4 and width w1 from the tangential edge closest to the center of the vortex chamber (See FIG. 2d), where h4≤h0/2 and w1<w0, yielding a cross section area of h0w0−h4w1≤πr02, to evenly distribute pressure in 31 as gases are compressed in 27 and 32. Below the vortex chamber 7, the cone 8 is held in a position aligned with the center of 7 by the helikon vortex exhaust/cone alignment base 9 which is attached to the bottom of 33. The position of 8 can be adjusted for calibration of the helikon vortex while remaining in alignment with the lower narrow vortex chamber 33. The top view (FIG. 2b) obstructs components below 31, but shows reinforcement for the tangential airflow 36, which is also visible on the right-side view (FIG. 2c). The interior volume of the Bilateral Compression Helikon Vortex as defined isV=πr12h1+πr22h2+2π(r12+r1r2+r22)h3/3. FIG. 3 is a Unilateral Compression Helikon Vortex Overview, with a front view (FIG. 3a), top view (FIG. 3b), and right-side view (FIG. 3c), and cross-section of the tangential airflow stabilizer (FIG. 3d). This assembly is one instantiation of the helikon vortex 1 in FIG. 1, and several components from FIG. 1 are recognizable here, including the airflow adapter 6, helikon vortex chamber 7, cone 8, and helikon vortex exhaust/cone alignment base 9. The vortex output adapter 24 is where CO2 with reduced 14C content is output, and this is attached to the wide vortex chamber cap/outlet 37, which is on top of 6. The vortex chamber consists of the lower narrow vortex chamber 33, and extends through the lower lateral vortex chamber adapter 32, and the center of the airflow adapter 6. The lower narrow vortex chamber has an interior radius of r1 and height of h1. The airflow adapter 6 consists of several components that are identifiable here, including the blower input connector 28, radial to tangential airflow adapter 29, tangential airflow stabilizer 30, and the wide vortex chamber with tangential input 31. The wide vortex chamber has an interior radius of r2 and height of h2, and is connected to the narrow vortex chamber 33 of interior radius r1 by 32, with a height h3. The blower input connector 28 is a circular adapter with an interior radius of r0 and thickness of to for an exterior radius of r0+t0, providing a cross-section area of πr02 for V0 per second of input from the high-speed blower 5. The radial to tangential airflow adapter 29 changes the radial airflow at 28 to a vertical stream at the tangential airflow stabilizer 30 with an interior stream height of h0, a maximum width of w0 where πr02≥h0w0. The stream cross-section 34 can be compressed to increase pressure in the vortex chamber or to achieve a higher input velocity based on the performance of 5. The stream can also be tapered or shaped at the bottom excluding a wedge from the tangential airflow 35 of height h4 and width w1 from the tangential edge closest to the center of the vortex chamber (See FIG. 3d), where h4≤h0/2 and w1<w0, yielding a cross section area of h0w0−h4w1/2≤πr02, to evenly distribute pressure in 31 as gases are compressed in 32. Below the vortex chamber 7, the cone 8 is held in a position aligned with the center of 7 by the helikon vortex exhaust/cone alignment base 9 which is attached to the bottom of 33. The position of 8 can be adjusted for calibration of the helikon vortex while remaining in alignment with the lower narrow vortex chamber 33. The top view (FIG. 3b) obstructs components below 31, but shows reinforcement for the tangential airflow 36, which is also visible on the right-side view (FIG. 3c). The interior volume of the Unilateral Compression Helikon Vortex as defined isV=πr12h1+πr22h2+π(r12+r1r2+r22)h3/3. FIG. 4 is a Perspective View of a Bilateral Compression Helikon Vortex (FIG. 4a) and a Perspective View of a Bilateral Compression Helikon Vortex (FIG. 4b). FIG. 5 is a Wide Vortex Chamber with Tangential Input Overview, with a front view (FIG. 5a), back view (FIG. 5b), top view (FIG. 5c), and right-side view (FIG. 5d). On all four views, the blower input connector 28, the radial to tangential airflow adapter 29, and the wide vortex chamber with tangential input 31 are visible. On all but the right-side view, the tangential airflow stabilizer 30 is visible. Cross-sections of 30 are provided in FIGS. 2d and 3d, detailing the interior cross-section area of the tangential airflow stabilizer 34 and variable exclusion wedges 35 detailed above, as related to the radius r0 of 28. The outer reinforcement for the tangential airflow 39 are clearly seen on FIG. 5b, FIG. 5c, and FIG. 5.d. These are evenly spaced vertically and centered around the input axis of 28, providing reinforcement for both 30 and 31 near the tangential input. The inner reinforcement for the tangential airflow 40 are seen on FIG. 5c and FIG. 5d, and are also evenly spaced vertically and centered around the input axis of 28, providing reinforcement for both 30 and 31 near the tangential input. FIG. 6 is a Perspective View of a Wide Vortex Chamber with Tangential Input. From this front-upper perspective view the tangential airflow vent 41 is visible inside 31, which was not visible from any of the four views on FIG. 5. As illustrated in FIG. 6, 41 has tangential dimensions with a height of h0 and width of w0 and is configured for either a bilateral or unilateral helikon vortex configuration with h4=0 and w1=0, omitting any exclusion wedges (i.e., 35) from the tangential airflow. The airflow adapter 6, as seen on FIGS. 1, 2, and 3, utilizes 28, 29, 30, and 35, as seen on FIGS. 2 and 3, to constitute a means to stabilize and shape the airflow of said atmospheric gases 2 into 34, as seen on FIGS. 2 and 3, prior to passing through 41 into 31, as seen here on FIG. 6. FIG. 7 is a Lateral Vortex Chamber Adapter Overview, with a front view (FIG. 7a), upper-front perspective view (FIG. 7b), and lower-front perspective view (FIG. 7c). The lateral vortex chamber adapter is utilized twice in the bilateral compression helikon vortex configuration 27 and 32, and once in the unilateral compression helikon vortex configuration 32. The lateral adapter 44 connects to a wide vortex chamber 32 with a wide vortex chamber connector 42 and connects to a narrow vortex chamber to a narrow vortex chamber 26 or 33 with a narrow vortex chamber connector 43. As illustrated in FIG. 7b, the interior of the narrow vortex chamber connector 45 has a radius equal to the outside radius of the narrow vortex chamber (See FIG. 8). The interior of the lateral adapter 47 is a smooth surface in the shape of a truncated cone and has a radius of r1 at the minimum radius at the edge shared with 45. The interior of the wide vortex chamber connector 46 has a radius equal to the outside radius of the wide vortex chamber 31. The maximum radius of 47 is equal to r2 at the edge shared with 46. Thereby, 47 provides a smooth surface inside the vortex chamber of height h3 between 45 and 46 for the compression of gases for separation by centrifugal acceleration while connecting wide and narrow vortex chamber components. FIG. 8 is a Narrow Vortex Chamber Overview, with a front view (FIG. 8a), top view (FIG. 8b), and upper-front perspective view (FIG. 8c). The narrow vortex chamber is utilized twice in the bilateral compression helikon vortex configuration 26 and 33, and once in the unilateral compression helikon vortex configuration 33. To reduce helikon vortex manufacturing costs, commercial pipe with standard inner and outer diameters can be utilized for narrow vortex chambers by sizing the connectors on all connecting components, including 9, 25, 27, and 32, to match the outer and inner diameters of standard commercial pipe(s). For instance, the interior diameter of narrow vortex chamber connector 45 must match the outer diameter of the exterior of the narrow vortex chamber 49, and the minimum interior diameter of 47 must match the interior diameter of 48. An example of adapting a commercial pipe would be a 3 inch Schedule 40 PVC pipe, in which case the outer diameter of 49 would be 88.9 mm and the interior diameter of 48 would be 76.2 mm. Any commercial pipes must be cleaned with solvents and in the case of plastic or related synthetic polymers (e.g., polyvinyl chloride), they must be rigid and the interior of the narrow vortex chamber 48 must be coated with an antistatic treatment prior to utilization. FIG. 9 is a Narrow Vortex Chamber Cap/Outlet Overview, with a front view (FIG. 9a), top view (FIG. 9b), top upper-front perspective view (FIG. 9c), and lower-front perspective view (FIG. 9d). The narrow vortex chamber cap/outlet 25 is utilized in the bilateral compression helikon vortex, and the vortex output adapter 24 is visible in FIG. 9a, FIG. 9b, and FIG. 9c. The top of the narrow vortex chamber cap 50 is visible on FIG. 9b and FIG. 9c. To reduce helikon vortex manufacturing costs, the interior dimensions of the vortex output adapter 24 are intended to connect to commercial pipe with standard inner and outer diameters. The interior of vortex output adapter 51, visible in FIG. 9b, FIG. 9c, and FIG. 9d, has a diameter matching the outer diameter of a commercial pipe, while the vortex chamber cap outlet 52, visible in FIG. 9b and FIG. 9.d, has a diameter matching the interior diameter of the same matching commercial pipe. E.g., when connecting 24 to a ½ inch Schedule 40 PVC pipe, the matching dimensions for 51 would be a diameter of 21.33 mm and 52 would be a diameter of 15.80 mm. The bottom of 50 is visible in FIG. 9d, which must be a smooth anti-static surface, like the other interior components of the helikon vortex. FIG. 10 is a Wide Vortex Chamber Cap/Outlet Overview, with a front view (FIG. 10a), top view (FIG. 10b), top upper-front perspective view (FIG. 10c), and lower-front perspective view (FIG. 10d). The wide vortex chamber cap/outlet 37 is utilized in the unilateral compression helikon vortex, and the vortex output adapter 24 is visible in FIG. 10a, FIG. 10b, and FIG. 10c. The top of the wide vortex chamber cap 53 is visible on FIG. 10b and FIG. 10c. To reduce helikon vortex manufacturing costs, the interior dimensions of the vortex output adapter 24 are intended to connect to commercial pipe with standard inner and outer diameters. The interior of vortex output adapter 51, visible in FIG. 10b, FIG. 10c, and FIG. 10d, has a diameter matching the outer diameter of a commercial pipe, while the vortex chamber cap outlet 52, visible in FIG. 10b and FIG. 10d, has a diameter matching the interior diameter of the matching commercial pipe. E.g., when connecting 24 to a ½ inch Schedule 40 PVC pipe, the matching dimensions for 51 would be a diameter of 21.33 mm and 52 would be a diameter of 15.80 mm. The bottom of 53 is visible in FIG. 10d, which must be a smooth anti-static surface, like the other interior components of the helikon vortex. FIG. 11 is a Manually Calibrated Helikon Vortex Cone Overview, with a front view (FIG. 11a), top view (FIG. 11b), and lower-front perspective view (FIG. 11c). The manually calibrated helikon vortex cone is one instantiation of 8 which can be utilized in either Bilateral or Unilateral Helikon Vortex configurations. The effective surface of the cone 54 is visible in FIG. 11a, FIG. 11b, and FIG. 11c. This surface must be a smooth anti-static surface, like the other interior components of the helikon vortex. The base of the cone 55 is visible in FIG. 11a and FIG. 11c. In the center of the base of the cone is the threaded core of the cone 56 which is visible in FIG. 11c. To reduce helikon vortex manufacturing costs, the threads are industry standard fine thread count and diameter so that the manually calibrated helikon vortex cone can be used with industry standard bold sizes. E.g., an industry standard ⅜″ bolt size has a fine thread count of 24 threads per inch (TPI). FIG. 12 is a Vertical Cross-Section View of the Manually Calibrated Helikon Vortex Cone (FIG. 12a), and a Horizontal Cross-Section View of the Manually Calibrated Helikon Vortex Cone (FIG. 12b). The effective surface of the cone 54 is visible in FIG. 12a on the upper external surface of the vertical cross-section, while the base of the cone 55 is visible on the bottom. The effective surface of the cone 54 is visible in FIG. 12b on the outer circumference of the horizontal cross-section. The threaded core of the cone 56 is visible on FIGS. 12a and 12b. To reduce helikon vortex manufacturing costs, the interior of the cone 57 is hollow, as seen on FIGS. 12a and 12b, precluding the utilization of unnecessary materials. The base of the cone is reinforced in three ways. First, a thick area of material reinforcement for the threaded core 58 is provided around 56, as seen on FIGS. 12a and 12b. Second, radial reinforcement structures 59 and 60 extend from 58 (i.e., near the center of the cone) to 54 (i.e., the outside of the cone), as seen on FIG. 12b. Third, and finally, a circular reinforcement structure 61 goes around the base of the cone and 56, as seen on FIGS. 12a and 12b, connecting the inner radial reinforcement structures 59 to the outer reinforcement structures 60. The inner and outer reinforcement structures, 59 and 60, are distributed at even intervals of angles around the central axis of the cone, but the angles separating structures for 59 and 60 are not necessarily equal, as seen on FIG. 12b, where six 59 are connected to 61 and eight 60 structures are connected to 61. Larger cones may have multiple circular reinforcements 61, in concentric circles, each connected by radial reinforcement structures, such as 59 or 60, while smaller cones may not require a circular reinforcement structure 61 and only a single set of radial reinforcement structures, such as 59, which would then directly connect 58 to 54. FIG. 13 is an Alternative Threaded Cone Overview, with a front view (FIG. 13a), bottom view (FIG. 13b), and lower-front perspective view (FIG. 13c). The alternative threaded cone differs from the manually calibrated helikon vortex cone in FIG. 11 in that it has no threaded core 56 and instead has a single threaded extrusion 62 and multiple axial alignment extrusions 63, as seen on FIGS. 13a, 13b, and 13c. The extrusions 62 and 63 are aligned with the central axis of the cone, with 62 being on the central axis as seen from the bottom view in FIG. 13b. One or more axial alignment extrusions, 63, appear around the central axis, with four visible on FIGS. 13b and 13c. The alternative threaded cone is intended for use with an electric motor 12 and the vortex exhaust/alternative threaded cone alignment base on FIGS. 15 and 16. FIG. 14 is a Vortex Exhaust/Cone Alignment Base Overview, with a front view (FIG. 14a), top view (FIG. 14b), and bottom view (FIG. 14c). The vortex exhaust/cone alignment base 9 is utilized with the cone 8 illustrated in FIG. 11 and has several critical functions. First, the bottom of the base 64, visible on FIGS. 14a, 14b and 14c, is held perpendicular to the central axis of the lower vortex chamber 7 via the connector to the vortex chamber 65, visible on FIGS. 14a and 14b, which attaches to the lower narrow vortex chamber 33. The inner diameter of 65 matches the outer diameter of 33 for alignment, and is large enough for the base of the cone 8 to be lowered into 9. Second, two or more vertical vent fins 66, visible on FIGS. 14a, 14b, and 14c, are symmetrically distributed around the central axis of 9, connecting 64 to 65, while being tangential to airflow from 33. The gaps between 66 permit exhaust to exit from the vortex chamber 9. Third, the bottom of the base 64 is structurally reinforced to hold the cone 8 in alignment with the central axis of the lower vortex chamber 7 with one or more circular reinforcements 67, visible on FIGS. 14a and 14b, symmetrically distributed radial reinforcements 68, visible on FIG. 14b, and a central reinforcement 69, visible on FIG. 14b, around the center of 64. The structural reinforcements 67, 68, and 69 support the alignment of the cone 8 while precluding the utilization of unnecessary materials. At the top of the base, 65 is contoured to maximize surface area with 66 to add structural strength. The cone is held in place by a commercial hex that is inserted from the bottom of 64 into the cylindrical hollow central shaft of the base 70, visible on FIGS. 14b and 14c. The hex head of the bolt fits into the base hex nut intrusion 71 which is visible on FIG. 14c. Therefore, the manually calibrated helikon vortex cone 8, in FIG. 11, can be attached to this vortex exhaust/cone alignment base 9, in FIG. 14, with a commercial hex bolt. The cone can be lowered by turning it clockwise, from the top view, down onto the threaded bolt, and raised by turning it counter-clockwise. When the cone is in a lower position there is a larger gap between the cone 8 and the lower narrow vortex chamber 33, allowing a larger volume of atmospheric gases to exhaust out of 7. These exhaust gases, which exit below 65 on FIG. 14a between the vent fins 66, are the densest atmospheric gases, being on the outside perimeter of 7 while under centrifugal acceleration. FIG. 15 is a Perspective View of the Vortex Exhaust/Cone Alignment Base, with an upper-front perspective view (FIG. 15a) and a lower-front view perspective view (FIG. 15b). All the reference numerals in FIG. 14 are visible in FIG. 15. On FIG. 15a, the circular and radial structural supports 67 and 68 can be seen to rise above the base 64, providing reinforcement to 69. The outermost circular structural support 67 also provides more surface area and structural support for 66 to attach to the base 64. The intrusion for the hex bolt 71 can be clearly seen on FIG. 15b in the center of the base 64. The variable outer diameter of 65 can also be seen on FIG. 15b, reducing materials required for construction while enhancing the surface are and structural support for 66 to attach to the connector 65. The vortex exhaust/cone alignment base 9 utilizes a hex bolt held stationary in axial alignment by 69, 70, and 71, and held in alignment with the lower narrow vortex chamber 33, as seen on FIGS. 2 and 3, by 65 and a plurality of 66, while said hex bolt is threaded into cone 8 holding 8 in axial alignment by 56 and 58, which are reinforced by 61 and a plurality of 59 and 60, as seen on FIG. 12, while 8 can be rotated clockwise and counter-clockwise to raise and lower position of 8 inside 33, constitutes a means to position said cone 8 inside said lower narrow vortex chamber 33. FIG. 16 is a Vortex Exhaust/Alternative Threaded Cone Alignment Base Overview, with a top view (FIG. 16a), and bottom view (FIG. 16b). The vortex exhaust/alternative cone alignment base 9 is utilized with the alternative threaded cone 8 illustrated in FIG. 13 and differs by the vortex exhaust/cone alignment base 9 illustrated in FIG. 14 in a few ways. First, instead of a smooth hollow central shaft 70, this base has a threaded central shaft 72, as seen on FIGS. 16a and 16b. Second, instead of the central reinforcement 69 being immediately around 70, there is a circular central shaft 73 that can rotate clockwise and counter-clockwise, as seen on FIGS. 16a and 16b. Third, the central reinforcement for the base 69 goes around 73 in this configuration, as seen on FIG. 16a. Fourth, there are axial alignment shafts 74 which extend through the radial reinforcements 68 and the base 64, as seen on FIGS. 16a and 16b. The front view of this configuration of 9 appears to be the same as FIG. 14a. The axial alignment extrusions 63 on the alternative threaded cone 8 extend through the axial alignment shafts 74 as the threaded extrusion 62 is threaded into 72. Together, the alignment extrusions 62 and shafts 74 align the cone 8 with the vortex chamber 7, as the cone position is raised and lowered by rotating 73 clockwise and counter-clockwise. Fifth, an axial alignment shaft reinforcement 75 is around each shaft 74 to reinforce the radial reinforcements 68, as seen on FIG. 16a. Finally, there is a motor attachment mount 76 on the bottom of 73, as seen on FIG. 16b. This is where an electrical motor 12 can be attached to rotate 73 to raise and lower the cone 8 via a control system 13 to automate the calibration process. FIG. 17 is a Perspective View of the Vortex Exhaust/Alternative Threaded Cone Alignment Base, with an upper-front perspective view (FIG. 17a) and a lower-front view perspective view (FIG. 17b). All the reference numerals in FIG. 16 are visible in FIG. 17. On FIG. 17a, the axial alignment shaft reinforcement 75 can be seen having a similar height to the radial, circular, and central reinforcement structures 67, 68, and 69. The circular central shaft 73 can be seen extending from the center of 69 in FIG. 17a to the center of 64 on FIG. 17b, where the motor attachment mount 76 is located. The other functions of 64, 65, 66, 67, 68, and 69 identified on FIG. 15 above are applicable here. The vortex exhaust/alternative threaded cone alignment base 9 utilizes a threaded central shaft 72 that is held in axial alignment by 69 and 73, and reinforced by a plurality of 68, and held in alignment with the lower narrow vortex chamber 33, as seen on FIGS. 2 and 3, by 65 and a plurality of 66, while 72 is threaded onto 62 of cone 8, as seen on FIG. 13, holding 8 in axial alignment by a plurality of extrusions 63 which are inserted into 74, which are reinforced by 68 and 75, while 76 can be rotated clockwise and counter-clockwise manually or by an electric motor 12 to raise and lower the position of 8 inside 33, constitutes a means to position said cone 8 inside said lower narrow vortex chamber 33. 1 helikon vortex 2 atmospheric gases 3 pressure sensor for atmospheric gases 4 CO2 sensor for atmospheric gases 5 high-speed blower 6 airflow adapter 7 helikon vortex chamber 8 helikon vortex cone 9 helikon vortex exhaust/cone alignment base 10 dense molecular gas (vortex chamber exhaust) 11 low density molecular gas (vortex chamber product) 12 electrical motor 13 control system 14 relief control valve 15 relief valve gas output 16 relief valve output CO2 sensor 17 vortex chamber control valve or controlled environment gaseous input control valve 18 vortex chamber control valve output 19 vortex chamber control valve output CO2 sensor 20 controlled environment 21 pressure sensor for controlled environment 22 controlled environment gaseous output control valve 23 controlled environment exhaust 24 vortex output adapter 25 narrow vortex chamber cap/outlet 26 upper narrow vortex chamber 27 upper lateral vortex chamber adapter 28 blower input connector 29 radial to tangential airflow adapter 30 tangential airflow stabilizer 31 wide vortex chamber with tangential input 32 lower lateral vortex chamber adapter 33 lower narrow vortex chamber 34 interior cross-section area of tangential airflow stabilizer 35 excluded wedge from tangential airflow 36 reinforcement for the tangential airflow 37 wide vortex chamber cap/outlet 39 outer reinforcement for the tangential airflow 40 inner reinforcement for the tangential airflow 41 tangential airflow vent 42 narrow vortex chamber connector 43 wide vortex chamber connector 44 lateral adapter 45 interior of narrow vortex chamber connector 46 interior of wide vortex chamber connector 47 interior of lateral adapter 48 interior of narrow vortex chamber 49 exterior of narrow vortex chamber 50 narrow vortex chamber cap 51 interior of vortex output adapter 52 vortex chamber cap outlet 53 wide vortex chamber cap 54 effective surface of cone 55 base of cone 56 threaded core of cone 57 hollow interior of cone 58 reinforcement for threaded core of cone 59 inner radial reinforcement structure for cone 60 outer radial reinforcement structure for cone 61 circular reinforcement for cone 62 threaded extrusion 63 axial alignment extrusion 64 bottom of base 65 connector to vortex chamber 66 vent fin 67 circular reinforcement for base 68 radial reinforcement for base 69 central reinforcement for base 70 hollow central shaft 71 base hex nut intrusion 72 threaded central shaft 73 circular central shaft 74 axial alignment shaft 75 axial alignment shaft reinforcement 76 motor attachment mount The operation for growing agricultural products with reduced 14C content requires a controlled environment 20 with filtered atmospheric gases 2 from which CO2 with 14C has been removed. 1. A filtration system comprising a blower 5 and a helikon vortex 1 constitutes a means to remove CO2 with 14C from atmospheric gases 2; blower 5 output velocity of 322 km per hour or greater is required for effective filtration with helicon vortex 1; 2. Control valves 17, 22 are required to control the flow of gases entering and exiting the controlled environment 20; 3. When the CO2 sensor 19 inside the controlled environment 20 detects a CO2 abundance lower than a predetermined amount, the said filtration system is turned on by the control system 13 and the relief control valve 14 is opened; 4. The CO2 sensor 16 at the relief output is monitored and compared to the CO2 sensor 4 for atmospheric gases 2 outside the controlled environment to ensure said filtration system removal of CO2 with 14C from atmospheric gases 2 is effective by detecting a predetermined delta which can be determined by said filtration system efficiency;5. Once effective filtration is verified, the control system 13 closes the relief control valve 14 and opens control valves 17, 22 which are connected to the controlled environment 20;6. When the CO2 sensor 19 inside the controlled environment 20 detects a CO2 abundance above a predetermined amount, the said filtration system is turn off and the control valves 17, 22 are closed by the control system 13;7. When the controlled environment input control valve 17 is open, the output control valve 22 is only opened by the control system 13 when the air pressure inside the controlled environment 20 as measured by the air pressure sensor 21 exceeds the atmospheric gas air pressure outside of the controlled environment by a predetermined amount as measured by air pressure sensor 3;8. Operation of said filtration system is initially required for a duration sufficient to replace the entire volume of air inside the controlled environment 20. Thereafter, continuous, periodic, or intermittent operation as determined by CO2 sensor 19, as detailed above, may be used to determine periods of operation for the filtration system to maintain sufficient CO2 levels inside the controlled environment 20;9. The control system 13 can either be programmed or configured to operate 5, 14, 17, and 22 utilizing electronic controls or switches with digital or analog signals, constituting a means to operate the blower and control valves. Similarly, 13 can either be programmed or configured to monitor digital or analog signals from 3, 4, 16, 19, and 21, constituting a means to monitor the sensors.10. Helikon vortex 1 above may comprise either a bilateral compression helikon vortex or a unilateral compression helikon vortex as detailed below; effective filtration has been demonstrated with centrifugal acceleration exceeding 16,000 g, a maximum narrow vortex chamber radius of 5.08 cm, and a maximum height of 1.94 m.11. Bilateral compression helikon vortex (FIG. 2) consists of an airflow adapter 6 (consisting of blower input connector 28, radial to tangential airflow adapter 29, tangential airflow stabilizer 30, and exclusion wedge 35), vortex chamber 7 (consisting of a wide vortex chamber 31, upper narrow vortex chamber 26, lower narrow vortex chamber 33, upper lateral adapter 27, and lower lateral adapter 32), cone 8, exhaust/cone alignment base 9, vortex output adapter 24, and narrow vortex chamber cap/outlet 25;12. Unilateral compression helikon vortex (FIG. 3) consists of an airflow adapter 6 (consisting of blower input connector 28, radial to tangential airflow adapter 29, tangential airflow stabilizer 30, and exclusion wedge 35), vortex chamber 7 (consisting of a wide vortex chamber 31, lower narrow vortex chamber 33, and lower lateral adapter 32), cone 8, exhaust/cone alignment base 9, vortex output adapter 24, and wide vortex chamber cap/outlet 37;13. During operation, the atmospheric gases 2 are accelerated by blower 5 and enter the airflow adapter 6 were they are stabilized and shaped prior to tangential injection into the wide vortex chamber 31; Centrifugal acceleration occurs while the atmospheric gases are separated by molecular density in vortex chamber 7; after separation, the high-density gases exit 7 between 33 and 8, while low-density gases exit 7 through 24;14. Calibration of the helikon vortex is essential prior to operation and this is accomplished by adjusting the position of the cone 8 inside the narrow vortex chamber 33 to ensure effective separation of CO2 with 14C. For manual calibration, the vortex exhaust/cone alignment base 9 utilizes a hex bolt held stationary in axial alignment by 69, 70, and 71 (FIG. 15), while cone 8 can be rotated clockwise and counter-clockwise to raise and lower the position of 8 inside 33. Alternatively, the calibration process can be automated with an electric motor 12. The vortex exhaust/alternative threaded cone alignment base 9 utilizes a threaded central shaft 72 that is held in axial alignment by 69 and 73 (FIG. 16), holding 8 in axial alignment by a plurality of extrusions 63 (FIG. 13) which are inserted into 74, while 76 can be rotated clockwise and counter-clockwise by an electric motor 12 to raise and lower the position of 8 inside 33. 3,004,158October 1961Steimel, K.3,421,334January 1969McKinney, et al.62-283,594,573July 1971Gerber, H.3,925,036December 1975Shacter, J.55/1583,939,354February 1976Janes, G. S.250/484 3,942,975March 1976Drummond, et al. 75/10 R4,070,171January 1978Wikdahl55/4194,311,674January 1982Janner, et al. 204/157.224,584,073April 1986Laboda, et al. 204/157.24,638,674October 1983Redmann 73/863.124,816,209July 1986Schweiger376/309 7,332,715 B2February 2008Russ, et al.250/288 8,460,434June 2013Turner, et al.95/1179,579,666 B2February 2017Mangadoddy, et al.B04C 5/04 Feiverson, H. A., Glaser, A., Mian, Z., & Von Hippel, F. N., Unmaking the Bomb: A Fissile Material Approach to Nuclear Disarmament and Nonproliferation, The MIT Press, Cambridge, Mass., London, England (2014). Genome Reference Consortium (GRC) Human Genome Assembly build 38 (GRCh38), 24 Dec. 2013. Lander, E. S. et al., Initial sequencing and analysis of the human genome, Nature 409, 860-921 (2001). Moore, J. D. L., South Africa and Nuclear Proliferation, Palgrave Macmillan, New York, N.Y. (1987). Patrick, A. D., & Patrick, B. E., Carbon 14 decay as a source of somatic point mutations in genes correlated with cancer diagnoses, Stable Isotope Foundation, Grants Pass, Oreg., USA (2017). Purdom, C. E., Biological hazards of carbon-14, New Sci. 298, 255-257 (1962). Sassi, M., et. al., Carbon-14 decay as a source of non-canonical bases in DNA, Biochimica et Biophysica Acta 1840 526-534 (2014). Sender, R., Fuchs, S., & Milo, R., Revised estimates for the number of human and bacteria cells in the body, PLoS Biol 14(8): e1002533 (2016).
	claims	1. A device for suspending an x-ray grid, the device comprising:a first rotating frame configured to support the x-ray grid therein or thereon; andtwo first flexible hinge elements connected to said first rotating frame and mounting said first rotating frame for reversible rotation about a first axis. 2. The device according to claim 1, wherein said first flexible hinge elements are plastic hinges. 3. The device according to claim 1, wherein said first rotating frame and said first flexible hinge elements are formed in one piece from sheet metal. 4. The device according to claim 1, wherein said first flexible hinge elements are spring hinges. 5. The device according to claim 1, wherein said first rotating frame and said first flexible elements are formed from plastic material. 6. The device according to claim 1, which further comprises:a second rotating frame rotatably mounted about a second axis, wherein said first rotating frame is rotatably mounted in said second rotating frame; andtwo second flexible hinge elements connected to said second rotating frame and aligned along the second axis, about which said second rotating frame is reversibly rotatably mounted. 7. The device according to claim 6, which further comprises a retaining frame, in which said second rotating frame is rotatably arranged. 8. The device according to claim 6, wherein the first axis and the second axis lie in a common plane and extend perpendicularly to one another, wherein a gimbal is formed. 9. The device according to claim 6, wherein said second flexible hinge elements are plastic hinges. 10. The device according to claim 6, wherein said second rotating frame and said second flexible hinge elements are formed in one piece and from sheet metal. 11. The device according to claim 6, wherein said second flexible hinge elements are spring hinges. 12. The device according to claim 6, wherein said second rotating frame and said second flexible hinge elements are formed from plastic material. 13. The device according to claim 6, which further comprises:a motor-driven first linear drive configured to exert a force on said first rotating frame so as to deflect said first rotating frame about the first axis; anda motor-driven second linear drive configured to exert a force on said second rotating frame so as to deflect said second rotating frame about the second axis. 14. The device according to claim 13, wherein each of said first linear drive and said second linear drive is a spindle drive. 15. The device according to claim 1, which further comprises a motor-driven first linear drive disposed to exert a force on said first rotating frame so as to deflect said first rotating frame about the first axis. 16. The device according to claim 15, wherein said first linear drive is a spindle drive. 17. An x-ray arrangement, comprising:an x-ray emitter and an x-ray detector; andat least one device according to claim 1 disposed between said x-ray emitter and said x-ray detector. 18. A method for operating an x-ray arrangement, the method comprising:providing an x-ray arrangement according to claim 17; andselectively tilting and canting the x-ray grid about predeterminable angles.
041707540	summary	CROSS REFERENCE TO RELATED APPLICATION This application is related to the co-filed, co-pending application Ser. No. 622,918, filed Oct. 16, 1978 entitled "A Flexible Position Probe Assembly", now U.S. Pat. No. 4,052,686 which is assigned to the assignee of the present invention. This application is directed to the mechanical components and design which, when assembled, form an elongated flexible probe. BACKGROUND OF THE INVENTION A major requirement for a versatile and stable position probe assembly exists in the nuclear reactor field where knowledge of the precise position of control rods affecting the position of nuclear fuel rods is essential. The critical requirements for control rod position indicators for use in nuclear reactor facilities extend to both the mechanical design as well as to the electrical readout. Typically, the probe assemblies required for monitoring control rod position can vary in length from two to twelve feet, thus requiring substantial overhead clearance for installation. In many facility designs the overhead clearance required to accommodate conventional rigid control rod position probe assembly does not exist. Efforts to date to design a suitable sectionalized or flexible position probe assembly capable of being installed and removed from a reactor facility having a clearance less than the overall length of the probe have not proven to be totally reliable over an extended period of time. In addition to the mechanical deficiencies of prior art position probe assemblies, the analog manifestation of the position indication developed by these probe assemblies have not proven satisfactory due to the adverse effects of temperature changes, line voltage and frequency variations, noise pick-up, and, interference from ambient magnetic fields. There is disclosed herein with reference to the accompanying drawings a unique probe design capable of flexing in two planes for ease of installation in facilities with limited overhead clearance, and developing a phase encoded electrical indication of the position of a magnetically permeable material, and signal processing circuitry for converting said phase encoded signal to a direct digital indication of the position of the magnetically permeable material. SUMMARY OF THE INVENTION A flat flexible elongated strip is inserted through slots machined in the tubular passages of each of a plurality of tubular spools, on each of which is wound primary and secondary windings to form a core section. The length of the flexible strip and the number of core sections, wherein each core section may measure 1/2 inch in length, positioned on the flexible band is a function of the required operational length of the probe assembly. The core sections are maintained in adjacent contacting relationship by a spring element at one end of the probe assembly. The spring element permits flexing of the probe assembly in either of two opposite directions defining a plane perpendicular to the surfaces of the flexible strip corresponding to the width of the flat flexible strip. The flat flexible strip effectively divides the tubular passages of each of the core sections in half, such that the alignment of adjacent core sections in contacting relationship about the flexible strip forms two longitudinal internal probe passages which are electrically isolated. Lead wires from the primary windings of the respective core sections extend through one of the internal passages to an external electrical terminal, while lead wires from the secondary windings of the respective core sections extend through the second internal probe passage to an external electrical terminal. The isolation of the primary and secondary leads within the longitudinal passages eliminates electrical interference, and further the routing of the electrical leads via the internal probe passages isolates electrical leads from adverse ambient conditions. Typically, the movement of a predetermined magnetically permeable element or object, such as a control rod in a nuclear reactor installation, or a magnetically permeable liquid, such as a liquid metal, is monitored indirectly as a function of the coupling and decoupling of one or more of the plurality of differential transformer core sections comprising the position probe assembly. Since the successive coupling or decoupling of adjacent differential transformer core sections in response to movement of the magnetically permeable material, follows the dimensional length of each core section, the accuracy of position indication is directly related to the length dimension of the core section. For instance, if the length of each core section is 1/2 inch, the resolution and accuracy of the position probe assembly will be approximately .+-.1/2 inch of the actual position of the predetermined element. The output signal developed at the secondary windings of the plurality of differential transformer core sections form a phase encoded digital signal in Gray code, which is not a function of analog signal strength. This output signal format is inherently independent of line voltage and frequency variations, as well as variations in nuclear reactor operating temperatures. A Gray code pattern, instead of the more common binary code, is desirable for two reasons: 1. A conventional binary code changes more than one bit as the code progresses, thus leading to large errors when the timing of the bits is not perfect; the unique characteristic of the Gray code is that only one bit changes as the code progresses, and therefore erroneous timing of a bit only produces a single increment error, which in the case of a core section length of 1/2 inch would correspond to a 1/2 inch increment error. 2. The Gray code requires only one secondary coil per primary and core section, while the binary code would require multiple secondaries for each core section. Signal processing circuitry converts the Gray code phase encoded digital output signals from the position probe assembly into a binary coded decimal format for numerical display of the position of the monitored magnetically permeable material.
055815863	summary	BACKGROUND OF THE INVENTION This invention relates to a drive device for control rod drive mechanisms using electric motor drive in an atomic power plant. With the revolution in reactor control technology in recent years, for control rod drive mechanisms in atomic power plants, use has come to be made of control rod drive mechanisms in which the position of the control rod is controlled using an electric motor, rather than control rod drive mechanisms as conventionally employed, in which insertion and withdrawal operation was performed by water pressure. In such control rod drive mechanisms, control of the direction of drive (insertion or withdrawal) of the control rod and/or the drive speed and position of the control rod can be effected by changing the voltage, frequency and drive time of the electric motor by operating switching elements in an inverter power source constituting the power source of the electric motor that drives the control rod. However, since, in such a control rod drive mechanism, the electric motor is controlled using an inverter power source, techniques for suppressing the adverse effects of noise generated from this inverter power source have become necessary. A conventional drive unit for a control rod drive mechanism comprises, as shown by the block diagram of FIG. 1, 205 control rod drive mechanisms 1, 205 inverter power sources 2 corresponding to the drive mechanisms, and 205 inverter controllers 3, as well as a control device 4 and a man-machine device 5. The plurality of control rod drive mechanisms 1 installed at the bottom of the reactor of the atomic power plant deliver output to control unit 4 in the form of control rod position signals S1 that indicate the positions of the control rods, not shown, in the reactor. When the control rods are driven, man-machine device 5 is used to select (1) a control rod selection mode (individual or ganged), (2) a control rod drive mode (step, notch, or continuous), or (3) a control rod insertion/withdrawal mode. The information regarding which selection has been made is output to control unit 4 in the form of drive information S2 indicating the target position obtained by calculation in accordance with the selection that was made. (1) The control rod selection mode is the mode that selects which control rods within the reactor are to be driven, and may be specified as either individual mode or ganged mode. Of these, the individual mode is a mode that is employed for surveillance of the control rods. In this mode, the control rods can be driven one at a time. Also, in the case where movement of the control rods is automated, the control rods are driven as groups of control rods comprising, for example, from 2 to a maximum of 26 control rods. This is called the ganged mode. When driving is effected by this ganged mode, up to a maximum of 26 inverter power sources 2 must be driven concurrently. (2) The control rod drive mode is the mode for determining how the control rods are to be driven, and may be specified as a step mode, a notch mode or a continuous mode. Step mode is a mode in which control rods are only driven through a fixed distance. This is employed when making fine adjustments, etc to the output of the reactor. The switching elements of inverter power sources 2 are turned ON or OFF so as to move control rod drive mechanisms 1 only through the width of the step. Notch mode is a mode in which driving is effected through a distance of four times the step. The switching elements of the inverter power sources 2 are turned ON and OFF so as to move the control rod drive mechanisms 1 through four times the step width. Continuous mode is a mode in which the control rods are driven continuously up to a target position that is input through man-machine device 5. Inverter power sources 2 output voltage continuously until the control rods reach the target position, whereupon their operation is stopped. (3) Control rod insertion/withdrawal mode is the mode for specifying whether the control rods are to be inserted or withdrawn. The switching elements of inverter power sources 2 are turned ON and OFF to give output voltages of opposite phase for insertion and withdrawal. In accordance with the control rod selection mode specified in the drive information S2 that is input from man-machine device 5, control unit 4 outputs the data of the control rod drive mode and control rod insertion/withdrawal mode as an inverter control signal S3 supplied to inverter controller 3 of the control rod drive mechanism 1 that effects drive. Also, depending on the control rod current-position signal S1 from the control rod drive mechanism 1 that is being driven and in the control rod drive mode, the target position to be reached by the control rod drive mechanism 1 that is driving the control rod is calculated, and the output of inverter control signal S3 is continued until the control rod drive mechanism 1 reaches this target position. Inverter controller 3 determines the direction of rotation of the motor in accordance with the control rod insertion/withdrawal mode of inverter control signal S3 that is input from control unit 4. Also, inverter controller 3 determines the motor drive time and voltage and frequency that are output by inverter power source 2 according to the control rod drive mode and outputs to inverter drive source 2 as inverter drive signal S4 the changeover timing information of the switching elements in the output unit of inverter power source 2. Inverter power source 2 supplies the power specified by control rod drive mechanism drive signal S5, in accordance with the control rod drive mode and the control rod insertion/withdrawal mode contained in drive information S2, to control rod drive mechanism 1 under the control of inverter drive signal S4 from inverter controller 3. Control rod drive mechanism 1 is fed with power from inverter power source 2, and the control rods are driven as long as this power is supplied. However, when reactor scram occurs, the control rods in control rod drive mechanisms 1 are temporarily separated from the motors and inserted at high speed by water pressure, in response to a full-insertion drive command from the reactor emergency shut-down system. As a backup system after the control rods of control rod drive mechanisms 1 have been fully inserted by water pressure, all 205 control rod drive mechanisms 1 are driven to a fully inserted position by operation of the electric motors in response to said full-insertion drive command. Since one inverter power source 2 is provided for each control rod drive mechanism 1, when there are 205 inverter power sources 2, an enormous number of control rod drive devices, i.e., inverter power source 2, and inverter controllers 3, of control rod drive mechanisms 1 are required. Also, in drive control of the control rods, even in the case of automatic control during reactor operation using a computer, the maximum number of ganged groups which can be operated without driving a plurality of control rods at the same time is 26. Furthermore, in emergency insertion such as reactor scram, the control rods are inserted by a water pressure unit, so there is no need for the motor drive using inverter power sources 2 to be able to actuate all the control rods urgently and simultaneously. For the above two reasons, the provision of one inverter power source 2 for each control rod drive mechanism 1 means that a large number of inverter power sources 2 are required and inverter power sources 2 generate a lot of noise. A further problem is that equipment cost is increased by the fact that the devices and wiring etc. are complicated, so that a lot of maintenance is required. SUMMARY OF THE INVENTION An object of this invention is to provide a drive device for control rod drive mechanisms which overcomes the above described deficiencies of the conventional drive unit. In order to achieve the above and other objects, there is provided a drive device for control rod drive mechanisms of an atomic power plant operated by electric motor drive, the device comprising: a control rod changeover device provided for each group of control rod drive mechanisms which are divided into a plurality of groups; an inverter power source which is constituted a drive power source of said electric motor drive; an inverter controller which outputs an inverter drive signal to said inverter power source; a control device which inputs control rod position signals from each of said control rod drive mechanisms and which outputs control signals to said control rod changeover device and inverter controller; and a man-machine device which is constituted an interface with an operator, the device outputting control rod drive information to said control device.
	description	Not Applicable. Not Applicable. Not Applicable. 1. Field of the Invention This invention relates generally to reducing the loss and/or damage caused by an earthquake, in particular, to artificially triggering a catastrophic earthquake to happen at a known time without having to rely on imminent earthquake prediction so that pre-earthquake evacuation and preparation can be implemented prior to the known time. 2. Description of the Related Art Catastrophic earthquakes occur almost every year in the world, causing tremendous loss and damage. Most catastrophic earthquakes are tectonic earthquakes that are caused by sudden releasing of underground elastic strain energy accumulated over time. People have been trying to develop techniques to predict the earthquakes, so that mitigation actions can be taken prior to the earthquake. Earthquake predictions may be roughly divided into four types: long-term predictions which are made in units of several tens of years; intermediate-term predictions made in units of several years; short-term predictions aiming at predicting an earthquake from several months to several tens of days before the occurrence; and imminent predictions aiming at predicting an earthquake from several days to several hours before the occurrence. While long-term predictions and intermediate-term prediction technologies have been making progress by researchers, effective pre-earthquake mitigation actions usually still have to rely on imminent earthquake predictions or at least short-term predictions. Unfortunately, due to the complexity of the nature of the tectonic earthquakes, many catastrophic earthquakes do not have any foreshocks or other indicators before occurring, imminent and short-term earthquake predictions have extremely low successful rate and many seismologists believe it is technically impossible in the foreseeable future. Therefore, there is a need in the art to work-around the technical difficulties of imminent and/or short-term earthquake predictions to effectively enable evacuation and other pre-earthquake mitigation actions. In one aspect, the invention provides a method of artificially triggering the catastrophic earthquake that is predicted by long-term and/or intermediate-term prediction techniques to release its energy at a predetermined time, so that pre-earthquake mitigation actions can be completed prior to the predetermined triggering time. In another aspect, at least one embodiment of the invention provides a system for artificially triggering the catastrophic earthquake through controlled underground explosion. Other aspects of the invention will become clear thereafter in the detailed description of the preferred embodiments and the claims. The same reference numerals are used in different Figs. to denote similar elements. It will be appreciated that in the description herein, numerous specific details are set forth in order to provide a thorough understanding of the invention. However, it will be understood by those of ordinary skill in the art that the invention may be practiced without these specific details. In other instances, well-known methods, procedures and components have not been described in detail so as not to obscure the invention. Furthermore, this description is not to be considered as limiting the scope of the invention, but rather as merely providing a particular preferred working embodiment thereof. FIG. 1 illustrates a fictional and simplified cross section diagram of stress accumulation along the edges of two plates where the strength of the rock is exceeded at certain time, the earth's crust may break and cause an earthquake. In the figure, compressional stress between the plate 1 and plate 2 is shown by the arrows, which is just an example of the plate tectonic relation. Other types of the strain relation are also possible, such as transform and extension. Once the built-up stress exceeds the strength limit of the rocky structures, which may be coincident with a constructive force of tide and/or the gravity of the moon and sun, a sudden fracture and movement of rocks along a fault happens, suddenly releasing the stored elastic energy. Some of the energy released is in the form of seismic waves, which cause the ground to shake with potential damages. As can be easily understood, the strength limit for the rocky structures, the amount of stress built up in a given localized underground position, the weakest underground position that will have the initial fracture and the combined environmental forces such as that caused by tide, gravity of the moon and sun, thermal expansion, water body, and so on are not a precisely measurable parameters. An imminent prediction of the earthquake usually has to rely on precise modeling of these parameters. The progress of the technologies, such as remote sensing technology, sensor technology, network technology, computer and signal processing technology have provided seismologists with new capabilities that make the long-term and intermediate-term earthquake prediction more and more practical. It becomes more and more feasible to predict the most likely location and scale of a big earthquake. Realizing the fact that, when the built-up stress is close to the limit that a fracture starts, a fracture will happen only with a relatively small incremental constructional force, such as mentioned earlier, that of the tide and/or the gravity of the moon and sun. Such incremental force can also be artificially created in a controlled fashion, such as through underground or surface explosion, or other types of impact. FIG. 2 illustrates an example of method and system to artificially trigger the earthquake. Based on long-term and intermediate-term predictions, we are able to determine the localized region of at least some of the catastrophic earthquakes. Once the region is determined, we can build underground explosion facilities 10 ahead of time in the region. Such underground explosion facilities 10 may consist of a single explosion spot, or more preferably an array of explosion cells that can be controlled to explode synchronously. One of the advantages of utilizing an array of explosive cells is that, a single accident of a cell among the plurality of cells will not jeopardize the overall plan. The synchronized explosion of the array of cells is preferably to form a lens focus effect of the shockwaves towards the weakest point of the stress built-up rocky structures. The depth of the underground explosive cells need to be chosen such that the explosion itself will not cause any damage to the surface buildings and other man-made facilities. The potential pollution that may be caused by the underground explosion must also be carefully controlled. The initial quake point to be triggered should be chosen to minimize the damage, such as to be far away from dense populated areas, dams, and other critical facilities. When the choices cannot be made to meet all the requirements with the given technological limitations, trade-offs must be made to minimize the damage. Generally speaking, the earthquake may be triggered early so that the elastic strain energy can be triggered to release in a smaller scale before it is accumulated by too much to cause larger overall damage. However, the energy needed to trigger the earthquake earlier is also higher. A trade-off between the needed triggering energy (and thus the cost) and the overall damage of the triggered earthquake need to be carefully made. Triggering too late may miss the opportunity and the explosives may even enhance the damaging energy when the earthquake happens naturally before the triggering action. The at least one explosive cells 10 need to be connected via reliable communication links 30 to a controller in a control office 20. After determining the triggering time of the earthquake, evacuations and other mitigation and relief preparations need to be carried out and completed prior to the triggering time. The planned triggering time may need to be announced to make sure all people who need to know are reached. Preferably, immediately prior to the predetermined triggering time, a final warning is also issued. At the predetermined triggering time, the control office 20 initiates the explosion of the cell 10 and the shockwaves 40 to propagate to the potential hypocentre. The earthquake monitoring stations (not shown in the drawing) closely monitor and determine whether the total release energy is higher than the explosion energy and in line with the expected values, and if the answer is positive, the artificial triggering of the earthquake is successful. The explosion devices installed in the at least one cells 10 may be regular explosive devices, and may also be a nuclear device. In the event of using a nuclear device, pollution control measures must be carefully implemented. The triggering process of the earthquake may be designed to be a single shock, or a sequential series of shocks. The latter may successfully divide the total accumulated earthquake energy into a plurality of lower scale earthquakes along a distance of fault line or in a distributed area, to reduce the damage. The triggering time may be chosen by synchronizing with certain constructive natural forces such as tide and gravity of the moon and sun, to save the energy produced by the artificial cells 10. Alternatively, the triggering time may be coincident with the predicted time of an imminent prediction, and in the event that the prediction is accurate, the triggering action is cancelled; in the event that the predicted time nearly passed and no earthquake had happened, then the triggering is activated to induce the earthquake to happen before the predicted time window ends. To allow sufficient time to build the underground cells 10, we may build more cells in a larger area based on early preliminary predictions, and when time becomes closer and more accurate predictions become available, we selectively load and activate some of the cells. When the initial activation of the cell begins, it is also possible to adaptively adjust the further activation of the rest prepared cells 10 real-time based on the feedback information received from of the sensors (not shown in the drawing). The real-time adaptive control based on the feedback may be implemented in a controller in the control office 20. The triggering force may also be implemented by other forces with predictable impact time than an explosion. FIG. 3 illustrates exemplary steps to artificially trigger the earthquake at a prepared time for minimizing the loss. The process begins at step 102. First, through long-term and/or intermediate-term prediction technology, a catastrophic earthquake is determined to happen, and the region of the upcoming earthquake is predicted at step 104. Since the predicted earthquake scale would be catastrophic, an artificial triggering is worth implementing. The next step 106 is to choose one or multiple suitable sites to install the earthquake triggering devices. Many factors need to be considered in choosing the sites, for the effectiveness of triggering the earthquake and for minimizing the loss and harmful side effects, such as pollution. Once the sites are chosen, at step 108, the triggering devices are built and installed. Next, at step 110, a triggering time need to be carefully chosen. The triggering time cannot be too early, because it needs higher triggering power to cause the earthquake energy to be released, increasing the risk of unsuccessful triggering. The triggering time cannot be chosen too late either, because it increases the risk that the earthquake occurs naturally and unprepared before the planned triggering time, causing high losses. Once the triggering time is determined, at step 112, evacuation and other preparation work need to be completed before the triggering time. When the planned triggering time arrives, at the step 114, the triggering action is initiated. If the implementation is based on correct prediction and calculation, the earthquake would be successfully triggered, and the earthquake energy begins to release at the triggering time. The step 116 confirms the success of triggering through monitoring the seismic activities. Once it is concluded that the major seismic energy has been released and the aftershocks would not cause any high risks, announcements can be made at step 118 to recover the normal life step by step, and the process ends 120. Certain terms are used to refer to particular components. As one skilled in the art will appreciate, people may refer to a component by different names. It is not intended to distinguish between components that differ in name but not in function. The terms “including” and “comprising” are used in an open-ended fashion, and thus should be interpreted to mean “including, but not limited to”. The terms “example” and “exemplary” are used simply to identify instances for illustrative purposes and should not be interpreted as limiting the scope of the invention to the stated instances. Also, the term “couple” in any form is intended to mean either a direct or indirect connection through other devices and connections. It should be understood that various modifications can be made to the embodiments described and illustrated herein, without departing from the invention, the scope of which is defined in the appended claims.
050698648	description	DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT The spring-and-spacer assembly of the present invention is provided for use in connection with a fuel assembly, such as that depicted in FIG. 1. The fuel assembly 20 includes a plurality of fuel elements or rods 21, supported between an upper tie plate 22 and a lower tie plate 23. The fuel rods 21 pass through a plurality of fuel rod spacers 24a, 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 longitudinally evenly spaced along the fuel assembly. Each of the fuel rods 21 is formed of an elongated tubular cladding material containing fissile fuel and other materials, such as burnable nuclear poison, inert material, or the like. The fuel and other materials are sealed in the tube by upper and lower end plugs 26, 27. The lower end plugs 27 are registered and positioned in cavities 29 formed in the lower tie plate 23. Similarly, the upper end plugs 26 fit into cavities 31 in the upper tie plate 22. Some of the fuel rods 21 may be provided with threaded lower and upper end plug extensions 27', 28'. The threaded lower end plugs screw into tapped holes in the lower tie plate. The threaded upper end plugs extend through the upper tie plate and receive retaining nuts 32. In this manner, the upper and lower tie plates and fuel rods are formed into a unitary structure. Typically, the fuel rod assembly includes a flow 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 24a and 24b, so that the channel 33 may readily be mounted and removed. The channel 33 is fastened to a post 36 of the upper tie plate 22 by means of a bolt 37 passing through a tab 34. 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 the pressurized coolant so that it can flow upwardly among the fuel rods. One or more of the fuel rods 21 may be replaced by a moderator tube 41 which contains a neutron moderator. The moderator tube 41 may be apertured, as shown at 42 and 43, and the upper and lower end plugs may be formed with passages 44, 46 to permit flow of water moderator therethrough. An understanding of the present invention is promoted by a brief description of the spacer and spring apparatus of previous devices. A prior art spacer 24, as depicted in FIGS. 2 and 3, is made up of a plurality of substantially cylindrical ferrules 51 joined to one another, for example by welding, at the upper and lower edges of abutting portions of adjacent ferrules. Each of the ferrules 51 provides a space for a fuel rod 21 or moderator 41. Preferably, each of the ferrules 51 is formed of a short section of tubing having circular cross-section. A peripheral band 56 surrounds and supports the plurality of ferrules 51. Two stops 62a, 62b are provided in each ferrule. Preferably, the stops 62a, 62b are formed integrally with each ferrule 51 as laterally spaced pairs of arched portions of the ferrule walls near the upper and lower edges of the ferrule. Preferably, the stops 62a, 62b are laterally oriented to minimize spacer projected area and coolant-flow resistance. As best seen in FIG. 3, each pair of ferrules 51 shares a continuous loop spring 70. As depicted in FIG. 4, the continuous loop spring 70, used in previous devices, includes first, second, third, and fourth legs 72a, 72b, 72c, 72d joined by mid-positioned ridges 74a, 74b and end-positioned arches 76a, 76b. The springs 70, depicted in FIG. 3, are in their substantially unstressed or unflexed condition. The springs in the flexed or stressed condition are depicted in FIGS. 5 and 6. In the previous devices, depicted in FIGS. 5 and 6, the springs 70 occupy the region 78 of rod-to-rod spacing. In previous devices, the rod-to-rod spacing 78 was approximately 0.125 to 0.16 inches (about 3 mm to about 4 mm). As seen in FIG. 6, the previous spring 70 contacted the fuel rods 82a, 82b at the ridge portion 74a, 74b, and provided a force tending to position the fuel rods 82a, 82b against the stops 62a, 62b of each ferrule. Each ferrule 51 was provided with a C-shaped slot 84, defining a tab 86 (See FIG. 7). In assembly, the spring 70 was inserted into the slot 84 and positioned over the tab 86. An adjacent ferrule was fitted to the initial ferrule, with the C-shaped slot of the second ferrule oriented with the tab 86 pointing in the opposite direction from the tab of the first ferrule. The tabs 86 of the first and second ferrules overlapped each other. The spring was then captured between the two ferrules, and formed a loop around the tabs 86. A number of difficulties have been noted in connection with the previous spring-and-spacer assembly, particularly when such assembly is intended for use in connection with a fuel assembly having a reduced rod-to-rod spacing. The spring which is used in a spring-and-spacer assembly must provide the required amount of force, preferably about 2.5 pounds (about 1 kg), but must also have sufficient flexibility to tolerate deflection beyond that normally needed for fuel positioning without substantially permanent deformation. Deflection of the spring beyond that normally needed for positioning the fuel rod can occur, for example, during assembly, particularly if the fuel rod 82 is encased, during assembly, in a protective plastic sheath (not shown). Because a single spring 70 of the previous device was provided for each pair of ferrules 51, special arrangements had to be made for fuel assemblies having an odd number of fuel rods 82. FIGS. 9 and 10 depict a portion of a spacer and the associated springs according to the present invention. A spring 92a is formed of a metallic strip having a width 94 (FIG. 16) and a thickness 96 (FIG. 15). The spring can be formed of a number of materials having suitable strength, corrosion resistance, and resilience characteristics. In one preferred embodiment, the spring is formed of a nickel alloy, such as Inconel.about., available from Huntington Alloy Products Division, International Nickel, Inc., Huntington, W.Va. In the preferred embodiment, the width 94 is about 0.1 to 0.15 inches (about 2.5 to 3.8 mm), and the thickness 96 is about 0.01 to 0.015 inches (about 0.25 to 0.38 mm). The ribbon has first and second ends 102, 104 which are spaced apart. The strip is formed into a spring having a leg portion 106. The leg portion 106 includes a rod-contacting arched portion 108, at the center of which is a rod-contacting point 110. Although an arch-shaped contact region 108 is a preferred method of forming a contact point 110, a contact point can also be formed by means of a cone-shaped dimple formed in the ribbon, a ridge, button, or other extension formed in or attached to the ribbon, or other similar expedients. The portions 112, 114 of the leg 106 on either side of the rod-contacting portion 108 are substantially flat. First and second loops 116, 118 are formed adjacent to the first and second ends 102, 104, positioned at each end of the leg 106. Preferably, the loops 116, 118 are formed by bending the ends of the ribbon to form a hairpin turn, and welding the ends 102, 104 where they contact the ribbon. FIG. 15 depicts a slightly modified configuration of the spring in which the strip passes on the near side of the ear, and then extends over the ear and down the far side of the ear, where it is welded. Only the upper half of the symmetrical spring is shown. Each loop 116 has a dimple 122 respectively, formed therein in the shape of a trough-like indentation extending toward the interior of the loops 116. The interior surfaces of the dimples 122 form points or lines of contact 126 with ears 132 formed in the ferrule as described more fully below. Although trough-shaped dimples 122, 124 are the preferred devices for providing contact with the ears 132, 134, contact can also be provided by, a cone-shaped dimple (forming a contact point), a ridge or button formed or attached to the interior surface of the loops 116, 118, or similar expedients. As depicted in FIG. 16, in one embodiment of the invention, the width 94 of the spring is the same in all portions of the spring. In another embodiment depicted in FIG. 17, the width 142 near the end portions, such as in the loop portions 116, 118, is less than the width 144 in the leg portion 106 of the spring. It is possible to provide a narrower or tapered region, as depicted in FIG. 17, at one or both ends of the spring because, as discussed more thoroughly below, stresses on the spring in the end portion 116 of the present invention (compared to stresses in other portions) are less than stresses of corresponding portions in previous designs. Because the stresses are reduced, less massive, i.e., narrower, structures can be used. By providing a narrowed or tapered width 142, a spring which has a lessened overall mass, as compared to the constant-width spring of FIG. 16, is possible. A spring having a lessened mass provides for a lower amount of neutron absorption, which is related to the mass of the spring. By reducing neutron capture, the deleterious effect of capturing neutrons (which can be otherwise used to sustain the nuclear reaction) is reduced. Furthermore, a spring with a tapered end portion, as depicted in FIG. 17, presents a smaller surface area in the end portion than the surface area presented by a constant-width spring, such as that depicted in FIG. 16. Such a smaller-surface area spring provides a decreased amount of flow resistance or blockage which is related to the surface area of an obstruction. The spring of the present invention is used in connection with a spacer comprising a number of ferrules 136 (FIGS. 11-13). The spacer can be formed of a number of materials having a suitably low neutron absorption cross-section, preferably a zirconium alloy, such as Zircaloy-4-.about.. The spacer in one embodiment is about 5.25 inches (about 13.3 cm) square, and the ferrules are about 0.64 inches (about 16.2 mm) in outside diameter with a wall thickness of about 0.020 inches (about 0.5 mm). Referring to FIGS. 11-14 the ferrule of the preferred embodiment includes two upper stops 146, 148 and two lower stops 152, 154 extending inwardly into the ferrule 136. Preferably, the stops 146, 148, 152, 154 are formed by indenting portions of the ferrule wall to produce inwardly-arching structures. As best seen in FIG. 9, the fuel rods 82a, 82b, 82c are abutted against the stops 146, 148, 152, 154 to place the fuel rods 82a, 82b, 82c in a preferred position within the ferrule 136, such as a position coaxial with the ferrule. The force to maintain the fuel rods 82a, 82 b, 82c in contact with the stops 146, 148, 152, 154 is provided by the spring 92a, 92b, 92c. In order to provide such force, the spring 92a is mounted on ears 132, 134 of the ferrule 136 (See FIG. 12). The ears 132, 134 are defined by an E-shaped slot 158 formed in the wall of the ferrule 136. The upper and lower legs 162, 164 of the E-shaped slot 158 are connected to narrow slots 166, 168. The region between the upper leg and slot and hole 162, 166 and the lower leg, and slot 164, 168 is generally in the form of a tab 176. The middle leg 178 of the E-shaped slot 158 serves to define the ears 132, 134. As seen in FIGS. 11 and 14, the tab 176 is curved outwardly from the circumference of the ferrule 136. As seen in FIG. 20, the spring 92a is attached to the ferrule 136 by slipping the upper and lower loops 116, 118 over the upper and lower ears 132, 134. The spring is retained in its position on the ears by engagement with a rectangular slot 178, formed in an adjacent ferrule 182. As seen in FIG. 20, after the adjacent ferrule 182 is attached to the first ferrule 136 the rectangular slot 178 of the adjacent ferrule 182 will prevent movement of the spring 92a off of the ears 132, 134. Preferably, each ferrule 134 contains both an E-shaped slot and, on substantially the opposite surface, a rectangular slot 178, as depicted in FIGS. 12--14. According to the depicted preferred spring-and spacer assembly, the spring 92a contacts the ferrule 136 at two points, i.e., the upper and lower ears 132, 134 (FIG. 10). By providing upper and lower loops 116, 118, which are continuous with the spring leg 106 for connection with the ears 132, 134, a single break in any portion of the spring will not permit the spring or any portion of the spring to become disengaged from the ferrule 136. The effective length 192 of the spring 92a is the distance between the contact points 126, 124. Because the spring 92a contacts the ears 132, 134 at points defined by the dimples 122, 124, the effective length of the spring 192 is not readily affected by small deformations of the spring, accumulation of corrosion products, and the like. Thus, accordingly to the present invention, the effective length of the spring 92a is invariable. In the preferred embodiment, the effective spring length 192 is about 0.8 inches (about 20 mm). In addition to providing for a stable and definable length 192, the dimples 122, 124 act to promote the flexibility of the spring 92a, as compared to previous spring designs. As noted above, previous spring designs, considering the portion loading a single fuel rod, act effectively as a constrained beam. As best shown in FIG. 15, the spring of the present invention acts as a simply supported beam. When the spring 92a is stressed, and the leg 106 moves in a direction away from the preferred-position fuel rod, it rotates about the ear contact point 126, as shown, in exaggerated form, by the phantom lines in FIG. 15. The increased flexibility of the spring of the present invention has been confirmed by computer modeling of stress and movement using a finite element technique. The increased flexibility (i.e., deflection at a given load) of the spring of the present invention, in addition to being useful for permitting smaller rod-to-rod spacing, is also useful in connection with an assembly technique used to avoid scratching the fuel rods. It is desired to avoid scratching the fuel rods because such scratching can conceivably contribute to crack initiation or propagation, and can detract from the appearance of the fuel assembly. According to the scratch-free assembly technique, the fuel rods, before insertion into the assembly, are encased in a plastic sheath. After the encased fuel rod is positioned as desired, the sheath is removed. Typically, the sheath has a thickness of about 0.004 inches (about 0.1 mm). The additional spring deflection (during assembly) caused by the thickness of the sheath must be accommodated by the spring without permanent deformation thereof. The spring of the present invention contains sufficient flexibility to avoid permanent deformation during scratch-free assembly, even with a lessened rod-to-rod spacing available for holding the spring. The spring of the present invention can be provided in connection with other types of spacers than that depicted in FIGS. 9-17. In one alternative spacer configuration, depicted in FIGS. 18 and 19, helically-twisted "swirl vanes" 237a, 237b, 237c, 237d are positioned around the periphery of a fuel rod 282, such as at four equally spaced positions on the perimeter of the fuel rod 282. Arms 239 connect the swirl vanes to each other to define a rectangular matrix for holding the fuel rods 282. A spring 293 includes two end loops 216, 218 attached to two of the arms 239. The spring 293 includes an arched rod contact portion 208 similar to the rod contact portion depicted above in connection with FIG. 10. Edges of the swirl vanes 237a, 237b, 237c, 237d are provided with protrusions 246, which act as stops for positioning the fuel rod 282. The upper and lower loops 216, 218 of the spring 293 contain dimples (not shown) similar to those depicted and described above in connection with the embodiment depicted in FIGS. 9 and 10, for contacting the arms 293b, 293a. In light of the above description, a number of advantages of the present invention are apparent. The spring can be provided in a smaller space, such as that available with the rod-to-rod spacing of only about 0.1 inches (about 2.5 mm), and yet be provided with the required force for fuel-rod loading of about one pound, preferably about 2.5 pounds (about 1 kg). The present spring is more flexible, and tends to rotate under stress in the manner of a simply supported beam, as opposed to having an endpoint fixed in the manner of a constrained beam. The effective length of the spring is a predeterminable, known, and relatively stable quantity. Since less strain is produced at the end portion of the spring, the end portion can be tapered or narrowed to provide for a lessened neutron absorption and a lessened flow obstruction. Certain configurations provide increased rod-to-spring spacing to provide for better coolant flow near the rod. By avoiding the need for using a single spring to load two fuel rods, there is no need to provide special configurations for spacers with an odd number of fuel rods. The increased flexibility of the spring permits the use of scratch-free assembly, even in configurations having small rod-to-rod spacing. The spring-and-spacer assembly provides for ease of construction, and the spring of the present invention is adaptable to a number of different spacer types. The spring of the present invention will not break free from its connection to the spacer as a result of a single, simple break in the spring. A number of variations and modifications of the present invention will be apparent to those skilled in the art. The spring and/or spacer can be made of materials other than those discussed herein. The general spring and spring-and-spacer assembly configuration can be used in connection with spacers having more or fewer fuel rod positions than those depicted herein, and can be used with configurations having components other than fuel rods passing therethrough, including tie rods, water (moderator) rods, and the like. The spring of the present invention can be attached to a spacer by means other than the ferrule slot depicted, such as by attachment using slots of other shapes. The spring can be retained on the ferrule by means other than an engaging slot in an adjacent ferrule, such as by bending the tab ears, interlocking resilient tabs and slots, and the like. Various aspects of the disclosed design can be used independently of other aspects, for example, a spring can be provided with an end loop but without a dimple. Although the description of the present invention has included a description of a preferred embodiment and various modifications thereof, other modifications and variations will be apparent to those skilled in the art, the present invention being described in the following claims.
047114363	abstract	A grid assembly fixture for use in assembling grid straps to form a grid comprises a generally flat plate having first and second sets of parallel grooves, with the grooves of the sets at right angles. A retention strap comprises first and second pairs of bars, the bars of each pair hinged together adjacent their ends, and releasable joining elements for joining the pair of bars.. A method of assembling the grid straps includes positioning a first set of straps in a grid assembly fixture, positioning a second set of straps at right angles to the first set to form a grid, placing outer straps on the grid, and placing a retention strap on the outer straps.
	abstract	A particle beam irradiation system comprising a first deflector having the maximum deflection amount which enables to move a particle beam in one direction to the maximum width of a target and a second deflector having the maximum deflection amount is less than the maximum deflection amount of the first deflector performs a control in which the particle beam is moved by increasing at least a deflection amount of the second deflector when the particle beam is moved, and performs a deflection substitution control in which a deflection of the second deflector is substituted to a deflection of the first deflector by decreasing the deflection amount of the second deflector and changing a deflection amount of the first deflector so as to make a position of the particle beam in the target dwell when the particle beam dwells.
	description	This application is a National Phase Application of PCT International Patent Application No. PCT/FR2008/051834, International Filing Date Oct. 9, 2008, which claims priority from French Patent Application, 0758301 filed Oct. 12, 2007, both of which are incorporated by reference herein. The invention concerns a method for establishing so-called “mixed IN-CORE mappings”. The present invention also refers to an application of said method to the calibration of fixed-type instrumentation. IN-CORE mappings are mappings illustrating a power distribution inside nuclear reactors, established by means of sensors that are placed, in either a fixed or mobile manner, either temporarily or permanently, inside the reactor core. Its essential purpose is to compensate a loss of density of a reference instrumentation, called “RIC instrumentation” (or “RIC system”), when a significant number of locations, initially used by the sensors of the RIC system, are occupied by fixed collectron-type rods. An obvious physical interest lies in the increase of the measurement density, and thus in the level of confidence associated with the operating results deduced from the processing of these measurements, The present document shall refer to a series of abbreviations or expressions, notably within the different equations and relations, a glossary of which being stated below: C/M: Calculation/Measurement gap μUN: uncertainty associated with the calculation of the distribution of local pencil power inside an assembly RU1N: uncertainty associated with the transposition of the C/M gaps on an “activity” type parameter over to a “power” type parameter RU2N: uncertainty associated with the spatial propagation of the C/M gaps MUN: uncertainty associated with the measurement system (detector and acquisition) EUN: standard overall uncertainty on the power reconstruction process C/PM: Calculation/Pseudo Measurement gap RU2pN: generalised extension uncertainty of the C/PM gaps EUpN: overall uncertainty, according to the RUN2p methodology, on the power reconstruction process SchX: any X-type instrumentation diagram REF: reference instrumentation diagram S(t): detector sensitivity upon completion of an irradiation t-time S(0): initial sensitivity of a new detector Q(t): integration of the current delivered by a detector upon completion of an irradiation t-time Q∞: initial load available for a new detector a: exponent of the experimental law of wear A1 and A2: distributions of activities associated with type 1 and type 2 detectors, respectively N1 and N2: number of acquisitions for type 1 and type 2 detectors, respectively A1MES and A2MES: distributions of activities measured by type 1 and 2 detectors, respectively A1CAL and A2CAL: distributions of equivalent activities calculated for type 1 and type 2 detectors, respectively A1←2CONV: conversion of an activity witnessed by a type 2 detector into an activity that would be witnessed by a type 1 detector brut: normally refers to a distribution whose elements have not undergone any normalization process relatif: refers to a distribution whose elements are linked by a normalization of series g: coefficient reporting the normalization differences between two distributions complet: characterises a distribution for which all the elements are used σ(TUN) and TUN: standard gap and uncertainty associated with the process for constructing a mixed mapping. σ(AUN) and AUN: standard gap and uncertainty associated with the calculation of the A activities r: linear correlation coefficient conversion: refers to the action for transforming an acquisition of a given type detector into an acquisition that could have been obtained at the same time and in the same place by a detector of another type. ACOLBRUT: gross activity, for XYZ 3D position, deduced from the acquisitions of a type 2 detector (here, COL). APRICBRUT: gross activity, for XYZ 3D position, initially of type 2 and converted into type 1 activities (here, PRIC). ARICEST: estimated activity, for XYZ 3D position, of type 1 (here, RIC). FCOR: 3D calibration or correction factor. ACOLCOR: corrected activity, for XYZ 3D position, of type 2 (here, COL). APRICCOR: corrected activity, for XYZ position, of type 1 after conversion (here, PRIC). CUN: uncertainty associated with the calibration process of a detector 2 based on the simultaneous acquisitions of a detector 1. The field of the invention is, generally-speaking, that of nuclear reactors. Nuclear reactors, such as pressurized water-cooled nuclear reactors, comprise a core constituted of fuel assemblies, each assembly being comprised of a plurality of fuel pencils, notably of uranium slightly enriched with isotope 235; the assemblies are placed vertically in juxtaposition with their longitudinal axes, i.e. by following the height of the core. As a general rule hereinafter, the longitudinal axes are thus identified as z-elevation, x-abscissae and y-coordinates, enabling determination of a nuclear reactor point within a horizontal plane. Hence, a nuclear reactor core can be considered as cut into sections, or axial grid cells, of a certain thickness, identified by the z-elevation; a nuclear reactor point is further identified by its azimuthal position, based on an angle defined in a horizontal plane, in relation to the z-axis of the orthogonal three-dimensional (x,y,z) markers, and by its radial position, defined by a distance, within a horizontal plane, between the point considered and the axis of the markers. The power released by the assemblies, power directly correlated with the neutron flux generated by the fuel present in said assemblies, is not uniformly distributed inside the volume of the reactor. There are areas where power is higher than in other areas, typically in the centre of the reactor when compared with the periphery. Hot spots are then referred to; in the region of these spots, in fact, the power supplied almost attains the design limits of the nuclear reactor core. Consequently, the power distribution inside a nuclear reactor core lacks consistency; the embodiment of a complete power mapping inside the core, referred to as 3D power distribution—a fundamental operation for obvious safety reasons, is thus a complex operation. Hence, the operating and securing of nuclear reactors requires determination of the energy supplied by the uranium 235 nuclei fissions, i.e. the nuclear power, in each spot of the nuclear reactor. For this purpose, measurements are performed in order to evaluate the power in the various spots of the nuclear core. In all cases, evaluation of such power involves measuring the radiation emitted by the reactor core, and more particularly the neutron flux. Measurement of a neutron flux is always achieved by way of a neutron/matter interaction, which in turn creates particles liable to produce a measurable electric current. After each neutron absorption, the atoms of the sensitive matter constituting the sensor will be transformed; the sensitive matter as such will thus gradually disappear. Such disappearance is performed at a speed that depends upon the intensity of the neutron flux and upon the probability of a reactive occurrence, itself directly linked to the cross-section absorption. The higher this probability and the stronger the current supplied, the faster on other hand the sensitive matter disappears, thereby requiring replacement of the sensor at very short intervals. Depletion of the sensitive matter thus becomes a crucial problem for a neutron sensor permanently located inside the core. In order to resolve this sensitive depletion problem for the sensors, many nuclear reactor manufacturers have chosen not to leave the sensors in a static measurement position inside the core and thus to dispatch them throughout the reactor for the sole purpose of taking intermittent readings. The term “mobile internal instrumentation” thus refers to traditionally-used sensors, which, hereinafter, shall be referred to as “RIC system” (Core Instrumentation Reactor). Other systems, for example the aeroball system, may also be considered as a mobile internal reference instrumentation system The function of the RIC system is to precisely measure the flux distribution inside the reactor core, with relatively minor constraints in terms of response time. In practice, therefore, the RIC system coexists with a control system called “NPR system” (nuclear protection reactor), placed outside the nuclear reactor core and responsible for measuring a few parameters of the power distribution (such as axial and azimuthal imbalance) and the power level with an excellent response time, though not quite as precise as the RIC system measurements. The NPR system is periodically calibrated, since the proportionality between the external measurement and the reactor's real power level depends upon the radial component of the power distribution, which itself varies as the fuel depletes. The data provided by the RIC system may be used to perform such calibration. In a more general manner, the RIC system is used in two separate situations: First of all, during the start-up test periods, after each assembly reloading, or during special test periods, the RIC system is used for: verifying that the power distribution at the start of a cycle complies with the design calculations and, in particular, that the value of the hot spots is in accordance with the design assumptions; calibrating the NPR system detectors; detecting a possible loading error; supplying data on the distribution of flux that are involved in the qualification of IT codes and of the methods used in the design calculations of the reactor core. Then, during a cycle and under normal operation, the RIC system is notably used for: verifying that the power distribution and, in particular, the hot spot factors, evolve according to time, such as provided in the design calculations; verifying and/or calibrating the NPR system detectors. In terms of precision, a compromise has been chosen in the past between the desire of wanting to measure the power in a vast number of assemblies, and a material reality residing in the fact that it is necessary to drill, for each instrumented position, a hole in the base of the nuclear reactor tank. Such compromise results in the penalising fact that a limited number of instrumented assemblies have been chosen—an economically and technologically advantageous solution, though consequently limiting the precision of the flux distribution measurement and necessitating the existence of certain leeway, provided by a subsequently detailed uncertainty calculation, having the purpose of compensating imperfect experimental knowledge of the 3D power distribution, notably in the region of the hot spots. In practice, six mobile neutron detectors are used. The mobile detectors are of the fission chamber type. This type of neutron sensor consists of a standard ionisation chamber and uses uranium as neutron-sensitive matter. The current supplied by the mobile detectors is proportional to the fission reaction rate in the detector and not directly to the power: activity rather than power is thus more often readily referred to; a phase for transposing the activity measurements into a power determination is subsequently introduced during analysis of the performed measurements. Such transposition gives rise to a particular uncertainty component, noted as RU1N. The mobile detectors are dispatched, via a switching device, into impervious tubes, called glove fingers, placed in an instrumentation tube of 60 fuel assemblies selected for the purpose. The selected fuel assemblies are called instrumented assemblies. Hence, each detector is designed to explore ten assemblies. In order to ensure the transfer of detectors from one assembly to another, mechanisms trigger off group selectors. It can be stated at this point that the acquisition process comprises one or several extra so-called inter-calibration passes. Indeed, the quantity of sensitive matter prone to interaction with the neutrons lessens as irradiation of the detector, or, to be more precise, of the particle fluence received by the latter, extends over time. The sensitivity, i.e. the ratio between the current emitted and the flux witnessed by the detector, will evolve over time: correction is thus necessary in the analysis stage in order to take account of such variation. Each mobile probe will evolve independently one from the other, since it receives its own particular fluence depending on the power of the assemblies that it explores. The function of the intercalibration passes is thus to allow the measurement of the relative sensitivities. Determination of the sensitivities should be performed before each complete flux map and is compulsory. Hence, calibration of the detectors is an operation which consists of acting on the electrical gain of the measurement chain, in order to compensate the reduction in current supplied by the sensor due to the depletion, and to maintain constant the indicated value. This operation also enables to correct the differences between detectors possibly occurring due to the fact that each of them has its own electronic acquisition system. In practice, it is performed in the following manner: All group selectors are directed towards a so-called “standby position”, which enables each of the probes to go and explore the assemblies normally measured by the probe located just above (except for probe 6 which, by circular permutation, goes and explores the assemblies normally allocated to probe 1). It is thus possible to compare the measurements obtained during the intercalibration passes for the purpose of determining the relative sensitivities of the probes, and to take account thereof during analysis of the measurements. The results of the measurement analysis by the mobile internal instrumentation system during examination of the 60 assemblies selected for the purpose, i.e. a partial distribution of the reaction rate in three dimensions across the core determined by the performed measurements, are referred to as the flux map. Hence, although measuring the flux distribution in a significant number of fuel assemblies—approximately 30% of the assemblies are instrumented—the RIC system does not radially cover the whole core. If the hot spot factor is located in a non-instrumented assembly, it fails to be measured. It is thus necessary to supplement the information supplied by the mobile detectors. Additional information is provided by theoretical calculation. The establishment of a 3D power distribution for a nuclear reactor core, detailed hereunder, thus always requires a combination of experimental data and calculated data. Instrumentation systems, other than the RIC, may equip industrial reactors. For example, the Aeroball system may be quoted here, which is an instrumentation system that brings into play mobile parts constituted of steel ball batches containing 1.5% of a sensitive isotope, such as Vanadium, and which circulate, driven by compressed nitrogen, inside pipes, and which penetrate into the tank through the cover. The neutron flux measurement depends on activation of the balls when the latter are placed under a neutron flux; the evaluation of their activity is performed by means of fixed detectors placed on racks located outside the tank, but inside the reactor construction. The collectron-type system may also be quoted, an electron-collecting system, which complies with the following physical principles: placed in a neutron flux, a body is able to emit electrons. The originality of a collectron lies in the fact that, under extremely reduced dimensions, the current supplied is quite high and that the electrons emitted are collected and measured in a continuous process without external polarisation tension. The collectrons, at the heart of the present invention, will be subsequently detailed hereunder. As a general rule, the data resulting from the power distribution calculation, a theoretical calculation, correspond to a power distribution calculated on the basis of a model that reproduces the operating conditions observed during embodiment of the flux map. This calculation is made in R&D bureaux during the planning stages. It observes the following principles: The signal resulting from the measurement by the fission detectors is proportional to a fission rate in the sensitive part of the detector, i.e. to the result produced between the fission and the flux cross-section. It is thus necessary to calculate the fission cross-section in order to be able to access the detector's activation rate. The theoretical models used explicitly represent the glove finger and the instrumentation tube in order to best approach the exact conditions of the measurement. The fission cross-section is calculated by taking account of the local conditions around the instrumentation tube and by explicitly representing the glove finger and the instrumentation tube for the calculation of the flux. This calculation is made for each instrumented assembly by a cell code, for example the code known by the person skilled in the art under the name of APOLLO 2F. The flux distribution is then calculated by a diffusion code, for example the code known by the person skilled in the art under the name of “SMART three-dimensional nodal code”. The data calculated are then as follows: 3D distribution of the mean powers per assembly. This MP CAL (x, y, z) power distribution is initiated in the transposition phase; all maximum pencil powers integrated over the core's active height. For each assembly, only a single pencil is retained, the one that carries the highest integrated power. This group, noted as P CAL DH (x, y), is used in a so-called superposition phase, which enables to calculate the enthalpy elevation factor of the core, noted as EEF; all local maximal powers. For each plane located at z-point, and for each assembly, only a single pencil is retained, that which carries the maximum local power. This group, noted as P CAL (x, y, z), is used in the superposition phase when calculating the hot spot factors of the FQ, FXY (z) cores. The reconstruction process of the measured power distribution, on its part, essentially involves three terms. The first term is the fission reaction rate in the detector, still called “activity”. The second term involves the ratio between the mean power of an instrumented assembly and the activity witnessed by a detector circulating inside the glove finger of said assembly. As already stated, it is not the power but the activity that is measured; it is thus necessary to adopt a method that enables to pass from activity to power, a method whose general principles are provided hereunder: the neutron absorption reaction by the detector's sensitive matter is produced inside an energy band characteristic of the latter. Gaining knowledge of the quantity of neutrons belonging to said energy band in relation to the total number of neutrons is a problem of neutron spectrum. The power/activity ratio is a parameter resulting from the 3D core calculations performed for all assemblies. These calculations take account, not only of the local spectrum effects through the intermediary of the neutron counter-reaction system, but also of the flux distribution. These ratios are updated according to the fuel depletion for the purpose of taking account of the evolution of the isotopic concentrations inside the assembly. In this connection, an assumption is made, consisting of recording that the ratios between the calculated values and the values reconstructed on the basis of experimental acquisitions are equal for both variables, activity and power alike. The third term is called “fine structure”; it enables to pass from the mean power of an assembly to the power of any pencil of said assembly. In order to do so, it is assumed that, for a given assembly, the ratio between the power of a pencil and the mean power of the assembly to which belongs such pencil is independent of the origin of said power, whether reconstructed or calculated. Moreover, a correction will be applied in accordance with the calculation/measurement gaps observed around the assembly. This correction leads to producing a plane-type two-dimensional linear interpolation. The interpolation is performed for each assembly and at each z-point. Moreover, in order to calculate the reconstructed power at all the non-instrumented points of the reactor, a method has been proposed enabling to estimate the calculation/measurement gaps at all other points of the core than those actually having been measured. This is the object of the error propagation method described in the following paragraphs. The error propagation process, which is explained hereunder, begins by an operation consisting, first of all, of calculating the gaps between the values actually measured and the values calculated for each assembly instrumented by the instrumentation system. Taking account of the existence of the theoretical calculation and of the previously described measurement process, for each of the instrumented assemblies, not only is known the value of the activity measured by the detectors, but also the corresponding value calculated under conditions very similar to the experimental conditions, and this for each of the axial grid cells. The error propagation process is, in general terms, conducted as follows: its aim is to determine, for each plane of the z-point, a degree-3 selected Sz surface in (x, y) for the complete maps, that is able to represent distribution of the gaps between the calculated activities and the measured activities throughout the core. It will be noted that the choice of said degree depends on the density of the available instrumentation. This method is described by the expression, “GSF error propagation method (Generalized Surfaces)”. As already stated, it is possible to calculate the gap between the measured activity and the theoretical activity at each instrumented position. It is then assumed that the distribution (x, y) of the gaps at the z-point, between the theoretical activity and the measured activity for all the assemblies, can be approached by a (x, y) Sz surface, being expressed analytically by a k-degree bi-dimensional polynomial, fixed by choice at the 3 value for the complete maps. The coefficients of the polynomial characterising this response surface are determined by minimising an F error function with several variables, each one being one of the polynomial coefficients. The minimisation process is a standard least squares' process performed on each axial measurement and reducing to a minimum the difference between the gaps previously obtained and the gaps calculated using the polynomial on all the instrumented assemblies. In practice, the extension process thus employs a standard gap minimising method, for the RIC system, on the 60 instrumented positions and, for each axial measurement, between the initial calculation/measurement (C/M) gap and the value given by the response surface. Hence, an analytical function in (x, y, z) is available, which enables calculation of the calculation/measurement gaps in all the reactor core positions. These gaps are then used to correct the theoretical values at all points. After normalization throughout the core, a reconstructed power distribution is obtained across the entire volume of the reactor. In the end, everything happens as though the calculation was being forced to approach the 60 measurement points as well as possible, the reconstructed power distribution being none other than the power distribution resulting from such forcing. By consequence, the error propagation process is associated with a particular uncertainty component, noted as RU2N, used in the calculation of an overall uncertainty, which in turn is used in a total margin report to be considered across the whole of the nuclear reactor considered. As a general rule, total uncertainty EUN is defined by the following relation, corresponding to a standard quadratic reassembly:EUN=√{square root over ((μUN)2+(RU1N)2+(RU2N)2+(MUN)2)}{square root over ((μUN)2+(RU1N)2+(RU2N)2+(MUN)2)}{square root over ((μUN)2+(RU1N)2+(RU2N)2+(MUN)2)}{square root over ((μUN)2+(RU1N)2+(RU2N)2+(MUN)2)} (Equation 1) The various components used in equation 1 are as follows: the distribution of 3D local pencil power in each assembly may only be deduced from the theoretical example simulating the experimental conditions. The μUN uncertainty calculation on this fine structure is thus the first component; the response of the detectors not being, as previously stated, of the power type, but of the activity or reaction rate type, an assumption should be made in that the Calculation/Measurement gaps of the activity type can be transposed to the power parameter. The RU1N uncertainty component is associated with this transposition assumption; the Calculation/Measurement gaps observed in the partial geometrical field and covered by the detectors are propagated throughout the core: the RU2N uncertainty component, or so-called “error propagation uncertainty component”, is associated with the corresponding algorithm; the last component characterises the detector, or the combination of detectors, whether from the signal's physical aspect or from that of the whole of the acquisition process. These different aspects are then covered by the MUN uncertainty component. This uncertainty component is referred to as “detector-intrinsic uncertainty component”. An outline of the calculation method for the error propagation uncertainty component, such as used in the prior art, is illustrated when referring to FIG. 1. This figure illustrates the fact that, for such a calculation, a real state 100 is used from the outset, which, by definition, presents an unknown power distribution, yet to be determined. As previously explained, a series of measurements 101 are taken—sixty in the case of the RIC system—across the whole of the reactor core. In parallel, as also already explained, a theoretical power distribution model 102 is available, such as prepared in R&D bureaux, which provides a complete mapping of the power distributions inside the reactor core. A step 103 is then undertaken, during which are calculated the gaps, or differences, noted as C/M, between the actually measured values and the values expected by the theoretical calculation, and this for all points of the reactor for which a measurement is available. Based on the gaps obtained, gaps, noted as (C/M), are then determined in a step 104, according to the aforementioned error propagation process, for all the points of the nuclear reactor. A generalised or extended gap is then obtained, deriving from the error propagation process, a gap to be applied on each calculated activity value, in order to obtain an estimated activity value for each point of the nuclear reactor. The extension uncertainty component (RU2N) is directly calculated, on its part, within a step 105, using the constituted residue, for each point having undergone an experimental measurement, by the difference between the extended gap (C/M) and the initial C/M gap corresponding to this point, for example, by taking a mean quadratic of said residue. Finally, in a step 106, following the aforementioned activity/power transposition step, a Pest estimated power is determined at every point of the nuclear reactor core, value Pest being specific to each point of the reactor core. The solution for determining the error propagation uncertainty component (RU2N) just explained is applicable to any nuclear reactor core for which measurements can effectively be performed, notably via the RIC system. But such a solution is not applicable to nuclear reactor cores which are about to be installed, for which no flux distribution measurement has been performed yet, nor for existing nuclear reactor cores for which a new instrumentation system is planned to be installed. Nevertheless, such changes are now in the process of becoming available. Indeed, the IT progress over recent years has enabled the generalization of 3D models for core calculations, not only in R&D bureaux, but also online, such models then being up-dated with the operating parameters of the relevant section in real time. Technological evolutions linked to sensors have also enabled the constant availability of signals supplied by the detectors placed in fixed positions inside the core. New instrumentation systems, having the purpose of monitoring operating margins online, can thus be defined. Nevertheless, the corresponding uncertainties, associated with such new systems, must of course undergo evaluation prior to industrial installation, i.e. in the absence of any operating feedback on said systems. In this context, determination of the RU2N error propagation uncertainty component thus becomes of interest for nuclear reactors and for which a new instrumentation system is liable to be used. Indeed, in such a case, a major problem comes to light for the determination of the RU2N uncertainty component: due to the measurement system for installation being relatively new, no operating measurements are available for determining said uncertainty component. Hence, a method is now proposed enabling to obtain an error propagation uncertainty component for any nuclear reactor, even those waiting to be equipped with a measurement instrumentation system and for which no operating feedback with regard to the relevant system is available. For this purpose, data originating from feedback acquired by way of a reference instrumentation system, for example the RIC system, is proposed. Such available feedback is then used for the purpose of applying disruptions to a theoretical power distribution model, the amplitude and the apportionment within space are such that the gaps observed between the disrupted theoretical model and the theoretical model directly deriving from the calculation are representative of those observed in reality. Hence, the problem posed by such lack of operating feedback with regard to a new measurement system can be overcome by using considerable feedback already acquired by way of a reference instrumentation. Such feedback essentially resulting in a 3D Calculation/Measurement gap database, theoretical disruption models are thus proposed for application, in the invention, whose amplitude and apportionment will be such that the 3D gaps, the noted gaps, such as subsequently explained, and the Calculation/Pseudo-Measurement, in relation to the initial models, are representative of those actually present in the nuclear reactor core on which is implemented the method according to the invention. Hence, for example, for nuclear reactor cores waiting to be equipped with measurement systems of the collectron type, for which feedback, having the characteristics required for the planned application, may be considered as inadequate, a disrupted theoretical model will be established using the measurements performed by means of the RIC systems, which have the advantage of offering significant feedback, enabling to precisely define the disruptions to be applied to a purely theoretical model. FIG. 2 illustrates an outline example for implementing the method, such as used for calculating the error propagation uncertainty component. In order to mark the difference, when determining this uncertainty component, between the method of FIG. 1 and this new method, the latter, when deriving from the new method, is noted as RU2pN. In this figure, it is illustrated that, in the new method, a so-called disrupted state 200 is used from the outset, which corresponds to a theoretical power distribution model 201, to which is applied, at each point of the nuclear reactor core, at least one physical disruption parameter. In a particular embodiment of the method, such disruption is applied to all the points of the nuclear reactor core. For example, the physical disruption to be applied corresponds to one or several physical parameters from among the following: misalignment of at least one control cluster in relation to the other control clusters of the considered nuclear reactor core; lack of precision on the position of the control clusters; these first two physical parameters are linked to the fact that the control clusters, which are traditionally introduced through the top of the reactor core, and have the task of controlling the power of the reactor core, or even to completely shut down the latter in the event of a serious incident, are activated by the complex mechanical systems of said control clusters, the precision of their movements and even their relative movements; lack of precision on the input temperature of the moderator; inhomogeneity of the boron concentration; inhomogeneity of fuel assembly irradiation; lack of precision on the nominal power of the reactor core; unbalance, whether azimuthal or radial, in the apportionment of nuclear power between quadrants of the reactor core. Appropriately, the values of the disruptions applied originate from a database deriving from experimental data obtained on nuclear reactor cores representing similarities with the reactor core on which the new method is implemented. The similarities presented essentially concern the spatial organisation of the fuel assemblies inside the reactor core with, for example, similarities in the observed apportionment symmetries. Conversely, it is not indispensable for the nuclear reactor core, on which the new method is implemented, to have the same type of measurement instrumentation. It is thus possible to use experimental results collected by means of a RIC system in order to determine the disruptions to be applied to the points of a nuclear reactor core which will be equipped with a different type of measurement instrumentation system, for example the aeroball or collectron type. In the illustrated new method, in a step 202, a series of activity values or reaction levels, referred to as pseudo-measurements, are selected from among the values defining the disrupted state of the nuclear reactor core; then, in a step 203, an initial gap is determined, noted as (C/PM), between the theoretical reaction rate and the corresponding pseudo-measurement, for each point of the nuclear reactor associated with a selected pseudo-measurement. In a step 204, using determined initial gaps, an error propagation process is then performed throughout the reactor core in order to associate an extended correction value, noted as (C/PM), with each point of the nuclear reactor core. In a step 205, estimated power is then determined for each point of the nuclear reactor, the extended correction value being used as a parameter in said estimated power determination. According to the new method, it is then possible, in a step 206, to calculate a plurality of residue by performing the difference, for at least a plurality of nuclear reactor core points, between the estimated power and the disrupted representation of said power for each point considered; the RU2pN error propagation uncertainty component then being established using the evaluated residue, for example, by performing their mean quadratic. Appropriately, the residue is calculated for all the points of the nuclear reactor. Hence, to resume, the new method for determining an uncertainty component, called “error propagation uncertainty component”, involved in the calculation of an overall uncertainty associated with the power distribution of a nuclear reactor core, is characterised by the different steps consisting of: establishing a three-dimensional mapping of a theoretical power distribution of the considered nuclear reactor core; appropriately, three-dimensional theoretical power distribution mappings are available for various configurations of the nuclear reactor core. establishing a disrupted representation of the nuclear reactor core, the disrupted representation consisting of applying at least one physical disturbance parameter to the theoretical power distribution for at least a plurality of points of the nuclear reactor core; selecting a series of activity values or reaction levels, referred to as pseudo-measurements, within the disrupted representation of the nuclear reactor core; determining, for each point of the nuclear reactor associated with a psuedo-measurement, an initial gap between a theoretical activity, deriving from the theoretical three-dimensional mapping of the nuclear reactor core, and the pseudo-measurement, deduced from the disrupted model, associated with the considered point in question; performing, using determined initial gaps, an error propagation method operation on the whole of the reactor core in order to associate an extended correction value with each point of the nuclear reactor core; determining, for each point of the nuclear reactor, an estimated power, the extended correction value being used as a parameter in said estimated power determination; calculating a plurality of residue by performing the difference, for this same plurality of points of the nuclear reactor core, between the estimated power and the disrupted representation of said power for each point considered; determining the error propagation uncertainty component using the residue just evaluated. The expression “point of the nuclear reactor core” refers to a nuclear reactor volume to which a power value, or a physical parameter value in relation to the power, should be assigned, in the context of preparing a 3D power distribution. Each point of the nuclear reactor core is thus associated with a single such value. Appropriately, the applied physical disruption parameter adopts a value deriving from measurements previously performed for comparable design nuclear reactor cores; the expression “comparable design nuclear reactor core” refers to nuclear reactor cores whose architecture, notably in terms of the general fuel assembly disposition, presents significant elements of resemblance with that of the nuclear reactor core on which is applied the method according to the invention. Hence, the method can be applied indifferently to 2-Loop (121 assemblies), 3-Loop (157 assemblies), 4-Loop (193 assemblies), N4 4-Loop (205 assemblies) and EPR (241 assemblies) cores. The ratio between the number of instrumented assemblies and the total number of assemblies for the reactor cores, other than those of the EPRs, is approximately 30% ( 30/121=0.25, 50/157=0.32, 58/193=0.30 and 60/205=0.29). In the case of EPRs, this ratio is 40/241=0.17. The method described is notably used, with identical instrumentation, for quantifying the impact of the significant reduction of this ratio on the extension factor. Such quantification has thus been performed in order to pass from 58 instrumented channels to 42 (in the context of a complementary RIC scheme, deriving from the introduction of 16 collectron rods into guide tubes, which were normally monitored by the mobile probes: 42/193=0.22 and 42/58=0.72), and from 58 channels to 16 (in the context of the aforementioned collectron scheme). Hence, equation 1, which defines the final reassembly of the EUN reconstruction uncertainty deriving from a process being applied to the triplet (actual configuration of the core, simulated theoretical configuration, C/M gaps), is then replaced by equation 2, defining the same reassembly using a new triplet (disturbed theoretical configuration, initial theoretical configuration, C/PM gaps). Equation 1 then becomes:EUpN=√{square root over ((μUN)2+(RU1N)2+(RU2pN)2+(MUN)2)}{square root over ((μUN)2+(RU1N)2+(RU2pN)2+(MUN)2)}{square root over ((μUN)2+(RU1N)2+(RU2pN)2+(MUN)2)}{square root over ((μUN)2+(RU1N)2+(RU2pN)2+(MUN)2)} Equation 2 The prime meaning of p index of this relation is: to essentially make a clear distinction in terms of the triplets, which are upstream of the final reassembly. The term (EUpN) of the equation 2 has the same meaning as the uncertainty (EUN) of the relation 1. It is thus composed of the same terms. The two factors assigned upon initial order via a change in the instrumentation system are of course the component (MUN), characterising the detector used, and the component (RU2N), covering the passage of the experimental data on a partial field towards 3D maximum local power at every point of the core. The component (RU2N) will always be concerned by a change in the instrumentation system. Its standard evaluation resides on a comparison between the extended (C/M) gap, via the retained error propagation algorithm, at a point monitored by the available instrumentation, and the initial C/M deviation, at an effectively instrumented point. Such comparison thus involves the existence of an experimental reference, said reference being partial in all cases. In order to compensate this partial character, the new method enables the making of such comparison on a complete whole. The RU2pN component is now evaluated by comparing the 3D local power distributions, reconstructed at every point of the core, and the equivalent distributions of reference, determined in the context of the new method. Additionally, it can be stated that, in order for the distributions of the C/PM gaps to be representative of the C/M gaps actually observed during the monitoring of the reactors under operation, the types and the amplitude of the disruptions applied to the generic models must have been correctly defined. Such definition passes through the construction of a real reference base, covering the maximum number of configurations, under the dual aspect of the assembly types loaded inside the operating reactors and the mode of managing the time spent by said assemblies inside the reactor. The definition of the series of Pseudo-Measurements is one of the aims assigned to the reference models. It is thus essential that such series are as close as possible to those actually observed on site for each of the analysed instrumentation systems. At the same time, it is thus necessary to take account of all the characteristics of these systems and of the impact of such characteristics in relation to the response of the RIC reference system. These impacts are linked: a) to the change in the radial density of the instrumented channels (58→42 channels for the complementary RIC schemes of a standard 4-Loop core and 58→16 for the collectron schemes of such same cores); b) to the type of detector (Uranium 235 in the case of the RIC, and Rhodium 103 in the case of the collectrons); c) to the change of the measurement points in the axial apportionment in the case of collectron-type detectors (65 continuous grid cells→8 discontinuous grid cells); hence the necessity for a grid-cell conversion; d) to the characteristics of MUN experimental uncertainty. In the case of RIC-type signals, such uncertainty only comprises a 3D local part, regardless of time. In the case of collectrons, it is important to take account of the 3D and 2D components (per rod) of such uncertainty and of its variability during the wear process. In order, on the one hand, to minimise the number of disrupted configurations to be constructed, and to further consolidate the link with the real experimental database on the other, a differential approach has been chosen for the first practical applications of implementing the new method in relation to the reference instrumentation. The internal instrumentation of MFC-type (Mobile Fission Chamber), used in the RIC system, is in fact considered as a reference instrumentation due to: 1. its axial resolution (1 acquisition/mm); 2. its self-calibration (several detectors are able to monitor the same channel); 3. its precision, regardless of time (negligible wear since the detectors are only irradiated approximately 1 hour per month); 4. almost complete coverage per quadrant in the cases of current 3-Loop and 4-Loop cores; 5. well-controlled final uncertainty (EUN) and relying on a considerable experimental database. Reassembly is thus carried out according to the following relation:(EUN)SchX=(EUN)REF+(ΔEUpN)SchXREF Equation 2a The SchX term refers to the expression “Scheme X”, applying to all instrumentation systems different than the reference instrumentation system (referred to by the term REF). The (ΔEU2pN)SchXREF corrective term of equation 2a, defining this differential reassembly, can thus be applied using: ( E Up N ) REF = ( μ U N ) 2 + ( R U ⁢ ⁢ 1 N ) REF 2 + ( R U ⁢ ⁢ 2 ⁢ p N ) REF 2 + ( M U N ) REF 2 ( equation ⁢ ⁢ 3 ) ⁢ And ( E Up N ) SchX = ( μ U N ) 2 + ( R U ⁢ ⁢ 1 N ) SchX 2 + ( R U ⁢ ⁢ 2 ⁢ p N ) SchX 2 + ( M U N ) SchX 2 ( equation ⁢ ⁢ 4 ) This corrective term contains not only the (ΔRU2pN)SchXREF difference, but also those that resulting from a change in detector or from a combination of detectors; hence, for example, the variations (ΔRU1N)SchXREF, (ΔMUN)SchXREF and/or (ΔXUN)SchXREF, X then referring to an uncertainty factor only existing for the SchX configuration. From the aspect of the distributions of the reconstructed power, the RU2pN component remains the characteristic indicator of any instrumentation system. The (ΔRU2pN)SchXREF difference is thus the determining parameter during the dimensioning of the EUN uncertainty and it has been analysed for all configurations of the disruption database. The variability observed on the (ΔRU2pN)SchXREF difference is, for the most part, a consequence of the MUN factor via the 3D sound-effect process of the Pseudo-Measurements. In practice, this difference is defined upstream of the final reassembly via the use of a statistical approach. As previously explained, in several reactor cores, sensors belonging to the RIC-type instrumentation system have been replaced by collectron-type sensors. Such reduction, which typically involves passing the number of sensors of the RIC system from 58 to 42, thus makes the number of collectrons fixed in the reactor core amount to 16. Hence, 16 guide tubes, which were normally monitored by mobile probes of the RIC system, are no longer of use when establishing a mapping deduced solely from the RIC sensors. Such a configuration results in a significant reduction in the number of instrumented channels monitored by the sensors of the RIC reference system, such sensors enabling to perform mappings that are normally required for checking the compliance of the reactor's core during start-up tests and for ensuring the periodic monitoring of the hot spot factors throughout the irradiation cycle. Moreover, in the invention, interest is particularly given to the collectron-type detectors. Collectrons are the detectors located at a fixed elevation inside the nuclear reactor core, and which are able to supply datum on a non-stop basis. The most widespread collectrons are the Rhodium-type collectrons. The measurements performed are directly processed online by an integrated calculator or by a section calculator. The system's response time essentially depends upon the performances of said calculator which determines the calculation time. The operating principles of the collectrons are henceforth known and are available in various literature. A major problem to be resolved when using the collectrons resides in the fact that the MUN uncertainty component increases significantly depending on the time passed inside the reactor and on the wear of the Rhodium emitter associated with the considered collectron. In order to take account of such wear of the Rhodium emitter, a correction law has been established after seven years of experimentation in a powerful reactor: the application of such law to the signal supplied by a collectron upon completion of an operating t-duration enables to relocate the signal that said collectron would have emitted at the outset. Such law, called “sensitivity law”, is stated as follows: S ⁡ ( t ) = S ⁡ ( 0 ) × ( 1 - Q ⁡ ( t ) Q ∞ ) a Relation ⁢ ⁢ 0 With: Q(t)=∫I(t′).dt′ and where I(t) is the gross current supplied by the detector to a t-instant. In practice, the initial signals supplied by the detector have had to be conditioned by deconvolution methods (in order regain lost time linked to the characteristics of the nuclear reactions in play) and by filtering (in order to reduce the noise induced by the deconvolution methods). The basic term should here be understood as “before correction of wear”. S(0) is the initial sensitivity of the detector and Q∞ is its total available load. The exponent a is a coefficient determined empirically as a result of experimentation. If the current supplied by a collectron at a given point in time of its irradiation is referred to as I(t), and the signal that should have had the same non-depleted collectron is referred to as I(0), then the sensitivity correction is performed according to the following relation: I ⁡ ( 0 ) = I ⁡ ( t ) S ⁡ ( t ) Relation ⁢ ⁢ 0 ⁢ bis A major consequence of the relation 0 is that, by implicating the integrated load, the application of said relation results in an accumulation of uncertainties on the debited current, whereby an increase in the overall error in accordance with the irradiation time. Hence, this uncertainty, or error, estimated at 2% at the start of life, attains between 4.3% and 68% of wear and exceeds 8% at collectron life end for 80% of wear, such as represented in FIG. 3, which illustrates the sensitivity law and the MUN uncertainty component for a Rhodium collectron in accordance with the wear of the considered detector, as well as, for comparison purposes, the uncertainty component MUN for a RIC-type detector. In comparison, it can be recalled that the MUN component of the RIC system is less than 2% and undergoes no increase during irradiation. One application of the method according to the invention concerns a collectron-type detector calibration method placed inside a nuclear reactor core. Such an application of the method according to the invention enables to benefit from a signal, supplied by a collectron-type detector, associated with an uncertainty component whose value is not too high, even after prolonged use of the collectron inside the nuclear reactor core. By the term “collectron calibration” is meant the fact of associating a signal, supplied by a collectron, representative of an activity inside a nuclear reactor with an uncertainty component associated with the considered collectron-type detector. The present invention provides a solution to the problems that have just been evoked. In the invention, a reduction in the spatial density of the reference measurements is proposed, as compensation, via the joint optimal use of measurements deriving from another system. The method used in order to ensure such compensation implements a means based on the mixed mapping principle, which will be developed hereunder. In the method according to the invention, the simultaneous presence of, on the one hand, sensors of the mobile reference instrumentation system (RIC system) and of, on the other hand, acquisitions constantly supplied by the sensors of the fixed instrumentation system (collectrons), is described for the benefit of the user. One immediate application of implementing the method according to the invention lies in initiating a calibration method of the collectron-type sensors, by using data from the supplementary RIC instrumentation system. Hence, the characteristic of the collectrons, according to which the MUN component attains high values relatively quickly, which significantly penalises, for example, all monitoring systems using the continuous acquisitions of collectrons, is resolved. The present invention thus essentially refers to a method for establishing a mapping representative of a power distribution inside a nuclear reactor core, the said mapping being established by means of detectors placed at least temporarily inside the reactor core, characterised in that it comprises the different steps consisting of: equipping, at least temporarily, a first fuel assembly unit of the nuclear reactor core by means of detectors from a first instrumentation system, called “reference instrumentation system”. equipping, at least temporarily, a second fuel assembly unit of the nuclear reactor core by means of detectors from a second instrumentation system; performing a first partial series of activity measurements by means of the reference system detectors; performing a second partial series of activity measurements by means of the detectors from the second instrumentation system; converting the activity measurements of the second series of measurements into activity measurements associated with the reference instrumentation system in order to obtain a series of converted measurements; establishing, by way of the first partial series of activity measurements, of the series of converted measurements, of a complete theoretical distribution of the theoretical activities associated with the reference instrumentation system, and of a complete theoretical distribution of the theoretical activities associated with the second instrumentation system, for every instrumented point of the nuclear reactor core, a final series of experimental reactor core activities, said data series only comprising the values relating to the activities associated with the reference instrumentation system; establishing (307), by way of the final series of experimental data and of theoretical data simulating a state of the reactor core at the time of the embodiment of the first series of measurements and of the second series of measurements, the mapping being representative of a power distribution mechanism inside a nuclear reactor core. Besides the main characteristics just mentioned, the method according to the invention may represent one or several additional characteristics from among the following: the step whereby conversion of the activity measurements of the second series of measurements into the activity measurements associated with the reference instrumentation system complies with the following relation: A 1 ← 2 CONV = g · ( A 1 A 2 ) CAL · ( A 2 MES ) relatif where: A2MES is the distribution of the activities measured by the detectors of the second instrumentation system, whose elements are linked by a normalization of series; A1CAL and A2CAL are the distributions of the equivalent activities calculated for the detectors of the first instrumentation system and of the second instrumentation system, respectively; A1←2CONV is the conversion of an activity measured by a detector of the second instrumentation system into an activity that would be detected by a detector of the first instrumentation system; g is a coefficient reporting the normalization differences between the two distributions. the step of converting the activity measurements of the second series of measurements into the activity measurements associated with the reference instrumentation system is followed by a normalization operation complying with the following relation: A 1 mixte = ∑ 1 N ⁢ ⁢ 1 ⁢ ( A 1 MES ) relatif + ∑ 1 N ⁢ ⁢ 2 ⁢ A 1 ← 2 CONV N ⁢ ⁢ 1 + N ⁢ ⁢ 2 where: N1 and N2 are the number of acquisitions for the detectors of the first instrumentation system and of the second instrumentation system, respectively; (A1MES)relatif is the distribution of the activities measured by the detectors of the first instrumentation system, whose elements are linked by a normalization of series; A1←2CONV is the conversion of an activity measured by a detector of the second instrumentation system into an activity that would be detected by a detector of the first instrumentation system. the reference instrumentation system is a system with mobile detectors. the second reference instrumentation system is a system with fixed detectors. the reference instrumentation system is a RIC-type system. the second reference instrumentation system uses collectron-type detectors. The present invention also refers to a method of correction for a component of intrinsic uncertainty associated with a collectron-type detector placed inside a nuclear power plant core, characterised in that it comprises the step consisting of performing a calibration operation on the considered collectron, the calibration operation being performed when the collectron-type detector has attained a given level of wear, the calibration operation consisting of performing a three-dimensional calibration using a mapping determined, with a reference instrumentation system implicating RIC-type detectors, by the method described above. This three-dimensional calibration resides on a particular mixed mapping methodology application, such application enabling to define the 3D factors that come to correct the values of the sensitivity law in the (XYZ) positions monitored by the collectrons. Besides the main characteristics just mentioned, the correction method may represent one or several additional characteristics from among the following: the calibration operation, consisting of performing a three-dimensional calibration using flux maps determined for the detectors of RIC-type detectors, provides, at the given wear level at which the calibration operation is performed, as value on the intrinsic uncertainty component of the considered collectron-type detector, the value of the determined intrinsic uncertainty component for a new detector, increased by a value called “calibration uncertainty”; the level of given wear, for which a calibration operation of the considered collectron is performed, is comprised between 50% and 60% of wear of the said collectron. The invention and its different applications will be better understood upon reading the following description and after studying the figures attached thereto. The method according to the invention uses algorithms developed in the context of the mixed flux maps, in which the method according to the invention finds its reasoning. A brief presentation of the mixed flux map principles is now recalled. First presented below are the algorithms for processing the mixed flux maps. For mechanical reasons, the fixed IN-CORE detectors may only be installed in positions normally monitored by mobiles IN-CORE detectors of the RIC system. The result causes a reduction in the data density available via said reference system. In order to compensate such reduction and therefore to avoid possible penalties, it thus became interesting to properly use all the available experimental data at the same time. The combination of the experimental distributions deduced from the two systems during an EP11-type measurement campaign enables to retrieve maximal density. The combination method of the mixed flux maps is briefly described below: Two experimental distributions of different types {A1} and {A2} and also comprising a different number of elements (N1, N2). Each of these units being normalized per unit, the elements A1 and A2 are linked by the following relations: ∑ 1 N ⁢ ⁢ 1 ⁢ A ⁢ ⁢ 1 = N ⁢ ⁢ 1 ⁢ ⁢ et ⁢ ⁢ ∑ 1 N ⁢ ⁢ 2 ⁢ A ⁢ ⁢ 2 = N ⁢ ⁢ 2 Relation ⁢ ⁢ 1 These separate initial normalizations are indispensable since the values deduced from the two acquisition systems are expressed in different units and, moreover, do not necessarily have the same physical nature. The aim is to obtain a new series, also normalized per unit and comprising elements N1+N2. In order to do so, it is first necessary to make the two units coherent under the aspect of their physical nature. These two distributions being able to be determined using theoretical models, the combination method is then based on the invariance assumption of the MES/CAL report: A 1 MES A 1 CAL ≈ A 2 MES A 2 CAL Relation ⁢ ⁢ 2 The symbol “≈” appearing in the relation 2 is rather a symbol of proportionality, implying that the corresponding equivalence is unable to be directly applied in order to convert type 2 data into type 1 data: it is in fact necessary to bear in mind that the initial normalizations refer to different units under the double aspect of the spatial apportionment and of the number of elements. The conversion of a type 2 element must be carried out based on gross values: ( A 1 ← 2 CONV ) brut = ( A 1 A 2 ) CAL · ( A 2 MES ) brut Relation ⁢ ⁢ 3 The gross label here implies “by identical normalization”. The A1/A2 conversion ratio of the relation 2 may nevertheless be used “as is”, since its two components have been obtained on the same unit: the entire active core. The aim being sought is thus to approach, as near as possible, the gross value that would have been obtained in the case of a homogenous instrumentation. This aim is recorded as:(A1←2CONV)brut=(A1MES)brut Relation 4 The indexed gross values not generally being available within a common unit, any comparison implies normalization and thus definition of the relative values: A relatif ensemble ⁢ ⁢ N = A brut A _ brut ensemble ⁢ ⁢ N Relation ⁢ ⁢ 5 The relation 3 is applied to the gross values. The use of the relative values, the only ones truly available, obliges addition of a g coefficient to said relation, in order to have knowledge of the normalization differences: A 1 ← 2 CONV = g · ( A 1 A 2 ) CAL · ( A 2 MES ) relatif Relation ⁢ ⁢ 6 the problem then being to determine the g factor. Due to the simultaneous unavailability of the real gross values of type 1 and type 2 data, the g factor shall be evaluated by considering that the values deduced from the complete theoretical distributions have the gross label, since such values originate from identical normalizations. Under such conditions, the relations 3 and 4 may be recorded in the following manner: ( A 1 ← 2 CONV ) brut = ( A 1 A 2 ) CAL · ( A 2 CAL ) complet Relation ⁢ ⁢ 7 ( A 1 ← 2 CONV ) brut = ( A 1 CAL ) complet Relation ⁢ ⁢ 8 The values effectively available being relative values defined by the relation 5, the relation 9 is thus recorded as follows: ( A 1 CAL ) relatif ⁢ ⁢ N ⁢ ⁢ 1 · ( A _ 1 CAL ) N ⁢ ⁢ 1 / complet = ( A 1 A 2 ) CAL · ( A 2 CAL ) relatif ⁢ ⁢ N ⁢ ⁢ 2 · ( A _ 2 CAL ) N ⁢ ⁢ 2 / complet The equivalence of the relations 6 and 9 provide the value of the g coefficient: g = ( A _ 2 CAL ) N ⁢ ⁢ 2 / complet ( A _ 1 CAL ) N ⁢ ⁢ 1 / complet Relation ⁢ ⁢ 10 The definition of the new unit is then based on the normalization of the initial activities (A1MES)relatif and the converted activities A1←2CONV via the relation 6. This final normalization is recorded as: A 1 mixte = ∑ 1 N ⁢ ⁢ 1 ⁢ ( A 1 MES ) relatif + ∑ 1 N ⁢ ⁢ 2 ⁢ A 1 ← 2 CONV N ⁢ ⁢ 1 + N ⁢ ⁢ 2 Relation ⁢ ⁢ 11 This ‘mixed’ unit is thus completely defined and possess all the necessary characteristics (coherence and standard). The relation 6 reveals ratio (A1/A2) enabling conversion of type 2 into type 1, with the g coefficient giving information on the normalization differences. Uncertainty on the A1/A2 ratio may be established in a similar manner to that already implemented for the P/A ratio, evoking the correlation existing between the activity in the instrumented channel and the mean power of the assembly containing said channel; it is recalled that the uncertainty associated with the use of this ratio is recorded as RU1N. The typical gaps characterising the uncertainties on the two terms of this ratio are designated by σA1 and σA2. Hence:σA1/A22=σA12+σA22−2·r·σA1·σA2 with r characterising the correlation between the terms A1 and A2. Feedback from the reactor known as the CATTENOM 1 reactor illustrate that the C/M gaps are almost the same as the units A1 (U5) and A2 (Rh). Hence, σA1=σA2=σA, which completely defines the conversion component of the TUN factor:(σ(TUN))conversion2=σA1/A22=2·(1−r)·σA2 Relation 12 The coefficient of correlation r may then be obtained by plotting A2 in accordance with A1 for symmetrical positions. In the case of the RU1N uncertainty factor, the correlation coefficient was found to be higher by 0.95 in more than 95% of cases. This is also the case for the couples (A1, A2): the digital value to be allocated to the (σ(TUN))conversion factor will thus be the same as the σ(RU1N) factor. The uncertainty associated with the g coefficient may, on its part, be evaluated by using the existing complete maps established using the RIC systems. Comparison of the power distributions reconstructed by said method (42+16) with the standard distributions reconstructed with the sketch 58 enables to quantify the (TUN)normalisation factor. The two effects, conversion type and normalization, being independent, hence:(TUN)2=(TUN)conversion2+(TUN)normalisation2 Relation 13 Now the algorithms for processing the mixed flux maps are applied to the calibration of the collectrons. This method is in fact an extension of the algorithms implemented for the mixed maps (RIC 42 channels+COL 16 rods). The departure point is the same: conversion of the collectron signals into Pseudo RIC signals. However, here the collectron signals used upstream of the process are non-corrected signals of the law of wear; they are only deconvoluated and filtered. The first relation of the method is then: A PRIC BRUT = g · [ A RIC A COL ] CAL · A COL BRUT Relation ⁢ ⁢ 14 Following the standard processing of a supplementary RIC flux map (42 channels), RIC activities are available, reconstructed in all points and notably in the positions occupied by the collectrons. Such distribution is thus recorded as ARICEST. A series of 30 corrective factors may now be defined: FCOR = A PRIC BRUT A RIC EST Relation ⁢ ⁢ 15 Although these 3D correction or calibration factors do not have the purpose of directly correcting the gross collectron activities, the relations hereafter illustrate the process for passing from an initial collectron signal, i.e. not corrected by the sensitivity law, to a signal having the same meaning as a RIC acquisition. Hence, with ACOLCOR, with the 3D value of a corrected collectron activity: A COL COR = A COL BRUT FCOR Relation ⁢ ⁢ 16 Still via application of the base relation of the ‘mixed map’ process, the following can be recorded: A PRIC COR = g · [ A RIC A COL ] CAL · A COL COR Relation ⁢ ⁢ 17 there again, through application of the relation 14: A PRIC COR = g · [ A RIC A COL ] CAL · A COL BRUT · A RIC EST A PRIC BRUT whereby:APRICCOR=ARICEST Relation 18 This last relation well illustrates that the collectron activities, after conversion of course into Pseudo RIC activities, are calibrated on the RIC-reconstructed activities. From a practical point of view, the calibration operation is more complex since the aim is then to correct the error committed on the integrated loads due to an accumulation of errors on the current. As soon as implementation of said calibration is decided, the FCOR factors are due to replace, via a process of mathematical equivalence, the S(t) sensitivity law presented in the relation 0 and used in the relation Obis for retrieving the signal that would have had the non-depleted collectron. The uncertainty associated with such calibration thus comprises, on the one hand, the TUN component, defined for the mixed maps (relation 12), and, on the other hand, the (RU2pN)PICSch42 component, defined in the context of the application of the RU2pN method described in the previous paragraph. Whereby:CUN=√{square root over ((TUN)2+[(RU2pN)RICSch42]2)}{square root over ((TUN)2+[(RU2pN)RICSch42]2)} Relation 19 The values of these two uncertainty factors are respectively 1.5% and 1.5%, whereby 2.1% for the CUN uncertainty. The MUN intrinsic uncertainty component at the start of the collectrons' life being by 1.93%, this implies that as of when calibration is activated, the collectrons' final uncertainty would be by 2.9%. This latest value is preliminary, but the fact that it is very clearly lower than the values than it would be necessary to use without calibration is a strong argument in favour of this approach. FIG. 4 illustrates an implementation example of the method according to the invention, by notably applying the mixed flux map combination method. In this Figure, the following abbreviations have been used: MFM=Mixed Flux Map; PREP=PREParation; DET=DETector; A=Activity; P=Power In this Figure, on the one hand is considered a first series of partial measurements 301, performed by means of a first type of sensors, or detectors, belonging to a first IN-CORE instrumentation system present inside a nuclear reactor core, and on the other hand a second series of partial measurements 302, performed by means of a second type of detectors, belonging to a second IN-CORE instrumentation system. By series of partial measurements is meant measurements performed for the sole points of the nuclear reactor equipped with detectors of the considered sensor type. It is further considered a first complete theoretical distribution 304, i.e. supplying an activity value for each point of the considered nuclear reactor core, available for the first-type detectors, and a second complete theoretical distribution 305, available for the second-type detectors. Based on these four data series, in a next step, a conversion of the activities associated with the second detector type into the activities associated with the first detector type is performed for the points of the instrumented nuclear reactor core. Such conversion, whose principle resides on the MES/CAL invariance assumption (relation 2), is entirely performed after application of the relation 11. Hence, a new series of experimental data 306 is obtained, for every instrumented point, whether this be by the first-type sensors or the second-type sensors, the new series of experimental data only comprising values relating to activities associated with the first-type sensors. Based on the new series of experimental data 306, and by implicating a series of theoretical data 303 obtained via a theoretical calculation simulating the state of the core at the time of the double mapping, a standard processing is thus initiated, previously recalled in reference to FIGS. 1 and 2, of a flux map that would have been exclusively obtained by using a satisfactory density. FIG. 5 illustrates the impact of the implementation of the mixed flux map combination method on the evolution of the uncertainty of the Rhodium detectors: in this figure is represented the case where a three-dimensional calibration of a collectron-type detector is performed using RIC flux maps when the wear of the considered collectron is in the region of 55%. Hence, when the calibration is performed, the collectron's MUN uncertainty component, when departing from a determined level via the CUN calibration uncertainty component, follows the evolution law dictated by the sensitivity law, by considering the latter as of its origin, i.e. by departing from the initial t instant (t=0). In between two calibrations, the standard correction law is applied using a change of origin. It is thus clear that this process enables to eliminate a vast share of the initial error propagation via the law of wear. Appropriately, a single calibration is planned: as of 50% of wear. This then allows operating, with a compatible uncertainty level and with the required operating constraints, for 3 to 4 years preceding the decennial of the section: during the corresponding stoppage, all the collections will then be replaced. This calibration may also be used at the beginning of each cycle, but in that case, without modifying the collectron signals in order to verify the axial positioning of the rods.
	summary
	description	Hereinafter, embodiments of the present invention will be described with reference to the drawings. A first embodiment of the present invention will be described with reference to FIGS. 1 to 6. A vertical sectional view showing the structure of a fuel assembly in this embodiment is shown in FIG. 2, and a transverse sectional view taken on line Axe2x80x94A of FIG. 2 is shown in FIG. 3. Referring to FIGS. 2 and 3, a fuel assembly 1 includes 74 fuel rods 2, filled with fuel pellets (not shown), which fuel rods are arranged in a square lattice array of 9 rowsxc3x979 columns; two water rods 3 arranged in a region in which seven of the fuel rods 2 are arrangeable; fuel spacers 4 for holding the fuel rods 2 and the water rods 3 with mutual radial intervals thereof kept immovable; an upper tie plate 5 and a lower tie plate 6 for holding the upper end portion and the lower end portion of a fuel bundle composed of the fuel rods 2 and the water rods 3, respectively; and a fuel channel box 8 for covering the outer peripheral portion of the above structure. The fuel rods 2 include normal-length fuel rods 2a, each having a normal fuel active length (filling length of fuel pellets), and short-length fuel rods 2b each having an effective length shorter than that of the normal-length fuel rods 2a. The short-length fuel rods 2b include four first short-length fuel rods 2b1 arranged in the outermost peripheral region of the square lattice array, and two second short-length fuel rods 2b2 arranged in a region adjacent to the water rods 3. The fuel spacers 4 are provided at a plurality of positions arranged in the axial direction. As shown in FIG. 3, there are 74 fuel rods 2, including the short-length fuel rods 2b, which are short in fuel active length and accordingly, the lattice positions, which are filled with the shortlength fuel rods 2b in the fuel spacer 4a positioned on the lower portion of the fuel assembly 1, become empty of fuel rods in the fuel spacer 4b positioned on the upper portion of the fuel assembly 1. For this reason, the structure of the fuel spacer 4b positioned above the upper ends of the short-length fuel rods 2b is designed to be slightly different from the structure of the fuel spacer 4a positioned below the upper ends of the short-length fuel rods 2b. Top views of the structures of these fuel spacers 4a and 4b are shown in FIGS. 4 and 1, respectively. Referring to FIGS. 4 and 1, each of the fuel spacers 4a and 4b includes a large number (74 cells for the spacer 4a, 70 cells for the spacer 4b) of cylindrical members (cells) 9 which are arranged in a square lattice array of 9 rows x 9 columns corresponding to the square lattice array of the fuel rods 2, and these cylindrical members are welded to each other and are of a size to permit the fuel rods 2 to be inserted therein, respectively; a square-shaped band member (band) 11 which surrounds the outer periphery of the joined cells 9; water rod holding members 12, each being formed into a xcexa9-shape in transverse cross-section, which are welded to those cells, which are arranged in the innermost peripheral region of the square lattice array, of the cells 9, for holding the water rods 3 in the radial and axial directions; approximately quarter-round water rod holding members 13; and water rod holding springs 14, provided on the water rod holding members 13, for imparting pressing forces to hold the water rods 3 in position. Each cell 9, which is formed into an approximately cylindrical shape, includes two projections 9a for holding a respective fuel rod 2; and a spring supporting portion (not shown), provided at the joined portion with the adjacent cell 9, for suitably supporting a loop-shaped spring 10 to press against the fuel rod 2 inserted in the cell 9. It should be noted that the structures of the loop-shaped spring and the spring supporting portion, while not shown particularly in detail in the figures, are known for example from Japanese Patent Laid-open No. Hei 6-273560. The band 11, having a square-shape, whose four sides are welded to each other, includes a large number of flow tabs 15 each of which is bent in such a manner as to project between the adjacent ones of the cells 9 in the outermost peripheral region of the square lattice array in order to introduce the flow of a coolant; and eight bathtubs 16 provided two for each side of the square-shape of the band 11, each tub projecting on the fuel channel box 8 side so as to be brought in contact with the inner surface of the fuel channel box 8. The feature of this embodiment lies in the structure of the fuel spacer 4b. That is to say, the fuel spacer 4b shown in FIG. 1 is different from the fuel spacer 4a shown in FIG. 4 in that the cells 9 located at first lattice positions 7a associated with the first short-length fuel rods 2b1 are omitted and instead the supporting members 17 are provided at the lattice positions 7a. The supporting member 17 connects the two cells 9A and 9B, adjacently located on both the sides of the first lattice position 7a associated with the first short-length fuel rod 2b1 in the outer peripheral region of the square lattice array, to the band 11. FIG. 5 is a perspective view showing the structure of the supporting member 17. Referring to FIG. 5, the supporting member 17 is formed into a shape similar to one of two halves obtained by vertically dividing a cylinder having an octagonal cross-section. While not shown in FIG. 5 to avoid complication in the drawing, as shown in FIG. 1, the supporting member 17 includes, at the joined portion with the adjacent cell 9A, a spring supporting portion for suitably supporting the loop-shaped spring 10, which operates to hold the fuel rod 2 inserted in the cell 9A by imparting a pressing force against the fuel rod 2. It should be noted that the structure of the spring supporting portion, while not shown particularly in detail in the figures, is known for example from Japanese Patent Laid-open No. Hei 2-163695. In the fuel spacer 4b, the cells 9 located at the lattice positions associated with the second short-length fuel rods 2b2 are left as they are; however, the loop-shaped springs 10, which are unnecessary for the cells 9, because the fuel rods are not inserted in the cells 9, are removed from the cells 9. The fuel assembly in this embodiment, which is configured as described above, exhibits the following effects: (1) Reduction in Pressure Loss This effect will be described with reference to a comparative example in which a fuel spacer having the same structure as that of the fuel spacer 4a, in which all of the cells 9 are located without any being omitted at all of the lattice positions, as shown in FIG. 4, is positioned above the upper ends of the first short-length fuel rods 2b1 of the fuel assembly 1. In this comparative example, the cells 9, which are not required to be provided at the lattice positions associated with the first short-length fuel rods 2b1 because the fuel rods are not present at the lattice positions, are provided at the lattice positions, and therefore, the pressure loss is correspondingly increased. On the contrary, in the fuel spacer 4b in this embodiment, the cells 9 at the first lattice positions 7a associated with the first short-length fuel rods 2b1 are omitted, and instead the supporting members 17 each being formed into a semi-octagonal cross-sectional shape, are provided at the first lattice positions 7a, as shown in FIG. 1. As a result, since the flow resistance of water as a coolant flowing upward in the fuel assembly 1 is made significantly smaller than that in the comparative example, it is possible to sufficiently reduce the pressure loss. (2) Attainment of Structural Strength This effect will be described in detail with reference to the above-described comparative example. As described above, in the fuel spacer (having the same structure as that of the fuel spacer 4a shown in FIG. 4) in the comparative example, all of the cells 9 in contact with the band 11 surrounding the outer periphery of the spacer are continuously in contact with each other, to thereby maintain the structural strength of the entire spacer. For example, if an external force is applied to the fuel spacer via the fuel channel box 8 in case of an earthquake or in handling the fuel assembly, the load is first transmitted to eight of the bath-tubs 16 provided on the band 11. After that, the load is transmitted, via the band 11, to the cells 9 in the outermost peripheral region of the square lattice array joined to the inner side of the band 11, and then the force is sequentially transmitted to the cells 9 arranged on the inner peripheral side of the square lattice array (see FIG. 4). In this way, for the fuel spacer in this comparative example, since the cells 9 arranged in succession in the path along which the load is transmitted are integrally formed and an integrity of ensuring the structural effect as a whole is obtained, it is possible to sufficiently ensure the structural strength of the entire fuel spacer. On the contrary, for the fuel spacer 4b in this embodiment, as shown in FIG. 1, since one cell 9 between the cells 9A and 9B at each side of the square lattice array is omitted, the arrangement of the cells 9 in the outermost peripheral region becomes discontinuous at the position between the cells 9A and 9B. In this embodiment, however, since the two cells 9A and 9B are joined to each other by means of the supporting member 17, the cells 9A and 9B are rigidly fixed to each other via the band 11. As a result, when a load is transmitted from the band 11 as described above, it can be received by the joined structure composed of the fixed two cells 9A and 9B and the supporting member 17, and accordingly, it is possible for the fuel spacer 4b to provide a structural strength substantially comparable to that of the fuel spacer in the comparative example in which a cell 9 is located at the lattice position between the cells 9A and 9B. (3) Attainment of Degree of Freedom in Design of Short-length Fuel Rod Arrangement As described in the paragraphs (1) and (2), the fuel spacer 4b in this embodiment is effective to reduce the pressure loss while ensuring the structural strength of the entire spacer. Such an effect can be obtained even if the lattice position associated with the first short-length fuel rod 2b1 is located at any position in the outermost peripheral region of the square lattice array. In other words, according to this embodiment, it is possible to ensure the degree of freedom in design. In the design of a fuel assembly including short-length fuel rods, as described above, various arrangements of the short-length fuel rods may be considered in accordance with the nuclear characteristics required for the fuel assembly. Therefore, for example, there may be considered an arrangement of the first short-length fuel rod 2b1 at a position between the two opposed bath-tubs 16 in the outermost peripheral region of the square lattice array. In this case, for example, in the fuel space having the prior art structure disclosed in Japanese Patent Laid-open No. Hei 6-3473, since the cells 9 at all of the lattice positions between the two opposed bath-tubs 16 cannot be omitted, the pressure loss cannot be sufficiently reduced. On the other hand, if the reduction in pressure loss takes precedence, the short-length fuel rods cannot be arranged at all of the lattice positions between the two opposed bath-tubs 16, and therefore, the degree of freedom in design of the fuel assembly is correspondingly limited. On the contrary, in such a case, the fuel spacer 4b in this embodiment can be modified, in accordance with the arrangement of the lattice positions associated with the first short-length fuel rods 2b1, for example, into a fuel spacer 4bA shown in FIG. 6 in which the cells 9 located at the lattice positions between the two opposed bath-tubs 16 are omitted and instead the supporting members 17 are provided at those lattice positions. Accordingly, unlike the fuel spacer having the prior art structure, even if the lattice positions associated with the first short-length fuel rods 2b1 are located between the two opposed bath-tubs 16 in the outermost peripheral region of the square lattice array, it is possible to sufficiently reduce the pressure loss while ensuring the strength of the fuel spacer. As described in the paragraphs (1) to (3), according to the fuel spacer 4b in this embodiment, it is possible to sufficiently reduce the pressure loss of the fuel spacer 4b positioned upward from the upper ends of the short-length fuel rods 2b1 while usually ensuring the structural strength of the fuel spacer 4b irrespective of the arrangement of the lattice positions associated with the first short-length fuel rods 2b1. (4) Attainment of Degree of Freedom in Design of Spring Arrangement As described above, the known loop-shaped spring 10 for pressing the fuel rods 2 is essentially disposed between the adjacent cells 9, and it functions to generate pressing forces when the fuel rods 2 are inserted in the cells 9, respectively. Accordingly, if the means for imparting a spring pressing force is not provided on the supporting member 17, which is additionally provided at the lattice position associated with the first short-length fuel rod 2b1, the supporting member 17 side of the loop-shaped spring 10 disposed at the joined portion between the supporting member 17 and the adjacent cell 9A comes into a free end, with a result that the loop-shaped spring 10 cannot achieve the function of pressing against the fuel rod 2 inserted in the cell 9A. Accordingly, to press the fuel rod 2 in the cell 9A, the loop-shaped spring 10 must be disposed between the cell 9A and the cell 9 which is adjacent to a portion, opposed to the supporting member 17, of the cell 9A. As a result, the spring arrangement in the entire spacer must be reviewed as a whole. This imposes a large limitation on the design. In this embodiment, however, the spring supporting portion provided on the supporting member 17 supports the loop-shaped spring 10 for pressing the fuel rods 2 and imparts a pressing force to the loop-shaped spring 10. As a result, since the loop-shaped spring 10 in the cell 9A is allowed to function just as in the fuel spacer 4a shown in FIG. 4, it is possible to increase the degree of freedom in arrangement of the loop-shaped springs 10 and hence to ensure a degree of freedom of design comparable to that in the fuel spacer 4a shown in FIG. 4. A second embodiment of the present invention will be described with reference to FIGS. 7 to 11). This embodiment has a feature such that, in the fuel spacer positioned above the upper ends of the first short-length fuel rods 2b1, the spring arrangement and the shape of the supporting member are changed. FIG. 7 is a top view showing the structure of a fuel spacer 204b in this embodiment. In the fuel spacer 204b, parts common to those in the fuel spacer 4b described in the first embodiment with reference to FIG. 1 are designated by the same symbols and an explanation thereof will be omitted. The fuel spacer 204b shown in FIG. 7 is different from the fuel spacer 4b shown in FIG. 1 in that the cells 9 at the second lattice positions 7b associated with the second short-length fuel rods 2b2 are omitted and instead supporting members 218 are provided at the second lattice positions 7b. The supporting member 218 connects the two cells 9C and 9D, located outwardly from and adjacently to the second lattice position 7b in the square lattice array, to the water rod holding member 12. The supporting member 218 is formed into an approximately polygonal cylindrical shape with an unnecessary side portion in terms of structure cut off for making the pressure loss as small as possible. In the fuel spacer 204b, a supporting member 217, having a structure in which the spring supporting portion is removed from the supporting member 17 shown in FIG. 1, is used as a supporting member for connecting the two cells 9A and 9B adjacently located on both sides of the lattice position associated with the first short-length fuel rod 2b1, to the band 11. With this configuration, since a loop-shaped spring 10 is not disposed on the supporting member 217, the spring arrangement in the entire fuel spacer is changed such that the supporting member 218 has two spring supporting portions (not shown) for suitably supporting the two loop-shaped springs 10 to press against the fuel rods 2 inserted in the cells 9C and 9D to impart pressing forces thereto. To be more specific, two of the known spring supporting portions having the same structure as that of the spring supporting portions used for the cells 9 are simply provided at a joined portion between the cells 9C and 9D of the supporting member 218. The remaining configuration of this embodiment is substantially the same as that of the first embodiment. According to this embodiment, in addition to the same effect as that of the first embodiment, there can be obtained an effect of simplifying the structure because the supporting member 217 has no spring supporting portion. While the supporting member 217 formed into a semi-octagonal cylindrical shape is used in the second embodiment, the present invention is not limited thereto. For example, the supporting member 217 may be formed into another shape, for example, a semi-cylindrical shape (with a partial peripheral length portion cut off) having the same thickness as that of the cell 9. This exhibits the following effect. In general, the cell 9 is manufactured by cutting a circular tube, having a specific outside diameter and a specific thickness, into a specific length, and processing the cut piece to form the projections 9a. and also cutouts for the spring supporting portion. Here, if a supporting member 217A formed into a semi-cylindrical shape having the same thickness as that of the cell 9 is used as the supporting member, such a supporting member 217A can be manufactured using the raw circular tube for forming the cell 9. This is effective to reduce the manufacturing cost by making the circular tube shareable between the-supporting member 217A and the cell 9. From the viewpoint of reduction in pressure loss, it may be Desirable that the peripheral length of the cylindrical shape of the supporting member 217A be made as short as possible within a length range required for welding the supporting member 217A to the adjacent cells 9 with no problem. Further, a supporting member 217A formed into an approximately cylindrical shape similar to that of the cell 9 may be used as the supporting member. FIG. 8 is a top view showing the structure of a fuel spacer 204bA including such a supporting member 217A. The supporting member 217A connects the cells 9A and 9B, adjacently located on both sides of the lattice position 7a associated with the first short-length fuel rod 2b1, to the band 11, and also the supporting member 217A is connected to the cell 9E located inwardly from and adjacently to the first lattice position 7a in the square lattice array. In addition, the supporting member 217 is made as thin as possible within an allowable thickness range in terms of the structural strength of the fuel spacer for making the cross-sectional area smaller than that of the cell 9 thereby reducing the pressure loss. The supporting member 217A formed into an approximately cylindrical shape is manufactured using a circular tube having a specific thickness, which tube is different from the raw circular tube for forming the cell 9, or using the raw circular tube for forming the cell 9, and grinding the inner surface of the tube to increase the inside diameter (that is, decrease the thickness). In the latter case, there can be obtained an effect of reducing the manufacturing cost by making the circular tube shareable between the supporting member 217A and the cell 9. It should be noted that the cross-sectional shape of the supporting member 217A in this modification is not limited to a cylindrical shape, but may be of course a polygonal shape insofar as it satisfies the requirement that the cross-section of the supporting member 217A is smaller than that of the cell 9. Further, the supporting member 217A may be configured as a member having the same cross-sectional shape as that of the supporting member 218 except that the spring supporting portions for supporting the loop-shaped springs 10 are not provided. In the manufacture of the supporting member 218, the member 217 (replaced from the supporting member 217A) having the same cross-sectional shape as that of the supporting member 218 can be manufactured by punching or bending the same raw material as that for the supporting member 218. This is effective to reduce the manufacturing cost by making the raw material shareable between the member (replaced from the supporting member 217A) and the supporting member 218. In this sharing of the raw material, the shape of the supporting member 218 is not limited to a polygonal cylindrical shape but may be of course a thin cylindrical shape or a cylindrical shape with a partial peripheral portion cut off. A third embodiment of the present invention will be described with reference to FIGS. 9 to 11. This embodiment has a feature such that, in the fuel spacer positioned above the upper ends of the first short-length fuel rods 2b1, the shape of the supporting member and the supporting structure are further changed. FIG. 9 is a top view showing the structure of a fuel spacer 304b in this embodiment. In the fuel spacer 304b, parts common to those of the fuel spacer 4b described in the first embodiment with reference to FIG. 1 are designated by the same symbols and an explanation thereof will be omitted. The fuel spacer 304b shown in FIG. 9 is different from the fuel spacer 4b shown in FIG. 1 in that the supporting member 17 located at the first lattice position (see FIG. 3) associated with the first short-length fuel rod 2b1 is replaced with a supporting member 317. The supporting member 317 connects the two cells 9A and 9B adjacently located on both sides of the first lattice position 7a associated with the first short-length fuel rod 2b1 in the outermost, peripheral region of the square lattice array, and the cell 9E located inwardly from and adjacently to the first lattice position 7a in the square lattice array, to each other. In summary, the supporting member 317, which is not joined to the band 11, joins the adjacent cells 9A, 9B and 9E to each other. With this configuration, the entire fuel spacer has a structure as shown in FIG. 9 in which each cell 9 in the second layer from the outermost periphery of the square lattice array is fixedly joined to four adjacent cells or three adjacent cells and one supporting member 317 at four positions spaced at intervals of 90xc2x0 in the circumferential direction thereof. Like the supporting member 17 in the first embodiment, the supporting member 317 is formed into one of two halves obtained by vertically dividing a cylinder having an octagonal cross-section. While not shown in detail, like the supporting member 17, the supporting member 317 includes, at the joined portion with the adjacent cell 9A, a spring supporting portion for suitably supporting a loop-shaped spring 10 to hold the fuel rod 2 inserted in the cell 9A to impart a pressing force thereto. The other configuration is substantially the same as that of the first embodiment. Like the fuel assembly in the first embodiment, the fuel assembly in this embodiment, configured as described above, has the following four effects: (1) Reduction in Pressure Loss In the fuel spacer 304b in this embodiment, the cells 9 at the first lattice positions 7a associated with the first short-length fuel rods 2b1 are omitted, and instead the supporting members 317 each being formed into a semi-octagonal cross-sectional shape are provided at the first lattice positions 7a. As a result, since the flow resistance of water as a coolant flowing upward in the fuel assembly 1 is significantly reduced, it is possible to sufficiently reduce the pressure loss. (2) Attainment of Structural Strength In the fuel spacer 304b in this embodiment, as shown in FIG. 9, since one cell 9 between the cells 9A and 9B at each side of the square lattice array is omitted, the arrangement of the cells 9 in the outermost peripheral region becomes discontinuous at the position between the cells 9A and 9B. In this embodiment, however, since the two cells 9A and 9B are connected to the adjacent cell 9E on the inner peripheral side by means of the supporting member 317, the two cells 9A and 9B are rigidly fixed to each other via the cell 9E. As a result, when a load is transmitted from the band 11, it can be received by the joined structure composed of the fixedly connected two cells 9A and 9B, the supporting member 317, and the cell 9E. Accordingly, it is possible for the fuel spacer 304b to provide a structural strength substantially comparable to that of the fuel spacer in which the cell 9 is located at the lattice position between the cells 9A and 9B. (3) Attainment of Degree of Freedom in Design of Short-length Fuel Rod Arrangement In the fuel spacer 304b in this embodiment, even if the lattice position associated with the first short-length fuel rod 2b1 is located at any position in the outermost peripheral region of the square lattice array, the cell 9 at the lattice position can be omitted and instead the supporting member 317 can be provided at the lattice position. Accordingly, unlike the fuel spacer disclosed for example in Japanese Patent Laid-open No. Hei 6-3473, even if the lattice positions associated with the first short-length fuel rods 2b1 are located between the two opposed bath-tubs 16 in the outer peripheral region of the square lattice array, it is possible to omit the cells 9 at the lattice positions, and hence to sufficiently reduce the pressure loss while ensuring the strength of the fuel spacer. (4) Attainment of Degree of Freedom in Design of Spring Arrangement In the fuel spacer 304b in this embodiment, the spring supporting portion provided on the supporting member 317 supports the loop-shaped spring 10 for pressing against the fuel rods 2 to impart a pressing force thereto. Accordingly, it is possible to ensure a degree of freedom in the spring arrangement comparable to that in the fuel spacer in which the cells 9 are located at the first lattice positions 7a associated with the first shortlength fuel rods 2b1. It should be noted that the third embodiment may be variously modified without departing from the basic configuration thereof. Some modifications will be described below. FIG. 10 is a top view showing the structure of a fuel spacer 304bA in which, like the second embodiment, supporting members 318 are provided at lattice positions associated with the second short-length fuel rods 2b2 and further the spring supporting portions are removed from the supporting members 317. As shown in FIG. 10, the supporting member 318 is provided to connect the two cells 9C and 9D, adjacently located on the outer peripheral side of the second lattice position 7b associated with the second short-length fuel rod 2b2 in the square lattice array, to the water rod holding member 12; and it is formed into an approximately cylindrical shape. Also, since supporting members 317A have no spring supporting portions, the spring arrangement in the entire spacer is changed such that the supporting member 318 has two spring supporting portions for suitably supporting the two loop-shaped springs 10 to press the fuel rods 2 inserted in the cells 9C and 9D for imparting pressing forces thereto (like the second embodiment, the known two spring supporting portions are simply provided). This modification is effective to simplify the structure because the supporting members 317 have no spring supporting portions. In the fuel spacer in the modification shown in FIG. 10, the supporting members 317A may be formed into a semi-cylindrical (with a partial peripheral length portion cut off) shape having the same thickness as that of the cell 9. In this case, the supporting member 317A can be manufactured using the raw circular tube for forming the cell 9. This is effective to reduce the manufacturing cost by making the circular tube shareable between the supporting member 317A and the cell 9. From the viewpoint of reduction in pressure loss, it may be desirable that the peripheral length of the cylindrical shape of the supporting member 317A be made as short as possible within a length range required for welding the supporting member 317A to the adjacent cells 9 with no problem. The shape of the supporting members 317A and 318A may be made similar to that of the supporting member 318. FIG. 11 is a top view showing the structure of a fuel spacer 304bB including supporting members 317B and 318A having the configuration described above. As shown in FIG. 11, the supporting members 317B and 318A, each being formed into a cylindrical shape with a partial peripheral length portion cut off, are identical in cross-sectional shape to each other. These supporting members 317B and 318A are substantially similar to each other except that the supporting member 317B has no spring supporting portions for supporting the loop-shaped springs 10 to press against the fuel rods 2 inserted in the cells 9C and 9D. As a result, in the manufacture of the supporting member 318A, the supporting member 317B having the same cross-sectional shape in this modification can be manufactured by punching or bending the same raw material element as that for the supporting member 318A. This is effective to reduce the manufacturing cost by making the raw material shareable between the supporting members 317B and 318A. A fourth embodiment of the present invention will be described with reference to FIG. 12. This embodiment has a feature such that, in the fuel spacer positioned above the upper ends of the first short-length fuel rods 2b1, the flow tabs are partially omitted. FIG. 12 is a top view showing the structure of a fuel spacer 404b in this embodiment. In the fuel spacer 404b, parts common to those in the fuel spacer 4b described in the first embodiment with reference to FIG. 1 are designated by the same symbols and an explanation thereof will be omitted. The fuel spacer 404b shown in FIG. 12 is different from the fuel spacer 4b shown in FIG. 1 in that, of the large number of the flow tabs 15 formed on the band 11 for introducing the flow of a coolant, those located at positions adjacent to the supporting member 17 for connecting the cells 9A and 9B to the band 11 at each first lattice position 7a associated with the first short-length fuel rod 2b1 are omitted. The remaining configuration is substantially the same as that in the first embodiment. This embodiment having the above configuration exhibits the following effects: The flow tabs 15 provided on the band 11 of the fuel spacer 404b have a function of directing the flow of a coolant in the fuel assembly 1 toward the fuel rod 2 side as much as possible, thereby improving the cooling effect of the fuel rods 2 and enhancing the critical power characteristic. The flow tabs 15, however, have an inconvenience in that, since the projecting shapes of the flow tabs 15 obstruct the flow of the coolant, the pressure loss is correspondingly increased. Incidentally, since the fuel spacer 404b is positioned above the upper ends of the short-length fuel rods 2b1, no fuel rods are present at the lattice positions associated with the short-length fuel rods 2b1 and instead the supporting members 17 are provided at the lattice positions. Accordingly, the provision of the flow tabs 15 in the vicinity of the lattice positions does not improve the critical power characteristic, but causes a problem in that it increases the pressure loss. For this reason, the flow tabs 15 adjacent to each supporting member 17 may be omitted. This is effective to eliminate an increase in pressure loss due to the projecting shapes of the flow tabs 15, and hence to further reduce the pressure loss. In the fourth embodiment, the lattice position associated with the first short-length fuel rod 2b1 is located at the intermediate position on each of the four sides of the outermost peripheral region of the square lattice array and the flow tabs 15 adjacent thereto are omitted; however, the present invention is not limited thereto. For example, the lattice position associated with the first short-length fuel rod 2b1 may be located at any position in the outermost peripheral region, preferably, except for four corners of the square lattice array. The reason for this will be described below. The effect of improving the critical power characteristic due to the flow tabs 15 becomes largest at the four corners in the outermost peripheral region of the square lattice array and becomes smaller at other positions in the outermost peripheral region. Accordingly, in the case where the lattice position associated with the first short-length fuel rod 2b1 is positioned in the outermost peripheral region except for the four corners and the supporting member 17 is provided at the lattice position, the effect of improving the critical power characteristic is not reduced so much even if the flow tabs 15 adjacent to the supporting member 17 are omitted; while the effect of reducing the pressure loss due to omission of the flow tabs 15 becomes large irrespective of the arrangement of the flow tabs 15. As a result, it may be desirable to locate the lattice position associated with the first short-length fuel rod 2b1 in the outermost peripheral region except for the four corners and omit the flow tabs 15 adjacent to the supporting member 17 located at the lattice position associated with the first short-length fuel rod 2b1. In the fourth embodiment, only the flow tabs 15 adjacent to the lattice position associated with the first short-length fuel rod 2b1 in the configuration of the first embodiment are omitted; however, the present invention is not limited thereto. For example, the flow tabs 15 at other positions may be omitted insofar as an effect exerted on the critical power characteristic due to omission of the flow tabs 15 is allowable. In this case, it is possible to further reduce the pressure loss. Further, the flow tabs 15 in each configuration of the second and third embodiments may be partially omitted. In each of the first to fourth embodiments, the present invention is applied to the fuel assembly in which the two water rods are arranged in the region in which the seven fuel rods 2 are arrangeable; however, the present invention is not limited thereto, but can be applied to a fuel assembly in which one or three or more water rods are arranged in a region in which six or less or eight or less of the fuel rods 2 are arrangeable. The present invention can be also applied to a fuel assembly including a square type water rod formed to have a square shape in transverse cross-section. FIG. 13 shows a fuel spacer 4bB of a fuel assembly including such a square type water rod, wherein the fuel spacer 4bB is modified from the fuel spacer 4b shown in FIG. 1. The fuel spacer 4bB shown in FIG. 13 is different from the fuel spacer 4b shown in FIG. 1 in that a square type water rod holding member 12A is provided at the central portion of the fuel spacer 4bB and the number of the cells 9 is correspondingly reduced to 68 cells. The remaining configuration is substantially the same as that shown in FIG. 1. The fuel spacer 4bB exhibits an effect similar to that obtained by the fuel spacer 4b. In each of the first to fourth embodiments, the present invention is applied to a fuel assembly including second short-length fuel rods 2b2 in addition to the first short-length fuel rods 2b1; however, the present invention is not limited thereto. For example, the present invention can be applied to a fuel assembly in which fuel rods each having the normal fuel active length are located at the second lattice positions in place of the second short-length fuel rods 2b2 or no fuel rods may be provided at the second lattice positions. In this case, the same effect can be obtained. In each of the first to four embodiments, the first short-length fuel rods 2b1 are dispersedly located in the outermost peripheral region of the square lattice array, that is, two or more of the first short-length fuel rods 2b1 are located so as to be not adjacent to each other in the outermost peripheral region, however, the present invention is not limited thereto. For example, even in the case where two or more of the short-length fuel rods are arranged in the outermost peripheral region as disclosed in Japanese Patent Laid-open No. Hei 6-2373, the cells at the lattice positions associated with the short-length fuel rods can be removed and instead members similar to the supporting members 17, 217 or 317 shown in the first to third embodiments can be provided at the lattice positions. Even in this case, there can be obtained a strength ensuring effect comparable to that obtained in each of the above-described embodiments. A fifth embodiment of the present invention will be described with reference to FIGS. 14 to 20. This embodiment has a feature such that the necessary minimum number of one kind of the loop-shaped springs are reasonably arranged over an entire fuel spacer. In this embodiment, parts common to those in the first embodiment are designated by the same symbols and explanation thereof will be omitted. FIG. 14 is a top view showing the structure of a fuel spacer 504b in this embodiment. As shown in FIG. 14, the fuel spacer 504b in this embodiment is configured such that the cells 9 at the first lattice positions 7a associated with the first short-length fuel rods 2b1 and the second lattice positions 7b associated with the second short-length fuel rods 2b2 are removed, and correspondingly, the arrangement of the loop-shaped springs 10 over the fuel spacer 504b is entirely changed from that in the fuel spacer 4a shown in FIG. 4. That is to say, with respect to the cell 9C adjacent in the same row to one of the two second lattice positions 7b adjacent to the water rods 3 and the cell 9D adjacent in the same column to the above second lattice position 7b, each of the cells 9C and 9D includes, at a portion on the second lattice position 7b side, a known spring supporting portion. A loop-shaped spring 10A (10B) for pressing against the fuel rod 2 inserted in the cell 9C (9D) is provided in the above spring supporting portion provided in the cell 9C (9D). The loop-shaped spring 10A (10B) is, as will be described later, supported by an approximately cylindrical spring pressing member 19 provided at the second lattice position 7b in such a manner as to generate a pressing force applied to the fuel rod 2 inserted in the cell 9C (9D). The spring pressing member 19 is made as thin as possible within an allowable range in terms of the structural strength of the fuel spacer. That is to say, to reduce the pressure loss the transverse cross-section of the spring pressing member 19 is made smaller than the transverse cross-section of the cell 9. In addition, the spring pressing member 19 can be manufactured commonly using the raw circular tube for forming the cell 9. This is effective to reduce the manufacturing cost by making the raw material shareable between the spring pressing member 19 and the cell 9. The water rod holding member 12 formed into the xcexa9-shape in transverse cross-section is joined to each of the two spring pressing members 19. The detailed structure near the two second lattice positions 7b, which forms the largest difference between the fuel spacer 504b and the fuel spacer 4a, will be described below. (1) Fuel Spacer 4a FIG. 15 is a transverse sectional view showing the structure near the two second lattice positions 7b of the fuel spacer 4a; and FIG. 16 is a perspective view showing the structure near the joined portions between the water rod holding member 12 and the cells 9. In addition, for convenience in description, the fuel rods 2 and the water rods 3 are additionally shown in FIG. 15. Referring to FIGS. 15 and 16, the water rod holding member 12 includes two spring holding projecting pieces 12a each of which projects in the shape of tongue in the loop of the loop-shaped spring 10 for holding the loop-shaped spring 10, and a spring pressing projecting piece 12b which projects in the shape of tongue in such a manner as to be brought in contact with the loop of the loop-shaped spring 10 from the outer peripheral side. The two spring holding projecting pieces 12a and the spring pressing projecting piece 12b all project in the same direction (leftward in FIG. 16), and the two spring holding projecting pieces 12a are provided above and below the spring pressing projecting piece 12b, respectively. These spring holding projecting pieces 12a and the spring pressing projecting piece 12b are manufactured by press-working a base portion of the water rod holding member 12 to form tongue-shaped cut pieces corresponding to the projecting pieces 12a and 12b and two windows 12c. The spring pressing projecting piece 12b is finished by bending the corresponding tongue-shaped cut piece from its root and then flattening it. The windows 12c are provided to provide spaces in which the loop-shaped spring 10 is inserted when the loop-shaped spring 10 is mounted. The cell 9 to be joined to the water rod holding member 12 has two windows 9b (only one is shown for simplicity) and a projecting piece 9c which projects in the direction opposed to the projecting direction of the projecting pieces 12a. The two spring holding projecting pieces 12a are in contact with the projecting piece 9c. The contact portions of the spring holding projecting pieces 12a with the cell projecting piece 9c are inserted in the loop of the loop-shaped spring 10, so that the vertical movement of the looped spring 10 may be restricted. In addition, both ends of the windows 12c and 9b form removal preventive portions 12c1 and 9b1 (only partially shown) for preventing the removal of the loop-shaped spring 10. The spring pressing projecting piece 12b functions to restrict the horizontal displacement of the loop-shaped spring 10 due to expansion/contraction thereof, and hence to generate a pressing force applied to the fuel rod 2 inserted in the cell 9. At this time, as shown in FIG. 15, the distance d1 between the spring pressing projecting piece 12b and the adjacent fuel rod 2 is made equal to the distance d2 between the two adjacent fuel rods 2. With this configuration, a pressing force of the loop-shaped spring 10 generated when the spring 10 is pressed by the spring pressing projecting piece 12b is made equal to a spring pressing force of the loop-shaped spring 10 generated when the spring 10 is held between the two fuel rods 2. As a result, the fuel rod 2 can be suitably fixed in the cell 9 at the lattice position 7b. (11) Fuel Spacer 504b FIG. 17 is a perspective view, with parts partially cutaway, showing the structure near the joined portions between the spring pressing member 19 and the cells 9 in the fuel spacer 504b, which is an essential portion of this embodiment. Referring to FIG. 17, the spring pressing member 19 includes two spring holding projecting pieces 19a and two spring holding projecting pieces 19b, which function as spring holding portions projecting in the shape of a tongue inserted in the loops of the loop-shaped springs 10A and 10B for holding the loop-shaped springs 10A and 10B, respectively; and two spring pressing projecting pieces 19C and 19D which function as spring pressing portions projecting in the shape of a tongue in such a manner as to be in contact with the loops of the loop-shaped springs 10A and 10B from the outer peripheral side, respectively. The structures and functions of these projecting pieces 19a (19b) and 19c (19d) are similar to those of the projecting pieces 12a and 12b of the water rod holding member 12 described in the paragraph (1), respectively. To be more specific, the two spring holding projecting pieces 19a are provided above and below the spring pressing projecting piece 19c, respectively; and the two spring holding projecting pieces 19b are provided above and below the spring pressing projecting piece 19d, respectively. These spring holding projecting pieces 19a and 19b and the spring pressing projecting pieces 19c and 19d are manufactured by press-working a cylindrical base portion 19g of the spring pressing member 19 to form tongue-shaped cut pieces corresponding to the projecting pieces 19a, 19b and 19c and 19d and four windows 19e and 19f. Each of the spring pressing projecting pieces 19c and 19d is finished by bending the corresponding tongue-shaped cut piece from its root and then flattening it. At this time, the two spring holding pieces 19a (19b) are in contact with a projecting piece 9Cc (9Dc) formed in the cell 9C (9D), and the contact portions of the spring holding projecting pieces 19a (19b),with the cell projecting piece 9Cc (9Dc) are inserted in the loop-shaped spring 10A (10B). In addition, to prevent the removal of the loop-shaped spring 10A (10B), the ends of windows 19e (19f) of the loop-shaped spring 10A (10B) have removal preventive portions 19e1 (19f1) and the ends of the windows of the cell 9C (9D) have removal preventive portions (not shown for simplicity). The spring pressing projecting piece 19c (19d) functions to restrict the horizontal displacement of the loop-shaped spring 10A (10B) due to expansion/contraction thereof, and hence to generate a suitable pressing force (similar to the pressing force described in the paragraph (1)) to the fuel rod 2 inserted in the cell 9C (9D). Here, there is a large structural difference between the projecting pieces 19a, 19b, 19c and 19d and the projecting pieces 12a and 12b of the water rod holding member 12 described in the paragraph (1) in that the spring holding projecting pieces 19a and 19b all project in the same direction (leftward in FIG. 17) and one spring pressing projecting piece 19d also projects in the same direction; however, the other spring pressing projecting piece 19c projects in the opposite direction (rightward in FIG. 17). The function of this embodiment having the above configuration will be described below. (1) Improvement in Reactivity Controllability In the fuel assembly 1 in this embodiment, since six of the short-length fuel rods 2b are included with the fuel rods 2 arranged in the square lattice array of 9 rowsxc3x979 columns, it is possible to equalize the H/U ratio by making use of a saturated water region formed on the upper side of the short-length fuel rods 2b. At this time, by arranging the short-length fuel rods 2b at the lattice positions in the outermost peripheral region of the square lattice array and at the lattice positions adjacent to the water rods, it is possible to more effectively improve the controllability of the reactivity by reducing the void coefficient as disclosed in Japanese Patent Laid-open No. Hei 5-232273. (2) Reduction in Pressure Loss Since the unnecessary cells 9 at the first and second lattice positions 7a and 7b in the fuel spacer 4b positioned above the upper ends of the short-length fuel rods 2b are omitted, the pressure loss can be correspondingly reduced. In addition, a spring pressing member 19 is provided in place of the cell 9 at each second lattice position 7b; however, since the transverse cross-section of the spring pressing member 19 is made smaller than that of the cell 9 as described above, it is possible to reduce the pressure loss. 1 (3) Reasonable Arrangement of Spring As a result of removing the cells 9 in the fuel spacer 504b for reducing the pressure loss (see the paragraph (2)), the arrangement of the loop-shaped springs 10 over the entire fuel spacer is necessarily changed such that the loop-shaped springs 10A and 10B are respectively provided on portions, on the second lattice position 7b side, of the cells 9C and 9D located at the lattice positions adjacent to each of the two second lattice positions 7b for pressing against the fuel rods 2 in the cells 9C and 9D. The usual loop-shaped spring 10 functions to generate pressing forces when the fuel rods 2 are inserted in a pair of the adjacent cells 9, and accordingly, if such usual loop-shaped springs are used for the loop-shaped springs 10A and 10B, the loop-shaped springs 10A and 10B are made free on the second lattice position 7b side and thereby cannot generate the pressing forces. Incidentally, the structure in which one loop-shaped spring which is free on one side is supported at one lattice position in such a manner as to generate a suitable pressing force has been known, for example, as represented by the structure shown in FIG. 16 or the structure disclosed in Japanese Patent Laid-open No. Hei 6-273560; however, the above-described structure in which the two loop-shaped springs 10A and 10B each being free on one side are supported at one lattice position has not been known. On the contrary, in the fuel spacer 504b in this embodiment, since the spring pressing member 19, which includes the four spring holding projecting pieces 19a and 19b and the two spring pressing projecting pieces 19c and 19d, is provided at one of two second lattice positions 7b, it is possible to support the loop-shaped springs 10A and 10B so that each is free on one side and hence to generate suitable pressing forces. This makes it possible to reasonably arrange the necessary minimum number (36 pieces) of the loop-shaped springs 10 over the entire fuel spacer without increasing the kinds of springs being used, more specifically, using only one kind of the springs. (4) Attainment of Rigidity/Strength of Spring Pressing Member to Pressing Force of Spring This function will be described with reference to a comparative example. FIG. 18 is a perspective view, with parts partially cutaway, showing the structure near the joined portions between a spring pressing member and the cells in the comparative example. In FIG. 18, parts common to those in FIG. 17 are designated by the same symbols. In the comparative example shown in FIG. 18, two of the structures shown in FIG. 16, each being similar to that disclosed in Japanese Patent Laid-open No. Hei 6-273560, are simply arranged for supporting the two loop-shaped springs 10A and 10B so that each is free on one side at one second lattice position 7b. The structure shown in FIG. 18 is different from that shown in FIG. 17 in that the spring holding projecting pieces 19a and 19b all project in the same direction (leftward in FIG. 18) and the two spring pressing projecting pieces 19c and 19d project in the same direction (leftward in FIG. 18). Such a structure as shown in FIG. 18 has the following inconvenience. That is to say, to bring the spring pressing projecting pieces 19c and 19d in contact with the loops of the loop-shaped springs 10A and 10B from the outer peripheral side, the spring pressing projecting pieces 19c and 19d are required to project on the inner side (toward the front in FIG. 18) of the spring pressing member 19 more than the spring holding projecting pieces 19a and 19b inserted in the loops. Accordingly, if the projecting direction of all of the spring pressing projecting pieces 19c and 19d is made identical to the projecting direction of the spring holding projecting pieces 19a and 19b, the size of the cutouts on both sides of the spring pressing projecting pieces 19c and 19d must be made larger, and correspondingly the width (or area) of a bridge 19g1, equivalent to the root portion of-the spring pressing projecting piece 19c, of the base plate portion 19g of the spring pressing member 19 becomes smaller. This makes it difficult to ensure a sufficient strength and rigidity against the pressing force of the loop-shaped spring 10A. On the contrary, according to the configuration of this embodiment as shown in FIG. 17, since the projecting direction of one spring pressing projecting piece 19c is reversed relative to the projecting direction of the spring holding projecting pieces 19a and 19b, the width (or area) of the bridge 19g1 can be made larger. This makes it possible to ensure a sufficient strength and rigidity. As described above, according to this embodiment, the fuel assembly 1 is configured such that the shortlength fuel rods 2b are arranged in the outermost peripheral region of the square lattice array of 9 rowsxc3x979 columns adjacent to the water rods 3, and at the fuel spacer 4b, the cells 9 at the lattice positions 7a and 7b associated with the short-length fuel rods 2b are removed to reduce the pressure loss. In this fuel assembly 1, the two loop-shaped springs 10A and 10B, each being free on one side, are supported by the spring pressing member 19 at one second lattice position 7b, so that the necessary minimum number of the loop-shaped springs 10 may be reasonably arranged over the entire fuel spacer without increasing kinds of the springs being used. Since the number of the loop-shaped springs 10 is selected at the necessary minimum value, there can be obtained an effect of further reducing the pressure loss, and since the width (or area) of the bridge 19g1 can be made larger, there can be obtained an effect of ensuring a sufficient strength and rigidity against the pressing forces of the loop-shaped springs 10A and 10B. In the fifth embodiment, the spring holding projecting pieces 19a and 19b all project in the same direction (leftward in FIG. 17) and one spring pressing projecting piece 19d also projects in the same direction; while the other spring pressing projecting piece 19c projects in the opposite direction (rightward in FIG. 17); however, the structure of the projecting pieces is not limited thereto. Hereinafter, two modifications will be described with reference to FIGS. 19 and 20. FIG. 19 is a perspective view, with parts partially cutaway, showing the structure near the joined portions between the spring pressing member and the cells according to the first modification. In FIG. 19, parts common to those in FIG. 17 are designated by the same symbols. The structure shown in FIG. 19 is different from that shown in FIG. 17 in that the spring holding projecting pieces 19a and 19b all project in the same direction (leftward in FIG. 19), while both spring pressing projecting pieces 19c and 19d project in the opposite direction (rightward in FIG. 19). With this structure, since the width (or area) of the bridge 19g1 can be made larger than that in the comparative example shown in FIG. 18, it is possible to ensure a sufficient strength and rigidity against the pressing forces of the loop-shaped springs 10A and 10B. That is to say, it becomes apparent that at least one of the spring pressing projecting pieces may project in the direction opposite to the projecting direction of the spring holding projecting pieces. FIG. 20 is a perspective view, with parts partially cutaway, showing the structure near the joined portions between the spring pressing member and the cells according to the second modification. In FIG. 20, parts common to those in FIG. 17 are designated by the same symbols. The structure shown in FIG. 20 is different from that shown in FIG. 17 in that the opposed free ends of the spring pressing projecting pieces 19c and 19d in the structure shown in FIG. 17 are connected to the base plate portion 19g of the spring pressing member 19, that is, as shown in FIG. 19, both right and left ends of each of the projecting pieces 19c and 19d are connected to the base plate portion 19g so as to be integrally formed therewith. In the second modification, when the pressing forces of the loop-shaped springs 10A and 10B are applied to the spring pressing projecting pieces 19c and 19d, they can be supported by the base plate portion 19g connected to both sides of each of the spring pressing projecting pieces 19c and 19d. This is effective to ensure a sufficient strength and rigidity. The second modification has another effect. In the structures shown in FIGS. 17 and 19, as described above, the spring pressing projecting pieces 19c and 19d are manufactured by forming tongue-shaped cut pieces in the base plate portion 19g of the spring pressing member 19, bending the cut pieces from the roots thereof and flattening them. In contrast, in this modification, the spring pressing projecting pieces 19c and 19d can be simply manufactured merely by forming cut lines corresponding to both side lines of the projecting pieces 19c and 19d in the base plate portion 19g of the spring pressing member 19 and pressing the portions surrounded by the cut lines such that the portions project inward of the spring pressing member 19. This is effective to reduce the manufacturing cost. While the spring pressing member 19 is formed into an approximately cylindrical shape in the fifth embodiment, the present invention is not limited thereto. For example, like the supporting member 218 in the second embodiment, the spring pressing member 19 may be formed into an approximately polygonal shape from which a partial side portion is removed for reducing the pressure loss. As shown in FIG. 21, flow tabs 17 having the same function as that of the flow tabs 15 provided on the band 11 may be provided on the spring pressing member 19. The flow tabs 17 project outward from the outer periphery of the spring pressing member 19 for introducing the flow of a coolant in the projecting direction of the flow tab 17. The structure including the flow tabs 17 exhibits the following effect. The spring pressing member 19 at the second lattice position 7b, which is at the level in which no fuel rod 2 is present is not required to be cooled. Accordingly, to direct the flow of a coolant passing through the spring pressing member 19 toward the other fuel rods 2 around the lattice position 7b as much as possible, the flow tabs 17 are provided on the spring pressing member 19. This makes effective use of the coolant and hence improves the effect of cooling the fuel rods 2. A sixth embodiment of the present invention will be described with reference to FIG. 22. In this embodiment, a spring pressing member having a shape different from that in the fuel spacer 504b of the fifth embodiment is provided. Parts common to those in FIG. 1 are designated by the same symbols and an explanation thereof will be omitted. FIG. 22 is a top view showing the structure of a fuel spacer 604b in this embodiment. In addition, the water rods 3 are also shown in the figure for more clearly showing the structure of the fuel spacer 604b. Like the fuel spacer 504b, the fuel spacer 604b shown in FIG. 22 is positioned above the upper ends of the short-length fuel rods 2b in the fuel assembly 1 shown in FIG. 2 or FIG. 3. In the fuel spacer 604b, the spring pressing member 19 of the fuel spacer 504b shown in FIG. 14 is replaced with a spring pressing member 619, formed into an umbrella shape in transverse cross-section, which has a function of the spring pressing member 19 in combination with the function of the water rod holding member 12. The structure of the spring pressing member 619 will be briefly described. Like the spring pressing member 19 shown in FIG. 14, the spring pressing member 619 includes spring holding projecting pieces (not shown) which are inserted in the loops of the loop-shaped springs 10A and 10B for holding the loop-shaped springs 10A and 10B; and spring pressing projecting pieces 619a and 619b which are brought in contact with the loops of the loop-shaped springs 10A and 10B from the outer peripheral side. The spring pressing member 619 also includes a water rod holding portion 619c which functions to hold the water rod 3 in the radial and axial directions, like, the water rod holding member 12 shown in FIG. 14. The remaining configuration is substantially the same as that of the fuel spacer 504b in the fifth embodiment. The fuel assembly including the fuel spacer 604b in this embodiment exhibits an effect comparable to that of the fuel assembly including the fuel spacer 504b in the fifth embodiment. Additionally, according to this embodiment, since the spring pressing member 19 and the water rod holding member 12 are replaced with the spring pressing member 619, it is possible to reduce the number of parts, and hence to lower the manufacturing cost; and also it is possible to reduce the transverse cross section of the portion adjacent to the water rods 3, and hence to further reduce the pressure loss. In the sixth embodiment, the structures of the spring pressing projecting pieces 619a and 619b and the spring pressing projecting pieces are not limited to those shown in FIG. 17, but may be similar to those shown in FIGS. 19 or 20. In the fifth and sixth embodiments, description is made by way of example of a fuel assembly in which the fuel rods 2 are located in a square lattice array of 9 rowsxc3x979 columns; however, the present invention can be applied to a fuel assembly having a square lattice array of 8 rowsxc3x978 columns or 10 rowsxc3x9710 columns, or a rectangular lattice array in which the rows and columns are different in number. Even in such a fuel assembly, if two of the loop-shaped springs are required to be supported at one lattice position, the concept of the present invention can be applied thereto, with a result that the same effect can be obtained. In a fuel assembly having a square lattice array of 9 rowsxc3x979 columns in which the arrangement of the loop-shaped springs 10 is different from that shown in FIG. 14 or FIG. 22, if two of the loop-shaped springs are required to be supported at one lattice position, the concept of the present invention can be applied thereto, with a result that the same effect can be obtained. Even in the case where three or more of the loop-shaped springs 10 are required to be supported at one lattice position, the concept of the present invention can be applied thereto. Such a modification will be described with reference to FIG. 23. FIG. 23 is a top view showing the structure of an essential portion of a fuel spacer 704b in this modification. Like the fuel spacers 504b and 604b, the fuel spacer 704b shown in FIG. 23 is positioned above the upper ends of the short-length fuel rods 2 in the fuel assembly 1 shown in FIG. 2 or 3. In the case of reviewing the arrangement of the loop-shaped springs 10 in accordance with removal of the cells 9 to reduce the pressure loss, there may occur a requirement in which four free loop-shaped springs 710A to 710D (each having the same structure as that of the loop-shaped spring 10) need to be supported at one lattice position, other than those in the outermost peripheral region of the square lattice array and those adjacent to the water rods 3. To meet such a requirement, the fuel spacer 704b is provided with a cylindrical spring pressing member 719 for supporting the springs 710A to 710D in such a manner that the springs can generate suitable pressing forces. The structure of the spring pressing member 719 will be briefly described. Like the spring pressing member 19 shown in FIG. 19, the spring pressing member 719 includes four spring holding projecting pieces (not shown) which are inserted in the loops of the four loop-shaped springs 710A to 710D surrounding the spring pressing member 719 for holding the loop-shaped springs 710A to 710D; and four spring pressing projecting pieces 719a to 719d which are brought in contact with the loops of the loop-shaped springs 710A to 710D from the outer peripheral side. The spring pressing member 719 is also made as thin as possible within an allowable thickness range in terms of the structural strength of the fuel spacer for making the transverse cross-section thereof smaller than that of the cell 9, thereby reducing the pressure loss. The remaining configuration is substantially the same as that of the fuel spacer 504b in the fifth embodiment. Like the fifth embodiment, the fuel assembly including the fuel spacer 704b in this embodiment exhibits an effect of reasonably arranging the necessary minimum number of the loop-shaped springs 710 over the fuel spacer without increasing the number of kinds of springs by supporting the four free loop-shaped springs 710A to 710D at one lattice position by means of the spring pressing member 719. In the above modification, description is made by way of. example of the case of providing four free loop-shaped springs 710A to 710D; however, the present invention is not limited thereto, but may be applied to the case of providing only three of the free loop-shaped springs. In this case, by providing three sets of the spring pressing projecting pieces 719a to 719c and the spring holding projecting pieces, an effect similar to that described above can be obtained. Since the variations in allowable arrangement of the loop-shaped springs can be further increased by making the spring pressing member 719 in the above modification in combination with the spring pressing members 19 and 619 in the fifth and sixth embodiments, this arrangement is expected to provide a more effective fuel spacer from the viewpoint of reduction in pressure loss. In each of the first to sixth embodiments, description is made by way of example of a fuel assembly in which the fuel rods 2 are located in a square lattice array of 9 rowsxc3x979 columns; however, the present invention is not limited thereto, but may be applied to another fuel assembly having a square lattice array of 8 rowsxc3x978 columns, 10 rowsxc3x9710 columns, or the like. In this case, an effect similar to that described above can be obtained. While the preferred embodiments of the present invention have been described using specific examples, such description is for illustrative purposes only, and it is to be understood that changes and variations may be made without departing from the spirit or scope of the following claims.
	claims	1. A fuel rod, comprising:a hollow cylindrical rod having an upper end having a water outlet and a lower end having a water inlet;an upper end plug connected to said upper end and defining an upper outlet channel disposed in fluid communication with said water outlet;a lower end plug connected to said lower end and defining a main inlet channel disposed in fluid communication with said water inlet; anda debris filter disposed in said main inlet channel, said debris filter comprises a plurality of intersecting pins extending across said main inlet channel, said plurality of intersecting pins are fitted within a plurality of radially disposed through-holes formed in said lower end plug, said upper end plug is provided with an upper inclined face on an exterior surface thereof, said upper end plug includes an upper handling groove circumferentially formed in an interior surface thereof to receive a fuel rod handling tool, said interior surface of the upper end plug defines the upper outlet channel, said lower end plug comprises an upper cylindrical portion above said debris filter and a lower cylindrical portion below said debris filter, said upper cylindrical portion includes a plurality of inner channel auxiliary inlets inclined at an upward angle with respect to a plane of the debris filter, wherein said plurality of inner channel auxiliary inlets form a cooling water fluid by-pass around said debris filter in the event the debris filter becomes blocked by debris. 2. The fuel rod as set forth in claim 1, wherein the pins of the debris filter have a circular cross section, intersect each other in a crisscross shape in the middle of each thereof. 3. The fuel rod as set forth in claim 1, wherein the pins of the debris filter have a circular cross section, intersect with each other as a grid shape in the middle of each thereof. 4. The fuel rod as set forth in claim 1, wherein the debris filter includes two pins with a circular cross section, one of which is disposed in a horizontal direction, and the other of which is disposed in a horizontal direction and bent downward to form a “V” shape, the two pins intersecting each other as a crisscross shape. 5. The fuel rod as set forth in claim 1, wherein the inner channel auxiliary inlets are radially formed in the lower end plug at predetermined intervals, and are vertically arranged in at least one row.
054065968	description	DESCRIPTION OF THE PREFERRED EMBODIMENT FIG. 1 shows part of a vessel head 1 of a pressurized-water nuclear reactor traversed by an opening 2 in which there is fixed, in a sealed manner by welding, an adapter 3 including a part projecting below the vessel head providing guidance for a thermocouple column 5, and a part projecting above the vessel head 1 constituting a flared part 4 with a threaded external surface 4a. On the top part 4 of the adapter 3 there is fixed, in the axial extension of the adapter 3, an assembly 6 for fixing and sealing the thermocouple column 5, by means of a bottom part 7 including a tapped bore which is engaged over the threaded part 4a of the flared part 4 of the adapter 3. The fixing and sealing assembly for the thermocouple column 5 is thus assembled at the end of the extension 3. The flared part 4 of the extension 3 and the bottom part 7 of the fixing and sealing assembly 6 includes circular seals 4', 7' which coincide when the part 7 is entirely screwed onto the flared part 4. The seals 4' and 7' are connected by welding so as to ensure sealing of the screwed connection between the components 4 and 7. The head and the port adapters 3 are produced at the factory and transported to the site where the nuclear reactor is installed. The fixing and sealing devices 6 for the thermocouple columns 5 are attached and fixed to the top ends of the adapters 3 on the site of the reactor. The bottom part 7 of the fixing and sealing assembly 6 is fixed onto the end of the extension so as to be able to be dismantled if need be by melting the junction zone of the seals 4' and 7'. This dismantling is only carried out for repairs or exceptional intervention on the thermocouple column port. The fixing and sealing assembly 6 includes a top part 8 which is assembled in a sealed manner to the bottom part 7 with the interposition of a metal seal 10 of a special shape, the parts 7 and 8 of the fixing and sealing assembly being assembled by means of a clamping flange 12, having two parts which can be joined together and clamped by means of screws inserted through openings 13 passing through opposing lugs located at the end of the two sector-shaped parts. The clamping flange includes frustoconical bearing surfaces on its inside, these surfaces coming into clamping contact with corresponding frustoconical bearing surfaces machined on end flared parts 7a and 8a respectively of the bottom part 7 and of the top part 8 of the fixing and sealing assembly. The adapter 3 and the fixing and sealing assembly 6 are tubular and are located in the axial extension of one another, so as to provide a passage for the thermocouple column 5, constituted by a tube for supporting and holding a set of thermocouples 15. The thermocouple column includes a support and sealing component 16 at its top end, inside the support assembly, this component 16 comprising a frustoconical sealing face 16a which interacts with a corresponding frustoconical shoulder inside the bore of the top part 8 of the support assembly. A metal sealing gasket is interposed between the two coinciding frustoconical bearing surfaces. Above the solid component 16, the thermocouple column includes a part which projects from the end of the top part 8 of the support assembly, in which is machined a groove allowing the engagement of a pull-ring 18 made up of two half-rings which can be engaged laterally in the annular groove. A pressure plate 19 includes lifting screws 20 engaged axially through the plate 19, and a lip engaging under the pull-ring 18. The end of the lifting screws 20 comes to bear directly on the top surface of the part 8 of the support assembly of the fixing and sealing device. By screwing the lifting screws 20, the pressure plate 19 is raised, and comes to bear on the pull-ring 18 of the thermocouple column 5. The thermocouple column is thus raised inside the bore of the support device 7, 8, then the sealing gasket associated with the frustoconical bearing surface 16a of the thermocouple column is made to bear on the corresponding shoulder machined in the bore of the top part 18 of the support assembly. Sealed fixing of the thermocouple column inside its support assembly is thus obtained, the pressure plate 19, the lifting screws 20 and the pull-ring 18 constituting the device for sealed clamping of the thermocouple column. The sealed clamping of the thermocouple column 5 is produced with the aid of a hand tool such as a torque wrench with which the screws 20 passing through the pressure plate 19 are successively tightened. An even number of lifting screws 20 is used, placed in positions which are diametrically opposed in pairs around the pressure plate 19, which is generally of circular shape. In general, six lifting screws located at 60.degree. from one another about the axis of the central opening of the plate 19 are used. The use of an even number of screws makes it difficult to balance the tightening of the various screws. Furthermore, the tightening of one screw to a certain torque interferes with the tightening of the adjacent screws, as indicated earlier. This results in it being very difficult to obtain effective and constant tightening of the lifting screws of the pressure plate. Furthermore, manual tightening and balancing of the six lifting screws of the pressure plate by an operator requires an operating time, per thermocouple column, which may be of the order of 10 to 15 minutes. When these operations are carried out within the scope of a maintenance and refuelling run of a nuclear reactor, which is the most frequent case, the operator is exposed to a dose of radiation coming from the adapter and from the bottom surface of the head, during the entire time necessary for sealed clamping of the thermocouple columns. Furthermore, bottom ends of the lifting screws 20 come to bear directly on the top part of the support assembly 8, which does not, strictly speaking, constitute an element of the clamping device. This top part of the support assembly 8 may be marked under the effect of the contact pressure of the screws, and may exhibit a shape or behavior which does not lend itself to producing effecting clamping. FIG. 2 shows the top end of the support assembly of a thermocouple column, and the part of a clamping device according to the invention remaining in place on the support assembly, after sealed clamping of the thermocouple column. The support assembly 21 and the thermocouple column 22 have a shape identical to the one which has been described with reference to FIG. 1 relating to a clamping device according to the prior art. The tubular support assembly 21 through which the thermocouple column 22 passes includes a frustoconical support surface 23, the thermocouple column 22 itself including a frustoconical bearing surface 24 in a corresponding position. In the sealed clamped position of the thermocouple column represented in FIG. 2, a frustoconical seal 25 is crushed between the support surfaces 23 and 24. The part of the clamping device according to the invention remaining in position on the support assembly 21 in the clamped position of the thermocouple column 22 includes a supporting endpiece 26 traversed by a two-part central opening 27, the bottom part 27a of the central opening 27a having a large diameter engaging practically without clearance on the top part of the support assembly 21. The top part 27b of the central opening, of smaller diameter, has a diameter which is greater than the usual diameter of the top part projecting above the support assembly 21 of the thermocouple column 22. The remaining part of the clamping device additionally includes a pressure plate 28 which can be seen in FIGS. 2 and 3, and which is traversed by a circular central opening 29 whose diameter is substantially equal to the diameter of the projecting part of the thermocouple column on which the plate 28 may be engaged, practically without clearance. As can be seen in FIG. 3, the pressure plate 28 has the shape of an equilateral triangle whose vertices are truncated and rounded. The endpiece 26 has a shape and dimensions which are identical to the shapes and dimensions of the pressure plate 28. The pressure plate 28 is traversed by three tapped circular holes 30a, 30b and 30c, in each one of which a lifting screw 31 is engaged. Each of the lifting screws 31 includes a first end 31a coming to bear on the top surface of the endpiece 26 and a second end 31b including a part which is shaped to receive a tightening tool. As can be seen in FIG. 2, the part of the device for clamping the thermocouple column, which remains in place, also includes a pull-ring 33 made up of two half-rings which can be engaged laterally in a groove machined in the thermocouple column, so as to come to bear on a projecting annular part 22a of the thermocouple column so as to provide upward thrust on the thermocouple column. The two half-rings constituting the pull-ring 33 have an external diameter which is greater than the diameter of the hole for passing through the pressure plate 28 which is substantially equal to the usual diameter of the thermocouple column, and an internal diameter which is less than the usual diameter of the thermocouple column 22. As can be seen in FIGS. 2 and 3, the pressure plate and the thermocouple column are lifted by three screws 31, each engaged in a tapped opening 30a, 30b or 30c of the pressure plate 28, this opening being located near a truncated corner of the equilateral triangle, and ends 31a of the bearings screws on the supporting end-piece 26. The axes of tapped holes 30a, 30b, 30c and the corresponding screws 31 are parallel to the axis of the central opening of the turning plate and of the thermocouple column 22, when the turning plate is in the clamped position, and are located equidistant and at 120.degree. from each other with respect to the axis of the opening 29 of the pressure plate which coincides with the axis of the thermocouple column 22. Perfectly balanced clamping is thus obtained, and the bearing of the screws on the end-piece 26, which is installed before clamping the thermocouple column, makes it possible to obtain a perfectly matched support surface. This support surface includes recesses 34 which may be constituted by frustoconical cavities in which the end parts 31a of the screws 31 engage, which end parts may have the shape of a spherical cap, as can be seen particularly in FIG. 5. Reference will now be made to FIGS. 4 to 6 for describing the removable part of the clamping device used for maneuvering the lifting screws, and the installation of the clamping elements remaining permanently on the support assembly of the adapter. As can be seen in FIGS. 4, to 6, the clamping device according to the invention, in addition to the pressure plate 28, the end-piece 26 and the pullring 33, includes a removable clamping assembly 35 and a manual tightening tool 40 for tightening the lifting screws 31 of the pressure plate 28. The installation of the various elements of the device according to the invention is performed in the following manner. The thermocouple column 22 is placed in the bottom position, represented in FIG. 5, inside the support assembly 21, with the support surfaces 23 and 24 being separated from one another by a certain vertical distance, and the sealing gasket 25 bearing on the shoulder 23 of the thermocouple column 22. The supporting endpiece 26 is engaged over the top part of the thermocouple column 22, then over the top part of the support assembly 21, on which it rests in a stable manner. The pressure plate 28 is engaged over the thermocouple column 22 via its top part, so as to come to rest via its bottom face on the top support face of the endpiece 26. The plate 28 is placed in an orientation which corresponds to the orientation of the endpiece 26, these two elements having the shape of equilateral triangles. The lifting screws 31 of the pressure plate 28 are placed in a top position, as represented in FIG. 5, in which the spherical-cap-shaped supporting bottom end 31a of the screws 31 projects slightly from the bottom face of the plate 28. When the plate 28 is installed on the endpiece 26, the bottom supporting parts 31a of the screws 31 become engaged in the frustoconical openings 34 of the support surface of the endpiece 26. A protection panel 36, on which three ferrules 37 are fixed, covers the top surface of the pressure plate 28, the screws 31 being engaged inside the protection ferrules 37. The panel 36, which has a contour identical to that of the plate 28, has a central opening whose dimensions are substantially greater than those of the central opening of the plate 28. The pull-ring 33 made in two parts 33a, 33b, is installed laterally inside the groove of the thermocouple column 22, below the projecting annular part 22a. The clamping assembly 35 is then engaged over the top part of the thermocouple column so as to engage with the screws 31 and tighten them by actuation of the tool 40. The clamping assembly 35 includes a mounting plate 38 made up of three superposed parts 38a, 38b and 38c of different thicknesses. The three panels constituting the mounting plate 38 have a U-shaped cross-section as represented in FIG. 4, which shows the top panel 38a. On the two opposite outer faces of the panel 38a, along the two branches of the U, there are fixed two handles 39a and 39b allowing handling of the clamping assembly 35, for engaging it over the thermocouple column and installing it above the pressure plate 28. The handles 39a and 39b may be fixed onto the top surface of the panel 38a of the mounting plate 38, in the event of operating in a restricted space requiring reduced lateral bulk of the clamping assembly 35. The clamping assembly 35 includes three actuation spindles 42a, 42b and 42c which are mounted rotationally in holes passing through the mounting plate 38 in a direction which is perpendicular to the faces of the superposed panels 38a, 38b, 38c. As can be seen in FIG. 5, the spindle 42a is mounted rotationally by means of two rolling-contact bearings 43a and 43b fixed respectively to the inside of the panel 38a and panels 38b, respectively, of the mounting plate 38. Between the rolling-contact bearings 43a and 43b, the spindle 42a is securely fastened to a gear 49a. The spindle 42a includes a first shaped part 45 of hexagonal cross-section projecting above the top panel 38a of the mounting plate 38, and a second shaped part 46a of hexagonal cross-section projecting below the bottom face of the bottom panel 38c of the mounting plate 38, inside an engagement and protection ferrule 47a. The spindles 42b and 42c are identical, so that the description of the spindle 42b, which will be given with reference to FIG. 6, applies equally to spindle 42c. The spindle 42b is mounted rotationally inside the mounting plate 38, in an opening whose direction is perpendicular to the faces of the panels making up the mounting plate 38. The spindle 42b is mounted rotationally by means of two rolling-contact bearings 48a and 48b mounted respectively inside the panel 38a and inside the panel 38b of the mounting plate 38. Between the rolling-contact bearing 48a and 48b, the spindle 42b is securely fastened to a gear 49b. The spindle 42c is securely fastened to a corresponding gear 49c referenced in FIG. 4. The spindle 42b includes a shaped part 46b of hexagonal cross-section projecting below the bottom face of the panel 38c of the mounting plate 38, inside an engagement and protection ferrule 47b. The shaped parts 46a of the spindle 42a and 46b of the spindle 42b, as well as the corresponding shaped end part of the spindle 42c, have a shape and dimensions allowing them to be engaged inside a six-head socket-shaped part inside the screw heads 31b, the engagement of the shaped parts such as 46a and 46b inside the screw heads 31b taking place when the clamping assembly 35 is installed above the pressure plate 28. The engagement ferrules such as 47a and 47b make it possible to facilitate the installation of the clamping assembly over the lifting screws of the pressure plate. Six mutually parallel shafts 50a, 50b, 50c, 50d, 50e, 50f are also mounted in openings inside the mounting plate 38, the parallel shafts 50a, . . . 50f being parallel to the spindles 42a, 42b, 42c. The shafts 50a, . . . 50f are all mounted in the same way inside the mounting plate 38 so that only the shaft 50a will be described, with reference to FIG. 6. The shaft 50a is mounted rotationally inside the mounting plate 38, by means of rolling-contact bearings 51a mounted inside the panel 38a and 51b mounted inside the panel 38b. Between the rolling-contact bearings 51a and 51b the shaft 50a carries a gear 52a. As can be seen in FIG. 4, the shafts 50a, 50b, 50c, 50d, 50e, 50f are respectively securely fastened to gears 52a, 52b, 52c, 52d, 52e, 52f which are all identical in diameter and diametral pitch. The gears 52a, . . . 52f are also identical, in diameter and diametric pitch, to the three gears 49a, 49b, 49c which are respectively securely fastened to the spindles 42a, 42b, 42c. As can be seen in FIG. 4, the gears 49a, 49b, 49c and 52a, . . . 52f mesh with one another in a sequential arrangement, the gears 52a, 52b, 52c which mesh with one another in this order being interposed between the gears 49c and 49a and the gears 52d, 52e, 52f which mesh with one another in this order being interposed between the gears 42a and 49b. When the clamping assembly 39 is arranged above the pressure plate, as represented in FIGS. 5 and 6, the shaped parts 46a, 46b, 46c of the shafts 42a, 42b, 42c being engaged in the corresponding hollow shaped parts of the screw heads 31b, a tool such as a box spanner 40 is engaged over the top shaped part 45 of the spindle 42a, and is rotated in the direction indicated by arrow 53, so as to turn the spindle 42a and the screw 31 in the direction corresponding to screwing the screw 31 into the tapped hole 30a of the pressure plate 28. By means of the gears 52c, 52b and 52a, the shaft 42c is rotationally driven at the same speed and in the same direction as the shaft 42a. Likewise, by means of the gears 52d, 52e, 52f, the shaft 42b, which is securely fastened to the gear 49b is rotationally driven at the same speed and in the same direction as the shaft 42a. The three screws 31 of the pressure plate 28 are therefore screwed in the same way and simultaneously into the tapped holes 30a, 30b, 30c. The plate 28 rises evenly, and comes into contact with the pull-ring 33 by means of which the thermocouple column 22 is raised until the moment at which the seal 25 is in its maximum crushed state between the shoulder 23 and the frustoconical bearing surface 24. This maximum crushed state of the seal is obtained in an even, simple and quick manner by actuating the spindle 42a by means of the tool 40. The time necessary for the sealed clamping of the thermocouple column is therefore considerably reduced by using the clamping device according to the invention. The displacement of the pressure plate and its clamping are produced identically at any point on the periphery of the plate in which the three screws 31, with balanced tightening, are engaged. When the clamping of the thermocouple column is completed, the clamping assembly 35 is lifted by means of the handles 39a and 39b so as to separate it from the pressure plate, then extracted through the top part of the thermocouple column. As indicated hereinabove, the pressure plate, the supporting endpiece and the pull-ring remain in place in order to hold the thermocouple column in the sealed clamped position. The clamping assembly 35 and the spanner 40 may be used to perform a new sealed clamping operation of a thermocouple column passing through the head via a second adapter. Preferably, the spanner 40 is a torque wrench, making it possible to tighten the lifting screws to a specific torque. The device according to the invention makes it possible to achieve, very effectively and quickly, the sealed clamping of an instrumentation column such as a thermocouple column passing through an adapter passing through the vessel head of a nuclear reactor. The pressure plate and the endpiece resting on the top part of the support assembly may have a shape other than the shape of an equilateral triangle with rounded corners. The number of lifting screws used may be greater than three, so long as their number is an odd number. A clamping assembly other than the assembly which has been described hereinabove may be used, this assembly being suited to the arrangement and to the number of lifting screws of the pressure plate. The synchronous rotation of the shaft driven by the tightening tool may be transmitted by means other than a chain of identical gears meshing with one another. The driving shaft of the clamping device may be rotationally driven by a means other than a torque wrench. Finally, the invention applies to the clamping of any instrumentation column passing through the head of the vessel of a nuclear reactor, inside an adapter.
	description	This application claims benefit under 35 U.S.C. § 119(e) from U.S. Provisional Patent Applications Ser. Nos. 60/265,353 and 60/265,354, both filed on Feb. 1, 2001, the entire contents of both being incorporated herein by reference. The invention was made with Government support under Grant Number 1 R43 CA76752-01, and under Grant Number 2 R44 CA76752-02, awarded by the National Institutes of Health, National Cancer Institute. The Government has certain rights in the invention. Related subject matter is disclosed in U.S. patent application Ser. No. 09/459,597, filed on Dec. 13, 1999, in U.S. patent application Ser. No. 09/734,761, filed Dec. 13, 2000, and in U.S. Pat. No. 5,949,850, the entire contents of all of these documents are expressly incorporated herein by reference. 1. Field of the Invention The present invention relates to a method and apparatus for making focused and unfocused grids and collimators that are stationary or movable to avoid grid shadows on an imager and which are adaptable for use in a wide range of electromagnetic radiation applications, such as x-ray and gamma-ray (γ-ray) imaging devices and the like. More particularly, the present invention relates to a method and apparatus for making focused and unfocused grids and collimators, such as air core grids and collimators, that can be constructed with a very high aspect ratio, defined as the ratio between the height of each absorbing grid or collimator wall and the thickness of the absorbing grid or collimator wall, and that are capable of permitting large primary radiation transmission there through. The present invention relates to a method and apparatus for making large area grids and collimators from a single piece or assembled from two or more pieces, For example, the grid and collimator can be assembled from two or more pieces in one layer, and there can be a plurality of layers, each of which includes thin metal walls defining the openings, and which can be stacked on top of each other to increase the overall thickness of the grid or collimator 2. Description of the Related Art Grids and collimators are used to let through the desirable electromagnetic radiation while eliminate the undesirable ones by absorption. Radiation can penetrate through thicker material as the radiation wavelength decreases or energy increases. The radiation decay length in the material decreases as the atomic number and the density of the materials increase, and according to other properties of the grid or collimator material. Grid and collimator walls, called the septa and/or lamellae, are usually made of metal because of their atomic number and density. Grids and collimators are used extensively in medical x-ray diagnostics, nuclear medicine, non-destructive testing, airport security, a variety of scientific and research applications, industrial instruments, x-ray astronomy and other devices to control, shape or otherwise manipulate beams of radiation. For the description below, the application related to medical diagnostics will be outlined, first for grids for x-ray and then collimators for γ-ray imaging. X-Ray Imaging: Conventional medical x-ray imaging systems consist of a point x-ray source and an image recording device (the imager). As x-rays pass through the object on the way to the imager, its intensity is reduced as the result of the internal structure of the object. Thus, x-rays are used in medical applications to differentiate healthy tissue, diseased tissue, bone, and organs from each other. As x-rays interact with tissue, the x-rays become attenuated as well as scattered by the tissue. X-rays propagating in a direct line from the x-ray source to the imager are desired. Contrast and the signal-to-noise ratio of image details are reduced by scatter. Anti-scatter grids are applied to most diagnostic x-ray imaging modality. For the description below, mammography is used as an example. Without intervention, both scattered and primary radiations from the subject are recorded in a radiographic image. For mammography, the typical scatter-to-primary ratios (S/P) at the imager range from 0.3 to 1.0. The presence of scatter can cause up to a 50% reduction in contrast, and up to a 55% reduction (for constant total light output from the screen) in signal-to-noise ratio as described in an article by R. Fahrig, J. Mainprize, N. Robert, A. Rogers and M. J. Yaffe entitled, “Performance of Glass Fiber Antiscatter Devices at Mammographic Energies”, Med. Phys. 21, 1277 (1994), the entire contents of both being incorporated herein by reference. The most common anti-scatter grids, called “one-dimensional” grid, or linear grid meaning that the projection of the lamellae walls on the imager are lines, are made by strips of lead lamella, sandwiched between more x-ray transparent spacer materials such as aluminum, carbon fiber or wood (see, e.g., the Fahrig et al article). This type of grid reduces scattered radiation by reducing scatter in one direction, the axis parallel to the strips. The typical grid ratio (height of grid wall divided by interspace length of the hole) is 4 to 5. The disadvantages associated with this type of one-dimensional grid are that it only reduces scattered x-rays parallel to the strips and that it requires an increase in x-ray dose because of absorption and scatter from the spacer materials. For scatter reduction applications, the grid walls preferably should be “two-dimensional,” meaning that the projection of the lamellae walls on the imager are not lines but two-dimensional patterns such as squares, rectangles, triangles or hexagonals, to eliminate scatter from all directions. For medical applications, the x-ray source is a point source close to the imager. In order to maximize the transmission of the primary radiation, all the grid openings have to point to the x-ray source. This kind of lamella geometry is called “focused.” Methods for fabricating and assembling focused and unfocused two-dimensional grids are described in U.S. Pat. No. 5,949,850, entitled “A Method and Apparatus for Making Large Area Two-dimensional Grids”, the entire content of which is incorporated herein by reference. When an anti-scatter grid is stationary during the acquisition of the image, the anti-scatter grid will cast a shadow on the imager. It is undesirable, since it can obstruct the image and make interpretation more difficult. The typical solution to eliminate the shadow of the grid is to move the grid during the period of exposure. The ideal anti-scatter grid with motion will produce uniform exposure on the imager, in the absence of an object being imaged. One-dimensional grids can be moved in a steady manner in one direction or in an oscillatory manner in the plane of the grid in the direction perpendicular to the parallel strips of highly absorbing lamellae. For two-dimensional grids, the motion can either be in one direction or oscillatory in the plane of the grid, but the grid shape needs to be chosen based on specific criteria. The following discussion pertains to a two-dimensional grid with regular square or rectangular patterns in the x-y plane, with the grid walls lined up in the x-direction and y-direction. If the grid is moving at a uniform speed in the x-direction, the film will show unexposed stripes along the x-direction, which repeat periodically in the y-direction. The width of the unexposed stripes is the same or essentially the same as the thickness of the grid walls. This grid pattern and associated motion are unacceptable. If the grid is moving at a uniform speed in the plane of the grid, but at a 45 degree angle from the x-axis, the image on the film or imager is significantly improved. However, strips of slightly overexposed images parallel to the direction of the motion at the intersection of the grid walls will still be present. As the grid moves in the x-direction at a uniform speed, the grid walls block the x-rays everywhere, except at the wall intersection, for the fraction of the time2d/D,where d is the thickness of the grid walls and D is the periodicity of the grid walls. At the wall intersection, the grid walls blocks the x-rays for the fraction of the time2d/D≦t≦d/D,depending on the location. Thus, stripes of slightly overexposed x-ray film are produced. Methods for attempting to eliminate the overexposed strips discussed above are disclosed in U.S. Pat. Nos. 5,606,589, 5,729,585 and 5,814,235 to Pellegrino et al., the entire contents of each patent being incorporated herein by reference. These methods attempt to eliminate the overexposed strips by rotating the grid by an angle A, where A=atan (n/m), and m and n are integers. However, these methods are unacceptable or not ideal for many applications. Not all x-ray imaging applications require focused grids. For example, the desirable x-rays for x-ray astronomy are from sources far away and they approach the detector as parallel rays. Anti-scatter grids are required to eliminate x-rays from different sources at different location in the sky. Thus, the walls of the grid should be parallel so that only x-ray from a very narrow angle can be detected. A grid with parallel walls is known as an unfocused grid. Also, there are variations of focused and unfocused grids, such as a) grids focused in one direction, but unfocused in the other direction; b) grids that are piecewise focused, and variations of these characteristic. Accordingly, the need exists for a method and apparatus to eliminate the overexposed strips associated with two-dimensional focused or unfocused grid intersections. γ-Ray Imaging Nuclear medicine utilizes radiotracers to diagnose disease in terms of physiology and biochemistry, rather than primarily in terms of anatomy, emphasizing function and chemistry rather than structure. Radiotracer studies usually measure three types of physiological activities: regional blood flow and other aspects of transport of matter through the body, bioenergetics, the provision of energy to body cells, cancer, effect of drugs, and intracellular and intercellular communication, the process by which molecular reactions are regulated.The typical γ-ray emissions are in the 80-500 keV energy range. These γ-rays can originate inside the body and emerge at the surface to be recorded by external radiation detectors. Nuclear imaging is able to examine the interactions for picomolar and lower quantities of molecules involved in biochemical interactions with macromolecular structures, such as recognition sites, enzymes, and substrates within different parts of the living body. Gamma cameras (γ-cameras) are used with collimators to capture the γ-rays emitted by the radionuclides. Unlike x-ray applications, γ-rays are emitted in all directions by the radioactive atoms, and they are distributed throughout large are of the body. Collimators are needed between the patient and the γ-camera to filter the γ-rays emitted from desirable locations, by selectively absorbing all but a few of the incident radiation. Gamma-rays that pass through the collimator have radiation propagation directions restricted to a small solid angle. In the absence of scattering within the patient, the photons propagate in a straight line from the point of emission to the point of detection in the γ-camera. Consequently, the collimator imposes a strong correlation between the position in the image and the point of origin of the photon within the patient. Because the collimator restricts the direction of the γ-ray propagation to a very small solid angle, the vast majority of the photons are absorbed by the collimator. This means that even minor improvements in collimator performance can significantly affect the number of detected events and reduce the statistical noise in the images. Collimators are typically made of lead. The conventional fabrication methods are pressing of thin lead foils and casting. Foil collimators can be made from foil as thin as 100 μm, but they are more susceptible to defects in foil misalignment, resulting in reduced resolution and uniformity of the image. Micro-cast collimators have more uniform septa thickness and good septa alignment, and are structurally stronger than foil collimators. However, micro-casting manufactures, such as Nuclear Fields, cannot make septa thinner than 150 μm. For small animal imaging, the main competitive technology is Tecomet's photochemically etched, stacked tungsten. This technology, however, is (a) limited in the septa thickness, (b) unable to fabricate focused cone beam collimators with smooth walls, and unable to fabricate collimators requiring large slant septa. Two-dimensional (2D) planar scintigraphy and three-dimensional (3D) single photon emission computed tomography (SPECT) imaging systems are used for visualization of in vivo biochemical processes, localization of disease, classification of disease, etc. SPECT provides information on three-dimensional in vivo distribution of radiotracers within the body, calculated from a set of 2D projectional images acquired from a number of γ-cameras surrounding the patient. An object of the present invention is to provide grids and collimators made from a variety of metals, where the walls focus to a point, where the walls focus to a line, the walls have varying focus, where the walls diverge from a point, where the walls diverge from a line, or where the walls are parallel (unfocused), that can be freestanding, released from substrate with hollow core or filled with scintillators, transparent, opaque, or other useful materials. Another objective of the present invention is to configure the grids to minimize shadow when the grid is moved during imaging. A further object of the present invention is to provide a method and apparatus for moving a focused or unfocused grid so that no perceptible shadow or area of variable density is cast by the grid onto the imager. Another objective of the present invention is to provide methods and apparatus for manufacturing grids and collimators. Another object of the present invention is to provide a method and apparatus for manufacturing focused and unfocused grids that are configured to minimize overexposure at wall intersections when a grid is moved during imaging. Grids and collimators can be made in one piece or by a plurality of pieces that can be combined to form an individual device. Tall grids and collimators can be made by stacking shorter pieces with precisely aligned walls. Large area grids and collimators can be made by assembling precisely matched pieces for each layer. These and other objects of the present invention are substantially achieved by providing a grid or collimator, adaptable for use with electromagnetic energy emitting devices. The grid or collimator comprises at least one solid metal layer. The solid metal layer comprises top and bottom surfaces, and a plurality of solid integrated intersecting walls, each of which extends from the top to the bottom surface, and having a plurality of side surfaces. The side surfaces of the walls are arranged to define a plurality of openings extending entirely through the layer, and at least some of the side surfaces have projections extending into the respective openings. The projections can be of various shapes and sizes, and are arranged so that a total amount of wall material intersected by a line propagating in a direction, for example, along an edge of the grid, for each period along the grid is substantially the same and is also substantially the same as another total amount of wall material intersected by another line for each period propagating in another direction substantially parallel to the edge of the grid at any distance from the edge. These and other objects are further substantially achieved by providing a method for minimizing scattering of radiation in a device to obtain an image of an object on an imager. The method includes placing a grid between radiation emitting source of the electromagnetic imaging device and the imager. The grid comprises at least one metal layer including top and bottom surfaces and a plurality of solid integrated, intersecting walls, each of which extending from the top to bottom surface and having a plurality of side surfaces, the side surfaces of the walls being arranged to define a plurality of openings extending entirely through the layer, and at least some of the side surface having projections extending into respective ones of the openings. The method further includes moving the grid in a grid moving pattern while the radiation source is emitting radiation toward the imager. In addition, the holes of one or more layers of a grid or collimator produced by the present invention can be filled with various materials that are transparent, opaque, or have other properties, such as scintillators. Examples of scintillator are phosphors, CsI, or the like. Since grids and collimators can be reproduced exactly, an air-core grid or collimator can be aligned precisely with the filled-core grid or collimator counterpart. The desired thickness of the filling can also be achieved precisely. This type of grid/scintillator or collimator/scintillator, therefore, can performs the functions of (1) eliminating detection of undesirable radiation, (2) conversion of x-rays or γ-rays to optical or UV signals or other forms of signals and (3) improving resolution of the image or (4) improve the structural strength or other properties of the device. Grid and collimator walls can be 5 μm or thicker. There is no inherent limitation on their height by stacking or their area by assembly. Methods to fabricate grids and collimators for a wide variety of materials and geometry are described in this patent. One method is to use ultra violet (UV) or x-ray lithography followed by electroplating/electroforming or micro casting methods. The UV or x-ray lithography/electroforming technology: Can produce metal septa as thin as 20 μm. Can make unfocused and focused grids and collimators that have parallel, converging fan-beam, cone-beam, diverging, or spatially varying focus walls, Allows septal thickness and opening geometry to vary with location in the horizontal plane, Allows grids and collimators to have non-uniform thickness in the vertical direction. The present invention provides designs, methods and apparatuses for making large area, two-dimensional, high-aspect-ratio, grids, collimators, grid/scintillators, collimator/scintillators, x-ray filters and other such devices, with focused walls, defocused walls, variable focus walls, parallel walls and other such orientations, as well as similar designs, methods and apparatuses for all electromagnetic radiation applications. Referring now to the drawings, FIG. 1 shows a schematic of a section of an example of a two-dimensional grid or collimator 30 produced in accordance with an embodiment of a method of the present invention. The method of grid manufacture described here is different from the embodiment of the invention, as described in more detail in U.S. Pat. No. 5,949,850 and 6,252,938 referenced above, the entire contents of both being incorporated herein by reference. In FIG. 1, the x-rays propagate out of a point source 61 with a conical spread 60. The x-ray imager 62, which may be an electronic detector or x-ray film, for example, is placed adjacent and parallel or substantially parallel to the bottom surface of the x-ray grid 30 with the object to be imaged (not shown) positioned between the x-ray source 61 and the x-ray grid 30. Typically, the top surface of the x-ray grid 30 is perpendicular or substantially perpendicular to the line 63 that extends between the x-ray source and the x-ray grid 30. The grid openings 31 that are defined by walls 32 are square in this example. However, the grid openings can be any practical shape as would be appreciated by one skilled in the methods of grid construction. The walls 32 are uniformly thick or substantially uniformly thick around each opening in this figure, but can vary in thickness as desired. The walls 32 are slanted at the same angle as the angle of the x-rays emanating from the point source, in order for the direct radiation to propagate through the holes to the imager without significant loss. This angle increases for grid walls further away from the x-ray point source. In other words, an imaginary line extending from each grid wall 32 along the x-axis 40 could intersect the x-ray point source 61. A similar scenario exists for the grid walls 32 along the y-axis 50. To facilitate the description below, a coordinate system in which the grid 30 is omitted will now be defined. The z-axis is line 63, which is perpendicular or substantially perpendicular to the anti-scatter grid, and intersects the point x-ray source 61. The z=0 coordinate is defined as the top surface of the anti-scatter grid. As further shown, the central ray 63 propagates to the center of the grid 30, which is marked by a virtual “+” sign 64. The grid openings 31 that are defined by walls 32 are square in this example. However, the grid openings can be any practical shape as would be appreciated by one skilled in the art of grid design and fabrication. The walls 32 are uniformly thick or substantially uniformly thick around each opening in this figure, but can vary in thickness as desired. The walls 32 are slanted at the angle that allows the x-rays from the point source to propagate through the holes to the imager without significant loss. That is, the directions in which the walls extend converge or substantially converge at the point source 61 of the x-ray. The angle at which each wall is slanted in the z direction is different from its adjacent wall as taken along the directions x and y. The desirable dimensions of the x-ray grids depend on the application in which the grid is used. For typical medical imaging applications, the area of the top view is large and the height of the grid is no more than a few millimeters. The variation in area and thickness depends on the x-ray energy, resolution, image size and the angle of the typical scattered radiation. For mammographic imaging, for example, the x-ray energy is in the range of about 17 kVp to about 35 kVp, but can be any level as would be necessary to form a suitable image. The distance between the x-ray source and the grid plane is usually in the range of 60 cm for mammography but, of course, could be different for other applications as would be appreciated by one skilled in the art. Without the grid, scatter blurs the image, reducing contrast and makes it difficult to distinguish between healthy and diseased tissues. Only the x-rays propagating in the line from the x-ray source to the detector are desired to produce a sharp image. For mammographic imaging, the dimensions of the grid are determined in the following manner. The field size is determined by the object to be imaged. Two field sizes are used for mammographies: 18 cm by 24 cm and 24 cm by 30 cm, but any suitable field size can be used. The field size depends on the imaging system in use and the medical procedure. For example, some procedures require only images over small areas as small as few cm2. The wall height is usually defined in terms of the grid ratio (grid height divided by the interspace length of the hole). Grid ratio in the range of 3.5 to 5.5 are typical for mammography. For the interspace length of 525 μm and a grid ratio of 5, the wall height is 2,625 μm. The wall thickness is determined by the x-ray energy and the material used to form the wall. The linear attenuation coefficient μ of copper (atomic number Z=29) is μ=303 cm−1 at 20 keV, as described in a book by H. E. Johns and J. R. Cunningham, The Physics of Radiology, Charles C. Thomas Publisher, Springfield, Ill. 1983, the entire contents of pages 134-166 and 734-736 being incorporated herein by reference. This means that the intensity of the x-rays decay by a factor of e in a distance of δ=1/μ=33 μm, and that scattered x-rays strike the grid walls will be absorbed. The interspace dimensions are to be determined by considerations such as the percentage of open area and the method of x-ray detection. The ratio of the open area is determined by (open area)/(open area+wall area). The percentage of open area should be as large as possible, in order to achieve the minimal practical Bucky factor. For interspace distance of 525 μm, and wall thickness of 25 μm, the percentage of the open area is 91%. For mammographic applications, the percentage of the ratio of the open area should be as close to 100% as possible, in order to produce a suitable image with the lowest possible radiation dose. For other medical x-ray imaging applications, the imaging systems are different, such as chest, heart and brain x-rays, computed tomography (CT) scans, etc. Anti-scatter grids for medical applications thus cover a wide range of sizes. The grid thickness can range from as little as 5 μm to any desirable thickness. The lower limit of the interspace length of the hole is on the order of a few μm and the upper limit is the size of substrates. However, there is a necessary relationship between wall thickness and hole sizes, the grid height and the absorption properties of the absorbent material. When the grid is made of copper, the following dimensions can significantly reduce scatter and improve mammography imaging: 550 μm holes, 25 μm thick walls, a grid height of 2000-3000 μm. As the hole size or wall thickness decreases, the layer height will have to be reduced. As stated, wall thickness can be varied, depending on the application in which the grid is used, and the walls do not need to be of uniform thickness. Also, the shape of the hole can be varied as long as it does not result in walls having extended sections thinner than about 5 μm. The shape of the holes does not have to be regular. Some hole shapes that may be practical for anti-scatter applications are rectangular, hexagonal, circular and so on. The walls can be made of any suitable absorbent material that can be fabricated in the desired structure, such as electroplating/electroforming, casting, injection molding, or other fabricating techniques. Materials with high atomic number Z and high density are desirable. For instance, the walls can include nickel, nickel-iron, copper, silver, gold, lead, tungsten, uranium, or any other common electroplating/electroforming or casting materials. FIGS. 2a and 2b show schematics of two air-core x-ray anti-scatter grids, such as grid 30 shown in FIG. 1, that are stacked on top of each other in a manner described in more detail below to form a grid assembly. These layers of the grid walls can achieve high aspect ratio such that they are structurally rigid. The stacked grids 30 or a grid made in a single layer can be moved steadily along a straight line (e.g., the x-axis 40) during imaging. As shown in these figures, the grids 30 have been oriented so that their walls extend at an angle of 45° or about 45° with respect to the x-axis 40. The top surface of the top grid 30 is in the x-y plane. The central ray 63 from the x-ray source 61 is perpendicular or substantially perpendicular to the top surface of the top grid 30. For mammographic applications, the central ray 63 propagates to the top grid 30 next to the chest wall at the edge or close to the edge of the grid on the x-axis 40, which is marked as location 65 in FIG. 2a. For general radiology, the central ray 63 is usually at the center of the top grid 30, which is marked as location 64 in FIG. 2b. In this example, the line of motion 70 of the grid assembly is parallel or substantially parallel to the x-axis 40. In the x-y plane, one set of the walls 32 (i.e., the septa) is at 45° with respect to the line of motion 70, and the shape of the grid openings 31 is nearly square. The grid assembly can move in a linear motion in one direction along the x-axis or it can oscillate along the x-axis in the x-y plane. During motion, the speed at which the grid moves should be constant or substantially constant. Two categories of grid patterns can be used with linear grid motion to eliminate non-uniform shadow of the grid. The description below pertains to portions of the grid not at the edges of the grid, so the border is not shown. For illustration purposes only, the dimensions of the drawings are not to scale, nor have they been optimized for specific applications. A.1. Grid Design Art Type I for Linear Motion As discussed above, the present invention provides a two-dimensional grid design and a method for moving the grid so that the image taken will leave no substantial artificial images for either focused or unfocused grids for some applications. In particular, as will now be described, the present invention provides methods for constructing grid designs that do not have square patterns. The rules of construction for these grids are discussed below. Essentially, Type 1 methods for eliminating grid shadows produced by the intersection of the grid walls are based on the assumptions that: (1) there is image blurring during the conversion of x-rays to visible photons or to electrical charge; and/or (2) the resolution of the imaging device is not perfect. A general method of grid design provides a grid pattern that is periodic in both parallel and perpendicular (or substantially parallel and perpendicular) directions to the direction of motion. The construction rules for the different grid variations are discussed below. A.1.a. Grid Design Variation I.1: A Set of Parallel Grid Walls Perpendicular to the Line of Motion FIG. 3 shows a top view of an exemplary grid layout that can be employed in a grid 30 as discussed above. The grid layout consists of a set of grid walls, A, that are perpendicular or substantially perpendicular to the direction of motion, and a set of grid walls, B, intersecting A. The thicknesses of grid walls A and B are a and b, respectively. The thicknesses a and b are equal in this figure, but they are not required to be equal. The angle θ is defined as the angle of the grid wall B with respect to the x-axis. The grid moves in the x-direction as indicated by 70. Px and Py are the periodicities of the intercepting grid wall pattern in the x- and y-directions, respectively. Dx and Dy represent the pitch of grid cells in the x- and y-directions, respectively. The periodicity of the grid pattern in the x-direction is Px=MDx, where M is a positive integer greater than 1. The periodicity of the grid pattern in the y-direction is Py=M(Dy/N), where N is a positive integer greater than or equal to 1, M≠N and Py=\|tan(θ)\|Px. For linear motion, the grid pattern can be generated given Dx,(θor Dy), (M or Px) and (N or Py). The parameter range for the angle θ is 0°<\|θ\|<90°. The best values for the angle θ are away from the two end limits, 0° and 90°. The grid intersections are spaced at intervals of Py/M in the y-direction. If Dx,θ, M and N are given, the parameters Px, Py, and Dy can be calculated FIG. 3 is a plot of a section of the grid for the following chosen parameters: θ=45°, M=3 and N=1. If the parameters Dx, Dy, M and N are chosen, the angle θ, Px and Py can be calculated: Px=MDx, Py=NDy and θ=±atan (Py/Px). FIG. 4 is a plot of a section of the grid for the parameters N=2, M=7 and θ=−atan (2Dy/7Dx). A.1.b. Grid Design Variation I.2: Grid Walls Not Perpendicular to the Line of Motion FIG. 5 is the top view of a section of the grid layout where neither grid walls A nor B are perpendicular to the direction of linear motion. The thicknesses of grid walls A and B are a and b, respectively. The thicknesses a and b are equal in this figure, but they are not required to be equal. The angles between the grid walls A and B relative to the x-axis are φ and θ, respectively. Choosing Dx, (M or Px), (N or Py), and angles (θor Dy) and φ, then Py=\|tan(θ)\|Px, N=Py/DY and (M=Px/Dx). The centers of grid intersections are separated by a distance Py/M in the y-direction. FIG. 5 shows an example where θ=−15°, φ=−80°, M=5 and N=1. FIG. 6 is the top view of a section of the grid layout where neither grid walls A or B are perpendicular to the direction of motion, but grid wall A is perpendicular to grid wall B, thus a special case of FIG. 5, where the grid openings are rectangular. The thicknesses of grid walls A and B are a and b, respectively. The thicknesses are equal in this figure, but again, they are not required to be equal. The angles between the grid walls A and B relative to the x-axis are φ and θ, respectively. By choosing Dx, (M or Px), (N or Dy), (θor Py) and φ, then Py=\|tan(θ)\|Px, Py=NDy, and Px=MDx. The centers of grid intersections are separated by a distance Py/M in the y-direction. FIG. 6 shows an example where θ=10°, φ=−80°, M=10 and N=1. A.1.c. Comments on the Grid Motion Associated with Grid Design I For all grid layout methods, the range of parameters for the grid can vary depending on many factors, such as whether film or digital detector is used, the type of phosphor used in film, the sensitivity and spatial resolution of the imager, the type of application, the radiation dose, and whether there is direct x-ray conversion or indirect x-ray conversion, etc. The ultimate criterion is that the overexposed strips caused by grid intersections are contiguous. Some general conditions can be given for the range of parameters for Grid Design Type I and associated motion. It is better for grid openings to be greater than the grid wall thicknesses a and b. For film, Py/M should be smaller than the x-ray to optical radiation conversion blurring effect produced by the phosphor. For digital imagers with direct x-ray conversion, it is preferable that pixel pitch in the y-direction is an integer multiple of the spacing, Py/M. Otherwise, the grid shadows will be unevenly distributed on all the pixels. The distance of linear travel, L, of the grid during the exposure should be many times the distance Px, where kPx>L>(kPx−δL), Dx>δL>a sin(φ), Dx>δL>b/sin(θ), δL/Px<<1, k >>1, and k is an integer. The ratio of δL/L should be small to minimize the effect of shadows caused by the start and stop. The distance L can be traversed in a steady motion in one direction, if it is not too long to affect the transmission of primary radiation. Assuming that the x-ray beam is uniform over time, the speed with which the grid traverses the distance L should be constant, but the direction can change. In general, the speed at which the grid moves should be proportional to the power of the x-ray source. If the required distance L to be traveled in any one direction is too long, that can cause reduction of primary radiation, then the distance can be traversed by steady linear motion that reverses direction. A.2. Grid Design Type II for Linear Motion The present invention provides further two-dimensional grid designs and methods of moving the grid such that the x-ray image will have no overexposed strips at the intersection of the grid walls A and B. The principle is based on adding additional cross-sectional areas to the grid to adjust for the increase of the primary radiation caused by the overlapping of the grid walls. This grid design and construction provides uniform x-ray exposure. Two illustrations of the concept are given below, followed by the generalized construction rules. This grid design is feasible for the SLIGA fabrication method described in U.S. Pat. No. 5,949,850 referenced above, because x-ray lithography is accurate to a fraction of a micron, even for a thick photoresist. A.2.a. Grid Design Variation II.1: Square Grid Shape with an Additional Square Piece FIG. 7 shows a section of a square patterned grid with uniform grid wall thickness a and b rotated at a 45° angle with respect to the direction of motion. When square pieces in the shape of the septa intersection are added to the grid next to the intersection, with one per intersection as shown in FIG. 8, the grid walls leave no shadow for a grid moving with linear motion 70. In the FIG. 8, Dx=Dy=Px=Py and θ=45°. The additional grid area is shown alone in FIG. 9. A.2.b. Grid Design Variation II.2: Square Grid Shape with Two Additional Triangular Pieces FIG. 10 shows another grid pattern, which has the same or essentially the same effect as the grid pattern in FIG. 8, by placing two additional triangular pieces at opposite sides of intersecting grid walls. In this FIG. 10 example, Dx=Dy=Px=Py and θ=45°. The additional grid area is shown alone in FIG. 11. With these modified corners added to the grid, there will not be any artificial patterns as the grid is moved in a straight line as indicated by 70 for a distance L, where kDx>L≧(kDx−δL), Dx>>δL>s, δL<<L, k>>1 and k is an integer. Along the x-axis, the grid wall thickness is s and the periodicity of the grid is Px=Dx. The distance of linear travel L should be as large as possible, while maintaining the maximum transmission of primary radiation. The condition for linear grid motion in just one direction is easier for grid Design Type II to achieve than grid Design Type I or the designs in U.S. Patents by Pellegrino et al. referenced above, because Px>Dx. for grid Design Type I. A.2.c. General Construction Methods for Quadrilateral Grid Design Type II for Linear Motion The exact technique for eliminating the effect of slight overexposure caused by the intersection of the grid walls with linear motion is to add additional grid area at each corner. Two special examples are shown in FIGS. 8 and 10 discussed above, and the general concept is described below and illustrated in FIGS. 12-16. The general rule is that the overlapping grid region C formed by grid walls A and B has to be “added back” to the grid intersecting region, so that the total amount of the wall material of the grid intersected by a line propagating along the x-direction remains constant at any point along the y axis. In other words, the total amount of wall material of the grid intersected by a line propagating in a direction parallel to the x-axis along the edge of a grid of the type shown, for example, in FIG. 8 or 10, is identical to the amount of wall material of the grid intersected by a line propagating in a direction parallel to the x-axis through any position, for example, the center of the grid. This concept can be applied to any grid layout that is constructed with intersecting grid walls A and B. The widths of the intersecting grid walls do not need to be the same, and the intersections do not have to be at 90°, but grid lines cannot be parallel to the x-axis. The width of the parallel walls B do not need to be identical to each other, nor do they need to be equidistant from one another, but they do need to be periodic along the x-axis with period Px. The widths of the parallel lines A do not need to be identical to each other, nor do they need to be equidistant from one another, but they do need to be periodic along the y-axis with period Py. The generalized construction rules are described using a single intersecting corner of walls A and B for illustration as shown in FIGS. 12-16. The top and bottom corners of parallelogram C are both designated as γ and the right and left corners of the parallelogram C as β1 and β2, respectively. Dashed lines, f, parallel to the x-axis, the direction of motion, are placed through points γ. The points where the dashed lines f intersect the edges of the grid lines are designated as α1, α2, α3 and α4. FIG. 12 shows the addition to the grid in the form of a parallelogram F formed by three predefined points: α1, α2, β1 , and δ, where δ is the fourth corner. This is the construction method used for the grid pattern shown in FIG. 8. FIG. 13 shows the addition of the grid area in the shape of two triangles, E1 and E2, formed by connecting the points α1, α2, β1 and α3, α4, β2, respectively. This is the construction method used to make the grid pattern shown in FIG. 10. There are an unlimited variety of shapes that would produce uniform exposure for linear motion. Samples of three other alternatives are shown in FIGS. 14-16. They produce uniform exposure because they satisfy the criteria that the lengths through the grid in the x-direction for any value y are identical. There is no or essentially no difference in performance of the grids if motion is implemented correctly. Additional grid areas of different designs can be mixed on any one grid without visible effect when steady linear motion is implemented. FIG. 17, for example, illustrates and arrangement where different combinations of grid corners are implemented in one grid. However, the choice of grid corners depends on the ease of implementation and practicality. Also, since it is desirable for the transmission of primary radiation to be as large as possible, the grid walls occupy only a small percentage of the cross-sectional area. A.2.d. General Construction Methods for Grid Design Type II for Linear Grid Motion It should be first noted that this concept does not limit grid openings to quadrilaterals. Rather, the grid opening shapes could be a wide range of shapes, as long as they are periodic in both x and y directions. The grid wall intercepts do not have to be defined by four straight line segments. Non-uniform shadow will not be introduced as long as the length of the lines through the grid in the x-direction are identical through any y coordinate. In addition to adding the corner pieces, the width of some sections of the grid walls would need to be adjusted for generalized grid openings. However, not every grid shape that is combined with steady linear motion produces uniform exposure without artificial images. The desirable grid patterns that produce uniform exposure need to satisfy, at a minimum, the following criteria: The grid pattern needs to be periodic in the direction of motion with periodicity Px. No segment of the grid wall is primarily along the direction of the grid motion. The grid walls block the x-ray everywhere for the same fraction of the time per spatial period Px. at any position perpendicular to the direction of motion. The grid walls do not need to have the same thickness. The grid patterns are not limited to quadrilaterals. These grid patterns need to be coupled with a steady linear motion such that the distance of the grid motion, L, satisfies the condition described in Sections Grid Design Type I and Type II for Linear Motion. If the walls are not continuous at the intersection or not identical in thickness through the intersection, the construction rule that must be maintained is that the length of the line through the grid in the x-direction is identical through any y-coordinate. Hexagons with modified corners are examples in this category. A.2.e. Implementation of the Grid Design Type II for Linear Grid Motion The additional grid area at the grid wall intersections can be implemented in a number of ways for focused or unfocused grids to obtain uniform exposure. The discussion will use FIGS. 8 and 10 as examples. 1. The grid patterns with the additional grid area, such as FIGS. 8, 10, 17, and so on, may have approximately the same cross-sectional pattern along the z-axis. 2. Since the additional pieces of the grid are for the adjustment of the primary radiation, these additional grid areas in FIGS. 8, 10, 17, and so on, only need to be high enough to block the primary radiation. This allows new alternatives in implementation. A portion of the grid layer needs to have the additional grid area, while the rest of the grid layer does not. For example, a layer of the grid is made with pattern shown in FIG. 8, while the other layers can have the pattern shown in FIG. 7. The portion of the grid with the shapes shown in FIGS. 8, 10, 17, and so on, can be released from the substrate for assembly or attached to a substrate composed of low atomic number material. The portion of the grid with the pattern shown in FIGS. 8, 10, 17, and so on, can be made from materials different from the rest of the grid. For example, these layers can be made of higher atomic number materials, while the rest of the grid can be made from the same or different material. The high atomic number material allows these parts to be thinner than if nickel were used. For gold, the height of the grid can be 20 to 50 μm for mammographic applications. The height of the additional grid areas depends on the x-ray energy, the grid material, the application and the tolerances for the transmission of primary radiation. The photoresist can be left in the grid openings to provide structure support, with little adverse impact on the transmission of primary radiation. 3. The additional grid areas shown in FIGS. 9, 11, and so on, can be fabricated separately from the rest of the grid. These areas can be fabricated on a substrate composed of low atomic number material and remain attached to the substrate. These areas can be fabricated along with the assembly posts, which are exemplified in FIGS. 16a and 16b of U.S. Pat. No. 5,949,850, referenced above. Patterns shown in FIGS. 9, 11, and so on, can be made of a material different from the rest of the grid. For example, these layers can be made from materials with higher atomic weight, while the rest of the grid can be made of nickel. The high atomic weight material allows these parts to be thinner than if nickel were used. For gold, the height of the grid can be 20 to 100 μm for mammographic applications. The height of the additional grid areas depends on the x-ray energy, the grid material, the application and the tolerances for the transmission of primary radiation. The photoresist can be removed from the fabricated grid or collimator or left in on substrate composed of low atomic number material to provide structural support.A.2.f. Grid Parameters and Design Examples of the parameter range for mammography application and definitions are given below. Grid Pitch is Px. Aspect Ratio is the ratio between the height of the absorbing grid wall and the thickness of the absorbing grid wall. Grid Ratio is the ratio between the height of the absorbing wall including all layers and the distance between the absorbing walls. Best Case:for x-ray anti-scatterRangegrid for mammographyGrid TypeType I or IIType II/FIG. 10Grid Opening ShapeQuadrilateralSquareThickness of Absorbing 10 μm-200 μm≈20-30 μmWall on the top plane ofthe gridGrid Pitch for Type I1000 μm-5000 μmGrid Pitch for Type II 100 μm-2000 μm ≈300-1000 μmAspect Ratio for a Layer1-100>15Number of Layers1-1001-5Grid Ratio3-10 5-8However, it should be noted that different parameter ranges are used for different applications, and for different radiation wavelengths. Imaging radioactive sources distributed throughout a volume requires collimators to localize the source by eliminating the γ-rays from undesirable locations. Gamma-ray imaging is utilized in nuclear medicine, basic research, national defense applications, etc. FIG. 18 shows a focused collimator 832, a gamma camera 862, and γ-rays 860. The most commonly used radionuclides for planar scintigraphy and SPECT are iodine-123, 123I, (13 hr half time and photon energy of 160 keV), technetium-99 m, 99 mTc, (6.0 hour half time, photon energy 140 keV), and indium-111, 111In, (2.8 days half time, photon energy 173 keV (50%), 247 keV (50%)), as described in a book by R. E. Henkin, et al., Nuclear Medicine, Mosby, St. Louis, 1996, the entire contents of both being incorporated herein by reference. The desirable materials for collimators would be tungsten, gold, lead and materials with the highest possible atomic number and density. For some research and defense applications, the γ-ray energies can be higher than those cited above. Typically, the periodicity, the wall thickness and the height of collimators are larger than that of the grid. The collimator parameters can vary widely depending on the radioactive material and the needs of a particular application. Table 1 gives the physical properties of tungsten, gold and lead at 140 keV and Table II gives a set of collimator design parameters. TABLE IPhysical properties of tungsten, gold and lead at 140 keV.DensityAttenuationAtomicρμ/ρCoefficientNumber(g/cm3)(cm2/g)μ (cm−1)Tungsten (W)7419.251.88236.23Gold (Au)7919.32.20942.63Lead (Pb)8211.362.3927.15 TABLE IIComparison of optimized collimator designs optimized for differentmaterials for 140 keV.HoleHoleHolePeriodicityDiameterSideSeptaThicknessOptimized(μm)(μm)(μm)(μm)(cm)Tungsten (W)380338300800.92Gold (Au)380343304760.82Lead (Pb)380329291881.13The distance d that the 140 keV γ-ray travels in the material and its intensity decreases by a factor e is d=1/μ. C.1. Grid and Collimator Joint Designs: Designs of grid joints were described in U.S. Pat. No. 5,949,850 and 6,252,938 referenced. FIG. 19 shows a grid to be assembled from two sections, using the pattern of FIG. 7 as an example. The curved corner interlocks in the shape of 110 and 111 shown in FIG. 19 are found to be more desirable structurally than other joints. Straight line boundaries are also acceptable as long as they retain their relative alignments. The details of the corner can vary. C.2 Grid and Collimator Wall Orientations: The are many possibilities for grid and collimator walls: (a) The walls can be all perpendicular to the substrate, FIG. 20a. (b) Only one set of walls is perpendicular to the substrate while the other set of walls is parallel to each other but are not perpendicular to the substrate, FIG. 20b. (c) Both set of walls are parallel to each other but are not perpendicular to the substrate. (d) One set of walls is focused to a line, FIG. 20c, and the other set of walls is parallel. (e) One set of walls is defocused from a line, FIG. 20d, and the other set of walls is parallel. (f) Both sets of walls are focused to a point, FIGS. 1 & 2. (g) Both set of walls are defocused to a point. (h) Walls do not have identical point focus or identical line focus, FIG. 20e. C.3. Stacking: The manner in which tall grids are made in accordance with the present invention will now be discussed. For many applications, it is possible to make a grid or collimator in one piece. When it is not possible to make it in one piece at the desirable height, two or more thinner pieces can be assembled in a stack. Stacking of 10 layers of 210 μm high grids has been demonstrated in accordance with the present invention, but as many as 100 layers or more can be stacked, if necessary, when the individual pieces are all fabricated with correct dimensions and assembled with adequate precision. An advantage of stacking is that the layers can be made of the same or similar material or of different materials. In the stacking arrangement, illustrated with parallel walls in FIG. 21a, layer 70, 80 and 90 can be made of same material, or of different materials. The materials within each layer do not have to be identical. For example, a grid that is fabricated by electroplating/electroforming can be composed of a layer of copper, followed by a layer of lead, and finished with a layer of copper, forming the structure shown in FIG. 21b. The advantages this structure is avoidance of planarizing lead surfaces, utilized the high absorption of x-rays and γ-rays, and stronger structure of copper than lead. C.4. Grid/Scintillators and Collimator/Scintillators: If desired, the holes of one or more layers of the grid or collimator can be filled with scintillators, solid, liquid, glue or any other material required for research or a specific application. Scintillators converts x-ray and γ-rays to optical or UV signal. Some examples of scintillators are phosphors, CsI, etc. In some applications, not all the holes need be filled. When the holes are filled with scintillator, the signal is confined to the hole avoiding blurring. The scintillator should only be in the lower portion of a layer or layers of the stack. FIG. 22a shows the side view of scintillator 33 filling the bottom of the holes for one layer of the grid or collimator. FIG. 22b shows the side view of two layers of anti-scatter grids with the scintillator 33 in all the holes of the bottom grid layer 32. The hole of the layer above 31 are not filled with scintillator. C.5. Attachment to Substrate: Grids and collimators can be free-standing pieces or attached to a substrate. The methods according to the present invention for manufacturing the grids and grid pieces discussed above (as shown, for example, in FIGS. 1, 2, 17, 18, and 19) will now be discussed. There are four general photoresist/substrate combinations for fabrication: (a) positive photoresist and silicon or similar substrate, (b) positive photoresist and graphite substrate, (c) negative photoresist and silicon or similar like substrate and (d) negative photoresist and graphite substrate. For positive photoresist, the part of the resist that is exposed to the x-rays or ultraviolet or other radiation is the part that is removed during development. The opposite is true for negative photoresist. All of the grids described above can be manufactured using the methodology that will now be described with reference to FIGS. 23a-23h and 24a-24f. D.1. Fabrication Using Positive Photoresist and Not Graphite Substrates The first fabrication method, using positive photoresist and silicon substrates, is based on the techniques developed by Prof. Henry Guckel at University of Wisconsin at Madison called SLIGA. The details of fabrication are shown in FIGS. 23a-23h, with the lettered paragraphs corresponding to the lettered figures (e.g., paragraph (a) describes FIG. 23a). This method can make free standing nickel grids, but it cannot make free standing copper or lead grids and collimators, because the etch used to release the electroformed parts also dissolves the copper and lead parts. (a) A substrate 720, such as a silicon wafer, is prepared by sputtering the plating base and releasing metal (titanium/copper/titanium) 721 onto it. Copper (Cu) is used as the electroplating/electroforming electrode, while titanium (Ti) is used to adhere copper with the photoresist 710, and to connect copper with the substrate. (b) A thin layer of the photoresist 710 is spun on the substrate 720 followed by gluing on a thicker layer of the photoresist. The photoresist 710 of choice for the LIGA process is polymethyl-methacrylate (PMMA) because of the highly prismatic structures, with low run-outs, that can be fabricated from it. (c) The x-ray mask 730 is aligned onto the photoresist 710 attached to the substrate 720. This assembly is then exposed to an x-ray source 700, which transfers the pattern on the mask 730 to the photoresist 710. Synchrotron radiation is usually used, because of its very high collimation, high flux, and short wavelength. Within the irradiated sections of the resist layer, the polymer chains are destroyed, reducing the molecular weight. The unexposed regions of the resist were covered by the gold absorbers on the mask during irradiation. (d) The exposed photoresist is then developed; the exposed resist is selectively dissolved by a solvent, while the unexposed resist 710 remains unchanged. The top layer of the Ti plating 721 has to be removed by wet etch before electroplating/electroforming, because Ti is not a good electroplating/electroforming contact. (e) Metal 740 is electroplated into the pattern. (f) The electroplated metal 740 is lapped and polished to the desired metal height with an accuracy of ±1 μm. (g) The photoresist mold 710 is then removed by dissolving it chemically. (h) The device is released from the substrate 720 by etching away the copper on the substrate.D.2 Fabrication Using Positive Photoresist with Graphite Substrate The fabrication method using positive photoresist and graphite substrate is shown in FIGS. 24a-24f, with the lettered paragraphs corresponding to the lettered figures (e.g., paragraph (a) describes FIG. 24a). (a) A thin layer of the photoresist 710 is spun on the graphite substrate 725 followed by gluing on a thicker layer of the photoresist. The sacrificial and electroplating layers (Ti/Cu/Ti), needed for FIG. 23a, are no longer required. (b) The x-ray mask 730 is aligned onto the substrate with the photoresist 710. This setup is then exposed by an x-ray source 700, which transfers the pattern on the mask 730 to the photoresist 710. Within the irradiated sections of the resist layer the polymer chains are destroyed, reducing the molecular weight. The unexposed regions of the resist were covered by the gold absorbers on the x-ray mask during irradiation. (c) The exposed photoresist 710 is then developed, the exposed resist is selectively dissolved while the unexposed resist remain unchanged. (d) Metal 740 is electroplated into the patterned photoresist 710. (e) Graphite substrate 725 is removed by abrasion. The grid or collimator is polished on both sides. (f) The remaining photoresist can then be left in place or removed by wet etch leaving the metal 740.D.3. Fabrication Using Negative Photoresist and Not Graphite Substrate The fabrication method using negative photoresist and silicon substrate is similar to that shown in FIGS. 23a-23h, except that the mask has the reverse pattern from the positive photoresist. An example of negative photoresist is SU-8. SU-8 can be exposed by x-rays or by ultraviolet radiation in the 350-400 nm wavelength regime. A separate release layer is required on the substrate and the releasing material is evolving. An example of a releasing material is manufactured by MicroChem Corp. D.4. Fabrication Using Negative Photoresist and Graphite Substrate The fabrication method using negative photoresist and graphite substrate is similar to that shown in FIGS. 24a-24f, except that the mask has the reverse pattern from the positive photoresist. The method to remove the negative photoresist, the step from FIG. 24e to FIG. 24f, is dependent on the material. Using SU-8 as an example of negative photoresist, the grid with the SU-8 has to be baked at a temperature of 500° C. after polishing on both sides. The SU-8 shrinks and releases the grid or collimator. D.5 Additional Advantages of Graphite as Substrate Beside the fact that graphite can be used to fabricate freestanding grids and collimators using copper, lead, or any material that can be electroplated/electroformed or cast, it has three other advantages for use as a substrate. Graphite has a low atomic number, so that it is transparent to x-ray radiation. Graphite is conducting, so that no electroplating/electroforming layer of Ti/Cu/Ti is required, simplifying the fabrication process. In addition, graphite has high porosity and microroughness, so that attachment of photoresist to the substrate is stronger than to the silicon substrate with the Ti/Cu/Ti layer. Focused grids and collimators of any pattern can be fabricated by the method described in U.S. Pat. No. 5,949,850, referenced above. For all grids or collimators that do not have parallel walls, methods for exposing the photoresist using a sheet of parallel x-ray beams and positive photoresist are described below. E.1. Exposure of Focused Grid Design Type I For Linear Motion or Focused Collimator in a Single Piece If the pattern of the focused grid or collimator in the x-y plane, consisting of quadrilateral shaped openings formed by two intersecting sets of parallel lines, can be made in one piece (not including the border and other assembly parts), the easiest method is to expose the photoresist twice with two masks. The pattern of FIG. 4 is used as an example to assist in the explanation below. 1. For illustration purposes, the case where the central ray is located at the center of the grid or collimator, as shown in FIG. 25, which is marked by a virtual “+” sign 100, will be considered. Two imaginary reference lines 201 and 101 are drawn running through the “+” sign, parallel to grid walls A and B, respectively. 2. The grid or collimator pattern requires double exposure using two separate masks. The desired patterns for the two masks are shown in FIG. 26a and 26b. 3. The photoresist exposure procedure by the sheet x-ray beam is shown in FIGS. 27a and 27b. For the first exposure, an x-ray mask 730, with pattern shown in FIG. 26a or 26b, is placed on top of the photoresist 710 and properly aligned, as follows. In FIG. 27a, the sheet x-ray beam 700 is oriented in the same plane as the paper, and the reference lines 101 in FIGS. 26a or 26b of the x-ray masks 730 are parallel to the sheet x-ray beam 700. In FIG. 27b, the sheet x-ray beam 700 is oriented perpendicular to the plane of the paper, as are the reference lines of x-ray mask 730. The x-ray mask 730, photoresist 710, and substrate 720 form an assembly 750. The assembly 750 is positioned in such a way that the line 740 connecting the virtual “+” sign 100 with the virtual point x-ray source 62 is perpendicular to the photoresist 710. The angle α is 0° when the reference line 101 is in the plane of the x-ray source 700. To obtain the focusing effect in the photoresist 710 by the sheet x-ray beam 700, the assembly 750 rotates around the virtual point x-ray source 62 in a circular arc 760. This method will produce focused grids with opening that are focused to a virtual point above the substrate.There are situations when one would like to produce a defocused grid or collimator, with walls focused to a virtual point below the substrate as shown in FIG. 27c. In FIG. 27c, the sheet x-ray beam 700 is oriented perpendicular to the plane of the paper, as are the reference lines of x-ray mask 730. The assembly 750 is positioned in such a way that the line 740 connecting the virtual “+” sign 100 with the virtual point x-ray source 62 is perpendicular to the photoresist 710. The angle a α is 0° when the reference line 101 is in the plane of the x-ray source 700. To obtain the defocusing effect in the photoresist 710 by the sheet x-ray beam 700, the assembly 750 rotates around the virtual point x-ray source 62 in a circular arc 770. 4. For the second exposure, the second x-ray mask is properly aligned with the photoresist 710 and the substrate 720. The exposure method is the same as in FIGS. 27a and 27b or 27c. 5. To facilitate assembly and handling of a grid, a border is desirable. The border can be part of FIG. 20a or 20b; or it can use a third mask. The grid border mask should be aligned with the photoresist 710 and its exposure consists of moving the assembly 750 such that the sheet x-ray beam 700 always remains perpendicular to the photoresist 710, as shown in FIG. 30. The assembly 750 moves along a direction 780. 6. The rest of the fabrication steps are the same as in described in U.S. Pat. No. 5,949,850, referenced above.E.2. Exposure of Positive Photoresist Using Sheet X-Ray Beam Unfocused grids and collimators, with two sets of parallel walls and at lease one set of parallel walls is perpendicular to the substrate of any design and orientation, can be easily fabricated with one mask using a sheet x-ray beam. Photoresist/substrate is to be oriented at the appropriate angle α as the x-ray beam sweeps across the mask as shown in FIGS. 27a and 27d. Unfocused grids and collimators with both sets of parallel walls not perpendicular to the substrate will require double exposure with two masks consisting of lines, exposing as shown in FIG. 27d with one mask and repeat the step shown in FIG. 27d with the second mask. When grid size is too large to be made in one piece, sections of grid parts can be made and assembled from a collection of grid pieces. E.3. Exposure of Focused Grid Design Type I For Linear Motion or Focused Collimator and Each Layer of the Grid or Collimator is Assembled from Two or More Pieces If two or more pieces of the grid or collimator are required to make a large device, the exposure is more complicated. In this case, at least three masks are required to obtain precise alignment of the pieces. The desired exposure of the photoresist is shown in FIG. 29, using pattern 115 shown on the right-hand-side of FIG. 19 as an example. The effect of the exposure on the photoresist outside the dashed lines 202 is not shown. The desirable exposure patterns are the black lines 120 for one surface of the photoresist, and are the dotted lines 130 for the other surface. The location of the central x-ray is marked by the virtual “+” sign at 200. The shape of the left border is preserved and all locations of the grid or collimator wall are exposed. Although the procedures discussed above with regard to FIGS. 27a and 27b are generally sufficient to obtain the correct exposure near the grid or collimator joint using two masks, one for wall A and one for wall B, incorrect exposure may occur from time to time. This problem is illustrated in FIG. 30. The masks are made so as to obtain correct photoresist exposure at the surface of the photoresist next to the mask. The dotted lines 130 denote the pattern of the exposure on the other surface of the photoresist. Some portions of the photoresist will not be exposed 140, but other portions that are exposed 141 should not be. The effect of the exposure on the photoresist outside the dashed lines 202 is not shown. At least three x-ray masks are required to alleviate this problem and obtain the correct exposure. Each edge joint boundary requires a separate mask. These are shown in FIGS. 31a-31c. FIG. 31a shows a portion of the grid lines B as lines 150, which do not extend all the way to the grid or collimator joint boundary on the left. FIG. 31b shows a portion of the grid lines A as items 160, which do not extend all the way to the grid joint boundary on the left. FIG. 31c shows the mask for the grid joint boundary on the left. The virtual “+” 200 shows the location of the central ray 63 in FIGS. 31a-31c. The distances from the joint border to be covered by each mask depend on the grid dimensions, the intended grid height, and the angle. The exposures of the photoresist 710 by all three masks shown in FIGS. 31a-31c follow the method described above with regard to FIGS. 29a and 29b or FIGS. 29a and 29c. The three masks have to be exposed sequentially after aligning each mask with the photoresist. If this pattern is next to the border of the grid or collimator as shown in FIG. 32, then the grid boundary 180 can be part of the mask of the grid joint boundary on the left, as shown in FIG. 33. At a minimum, the grid border 180 consists of a wide grid border for structural support, may also include patterned outside edge for packaging, interlocks and peg holes for assembly and stacking. The procedure would be to expose the photoresist 710 by masks shown in FIGS. 31a and 31b following the method described in FIGS. 29a and 29b or FIGS. 29a and 29c. The exposure of the joint boundary section 170 in FIG. 33 follows the method described in FIGS. 29a and 29b or FIGS. 29a and 29c while the exposure of the grid border section 180 in FIG. 33 follows the method described in FIG. 30. The location of the joint of the two pieces can have many variation other than that is shown in FIG. 19. The masks, boarders and exposure methods have to be adjusted accordingly, but the concept remains the same. E.4. Exposure of Focused Grid Design Type II For Linear Motion The exposure of the photoresist for a “tall” type II grid pattern design for linear grid motion, such as those grid patterns illustrated in FIGS. 8, 10, 17, and so on, can be implemented based on the methods described in U.S. Pat. No. 5,949,850, referenced above. The grid is considered “tall” whenH sin (Φmax)>>s,where H is the height of a single layer of the grid, Φmax is the maximum angle for a grid as shown in FIGS. 2 and 3, and s is related to the thickness of the grid wall as shown in FIGS. 7, 8, 10 and 17. “High” grids are not easy to expose using long sheet x-ray beams when the same grid pattern is implement from top to bottom on the grid. As described in an earlier section, the grid shape shown in FIGS. 8, 10, 17, and so on, need only be just high enough to block the primary radiation without causing undesirable exposure. Using the grid pattern shown in FIG. 10 as an example, three x-ray masks, FIGS. 34a, 34b and 34c can be used for the exposure. Additional x-ray masks might be required for edge joints and borders. The exposure of the photoresist for the joints and borders would be the same as for that describing FIG. 33. The virtual “+” 210 shows the location of the central ray 63 in FIGS. 34a, 34b and 34c. The dashed lines 211 denote the reference line used in the exposure of the photoresist by sheet x-ray beam as described in FIGS. 29a and 29b or FIGS. 29a and 29c. The three masks have to be exposed sequentially after aligning each mask with the photoresist.E.5. Exposure of the Focused or Unfocused Grids and Collimators Using a Point Source The method to expose photoresist to obtain a focused or unfocused grid or collimator can be achieved using point, parallel UV or x-ray source. To obtain the correct exposure at each location on the photoresist, the photoresist/substrate has to be properly oriented with respect to the source by moving the photoresist/substrate. A description to obtain focused grid or collimator using point, parallel UV or x-ray source 703 is shown in FIGS. 35a and 35b. An optical mask can be used for UV exposure. An x-ray mask is needed for x-ray exposure. The layout of the mask can be the pattern needed for the grid or collimator, and the assembly of mask 731 and the photoresist/substrate have to be moved appropriately during the exposure. For unfocused grids and collimators, the orientation of the UV or x-ray source respect to the photoresist/substrate remains the same as the source sweeps across its surface. For focused grids and collimators, the assembly of mask and photoresist/substrate are moved in an arc to simulate the cone shape of the source located at a fixed imaginary point 64. E.6. Exposure of the Focused Grids and Collimators Using a Cone Beam Source The UV photoresist exposure method to obtain a focused grid or collimator with a cone beam UV source or a point parallel UV source that sweeps across the optical/resist simulating a cone beam is shown in FIG. 36. The assembly of the mask and the photoresist/substrate do not need to be moved during the exposure. F.1. Other Methods of Fabrication of Mold on Graphite for Electroplating/Electroforming for General Applications, as Well as for Grids and Collimators. For some grid and collimator applications the mold structure shown in FIG. 24c can be achieved by means other than lithography. The trenches, shown in FIG. 24c can sometimes be produced by mechanical machining, laser ablation, reactive ion etching, or other means. All the fabrication steps are the same as FIGS. 24a-24f, except step 24b. The mold material can be a photoresist or any other material that can be attached to the graphite. When the trenches are cut all the way through to the graphite looking like FIG. 24c, then the grid, collimator, or any other device can be fabricated by electroplating/electroforming following the same procedures as FIGS. 24d-24f. This is made possible by the conducting property of graphite substrate. F.2. Fabrication of Molds on Graphite for Casting With the appropriate choice of the mold material on graphite substrate and any appropriate methods to fabricate the trenches, the mold can be used to cast structures for general applications as well as for grids and collimator. The graphite substrate can be removed abrasively to release the grid or collimator. This would be possible for low melting temperature metals such as lead. A freestanding copper grid appropriate for mammography x-ray energies with parallel walls was made using deep x-ray lithography and copper electroplating/electroforming on graphite substrate. The exposure is performed using x-rays from the bending magnet beaniline 2-BM at the Advanced Photon Source of Argonne National Laboratory. A scanning electron micrograph (SEM) of the copper grid is shown in FIG. 37. The parameters of the grid are: 25 μm lamellae, 550 μm period, 1 mm high and 60×60 mm2 area including a 2.5 mm boarder. The results are described in the paper: O. V. Makarova, C.-M. Tang, D. C. Mancini, N. Moldovan, R. Divan, D. G. Ryding, and R. H. Lee, “Micorfabrication of Freestanding Metal Structures Released from Graphite Substrates,” Technical Digest of The Fifteenth IEEE International Conference on Micro Electro Mechanical Systems, Las Vegas, Nev., USA, Jan. 20-24, 2002, IEEE Catalog Number 02CH37266, ISBN: 0-7803-7185-2, pp. 400-402, and the entire contents is incorporated herein by reference. Although only a few exemplary embodiments of this invention have been described in detail above, those skilled in the art will readily appreciate that many modifications are possible in the exemplary embodiments without materially departing from the novel teachings and advantages of this invention. Accordingly, all such modifications are intended to be included within the scope of this invention as defined in the following claims.
	description	Shown in FIGS. 1 and 2 are two embodiments of the inventive device, differing from each other by the fact that the device shown in FIG. 1 defines an open circuit for the transportation of a cleaning fluid, while the device shown in FIG. 2 defines a closed circuit for said transportation. The device is arranged for the removal of a radioactive deposition from a fuel assembly 1 in a nuclear plant. For this purpose, the device comprises a container 2 accommodating the fuel assembly 1. Moreover, it comprises first means 3, 4, 5 which are arranged to feed a fluid into and through the container 2, the fluid being provided to release the deposition from the fuel assembly 1 by means of abrasion, and transport the radioactive deposition material thus released out of the container 2. The device also comprises second means 6, 7 which are arranged to receive the radioactive material transported out of the container 2 by the fluid. These second means 6, 7 thereby comprise a filtering device 6 by which a filter is arranged to separate the released radioactive material from the fluid. Moreover, they comprise a tube conduit 7 extending from the container 2 to the filter 6 and arranged for the transportation of the fluid and the released, radioactive material. The first means 3, 4, 5 comprise a feeding funnel 3 for solid particles, here ice particles, a first pump 4 in which the solid particles are mixed with a liquid, here water, and a second pump which pumps the liquid to the first pump 4. The first pump 4 is thereby arranged to suck into it the solid particles that are fed from the feeding funnel 3. The device is arranged in a basin 8, which is filled with a xcex3-radiation-dampening medium, here water. In particular, the container 2 which accommodates the fuel assembly 1, and the second means 6, 7 are arranged at a considerable depth in order to avoid any larger amount of radiation, mainly gamma radiation (xcex3-radiation), from leaking out from the basin via the liquid surface 9. According to FIG. 1, the second pump 5 providing the first pump 4 and the container 2 with liquid is connected to an external liquid source. Thereby, also the second means 6, 7 are connected, via a conduit 11, to an arrangement or the like (not shown) for the reception of the liquid fluid that has passed the filter 6 and thereby been separated from radioactive material. Accordingly, an open circuit exists. However, in the embodiment according to FIG. 2, the second pump 5 is connected to the filter 6 via a conduit 12, the pump 5 being arranged to pump the liquid which has passed through the filter 6 back to the first pump 4 and the container 2. Accordingly, FIG. 2 describes a closed system. It should however be mentioned, that the closed circuit according to FIG. 2, at least after having been used for a certain time, requires some kind of liquid draining as the provision of solid particles which then melt and form a liquid phase results in a continuous increase of the amount of liquid in the circuit according to FIG. 2. The first pump 4 is also arranged to suck into it a gas, in this case air, which then, together with the mixture of water and solid articles, is pumped into and through the container 2. Preferably, the first pump 4 is arranged to suck into it the gas and the solid particles. The fluid, thus comprising a liquid, here water, a gas, here air, and solid particles, here ice, is pumped into and through the container 2 which encloses the fuel assembly 1 which in its turn, in a way known per se, comprises a plurality of cladding tubes for nuclear fuel. Thanks to the effect of the gas (amongst others), giving rise to turbulence, a relatively rapid flushing through the container 2 can take place, with an abrasive action on the crud deposit which is present at the outside of the cladding tubes. Above all, loose crud is removed, and, for operational technical and economical reasons, this treatment should go on for a period as short as possible in order to make it possible to put the fuel assembly 1 into operation as soon as possible. Thereby, basin 8 is other than the basin in which the fuel assembly is arranged during operation. The time of treatment is approximately 15 minutes or less. The device also comprises means 14 arranged to transmit infrasounds towards the fuel assembly 1. Those means 14 suitably comprise an infrasound source of a conventional sort and are only schematically shown in FIGS. 1 and 2. Moreover, the fuel assembly 1 is rotatably arranged in the container 2, and the device comprises a member 15, here a schematically shown motor, arranged to rotate the assembly 1. The rotational axis of the fuel assembly 1 thereby coincides with the longitudinal axis thereof. The device also comprises means 16, schematically shown in FIG. 1 and 2, arranged to heat at least a part of the fluid which is fed into the container 2. The means 16 are preferably arranged to periodically heat the water which is delivered to the container 2. Suitably, when a flow of hot water into the container 2 is generated, no ice is fed into the fluid. In order to obtain a treatment time which is as short as possible, a suitable mixture of liquid, solid particles and gas needs to be guided into the container 2. During the treatment of a boiling water reactor assembly, the flow of liquid should be approximately 10-50 kg/s, preferably 15-35 kg/s and most preferably 20-25 kg/s. For a pressure water reactor assembly, the corresponding amounts are 50-200 kg/s, 75-150 kg/s, and 100-125 kg/s. Approximately 1 kg air is added for each 20 liter of liquid, here water. The ice flow is approximately 30-70% of the liquid flow. The amount of ice added is preferably as low as possible within this interval, and, if possible, even lower. The size of the solid particles, that is the ice particles, is in the range of 5xc3x975xc3x975 mm. Of course, a plurality of variants and embodiments will be obvious for the man skilled in the art without thereby leaving the very scope of the invention. The invention is also to be defined only by what is stated in the appended claims, the drawing and the rest of the description.
	claims	1. A container device for the long-term storage of hazardous material, particularly for the ultimate disposal of nuclear fuel, comprisingat least one elongate, cylindrical first containment body having a casing wall and end walls, the casing wall and the end walls defining a first compartment for accommodating at least one hazardous-material body formed by the hazardous material or containing or supporting the hazardous material, the first compartment comprising support means for supporting the hazardous-material body centrally in the first compartment and spaced from the casing wall and the end walls,an elongate, cylindrical second containment body having a casing wall and end walls, the casing wall and the end walls defining a cylindrical second compartment, the second compartment comprising support means for supporting the first containment body centrally in the second containment body and spaced from the casing wall and the end walls of the second containment body, andpassages provided in at least one of the end walls of each of the first and second containment bodies for the introduction of wet concrete in the first and second compartments for filling the space between, as regards the first containment body, the hazardous-material body and the walls defining the first compartment, and, as regards the second containment body, the space between the first containment body and the walls defining the second compartment. 2. A container device according to claim 1, comprisingan elongate, cylindrical third containment body having a casing wall and end walls, the casing wall and the end walls defining a cylindrical third compartment, the third compartment comprising support means for supporting the second containment body centrally in the third containment body and spaced from the casing wall and the end walls of the third containment body, andpassages provided in at least one of the end walls of the third containment body for the introduction of wet concrete in the third compartment for filling the space between the second containment body and the walls defining the third compartment. 3. A container device according to claim 2, comprisingan elongate, cylindrical fourth containment body having a casing wall and end walls, the casing wall and the end walls defining a cylindrical fourth compartment, the fourth compartment comprising support means for supporting the third containment body centrally in the fourth containment body and spaced from the casing wall and the end walls of the fourth containment body, andpassages provided in at least one of the end walls of the fourth containment body for the introduction of wet concrete in the fourth containment body for filling the space between the third containment body and the walls defining the fourth compartment. 4. A method for manufacturing a container device for the ultimate disposal of nuclear fuel elements arranged in a fuel assembly, comprising the steps of:introducing and fixing the nuclear fuel elements in a defined position in an essentially cylindrical container, wherein a length of the cylindrical container is substantially larger than a length of the nuclear fuel elements, and wherein a space is provided between the nuclear fuel elements and between a side and end walls of the cylindrical container, andembedding the nuclear fuel elements throughout the length thereof and at ends thereof in a casting compound which casting compound fills completely the space between the nuclear fuel elements and the side and end walls of the cylindrical container and between the individual nuclear fuel elements, wherein the embedding step includes the steps of forcing the casting compound into the container under a pressure in the range of 10 to 50 bar through one of the end walls, and discharging excess casting compound through one of an opposite end wall or the same end wall. 5. A method according to claim 4, in which the container is in an underwater position during the introduction of the nuclear fuel elements in the container and during the embedding of the nuclear fuel elements in the casting compound. 6. A method for manufacturing a container device for the long-term storage of hazardous material included in an elongate hazardous-material body, comprising the steps of:placing the hazardous-material body in an elongate, cylindrical first containment body having a casing wall and end walls, and fixing the hazardous-material body in a defined central position in the containment body which is spaced from the casing walls and the end walls of the containment body,embedding the hazardous-material body in the first containment body throughout a length thereof and at ends thereof in concrete, including the steps of introducing the concrete through one of the end walls and causing the concrete to completely fill the space between the hazardous-material body and the inside of the first containment body,placing the first containment body with the embedded hazardous-material body embedded therein in an elongate, cylindrical second containment body having a casing wall and end walls, and fixing the first containment body in a defined central position in the second containment body which is spaced from the casing and the end walls of the second containment body, andembedding the first containment body throughout a length thereof and at ends thereof in concrete, including the steps of introducing the concrete through one of the end walls of the second containment body and causing the concrete to fill completely the space between the first containment body and the inside of the second containment body. 7. A method according to claim 6, comprising the further steps of:placing the second containment body with the embedded first containment body therein in an elongate, cylindrical third containment body having a casing wall and end walls, and fixing the second containment body in a defined central position in the third containment body which is spaced from the casing and the end walls of the third containment body, andembedding the second containment body in the third containment body throughout a length thereof and at ends thereof in concrete, including the steps of introducing the concrete through one of the end walls of the third containment body and causing the concrete to fill completely the space between the second containment body and the inside of the third containment body. 8. A method according to claim 6, in which the embedding takes place under water. 9. A method according to claim 6, wherein the introducing the concrete into the first containment body through one of the end walls thereof takes place at a pressure of the concrete in the range of 10 to 50 bar. 10. A method according to claim 6, wherein the introducing the concrete into the first containment body through one of the end walls thereof takes place at a pressure of the concrete in the range of 10 to 50 bar, and wherein the introducing the concrete into the second containment body through one of the end walls thereof takes place at a pressure of the concrete in the range of 10 to 50 bar. 11. A method according to claim 7, wherein the introducing the concrete into the first containment body through one of the end walls thereof takes place at a pressure of the concrete in the range of 10 to 50 bar, and wherein the introducing the concrete into the third containment body through one of the end walls thereof takes place at a pressure of the concrete in the range of 10 to 50 bar. 12. A method according to claim 4, in which the casting compound is concrete.