Split (2)

train · 33.5k rows

patent_number stringlengths 0 9	section stringclasses 4 values	raw_text stringlengths 0 954k
051788233	summary	SUMMARY OF THE INVENTION The apparatus of this invention is primarily designed for use by the nuclear industry for decontamination of radioactive contaminated surfaces. The apparatus consists of a number of operational stages, one of which provides a continuous decontaminating liquid flow ranging from ambient temperatures up to +500 degree F. Suitable flow, temperature and pressure valves and gauges are provided for permitting the operator to select the optimum parameters for the clean up being performed. Included in the operational stages is a single vacuum power unit for creating an operational controlled recovery vacuum flow throughout the entire apparatus. Further stages of the recovery and discharge system including a critical mass collector and separator, demister, filters and absorbers, all having the construction and configuration necessary for performing the specific cleaning application as required. Simultaneously with the remote cleaning activity the recovery vacuum flow induced throughout the operational stages will pick up the liquid laden contamination removed from the surface being cleaned and transfer it to the collector and filtering units that in turn separates the air and liquid mixture with each being separately filtered and contained for disposal. PRIOR ART Past devices of a similar type to that described in this application have included a pressurized heat cleaning liquid and dispensing means associated with a vacuum recovery system such as is described and shown in U.S. Pat. No. 2,908,030, dated Oct. 13, 1959. While these prior devices also include operational controls, such controls are activated by the inherent functioning of the associated parts, such as a pressure relief valve controlling the heating and pressurizing of the cleaning liquid. The object of the present invention is to provide operational controls which are individually and collectively associated with each segment of the system and which are responsive to the results achieved thereby. Considering the decontaminating of radioactive material, the controls for the accumulation of a critical mass, which is the product of the operation of the system, is the principal object of this invention. Thus the collector of the decontaminated material is provided with a operational control responsive to a critical mass volume. The filter and demisting of the collected contaminated material is provided with a operational control responsive to a critical mass volume. Each of these controls, activated by the results of the operation of the system, will disable the system by interrupting the recovery vacuum source. The above object is achieved by the new and novel arrangement and association of structure hereinafter described. Other objects of the invention will be hereinafter made apparent.
	claims	1. Radiation therapy equipment including a radiation collimator unit for setting a desired irradiation field onto a target treating region by shielding radiation on unnecessary regions, the radiation therapy equipment comprising:a pair of movable collimator blocks, each block including a plurality of movable leaf plates, each leaf plate having a toothed configuration on an arc shaped tracking surface on the leaf plate;a plurality of driving gears configured to independently drive each of the plurality of leaf plates by engaging with the toothed configuration provided on each tracking surface of the plurality of leaf plates;a plurality of constant force spring units coupled to each of the plurality of driving gears configured to constantly apply a direct force to each of the plurality of driving gears so as to independently move each of the plurality of leaf plates in closing directions;a plurality of wires connected to each of the plurality of leaf plates; anda plurality of roll-up units configured to independently roll-up each of the plurality of wires so as to independently move each of the leaf plates in an opening direction against each of the closing forces of the plurality of constant force spring units. 2. In the radiation therapy equipment according to claim 1, wherein the radiation collimator unit comprises:a pair of upper collimator blocks provided at a near position of the radiation source, each block having a first toothed configuration on a first arc shaped tracking surface so as to move along a first moving track;a pair of driving gear units configured to drive the pair of upper collimator blocks by engaging with the toothed configurations provided on each of the tracking surfaces;a pair of lower collimator blocks provided under the pair of upper collimator blocks, each block being formed with a plurality of leaf plates, and each plate having a second toothed configuration on a second arc shaped tracking surface so as to orthogonally move crossing the first moving track;wherein each of the lower collimator blocks comprises:a plurality of driving gears configured to independently move each of the plurality of leaf plates engaging with the second toothed configuration;a plurality of constant force spring units coupled to each of the plurality of driving gears so as to independently apply a constant force in a closing direction;a plurality of wires connected to the plurality of leaf plates; anda plurality of roll-up units configured to independently roll-up the plurality of wires so as to independently move each of the plurality of leaf plates in an opening direction against the force applied by the plurality of constant force spring units. 3. The radiation therapy equipment according to claim 2, wherein each of the plurality of constant force spring units is comprised of two drums for winding up a plate spring, andeach of the plurality of constant force spring units is supported by a pair of rotation shafts; and each shaft penetrates each of center apertures of the two drums. 4. The radiation therapy equipment according to claim 3, wherein the plurality of constant force springs is supported by a plurality of pairs of rotation shafts arranged in a direction of the first arc shaped track. 5. The radiation therapy equipment according to claim 4, wherein first ones of the plurality of driving gears are coaxially supported on one pair of rotation shafts and second ones of the plurality of driving gears are coaxially supported on another pair of rotation shafts, and the first and second ones of the plurality of driving gears are arranged in staggered positions. 6. The radiation therapy equipment according to claim 3, wherein each of the plurality of driving gears engaged to each toothed configuration of the plurality of leaf plates is connected to one of the two drums forming the constant force spring. 7. The radiation therapy equipment according to claim 6, further comprising a control unit configured to independently control a moving amount of each of the leaf plates in the opening direction by each of the plurality of roll-up units so as to set an irradiation field of radiation in accordance with a position of each of the leaf plates and a target shape of a treatment region. 8. The radiation therapy equipment according to claim 1, wherein the plurality of leaf plates is respectively attached to a plurality of wires, andthe plurality of roll-up units independently roll-up the plurality of wires so as to move each of the leaf plates in an opening direction against the force of the spring unit in a closing direction. 9. The radiation therapy equipment according to claim 1, wherein each of the plurality of roll-up units includes a motor for winding up the wire so as to move the each of the leaf plates at a desired position in order to form the irradiation field. 10. The radiation therapy equipment according to claim 9, wherein each of the plurality of roll-up units winds up the wire to a corresponding one of the motors through a pulley and the plurality of roll-up units is freely installed in a space in the radiation collimator unit. 11. The radiation therapy equipment according to claim 1, further comprising a detection unit configured to detect a position or a moving amount of each of the leaf plates by a non-contacted detection of a specific pattern fixed on a tracking surface of each of the leaf plates. 12. The radiation therapy equipment according to claim 1, further comprising a magnetic sensor provided near the driving gears for acquiring detection data on a multi-pole magnetic pattern attached on a peripheral surface of the driving gear; anda moving amount of each of the plurality of leaf plates is controlled based on the detection data. 13. A radiation therapy equipment comprising:a radiation source;an upper irradiation collimator unit provided near the radiation source along an irradiation axis of the radiation source; anda lower irradiation collimator unit provided under the upper irradiation collimator unit along the irradiation axis so as to orthogonally cross a moving direction of the upper irradiation collimator unit; the upper irradiation collimator unit and the lower irradiation collimator unit are moved so as to determine an irradiation aperture for irradiating an object;wherein the lower irradiation collimator unit is comprised of:a plurality of movable leaf plates, each having a toothed configuration on one curved edge of the leaf plate;a plurality of driving gears for respectively engaging to each of the toothed configurations of the plurality of movable leaf plates;a plurality of spring units coupled to each of the plurality of driving gears for respectively applying a direct force to each of the plurality of movable leaf plates through each of the plurality of the driving gears so as to independently move in a closing direction of the irradiation aperture;a plurality of wires connected to each of the plurality of movable leaf plates; anda plurality of wire drive units configured to drive each of the plurality of wires so as to move in an opening direction of the irradiation aperture.
053501616	summary	BACKGROUND OF THE INVENTION The present invention relates generally to metal components and more particularly relates to metal components suitable for use in or near nuclear reactors. The properties of metal components in a nuclear reactor are affected by radiation exposure. For safety reasons, the extent of irradiation-induced change in reactor parts can be a significant factor in reactor design. One component of a nuclear reactor system in which consideration of radiation-induced changes is particularly important is in the fuel assembly. In a conventional reactor, the fuel is contained in rods which are grouped together and held in place by fuel assembly grids. The grids are structured to provide an individual channel for each rod. Coolant is circulated through the channels along the outer surface of the fuel rods. Adequate flow of coolant is needed in order to keep the fuel rods from overheating. Each rod in a fuel assembly grid is held in place within a particular channel by springs, usually cantilever or arched springs. The springs are specially designed to result in minimal disruption to the flow of coolant around the rods, while supporting the rods strongly enough to prevent vibration or longitudinal displacement due to flow forces. Cantilever springs frequently are preferred over arched springs because grids containing cantilever springs have less blockage of coolant flow, and can be made shorter, than grids containing arched springs. Both of these factors contribute to the desirable result of a relatively low pressure drop across grids having cantilever springs. The springs often are made of zircaloy, a zirconium-tin alloy. Cantilever springs made from zircaloy have a disadvantage, however, in that they tend to relax after only a short period of irradiation. Frequently, the load on the springs decreases to zero during one operational cycle of the fuel assembly. While the use of a high initial load may result in a somewhat longer retention period for a positive spring force, the degree of improvement is not substantial. In any event, the use of springs having a high initial load can be disadvantageous because the fuel rods must be inserted very carefully in order to avoid scoring the cladding of the rods. SUMMARY OF THE INVENTION An object of the invention is to provide a metal component which will undergo differential growth upon exposure to a neutron flux. Another object of the invention is to provide a metal component with improved fatigue resistance and a reduced tendency to crack due to radiation exposure. Another object of the invention is to provide a treated metal spring which has a smaller loss in spring force when subjected to a neutron flux than an untreated spring of otherwise similar structure and composition. Yet another object of the invention is to provide a nuclear fuel assembly grid spring made from zircaloy which will allow for relatively easy insertion of a fuel rod against an initial preload, and will maintain sufficient spring force during in-reactor use. Broadly stated, the present invention is a metal component having non-uniform material characteristics which cause the component to undergo differential growth upon exposure to a neutron flux. The metal component can be straight or have any number of folds or curves. The component preferably is a curved or bent sheet or rod suitable for use in a nuclear reactor, and more preferably is a cantilever spring. The type of metal and the thickness or cross section of the component are selected such that the shape of the component will change as a result of its differential growth. "Non-uniform material characteristics" of the component include differences in physical and/or chemical properties of various portions or layers of the component. These differences may be incorporated deliberately in the component at the time of manufacture, or may be brought about by physically and/or chemically treating selected portions of a component which previously had uniform material properties. Preferably, the non-uniformity is obtained by cold working selected portions of a component which previously had uniform material characteristics. The invention is based upon the inventor's recognition of the potential usefulness of the phenomenon of radiation-induced differential growth, i.e., that portions or layers of a metal component which have different material characteristics undergo different rates of growth upon exposure to a neutron flux. This differential growth occurs whether or not irradiation occurs at a constant temperature. By selecting an appropriate pattern of material characteristics for a particular component, the size and shape of the component can be caused to change in a desired way upon exposure to radiation. The change usually is gradual, and often begins immediately when irradiation commences. For example, when a selectively cold worked metal component is exposed to radiation, the portions of the metal component that have been cold worked "grow" faster than the portions that have not been cold worked. Thus, when a layer on the convex side of a curved metal component is cold worked, but a layer on the directly opposite concave side is not cold worked, the degree of curvature of the component increases upon exposure to radiation. On the other hand, if a layer on the concave side of the curved metal component according to the invention is cold worked, a layer on the directly opposite convex side is not cold worked, the degree of curvature of the component decreases upon exposure to radiation. Changes in shape other than changes in degree of curvature also can be effected according to the invention. In a preferred embodiment, a conventional cantilever spring is cold worked along the convex side of the curve at the base of the spring, while the directly opposite concave side of the curve is not cold worked. If the selectively cold worked spring is in an unloaded state and is exposed to radiation, the degree of curvature of the spring increases (the radius of curvature decreases) because the cold worked layer on the convex side of the spring "grows" faster than the untreated layer on the concave side of the spring. Along these same lines, if the selectively cold worked spring is loaded with a fuel rod and is exposed to radiation, the spring will exert a force against the loaded object which is due to the differential growth between the adjacent cold worked and untreated layers. This force will at least partially compensate for any relaxation of the spring which may occur due to radiation exposure. Another feature of the invention is that as a result of the cold working, the fatigue resistance of the component is increased, and cracking of the metal is substantially prevented.
	summary
053902181	summary	BACKGROUND OF THE INVENTION (a) Field of the Invention The present invention relates to a process for preparing a fuel pellet for nuclear reactor. (b) Description of the Prior Art The prior main process for preparing a fuel pellet for nuclear reactor is to make a powder of oxide, carbide or nitride having the same chemical form as the fuel by any means and press mold the power and sinter it. In this case, the contamination of manufacturing institution by finely divided powder is a problem because of handling powder. In particular, in case of handling plutonium, or in case of handling a recycling fuel in thorium cycle, as a substitute method for avoiding the handling of powder, pellet manufacturing method by the above described "gel particle pressing method" which is called "Sphere-Cal" [Literature 1, see below], "COGEPEL" (COnversion using GElation to produce PELlets) [Literature 2] or "SGMP" (Sol-Gel Microsphere Pellatization) [Literature 3], using a gel particle manufactured by a sol-gel method developed in the field of manufacturing of high-temperature gas-cooled reactor fuel sphere-pac fuel as a pressing raw material has been researched. In case of uranium dioxide or a mixed oxide containing uranium as a base, a high density of pellet as 95% T.D. was obtained by 1982 [Literature 2]. Herein, % T.D. is relative density for theoretical density. On the other hand, in case of thorium oxide or a mixed oxide containing thorium as a base, a high density of pellet was not obtained in that time because a dry gel particle obtained by the sol-gel method is in the glass state, hard and wrong in press characteristics [Literature 1 and 4) . In case of this thorium-base oxide, a high density of pellet could not be obtained until the press characteristics has been improved in the joint work of Germany and Brasil (1979-1983), by mixing finely divided powder of carbon and poly vinylalcohol in the sol and making the particle soft by oxidatively removing carbon by high temperature heat treatment of gel particle (in this case, as a sintering aid, calcium was previously added in the sol.) [Literature 5]. Thereafter it has been published 1986 that this method is successful even in the condition of not using polyvinyl alcohol [Literature 3]. The above literatures are as shown in the following Table 1. TABLE I ______________________________________ Literature 1: ORNL/TM-6906(1979) Literature 2: Trans. Am. Nucl. Soc., 40(1982)52-54 Literature 3: Nucl. Technol., 73(1986)84-95 Literature 4: J. Nucl. Mater., 92(1980)207-216 Literature 5: JUL-SPEZ-266(1984)86-121 Literature 6: J. Nucl. Sci. Technol., 25(1988)848-856 ______________________________________ SUMMARY OF THE INVENTION An object of the present invention is to provide a process by which a high density of sintered pellet of thorium oxide or a mixed oxide containing thorium as a base can be prepared without adding a sintering aid by developing a method of obtaining an easily press moldable soft gel particle without using an additive and a method of elevating a lubrication effect for further improving a press moldability.
	summary
053655613	abstract	Exposure control method and apparatus particularly suitably usable in an X-ray exposure apparatus, for exposing a mask and a wafer to radiation (X-rays) from a synchrotron to transfer a pattern formed on a mask onto the wafer, is disclosed. A shutter device for controlling passage and interception of the synchrotron radiation is provided between the synchrotron and the wafer. The shutter device includes a blade member having a leading edge and a trailing edge which are rectilinearly movable in a direction in which there exists non-uniformness in illuminance of the radiation. The leading edge is used to determine the timing of start of passage of the radiation to the wafer, while the trailing edge is used to determine the timing of interception of the radiation. The moving speeds of the leading edge and the trailing edge are controlled independently of each other to provide different exposure times for different portions of an exposure region on the wafer, in accordance with the non-uniformness in illuminance. By this, the amount of absorption of radiation by a resist material on the wafer can be make uniform throughout the exposure region.
047524413	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention This invention relates to pressurized water reactors and, more particularly, to a modular former for use in the inner barrel assembly of such reactors. 2. State of the Prior Art Baffle type structures have been known for use in the portion of a reactor which surrounds the reactor core, usually termed the lower barrel assembly. Such baffle structures typically are comprised of plates which are bolted together to form a unitary baffle structure. Since not positioned in the core outlet flow, such known baffle structures are not subjected to the same type of thermal transients and flow loads as can exist within the inner barrel assembly of a reactor, and particularly a reactor of the advanced design to which the present invention primarily is directed. Specifically, in a reactor of the type with which the modular former of the present invention is employed, the reactor internals of the inner barrel assembly includes several hundred rods, or rodlets, which are selectively movable in an axial, vertically oriented direction, into and out of the lower barrel assembly, to control the activity of the reactor core. The rodlets thus are positioned directly in the core outlet flow, rendering it critical to maintain substantially uniform distribution of the outlet flow from the reactor core in an axial direction as the flow passes along the rodlets and through the upper barrel assembly. The core outlet flow, moreover, at least potentially presents significant thermal stresses due to the core outlet flow transients and induces vibrations. Since these conditions and corresponding flow control requirements do not exist in conventional reactors, no corresponding structures are known in the prior art for performing the functions of the modular formers of the present invention. SUMMARY OF THE INVENTION A pressurized water nuclear reactor of the type with which the modular formers of the present invention are intended for use employs a large number of reactor control rods or rodlets, typically arranged in what are termed reactor control clusters (RCC) and, additionally, a large number of water displacer rods, or rodlets, similarly arranged in water displacer rodlet clusters (WDRC). For example, in one such reactor, an array of 185 such clusters containing a total of 2800 rodlets (i.e., the total of reactor control rods and water displacer rods) are mounted in parallel axial relationship within the inner barrel assembly. Each of these clusters moreover is received within a corresponding rod guide structure. In operation, it is desired to maintain the core outlet flow in an axial flow condition and in a substantially uniform distribution throughout the cross-section of the inner barrel assembly, as it passes through the inner barrel assembly, and thus prevents cross-flow conditions (i.e., core flow in a direction transverse of the rod guides). This is a critical requirement in reactors of such advanced designs in which the inner barrel is densely loaded with rodlets, as before noted. The geometry of the reactor vessel itself introduces a structural anomaly which is contrary to maintaining the desired, substantially uniform axial flow condition. Particularly, the circular configuration of the reactor vessel, including the inner barrel assembly, is geometrically incompatible with the generally rectangular or square cross-sectional configuration of the individual rod guides, and of an array thereof as stacked in closely adjacent relationship within the inner barrel. Thus, in the peripheral regions between the inside diameter of the cylindrical inner barrel assembly and the outer periphery of the array of rod guides, no rodlets are present, resulting in a nonuniform flow distribution and presenting at least the potential of turbulence and cross-flow conditions with attendant problems of vibration. The modular formers of the present invention thus are configured to be received and rigidly supported in these peripheral regions, to provide hydraulic resistance and thereby to maintain a primarily axial direction, and substantially uniform distribution, of the core outlet flow throughout the length of the rod guides within the inner barrel assembly. The formers thus are directly exposed to the core outlet flow and are potentially subjected to flow induced vibrations and significant thermal stresses due to core outlet flow transients. To accommodate these stringent operating and environmental conditions, the formers are of a modular configuration, each including upper and lower, horizontally and radially inwardly extending former plates interconnected by vertically (i.e., axially) extending corrugated columns which are welded at their opposite ends to the respective upper and lower former plates. In a specific embodiment herein disclosed, the modular formers are of two different configurations, respectively corresponding to the two different spacings, or shapes, of the peripheral regions, each extending for only a limited arcuate segment of the circumferential distance about the inside diameter of the inner barrel assembly. In the disclosed design, eight such modules, four of each of the two types, are disposed in a common horizontal rank, and three such ranks are disposed in vertically displaced positions, or elevations, within the inner barrel, thereby to obtain the proper pressure drop for assuring that the axial flow and uniform distribution conditions in the rod guide region are achieved. Each module is fabricated by welding so as to form a unitary structure prior to being mounted within the upper barrel assembly. Cantilever attachment elements are welded, in advance, onto the remote surfaces of the upper and lower former plates and include parallel-extending shanks. The shanks are inserted through corresponding holes provided therefore in the sidewall of the inner barrel, and then welded to the sidewall from the exterior of the inner barrel sidewall. The structural configuration and assembly of the modules, including the mounting of the cantilever attachment means thereon, readily adapts same to efficient, automated production, permitting complete assembly of the modules in advance of positioning same within the inner barrel; as well, installation of the modules can be performed quickly and easily, in view of the capability of the exterior welding of the cantilever attachment elements to the sidewall of the inner barrel. The welded, unitary construction of each modular former and the welded attachment to the inner barrel sidewall furthermore eliminates the use of bolts, such as are employed in prior art core baffle structures and the problem of maintaining preloads on such bolts, which problem is far more severe in the core outlet flow environment which exists within the inner barrel than the environment which exists within the region of the core itself. These and other objects and advantages of the modular formers for the inner barrel assembly of a pressurized water reactor in accordance with the invention will be more apparent from the following detailed description of the invention and the accompanying drawings.
	summary
	claims	1. A nuclear fuel comprising:a pellet comprised of compressed and densified grains of a fissile material selected from the group consisting of UN and U3Si2, the grains having grain boundaries; andan oxidation resistant additive coating at least a portion of the grain boundaries of the fissile material;wherein the additive has a melting point greater than the sintering temperature of the fissile material, the additive is melted into at least a portion of the grain boundaries, and the additive has a median particle size less than that of UN or U3Si2. 2. The nuclear fuel recited in claim 1 wherein the fissile material is UN, and the additive is selected from the group consisting of tungsten and alloys containing at least 50 atomic % thereof, and UO2. 3. The nuclear fuel recited in claim 1 wherein the additive is present in amounts less than 20 weight percent of the fissile material. 4. The nuclear fuel recited in claim 1 wherein the median particle size of the additive is less than 10% of the median particle size of the UN or U3Si2. 5. The nuclear fuel recited in claim 1 wherein the median particle size of the additive is less than 1% of the median particle size of the UN or U3Si2. 6. The nuclear fuel recited in claim 1 wherein the powdered fissile material is U3Si2, and the additive is selected from the group consisting of molybdenum, titanium, chromium, thorium, tungsten, niobium, and zirconium, alloys containing at least 50 atomic % thereof, BeO, and UO2. 7. A nuclear fuel comprising:a pellet comprised of compressed and densified grains of a fissile material UN, the grains having grain boundaries; andan oxidation resistant additive coating at least a portion of the grain boundaries of the fissile material;wherein the additive has a melting point lower than the sintering temperature of the fissile material, the additive is melted into at least a portion of the grain boundaries, and the additive is a borosilicate glass. 8. The nuclear fuel recited in claim 7 wherein the additive is present in amounts less than 20 weight percent of the fissile material. 9. A nuclear fuel comprising:a pellet comprised of compressed and densified grains of a fissile material U3Si2, the grains having grain boundaries; andan oxidation resistant additive coating at least a portion of the grain boundaries of the fissile material;wherein the additive has a melting point lower than the sintering temperature of the fissile material, the additive is melted into at least a portion of the grain boundaries, and the additive is a borosilicate glass. 10. The nuclear fuel recited in claim 9 wherein the additive is present in amounts less than 20 weight percent of the fissile material.
	summary
	abstract	This invention relates to a process and apparatus for growing agricultural products with a reduced abundance of radioactive carbon-14 (14C) by employing centrifugal separation of atmospheric gases to selectively remove carbon dioxide (CO2) with 14C. Agricultural products with reduced 14C content can be grown in controlled environments with filtered atmospheric gases 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. Bilateral and unilateral compression helikon vortex apparatus provide efficient and economical removal of CO2 with 14C from atmospheric gases with a single filtration pass, which is ideally suited for large scale agricultural production.
046474242	claims	1. A refueling system for a nuclear reactor facility, comprising: a refueling machine, comprising an outer stationary mast and a vertically movable inner mast, for removing and inserting fuel assemblies from and into the reactor core of said facility; means mounted upon each of said fuel assemblies for engaging and disengaging support structure within said nuclear reactor core so as to fixedly connect and disconnect said each of said fuel assemblies to and from said nuclear reactor core upon insertion and removal of said each of said fuel assemblies within and out of said nuclear reactor core; means mounted upon said vertically movable inner mast of said refueling machine for gripping said each of said fuel assemblies so as to facilitate the transportation of said each of said fuel assemblies during said insertion and removal of said each of said fuel assemblies into and out of said nuclear reactor core; first rotary drive means mounted upon said outer stationary mast of said refueling machine; and second rotary drive means, separate and distinct from said gripping means, mounted upon said vertically movable inner mast of said refueling machine for engaging said first rotary drive means of said outer stationary mast and actuating said connecting and disconnecting means mounted upon said each of said fuel assemblies independently of the operation of said gripping means mounted upon said vertically movable inner mast of said refueling machine. said connecting and disconnecting means of said any one of said fuel assemblies comprises a threaded screw. said threaded screw has hexagonal socket means defined therein; and said second rotary drive means includes a hexagonally headed rod for engagement within said hexagonal socket means of said threaded screw. said second rotary drive means comprises gear means operatively connected to said rod. said first rotary drive means comprises a first gear rotatably mounted upon said stationary mast, and said second rotary drive means comprises a second gear operatively connected to said rod and said first gear for transmitting torque from said first gear to said rod. tube means connected to said gear means and operatively connected to said rod for transmitting torque from said gear means to said rod. said rod is disposed within a housing having substantially square-shaped regions in cross-section; and said tube means is substantially square in cross-section whereby torque can be transmitted therebetween. said connecting and disconnecting means is co-axially disposed with respect to said stationary mast and said inner mast. first winch drive means for vertically controlling the displacement of said inner mast; and second winch drive means for vertically controlling the displacement of said connecting and disconnecting means independently of said first winch drive means. means disposed upon said any one of said fuel assemblies for preventing inadvertent disconnection of said any one of said fuel assemblies from said support structure. first ratchet means disposed upon said any one of said fuel assemblies; said connecting means of said any one of said fuel assemblies comprises a threaded screw; and second ratchet means disposed upon said threaded screw for cooperation with said first ratchet means. slot means defined within said threaded screw; and sleeve means slideably disposed upon said threaded screw and having means engaged within said slot means for preventing relative rotation between said screw and said sleeve means; said second ratchet means being disposed upon said sleeve means whereupon said threaded screw is prevented from rotating relative to said any one of said fuel assemblies when said first and second ratchet means are engaged and when torque is applied to said screw in a direction tending to force said ratchet teeth into relative engagement. spring biasing means for tending to maintain said ratchet teeth engaged. a refueling machine, comprising an outer stationary mast and an inner mast vertically movable between a raised post-removal or pre-insertion position and a lowered insertion or pre-removal position, for removing and inserting fuel assemblies from and into the reactor core of said facility; means rotatably mounted upon each of said fuel assemblies for engaging and disengaging support structure within said nuclear reactor core so as to fixedly connect and disconnect said each of said fuel assemblies to and from said nuclear reactor core upon insertion and removal of said each of said fuel assemblies within and out of said nuclear reactor core; means mounted upon said vertically movable inner mast of said refueling machine for gripping said each of said fuel assemblies so as to facilitate transportation of said each of said fuel assemblies during said insertion and removal of said each of said fuel assemblies into and out of said nuclear reactor core; first rotary drive means mounted upon said outer stationary mast of said refueling machine; and second rotary drive means, separate and distinct from said gripping means, rotatably mounted upon said vertically movable inner mast of said refueling machine for rotatably engaging said first rotary drive means of said outer stationary mast when said inner mast is disposed at said lowered position and actuating said connecting and disconnecting means rotatably mounted upon said each of said fuel assemblies independently of the operation of said gripping means mounted upon said vertically movable inner mast of said refueling machine. said connecting and disconnecting means of said each one of said fuel assemblies comprises a theaded screw. said first rotary drive means comprises a first gear; and said second rotary drive means comprises a second gear operatively engaged with said first gear. said threaded screw is provided with a hexagonal socket; and said second rotary drive means comprises a torque tube fixed to said second gear and having a square-shaped configuration, and a rod assembly the lower portion of which comprises a rod having a hexagonally-shaped head portion for engagement with said hexagonal socket of said threaded screw, and the upper portion of which has a square-shaped means for engagement with said torque tube so as to transmit torque from said second gear and said torque tube to said rod and said threaded screw. said outer stationary mast, said inner vertically movable mast, said torque tube, said rod, and said threaded screw are all substantially co-axially disposed with respect to each other. slot means defined within a sidewall portion of said outer stationary mast; and said first gear is mounted substantially externally of said outer stationary mast with a radial portion of said first gear projecting through said slot means of said outer stationary mast so as to be engageable with said second gear of said second rotary drive means. a refueling machine, comprising an outer stationary mast and a vertically movable inner mast, for removing and inserting fuel assemblies from and into the reactor core of said facility; means mounted upon each of said fuel assemblies for engaging and disengaging support structure within said nuclear reactor core so as to fixedly connect and disconnect said each of said fuel assemblies to and from said nuclear reactor core upon insertion and removal of said each of said fuel assemblies within and out of said nuclear reactor core; means mounted upon said vertically movable inner mast of said refueling machine for gripping said each of said fuel assemblies so as to facilitate transportation of said each of said fuel assemblies during said insertion and removal of said each of said fuel assemblies into and out of said nuclear reactor core; first rotary drive means mounted upon said outer stationary mast of said refueling machine; and second rotary drive means, separate and distinct from said gripping means yet coaxially mounted therewith, upon said vertically movable inner mast of said refueling machine for engaging said first rotary drive means of said outer stationary mast and actuating said connecting and disconnecting means mounted upon said each of said fuel assemblies independently of the operation of said gripping means mounted upon said vertically movable inner mast of said refueling machine. 2. A refueling system as set forth in claim 1, wherein: 3. A refueling system as set forth in claim 2, wherein: 4. A refueling system as set forth in claim 3, wherein: 5. A refueling system as set forth in claim 4, wherein: 6. A refueling system as set forth in claim 4, further comprising: 7. A refueling system as set forth in claim 6, wherein: 8. A refueling system as set forth in claim 1, wherein: 9. A refueling system as set forth in claim 1, further comprising: 10. A refueling system as set forth in claim 1, further comprising: 11. A refueling system as set forth in claim 10, wherein said means for preventing inadvertent disconnection, comprises: 12. A refueling system as set forth in claim 11, further comprising: 13. A refueling system as set forth in claim 11, further comprising: 14. A refueling system for a nuclear reactor facility, comprising: 15. A refueling system as set forth in claim 14, wherein: 16. A refueling system as set forth in claim 15, wherein: 17. A refueling system as set forth in claim 16, wherein: 18. A refueling system as set forth in claim 17, wherein: 19. A refueling system as set forth in claim 16, further comprising: 20. A refueling system for a nuclear reactor facility, comprising:
	abstract	A system and method are used to recycle gases in a lithography tool. A first chamber includes an element that emits light based on a first gas. A second chamber uses the emitted light to perform a process and includes the second gas. The first and second gases converge between the two chambers, and at least one of the gases is pumped to a storage device. From the storage device, at least one of the two gases is recycled either within the system or remote from the system and possibly reused within the system. A gaslock can couple the first chamber to the second chamber. A gas source supplies a third gas between the first and the second gas in the gaslock, such that the first gas is isolated from the second gas in the gaslock. The first, second, and/or third gas can be pumped to the storage device and routed to the recycling device. The first, second, and/or third gas can be recycled for reuse to form the emitting light.
	claims	1. An apparatus for inspecting an underwater pipeline to determine wall thickness or information about contents of the underwater pipeline, the apparatus comprising:a gamma radiation source;an array of detector units;a data processor, and,a buoyancy material,wherein the gamma radiation source and the array of detector units are mounted to enable an underwater pipeline to be interposed between the gamma radiation source and the array of detector units so that gamma radiation emitted by the gamma radiation source passes along a plurality of paths through a portion of the underwater pipeline and impinges upon the array of detector units,wherein the data processor is configured to acquire data at a plurality of radially offset positions around the underwater pipeline to acquire density data at a variety of angles through the underwater pipeline when the underwater pipeline is interposed between the gamma radiation source and the array of detector units in order to produce a representation of the underwater pipeline or contents of the underwater pipeline using the density data. 2. The apparatus according to claim 1, wherein at least one of the gamma radiation source and the array of detector units are rotatable around a circumference of the underwater pipeline when acquiring data. 3. The apparatus according to claim 1, further comprising a hinged housing that is configured to be opened and closed around the underwater pipeline. 4. The apparatus according to claim 1, wherein the representation is a representation of a composition of the underwater pipeline or the contents of the underwater pipeline. 5. The apparatus according to claim 1, further comprising a tomography algorithm for building the representation of the underwater pipeline or the contents of the underwater pipeline.
	claims	1. An apparatus comprising:an input interface configured to receive an input signal associated with at least one stage of an impeller and with two or more frequencies associated with operation of the impeller;a processor configured to identify a specific failure mode in the impeller using the input signal; andan output interface configured to provide an indicator identifying a health of the impeller;wherein the processor is configured to identify the specific failure mode by:determining a family of frequencies related to at least one of the two or more frequencies associated with operation of the impeller, the family of frequencies including a vane pass frequency and its harmonics;decomposing the input signal by performing a transform of the input signal and isolating components of the transformed input signal into selected frequency bands;reconstructing an impeller signal using portions of the decomposed input signal within the selected frequency bands associated with the family of frequencies;comparing the reconstructed impeller signal to a baseline signal; andcategorizing the comparison based on failure mode rules including effects that the specific failure mode has on the two or more frequencies associated with operation of the impeller. 2. The apparatus of claim 1, wherein:the two or more frequencies associated with operation of the impeller comprise the vane pass frequency, a rotating shaft speed frequency, a shaft sideband frequency, and a background noise frequency; andthe reconstructed impeller signal comprises at least one of:a signal associated with the vane pass frequency and its harmonics;a signal associated with the rotating shaft speed frequency;a signal associated with the shaft sideband frequency; anda signal associated with the background noise frequency. 3. The apparatus of claim 1, wherein the processor is configured to decompose the input signal through a plurality of band-pass filters. 4. The apparatus of claim 3, wherein the processor is configured to reconstruct the impeller signal by:determining maximum and minimum amplitudes in the selected frequency bands in outputs of the band-pass filters;combining the maximum amplitudes to produce a first matrix;combining the minimum amplitudes to produce a second matrix; andreconstructing multiple impeller signals using the first and second matrices. 5. The apparatus of claim 1, wherein the processor further is configured to:normalize the reconstructed signal with the baseline signal; andapply a feature fusion technique to obtain a value for use by the indicator. 6. The apparatus of claim 1, wherein the processor is further configured to store a portion of the input signal, corresponding to normal operation of the impeller, as the baseline signal. 7. The apparatus of claim 1, wherein the indicator identifying the health of the impeller comprises at least one of: an impeller wear indicator, an impeller crack indicator, a cavitation indicator, and an impeller health indicator. 8. The apparatus of claim 1, wherein:the input signal comprises at least one of: vibration information in a time domain and speed information in the time domain associated with the impeller;the input interface comprises multiple input interfaces; andthe processor comprises an artificial intelligence portion, andthe reconstructed impeller signal comprises a time domain signal. 9. A system comprising:a plurality of sensors configured to measure one or more characteristics of an impeller; andan impeller condition indicator device comprising:an input interface configured to receive input signals from the sensors, each of the input signals associated with at least one stage of the impeller and with two or more frequencies associated with operation of the impeller;a processor configured to identify a specific failure mode in the impeller using the input signals; andan output interface configured to provide an indicator identifying a health of the impeller;wherein the processor is configured to identify the specific failure mode by, for each of the input signals:determining a family of frequencies related to at least one of the two or more frequencies associated with operation of the impeller, the family of frequencies including a vane pass frequency and its harmonics;decomposing the input signal by performing a transform of the input signal and isolating components of the transformed input signal into selected frequency bands;reconstructing an impeller signal using portions of the decomposed input signal within the selected frequency bands associated with the family of frequencies;comparing the reconstructed impeller signal to a baseline signal; andcategorizing the comparison based on failure mode rules including effects that the specific failure mode has on the two or more frequencies associated with operation of the impeller. 10. The system of claim 9, wherein:the two or more frequencies associated with operation of the impeller comprise the vane pass frequency, a rotating shaft speed frequency, a shaft sideband frequency, and a background noise frequency; andthe reconstructed impeller signal comprises at least one of:a signal associated with the vane pass frequency and its harmonics;a signal associated with the rotating shaft speed frequency;a signal associated with the shaft sideband frequency; anda signal associated with the background noise frequency. 11. The system of claim 9, wherein the processor is configured to, for each of the input signals, decompose the input signal through a plurality of band-pass filters. 12. The system of claim 11, wherein the processor is configured to reconstruct the impeller signal by:determining maximum and minimum amplitudes in the selected frequency bands in outputs of the band-pass filters;combining the maximum amplitudes to produce a first matrix;combining the minimum amplitudes to produce a second matrix; andreconstructing multiple impeller signals using the first and second matrices. 13. The system of claim 9, wherein the processor is further configured to store a portion of each of the input signals, corresponding to normal operation of the impeller, as the baseline signal. 14. The system of claim 9, wherein the indicator identifying the health of the impeller comprises at least one of: an impeller wear indicator, an impeller crack indicator, a cavitation indicator, and an impeller health indicator. 15. A method for identifying a health of an impeller corresponding to a specific failure mode of the impeller, the method comprising:receiving at one or more processing devices an input signal comprising at least one of vibration and speed infonnation corresponding to at least one stage of the impeller, the input signal associated with two or more frequencies associated with operation of the impeller;determining a family of frequencies corresponding to at least one of the two or more frequencies associated with operation of the impeller, the family of frequencies including a vane pass frequency and its harmonics;decomposing the input signal by performing a transform of the input signal and isolating components of the transformed input signal into selected frequency bands;reconstructing an impeller signal using portions of the decomposed input signal within the selected frequency bands associated with the family of frequencies;comparing the reconstructed impeller signal to a baseline signal;categorizing the comparison based on failure mode rules including effects that the specific failure mode has on the two or more frequencies associated with operation of the impeller;outputting an indicator identifying the health of the impeller; andin response to determining the reconstructed impeller signal differs from the baseline signal by a threshold amount, outputting an indicator identifying a fault. 16. The method of claim 15, wherein:the two or more frequencies associated with operation of the impeller comprise the vane pass frequency, a rotating shaft speed frequency, a shaft sideband frequency, and a background noise frequency; andthe reconstructed impeller signal comprises at least one of:a signal associated with the vane pass frequency and its harmonics;a signal associated with the rotating shaft speed frequency;a signal associated with the shaft sideband frequency; anda signal associated with the background noise frequency. 17. The method of claim 15, wherein decomposing the input signal comprises decomposing the input signal through a plurality of band-pass filters; andwherein reconstructing the impeller signal comprises:determining maximum and minimum amplitudes in the selected frequency bands in outputs of the band-pass filters;combining the maximum amplitudes to produce a first matrix;combining the minimum amplitudes to produce a second matrix; andreconstructing multiple impeller signals using the first and second matrices. 18. The method of claim 15, further comprising:normalizing the reconstructed signal with the baseline signal; andapplying a feature fusion technique to obtain a value for use by the indicator. 19. The method of claim 15, further comprising:storing a portion of the at least one of vibration and speed information corresponding to normal operation of the impeller as the baseline signal. 20. The method of claim 15, wherein the threshold amount comprises a first amount associated with a warning threshold and a second amount associated with an alarm threshold. 21. The method of claim 16, wherein the failure mode rules identify at least one of:a high flow cavitation failure mode having an increasing effect on the background noise frequency and a decreasing effect on the vane pass frequency;an impeller wear failure mode having an increasing effect on the shaft sideband frequency and an increasing effect on the vane pass frequency; andan impeller crack failure mode having an increasing effect on the shaft sideband frequency and a decreasing effect on the vane pass frequency.
	description	The present invention relates to a buffer body attached to a cask. A spent nuclear fuel assembly having burnt at the end of a nuclear fuel cycle is referred to as “recycle fuel”. Since the recycle fuel contains highly radioactive materials such as FPs and needs to be thermally cooled, it is cooled in a cooling pit in a nuclear power plant for a predetermined period of time. The cooled recycle fuel is then contained in a cask, which is a shielding container, and the cask is transported and stored in reprocessing facilities or intermediate storage facilities by truck, ship or the like. If the cask is transported to the reprocessing or intermediate storage facilities, the recycle fuel containing highly radioactive materials is contained in the cask. Therefore, the cask should be kept shielded and hermetically sealed as much as possible unless it is unnecessary to do so. To do so, during transport of the cask, the cask is protected by covering both ends of a cask main body with cask buffer bodies. By doing so, even if the cask falls, for example, the cask is kept shielded and hermetically sealed. As an example of the cask buffer body of this type, Patent Document 1 discloses a cask buffer body having an interior filled with a wood material. Patent Document 1: Japanese Patent Application Laid-open No. 2003-315493. The cask buffer body disclosed in the Patent Document 1 uses the wood material as a shock absorber that absorbs a shock energy by crashing the wood material. Since the wood material is a natural material and a fiber assembly, reproducibility of a crash behavior of the wood material is poor and it is difficult for the wood material to exhibit a stable shock absorbing performance. The present invention has been achieved to solve the conventional problems. It is an object of the present invention to provide a cask buffer body capable of exhibiting a stable shock absorbing performance. To solve the above problems and to achieve the goal, a cask buffer body according to one aspect of the present invention includes a shock absorber configured to be attached to a cask that stores a recycle fuel. The shock absorber absorbs a shock against the cask by being deformed, and includes a space for adjusting a shock absorbing capability. In this cask buffer body, the space for adjusting the shock absorbing characteristic is provided in the shock absorber that constitutes the buffer body, and the shock absorbing characteristic of the shock absorber is adjusted. It is thereby possible for the shock absorber to include the shock absorbing characteristic and to exhibit a stable shock absorbing performance. The shock absorbing characteristic is a shock energy absorbing characteristic relative to a compression amount of the shock absorber. According to the present invention, the space is a hole formed in each of the shock absorber blocks. By doing so, shearing, cracking, and crashing can be generated in the shock absorber with this hole set as a point of origin, and absorption of the shock energy by the shock absorber can be accelerated. Furthermore, since the space can reduce a rigidity of an entire shock absorber, lockup of the shock absorber, i.e., a sudden rise of a reaction force of the shock absorber can be delayed. According to the present invention, a cross-sectional shape of the hole includes an angular portion. The cross-sectional shape of the hole refers to a shape within a cross section orthogonal to a formation direction of this hole. According to the present invention, a dimension of the hole is changed toward a direction in which the shock is input to the shock absorber. Accordingly, right after the shock is input to the shock absorber, it is possible to promptly crash the shock absorber, and sufficiently absorb a shock energy. In addition, by making it more difficult to crash the shock absorber as crashing progresses, a motion of the cask can be effectively stopped. According to the present invention, the space is a wedge notch, and the wedge notch is formed at least on a side of the shock absorber on which the shock is input to the shock absorber. As seen in a cask buffer body according to the next invention, the space can be a notch formed in the shock absorber block. According to the present invention, the shock absorber is formed by combining a plurality of shock absorber blocks made of a wood material. The shock absorber constituting this buffer body is constituted by combining a plurality of shock absorber blocks each consisting of the wood material. In addition, a space is provided in each of these shock absorber blocks so as to adjust the shock absorbing characteristic of the shock absorber. By doing so, even if the shock absorber consists of the material, or particularly the wood material, to which an excessive initial stress occurs at the moment of shock or the crash behavior of which has poor reproducibility, the shock absorber can include the shock absorbing characteristic and exhibit a stable shock absorbing performance. According to the present invention, the shock absorber is formed by combining a plurality of shock absorber blocks made of a wood material, in an annular shape, and the shock absorber blocks are integrated by winding a block binding unit around a circumferential groove formed on an outer circumference of the shock absorber in the annular shape. By thus constraining the shock absorber blocks constituting the cask buffer body, the respective shock absorber blocks can be firmly fixed by a tensile force of the block binding unit. In addition, shearing, cracking, and crashing can be generated in the shock absorber blocks with grooves set as points of origin, and absorption of the shock energy by the shock absorber blocks can be accelerated. According to the present invention, the shock absorber is formed by combining a plurality of shock absorber blocks made of a wood material, in an annular shape. Each of the shock absorber bocks includes a shock absorber block A having a diametral outside dimension smaller than a diametral inside dimension; and a shock absorber block B having a diametral outside dimension larger than a diametral inside dimension. A compressive strength of the shock absorber block A is stronger than a compressive strength of the shock absorber block B. By thus making the shock absorber block A, which consists of a material (e.g., oak) having a high compressive strength, have the diametral outside larger in area than the diametral inside, a reaction force within the shock absorber block A gradually rises if a shock load is applied to the block A. The shock absorber block B can suppress a motion of the shock absorber block A toward a circumferential direction of the shock absorber. As a result, a peak load that tends to occur in an initial period of the shock can be suppressed to be low and the shock load can be absorbed by a predetermined crash margin. According to the present invention, the space is provided in such a manner that the space divides or passes through fibers of the wood material constituting each of the shock absorber blocks. Since the space that divides or penetrates a formation direction of fibers that has a great influence on the crash characteristic of the wood material is thus provided, it is possible to make the shock absorbing characteristics uniform and exhibit a stable shock absorbing performance. According to the present invention, the space is provided substantially in parallel to fibers of the wood materials constituting each of the shock absorber blocks. By thus providing the space in parallel to the fiber direction of the wood material, crashing of the shock absorbing block can be generated more easily in response to the compressive load. The shock load can be thereby absorbed more easily if the shock load acts as the compressive load. According to the present invention, the space is a hole formed in each of the shock absorber blocks. By doing so, shearing, cracking, and crashing can be generated in the shock absorber with this hole set as a point of origin, and absorption of the shock energy by the shock absorber can be accelerated. According to the present invention, a cross-sectional shape of the hole includes an angular portion. By generating shearing, cracking, and crashing in the shock absorber consisting of the wood material with the angular portion included in the hole set as a point of origin, this cask buffer body can accelerate absorption of the shock energy by the shock absorber. According to the present invention, the angular portion is formed on a side of the shock absorber on which the shock is input to the shock absorber. Since the angular portion is thus formed on the side of the shock absorber on which the shock is input to the shock absorber, it is possible to effectively generate shearing, cracking, and crashing in the shock absorber consisting of the wood material with this angular portion set as a point of origin, and accelerate absorption of the shock energy by the shock absorber. According to the present invention, the space is a wedge notch, and the wedge notch is formed at least on a side of the shock absorber on which the shock is input to the shock absorber, in such a manner that a top of the wedge notch is oriented to a direction in which the shock is input to the shock absorber. By thus forming the top of the notch to be oriented toward the direction in which the shock is input to the shock absorber, it is possible to effectively generate shearing, cracking, and crashing in the shock absorber consisting of the wood material with the top of this notch set as a point of origin, and accelerate absorption of the shock energy by the shock absorber. According to the present invention, the space is a notch formed toward a direction in which the shock is input to the shock absorber. This notch can reduce an apparent cross-sectional area of the shock absorber and reduce an initial peak load when the shock acts on the shock absorber. According to the present invention, the space is a notch formed perpendicular to a fiber direction of the wood material. Since this notch can reduce the rigidity of the entire shock absorber, lockup of the shock absorber, that is, a sudden increase of the reaction force within the shock absorber can be delayed. According to the present invention, the shock absorber includes a first shock absorber group that is obtained by combining the shock absorber blocks in such a manner that a fiber direction of the wood material is parallel to a shock input direction, that absorbs the shock in a direction parallel to an end surface of the cask, and that consists of a first material; a second shock absorber group that absorbs the shock in a direction perpendicular to or oblique with respect to the end surface of the cask, and that consists of a second material of which a compressive strength is weaker than a compressive strength of the first material; and a third shock absorber group that absorbs the shock in a direction perpendicular to the end surface of the cask, and that consists of a third material of which a compressive strength is weaker than a compressive strength of the second material. The space is provided at least in the first shock absorber group. In this cask buffer body, the hole, the notch or the other space is provided in the first shock absorber group consisting of the first material (wood material) having the highest compressive strength. It is thereby possible to adjust the shock absorbing characteristic of the first shock absorber group and stably exhibit the shock absorbing performance. The cask buffer body according to the present invention can stably exhibit the shock absorbing performance. 1 Cask 1t End 1tp End surface 1b Barrel main body 4tp Secondary lid end surface (Occasionally, third lid end surface) 6w Outer plate 6 Buffer body 6o Opening 7 Attachment hole 10p Veneer 10h1 Plate material 10s Plate piece 10, 10a, 10b, 10c, 10d, 10e, 10f, 10g, 10h, 10i, 10j, 10k, 10k′, 10l, 10m, 10n, 10o, 10p, 10q, 10r, 10s(10s1, 10s2), 10t, 10x, 10y, 10z, 10A, 10B First shock absorber block 11 Second shock absorber block 12 Third shock absorber block 13 Fourth shock absorber block 14 Fifth shock absorber block 15 Sixth shock absorber block 16 Seventh shock absorber block 17 Eighth shock absorber block 20, 22, 23 Hole 21 Bottomed hole 24 Angular hole 25 Notch 26 Groove 27 Slot B1 First shock absorber B2 Second shock absorber B3 Third shock absorber B4 Fourth shock absorber B5 Fifth shock absorber B6 Sixth shock absorber B7 Seventh shock absorber B8 Eighth shock absorber The present invention will be explained below with reference to the accompanying drawings. This invention is not limited by best modes for carrying out the invention. Furthermore, constituent elements in embodiments described below include elements easily ascertained by those skilled in the art, or substantially the same elements. The present invention is suitable particularly for a case of using a wood material as a shock absorber of a cask buffer body. However, the invention is not limited to the case. For example, the present invention is also applicable to a case of using a metal material, FRP or the like for the shock absorber of the cask buffer body. The present invention is applicable even to a case that a fiber direction of the wood material that constitutes a second shock absorber assembly is parallel to or orthogonal to an oblique falling direction of the cask. FIG. 1 is an explanatory view of a configuration of a cask according to a first embodiment of the present invention. A cask 1 is employed to contain a recycle fuel inside, and transported and stored. A space called “cavity” 1c is formed inside a barrel main body 1b of the cask 1 and a basket 2 is stored in the cavity 1c. The basket 2 is constituted by, for example, bundling square pipe steels each having square cross-sectional internal and external shapes and includes a plurality of lattice cells. A recycle fuel assembly 5 is stored in each lattice cell of the basket 2. The barrel main body 1b is a forged part consisting of carbon steel that exhibits a gamma ray shielding function. Alternatively, stainless steel can be employed instead of the carbon steel. After storing the basket 2 in which the recycle fuel assembly 5 is contained, in the cavity 1c, a primary lid 3 and a secondary lid 4 are attached to an opening of the barrel main body 1b, thereby hermetically sealing the cavity 1c. At this time, gaskets are provided between the barrel main body 1b and the primary lid 3 and between the barrel main body 1b and the secondary lid 4, respectively, so as to ensure a hermetically sealing performance. Furthermore, a ternary lid is often attached to the opening depending on the type of the cask. FIGS. 2A and 2B are perspective views of a form of the cask during transport. FIG. 3 is an explanatory view of an example of transporting the cask by train. As shown in FIGS. 2A and 2B, at the time of transporting the cask 1, cask buffer bodies (hereinafter, “buffer bodies”) 6 are attached to both ends of the cask 1, respectively, so as to prevent possible falling, collision or the like during the transport. If the cask 1 is transported by train, the cask 1 having the buffer bodies 6 attached to the respective ends thereof is mounted on a transport stand 9 and installed in a dedicated freight car. The cask 1 is transported with trunnions 8 provided at the cask 1 fixed to the transport stand 9. As the buffer bodies 6, buffer bodies having each square corner formed into a circular arc as shown in FIG. 2A are used. Alternatively, as shown in FIG. 2B, circular buffer bodies 6′ are used. Furthermore, buffer bodies of various shapes can be used according to specifications of the cask 1. FIG. 4A is an explanatory view of definition of a central axis of the cask. In the first embodiment, the central axis Z of the cask 1 is parallel to a longitudinal direction of the cask 1 (that is, a longitudinal direction of the recycle fuel in a state where the fuel is stored in the cask 1), and orthogonal to an end surface 1tp of the cask 1. The central axis Z passes a center within a cross section perpendicular to the longitudinal direction of the cask 1. Forms of falling or collision of the cask 1 will be explained next. FIGS. 4B to 4D are explanatory views of forms of falling or collision of the cask 1. The forms of falling or collision of the cask 1 mainly include three forms. The form of falling or collision shown in FIG. 4B is horizontal falling or horizontal collision. This is a form in which the cask 1 falls down on or collides against a ground L a collision target surface while the central axis Z of the cask 1 is almost parallel to the ground L or the collision target surface. The form of falling or collision shown in FIG. 4C is vertical falling or vertical collision while the central axis Z of the cask 1 is almost orthogonal to the ground L or the collision target surface. The form of falling or collision shown in FIG. 4D is oblique falling or oblique collision while the central axis Z of the cask 1 is oblique relative to the ground L or the collision target surface. An oblique angle is denoted by θ. At the oblique angle θ of about 90 degrees, the form of falling or collision is the vertical falling or collision. At the oblique angle θ of about 0 degrees, the form of falling or collision is the horizontal falling or collision. FIG. 5A is an overall front view of the buffer body according to the first embodiment. FIG. 5B is an overall side view of the buffer body according to the first embodiment. As shown in FIGS. 5A and 5B, the buffer body 6 is constituted by containing a shock absorber, to be explained later, in an outer plate 6w, which consists of stainless steel, carbon steel or the like. In the front view, that is, if viewed from a direction parallel to a central axis Z1 of the buffer body 6 (hereinafter, “buffer body central axis Z1”), the buffer body 6 is of a disk shape configured by four circular arcs and four lines. Namely, the four square corners of the buffer body 6 are circular arc-shaped. By so forming, it is possible to make a distance between opposing sides of the buffer body 6 smaller than a distance between opposing circular arcs, and to thereby reduce an external size of the buffer body 6. It is noted that the buffer body central axis Z1 is equal to the central axis Z of the cask 1 and orthogonal to the end surface 1tp of the cask 1 shown in FIG. 5B (an end surface 4tp of the secondary lid 4 in FIG. 5B). According to the present invention, the shape of the buffer body is not limited to that of the buffer body 6 shown in FIG. 5A. The present invention is also applicable to the buffer bodies of various shapes including the circular shape viewed from the direction parallel to the buffer body central axis Z1, according to the specifications of the cask 1. Furthermore, the shape of the buffer body 6 viewed from the direction parallel to the buffer body central axis Z1 is not limited to the circular shape. For example, if the shape of the buffer body 6 viewed from the direction parallel to the buffer body central axis Z1 is the circular shape, various shapes such as a shape partially having a linear portion (that is, having a flat surface) can be selectively adopted according to the specifications of the cask 1. As shown in FIG. 5A, the buffer body 6 according to the first embodiment is provided with a plurality of attachment holes 7 parallel to the buffer body central axis Z1 and formed on a circumference about the buffer body central axis Z1. As shown in FIGS. 5A and 5B, the buffer body 6 according to the first embodiment is provided with an opening 6o, and the opening 6o is covered on an end it of the cask 1 (the secondary lid 4 in FIGS. 5A and 5B). Fastening units (such as bolts) are inserted into the respective attachment holes 7 and screwed into the end 1t of the cask 1, thereby attaching the buffer body 6 to the end 1t of the cask 1. In the first embodiment, the buffer body 6 is fastened to the secondary lid 4. Alternatively, the buffer body 6 can be fastened or secured to the barrel main body 1b of the cask 1. In addition, the buffer body 6 can be attached to the cask 1 not only by being directly attached to the end it of the cask 1 by the fastening units but also by interposing an attachment member such as an attachment plate between the buffer body 6 and the cask 1. Furthermore, the buffer body 6 can be attached to the buffer body 6 by interposing a shim between an outside of the end it of the cask 1 and an inside of the opening 6o of the buffer body 6 so as to make a gap between the outside of the end 1 and the inside of the opening 6o as small as possible. An internal structure of the buffer body according to the first embodiment will be explained next. FIG. 6 is an explanatory view of the internal structure of the buffer body according to the first embodiment. FIG. 7 is a cross-section taken along a line X-X of FIG. 6. FIG. 8 is a cross-section taken along a line A-A of FIG. 7. FIG. 9 is a cross-section taken along a line B-B of FIG. 7. FIG. 10 is a view from line C-C of FIG. 7. The buffer body 6 according to the first embodiment employs wood materials as shock absorbers. Arrows in FIGS. 6 to 10 indicate directions of fibers of the wood materials that constitute the shock absorbers. As can be seen from FIGS. 7 and 8, the buffer body 6 according to the first embodiment is configured so that the shock absorbers that absorb a shock generated when the cask 1 falls or collides are arranged within the outer plate (see FIGS. 5A and 5B). As explained above, the shock absorbers consist of the wood materials and are arranged by changing the types of the shock absorbers or the directions of the fibers of the wood materials. By doing so, the buffer body 6 according to the first embodiment can exhibit the function required as the buffer body of the cask 1. As shown in FIG. 7, the buffer body 6 is configured by combining a first shock absorber B1, a second shock absorber B2, a third shock absorber B3, a fourth shock absorber B4, a fifth shock absorber B5, a sixth shock absorber B6, a seventh shock absorber B7, and an eighth shock absorber B8. In the first embodiment, the first shock absorber B1 corresponds to “a first shock absorber group”, the second to the fourth shock absorbers B2 to B4 correspond to “a second shock absorber group”, and the fifth to the eighth shock absorbers B5 to B8 correspond to “a third shock absorber group”. These shock absorbers are constituted by a combination of a plurality of shock absorber blocks. In addition, the shock absorber 6 is attached to each of the both ends of the cask 1 or each of the end plates 1p (see FIG. 1) of the cask 1 by inserting a bolt 50, which is the fastening unit, into the attachment hole 7 and screwing the bolt 50 into a bolt hole formed in the cask 1. The bolt hole formed in the cask 1 is provided in, for example, the barrel main body 1b (see FIG. 1) of the cask 1 or the end surface 1tp of the cask 1 (the secondary lid end surface 4tp in FIG. 7). FIG. 11A is an enlarged cross-section of the attachment hole. FIG. 11B is an enlarged cross-section of another configuration of the attachment hole. Both FIGS. 11A and 11B show a region D shown in FIG. 7. The attachment hole 7 of the buffer body 6 according to the first embodiment consists of a bellows 7s so as to be contractible and expandable in the direction of buffer body central axis Z1. The bellows 7s enables the attachment hole 7 to be deformed in the direction of the buffer body central axis Z1 without few resistances when the cask 1 vertically falls or collides. In addition, the bellows 7s can suppress sudden increase of a shock load due to deformation of the attachment hole 7 when the buffer body 6 starts to be deformed during the vertical falling or collision of the cask 1. As a result, it is possible to suppress an excessive force from acting on the bolts interposed between the primary lid 3 and the barrel main body 1b (see FIG. 1) and between the secondary lid 4 and the barrel main body 1b for securing the primary lid 3 and the secondary lid 4 during the vertical falling or collision of the cask 1, and to keep the cask 1 hermetically sealed by the gaskets. As shown in FIG. 11B, the attachment hole 7 can be configured so that ends of two cylindrical members 7s1 and 7s2 having different diameters are fitted into each other, and so that an entire length of the attachment hole 7 is reduced by the load in the direction of the buffer body central axis Z1. The first shock absorber B1 absorbs a shock generated by the horizontal falling or collision of the cask 1. A part of an outer circumference of the buffer body 6 collides against the ground or the like when the cask 1 falls horizontally or collides against the ground or the like. Due to this, an area of the first shock absorber B1 contributing to absorbing the shock is made small. For this reason, the first shock absorber B is made of a first material having a highest compressive strength among all of the first to the eighth shock absorbers B1 to B8 constituting the buffer body 6 according to the first embodiment. If the wood material is used, oak, for example, is used as the first material. The “compressive strength” means herein a Young's modulus, a compression strength or the like when the shock absorber is compressed. The second to the fourth shock absorbers B2 to B4 absorb a shock when the cask 1 vertically falls or collides or obliquely falls or collides. During the vertical falling or the like, surfaces of the second to the fourth shock absorbers B2 to B4 perpendicular to the buffer body central axis Z1 absorb the shock generated by the vertical falling or the like. Namely, when the cask 1 vertically falls, the buffer body 6 collides against the ground L or the like by a wider area than that during the horizontal falling and absorbs the shock. Therefore, the second to the fourth shock absorbers B2 to B4 contributing to absorbing the shock are larger in area than the first shock absorber B1. For this reason, the second to the fourth shock absorbers B2 to B4 are made of a second material lower in compressive strength than the first shock absorber B1. If the wood material is used, red cedar (western cedar), for example, is used as the second material. The fifth to the eighth shock absorbers B5 to B8 absorb a shock generated when the cask 1 vertically falls or collides or obliquely falls or collides. The fifth to the eighth shock absorbers B5 to B8 sufficiently relax a shock force transmitted to the primary lid 3 and the secondary lid 4 (see FIG. 1). The cask 1 is kept hermetically sealed by interposing the gaskets between the primary lid 3 and the barrel main body 1b and between the secondary lid 4 and the barrel main body 1b (see FIG. 1), respectively. Therefore, the fifth to the eighth shock absorbers B5 to B8 sufficiently relax the shock such as the falling so as not to disturb the hermetic sealing. For this reason, the fifth to the eighth shock absorbers B5 to B8 are made of a third material lower in compressive strength than the second to the fourth shock absorbers B2 to B4. If the wood material is used, balsa, for example, is used as the third material. If the first to the third materials are other than the wood materials, for example, resin materials or metal materials, they can be arbitrarily selected so as to satisfy the relationship of (the compressive strength of the first material)>(the compressive strength of the second material)>(the compressive strength of the third material). Each shock absorber will be explained next. The shock absorbers consisting of the second material will be explained. The second, the third, and the fourth shock absorbers B2 to B4 consist of the second material. As shown in FIG. 7, the third shock absorber B3 and the fourth shock absorber B4 are arranged on a shock load (shock) input side in the direction of the buffer body central axis Z1, that is, on an opposite side to the opening 6o in the direction of the buffer body central axis Z1. As shown in FIG. 6, the third shock absorber B3 and the fourth shock absorber B4 are arranged around the buffer body central axis Z1 in the proximity order of the fourth shock absorber B4 and the third shock absorber B3 to the central axis Z1. A shown in FIGS. 7 and 8, if viewed from within the cross section perpendicular to the buffer body central axis Z1, the second shock absorber B2 is arranged around the central axis Z1 and on an outermost circumference of the buffer body 6. In addition, the second shock absorber B2 is arranged between the first shock absorber B1 and the third and the fourth shock absorbers B3 and B4. The second shock absorber B2 is constituted by a plurality of second shock absorber blocks 11. The third shock absorber B3 is constituted by a plurality of third shock absorber blocks 12. The fourth shock absorber B4 is constituted by a plurality of fourth shock absorber blocks 13. These shock absorber blocks are formed by, for example, superimposing wood materials. As shown in FIGS. 6 and 7, the second, the third, and the fourth shock absorbers B2, B3, and B4 are arranged so that directions of fibers are orthogonal to the buffer body central axis Z1. When the cask 1 vertically falls or collides, the shock load is input to the second, the third, and the fourth shock absorbers B2, B3, and B4 perpendicularly to the fiber directions. This shock load is absorbed by the second, the third, and the fourth shock absorbers B2, B3, and B4 by crashing the absorbers B2, B3, and B4 perpendicularly to the fiber directions. The shock absorbers consisting of the third material will be explained. The fifth to the eighth shock absorbers B5 to B8 consist of the third material. As shown in FIG. 7, the fifth shock absorber B5 and the sixth shock absorber B6 are arranged on the shock load input side in the direction of the buffer body central axis Z1, that is, on the opposite side to the opening 6o in the direction of the central axis Z1. The fifth shock absorber B5 and the sixth shock absorber B6 are arranged around the central axis Z1 in the proximity order of the sixth shock absorber B6 and the fifth shock absorber B5 to the buffer body central axis Z1. As shown in FIG. 7, the fifth shock absorber B5 is arranged so that the fiber direction is parallel to the buffer body central axis Z1. The sixth shock absorber B6 is arranged so that, for example, the fiber direction is orthogonal to the buffer body central axis Z1. The fifth and the sixth shock absorbers B6 can be constituted by combinations of a plurality of fifth and sixth shock absorber blocks 14 and 15 each of which is a fan-shaped block, respectively. As shown in FIG. 7, the seventh shock absorber B7 and the eighth shock absorber 8 are arranged in this order in the direction in which the shock load is input to the buffer body 6, that is, from the load input side of the shock absorber 6 toward the opening 6o thereof. As shown in FIGS. 7, 8, and 9, the sixth and the seventh shock absorbers B6 and B7 are cylindrical absorbers around the buffer body central axis Z1. As shown in FIGS. 7 and 8, the seventh shock absorber B7 is arranged so that the fiber direction is orthogonal to the buffer body central axis Z1 (it is noted that the seventh shock absorber B7 is not shown in FIG. 8). As shown in FIGS. 7 and 9, the eighth shock absorber B8 is arranged so that the fiber direction is parallel to the buffer body central axis Z1 (it is noted that the eighth shock absorber B8 is not shown in FIG. 9). As shown in FIGS. 8 and 9, the seventh and the eighth shock absorbers B7 and B8 are constituted by combinations of a plurality of seventh and eighth shock absorber blocks 16 and 17, each of which is a fan-shaped block, respectively. If the cask 1 vertically falls or collides, the shock load is input to the fifth to the eighth shock absorbers B5 to B8. This shock load is absorbed by the fifth to the eighth shock absorbers B5 to B8 by crashing the fifth to the eighth shock absorbers B5 to B8 in an input direction of the shock load. It is thereby possible to keep the barrel main body 1b of the cask 1 hermetically sealed from the primary lid 3 and the secondary lid 4, respectively. The shock absorber consisting of the first material will be explained. The first shock absorber B1 consists of the first material. As shown in FIGS. 7, 9, and 10, the first shock absorber B1 is arranged on the opening 6o side of the buffer body 6 in the direction of the buffer body central axis Z1. As shown in FIGS. 7, 9, and 10, if viewed from the opening 6o side of the buffer body 6, the first shock absorber B1 is arranged around the buffer body central axis Z1 on the outermost circumference of the buffer body 6. The first shock absorber B1 is thereby arranged to be superimposed on the end 1t of the cask 1 (the secondary lid 4 in FIGS. 7 and 10). By thus arranging the first shock absorber B1, the shock generated when the cask 1 horizontally falls or collides can be absorbed by the first shock absorber B1. As shown in FIG. 10, the first shock absorber B1 is constituted by a combination of a plurality of first shock absorber blocks 10. These shock absorber blocks are formed by, for example, superimposing wood materials. As shown in FIGS. 7, 9, and 10, the first shock absorber B1 is arranged so that the fiber direction is orthogonal to the buffer body central axis Z1. During the vertical falling or collision of the cask 1, the shock load is input in parallel to the fiber direction of the first shock absorber B1. This shock absorber is absorbed by the first shock absorber B1 by crashing the first shock absorber B1 in parallel to the fiber direction. The shock load generated when the cask 1 horizontally falls or collides is input from a direction orthogonal to the buffer body central axis Z1 as shown in FIG. 10. In this case, as evident from FIG. 10, the first shock absorber B1 that can contribute to absorbing the shock is a part of the first shock absorber B1 annularly arranged around the end it of the cask 1. Therefore, the first shock absorber B1 consists of the first material having a highest compressive strength among all the shock absorbers, and is arranged so that the fiber direction is parallel to the input direction of the shock load. By doing so, when the cask 1 horizontally falls or collides, the shock load can be sufficiently absorbed by part of the first shock absorber B1. Instances of constituting the shock absorber block that constitutes each shock absorber by superimposing wood materials will be explained. FIG. 12A is an explanatory view of an example of the first shock absorber block 10 constituted by superimposing wood materials. FIG. 12B is an explanatory view of an example of the second shock absorber block 11 constituted by superimposing wood materials. In both FIGS. 12A and 12B, arrows indicate fiber directions. As shown in FIG. 12A, the first shock absorber block 10 is manufactured by adhesively bonding and superimposing three veneers 10p each manufactured by adhesively bonding and superimposing plate pieces 10s. In the manufacturing, the plate pieces 10s are arranged so as to be parallel to one another in the fiber direction. As shown in FIG. 12B, the second shock absorber block 11 is manufactured by adhesively bonding and superimposing three plate pieces 11s. In the manufacturing, the plate pieces 11s are arranged so as to be parallel to one another in the fiber direction. The third and the fourth shock absorber blocks 12 and 13 are manufactured similarly to the second shock absorber block 11. By thus changing the manner of superimposing the wood materials according to a location of the buffer body 6, the shock absorbing characteristics according to the specifications of the buffer body 6 can be obtained. The first shock absorber blocks 10 are arranged annularly toward the direction in which the plate pieces 10s are superimposed, and constitute the first shock absorber B1. Therefore, the plate pieces 10s that constitute each first shock absorber block 10 are superimposed toward a circumferential direction Co (see FIGS. 10 and 12A). The shock load is input to the first shock absorber blocks 10 substantially in parallel to the fiber direction, and a force in a direction in which the superimposed plate pieces 10s are peeled off (circumferential direction Co of the first shock absorber B1) acts on each first shock absorber block 10. In the first shock absorber B1 according to the first embodiment, the force of peeling off the superimposed plate pieces 10s is suppressed by the first shock absorber blocks arranged adjacently. Therefore, even if the shock load is input to the first shock absorber block 10, it is possible to suppress the superimposed plate pieces 10s from being peeled off. FIG. 13 is a stress-strain diagram of one example of the relationship between a stress and a strain of the wood material. The stress a shown in FIG. 13 corresponds to a compressive stress. FIGS. 14A and 14B are explanatory views of examples of a method for forming holes in the shock absorber block. The wood material constituting the first shock absorber block 10 is a fiber assembly. If the shock absorber block 10 is thus constituted by the wood material, a shock load energy is absorbed by the shock absorber block 10 by causing wood fibers to be sheared or locally crashed. As a result, as shown in FIG. 13, the wood material shows a crash behavior in which the stress a rises according to a rise of the strain. Specifically, if the stress σ rises and the strain ε rises, accordingly, the stress σ rapidly increases at a certain strain εc (indicated by a one-dot chain line in FIG. 13). In addition, the moment the shock load acts on the cask 1, an excessive initial stress is generated (indicated by a dot line in FIG. 13). Thus, the shock absorbing characteristic required of the buffer body 6 cannot be obtained unless the wood material shows a uniform crash (strain) behavior when the shock load acts on the cask 1. As a result, an excessive shock load sometimes acts on the cask 1. The cask 1 has an entire length as large as several meters. Due to this, in a test of falling or collision of the cask 1, scaled-down models of the cask 1 and the buffer body 6 are employed. In the test, if the wood materials are used for the shock absorbers of the buffer body 6, sizes of the shock absorbers differ between the scaled-down model of the buffer body and the actual buffer body 6 while fiber widths of the shock absorbers are equal. Namely, the shock absorbers of the scaled-down model of the buffer body are relatively larger in fiber width than those of the actual buffer body 6. The first shock absorber block 10 according to the first embodiment is provided with a plurality of holes 20 serving as spaces to cross the fibers. Thanks to the holes 20, if the shock load is input to the first shock absorber block 10, the first shock absorber block 10 can be stably crashed over entire regions of the block 10. As a result, even if the first shock absorber block 10 consists of the wood material, it can be dealt with as a uniform material. If the holes 20 are provided, a stress (indicated by symbol x on a solid line of FIG. 13) can be made lower than a crash stress (indicated by symbol x on a one-dot chain line of FIG. 13) by which the first shock absorber block 10 is crashed. Therefore, during absorption of the shock load, it is possible to suppress the excessive shock load from acting on the cask 1. In addition, even the buffer body different in size from the actual buffer body 6 (e.g., the scaled-down model of the buffer body or a scaled-up model thereof) can sufficiently ensure reproducibility of the shock absorbing performance. These actions make it unnecessary to give an excessive margin to the shock absorbing performance of the buffer body when the buffer body 6 is made large in size. It is, therefore, possible to make the size of the buffer body 6 a necessary and sufficient size and decrease shock acceleration. As a result, it is unnecessary to considerably increase the shock absorbing performance of the cask 1 main body, and the number of recycle fuels stored in the cask 1 can be, therefore, increased. Besides, since the buffer body 6 can exhibit the sufficiently high shock absorbing performance with the necessary and sufficient size, the present invention can be applied to even a case that strict size limitation is imposed on transportation of the cask 1. Furthermore, a pitch Pt (shown in FIG. 14A) of the holes in the fiber direction which has a great effect on the crash characteristics of the wood material is conserved between the scaled-down model of the buffer body 6 and the actual buffer body 6. It is thereby possible to make the shock absorbing characteristics of the entire shock absorber block 10 uniform between the scaled down model of the buffer body 6 and the actual buffer body 6. Differences in characteristics of the wood material according to a difference in size can be thereby reduced. This can facilitate predicting the shock absorbing characteristics of the actual buffer body 6 from the shock absorbing characteristics obtained by the scaled-down model of the buffer body 6. This can facilitate designing the actual buffer body 6 and enables the buffer body 6 to exhibit the shock absorbing characteristics as designed. Besides, the reproducibility of the shock absorbing performance can be sufficiently ensured even among different buffer bodies. In the first embodiment, the holes 20 are provided only in the first shock absorber block 10 consisting of the first material having the highest compressive strength for the following reason. The oak used as the first material in the first embodiment is high in compressive strength and an excessive initial stress is often generated at the moment of the shock (the dot line part of FIG. 13). The holes 20 are intended to avoid this excessive initial stress and stably crash the absorbers in all regions. Depending on the shock absorbing performances required of the second, the third, and other shock absorbers B2, B3, holes can be formed in the second, the third, and other shock absorber blocks 11, 12, and the like constituting the second, the third, and other shock absorbers B2, B3, and the like, respectively. In forming the holes 20 in the first shock absorber block 10, it is necessary to consider the fiber direction having a great effect on the crash characteristics of the wood material. To do so, it is preferable to provide the holes 20 so as to divide the fibers as shown in FIG. 14A. Alternatively, as shown in FIG. 14B, the holes 20 can be provided so as to divide the fibers within the first shock absorber block 10. The both methods can be combined. Depending on the specifications of the buffer body 6, the number of holes 20 and the pitch Pt can be set in combination. Another example of the shock absorber blocks according to the first embodiment will be explained. While the first shock absorber block 10 constituting the first shock absorber B1 will be explained herein by way of example, the second and the third shock absorber blocks 11 and 12 constituting the second, the third, and the other shock absorbers B2, B3, and the like can be similarly applied (the same thing will apply hereafter). FIGS. 15A to 15H are explanatory views of examples of providing holes as spaces formed in the shock absorber block. In FIGS. 15A to 15H, arrow directions indicate fiber directions and symbol P denotes the shock load input to the first shock absorber block. A first shock absorber block 10a shown in FIG. 15A is configured to provide the holes 20 penetrating the first shock absorber block 10a so as to divide the fibers and to be orthogonal to an input direction of the shock load P. By so configuring, a strength and a rigidity of the first shock absorber block 10a can be adjusted. By providing the holes 20 in the first shock absorber block 10a, cracking and crashing can be generated in sheared parts of the first shock absorber block 10a with the holes 20 set as points of origin. It is thereby possible to accelerate absorption of the shock energy by the first shock absorber block 10a. A first shock absorber block 10b shown in FIG. 15B is configured to provide the holes 20 penetrating the first shock absorber block 10b so as to divide the fibers and to be parallel to the input direction of the shock load P. By so configuring, a strength and a rigidity of the first shock absorber block 10b can be adjusted. In addition, the holes 20 can reduce an apparent cross-sectional area of the first shock absorber block 10b, and reduce an initial stress generated in the first shock absorber block 10b right after the shock load P is input to the first shock absorber block 10b. A first shock absorber block 10c shown in FIG. 15C differs from the first shock absorber block 10b shown in FIG. 15B in that bottomed holes 21 that do not penetrate the first shock absorber block 10c are provided. Even with this configuration, the first shock absorber block 10c can attain the same functions and advantages as those of the first shock absorber block 10b shown in FIG. 15B. In addition, by adjusting a depth of each bottomed hole 21, a strength and a rigidity of the first shock absorber block 10c can be adjusted. A first shock absorber block 10d shown in FIG. 15D is configured to provide first holes 201 so as to divide the fibers and to be orthogonal to the input direction of the shock load P, and provide second holes 202 so as to divide the fibers and to be parallel to the input direction of the shock load P. By so configuring, cracking and crashing can be generated in sheared parts of the first shock absorber block 10d with the first holes 201 set as points of origin. It is thereby possible to accelerate absorption of the shock energy by the first shock absorber block 10d. At the same time, the second holes 202 can reduce an apparent cross-sectional area of the first shock absorber block 10d, and reduce an initial stress generated in the first shock absorber block 10d right after the shock load P is input to the first shock absorber block 10d. In this example, the first holes 201 and the second holes 202 are located in a distorted relationship therebetween. Alternatively, the first holes 201 and the second holes 202 can be located to cross one another. Furthermore, at least either the first holes 201 or the second holes 202 can be bottomed holes. It is noted that diameters d of the first holes 201 and the second holes 202 are appropriately changed according to the specifications of the buffer body 6. First shock absorber blocks 10e and 10f shown in FIGS. 15E and 15F are configured to provide holes 22 and 23 so as to divide the fibers and to be parallel to the input direction of the shock load P, and so as to reduce cross-sectional areas of the holes 22 and 23 toward the input direction of the shock load P (d12/4>d22/4), respectively. By so configuring, the first shock absorber block 10e or 10f can be promptly crashed and the shock energy can be sufficiently absorbed by the first shock absorber block 10e or 10f right after the shock load is input to the block 10e or 10f. In addition, as the crashing of the shock absorber block 10e or 10f progresses, the shock absorber block 10e or 10f is more difficult to crash, thereby making it possible to effectively stop a motion of the cask 1. Cross-sectional areas of the holes 22 can be gradually reduced toward the input direction of the shock load P as shown in FIG. 15E. Alternatively, as shown in FIG. 15F, cross-sectional areas of the holes 23 can be reduced step by step. In the latter case, the holes 23 can be formed relatively easily. A first shock absorber block 10g shown in FIG. 15G differs from the first shock absorber block 10a shown in FIG. 15A only in that angular holes (rectangular holes in this embodiment) 24 are provided. As shown in FIG. 15G, the circular holes can be replaced by the angular holes 24. By doing so, cracking and crashing are generated in sheared parts of the first shock absorber block 10g with corners 24t of the angular holes 24 set as points of origin. It is thereby possible to accelerate absorption of the shock energy by the first shock absorber block 10g. From these viewpoints, it is preferable that the corners 24t of the angular holes 24 are formed on the input side of the shock load P. A first shock absorber block 10h shown in FIG. 15H is similar to the first shock absorber block 10g shown in FIG. 15G except that angular holes 24 are formed by adhesively bonding and superimposing plate materials 10h1, 10h2, and 10h3 having grooves 24s having a generally triangular cross section, and by combining the grooves 24s. The angular holes 24 provided in the first shock absorber block 10g shown in FIG. 15G can be formed by a dedicated tool. Alternatively, they can be formed by adhesively bonding and superimposing the plate materials 10h1 and the like having the grooves 24s formed therein in advance similarly to the first shock absorber block 10h shown in FIG. 15H. FIGS. 15I and 15J are explanatory views of examples of providing notches as spaces provided in the shock absorber block. Each of first shock absorber blocks 10i and 10j is configured so that an acting direction of the shock load P is parallel to the fiber direction. As shown in FIG. 15I, the shock absorber block 10i is provided with notches 25 in parallel to the fiber direction. By so configuring, a strength and a rigidity of the first shock absorber block 10i can be adjusted, an apparent cross-sectional area of the first shock absorber block 10i can be reduced, and an initial stress generated when the shock load P acts on the shock absorber block 10i can be reduced. The notches 25 can be formed either to be parallel to the fibers or to divide the fibers. In the latter case, the effect of reducing the initial stress, when the shock load P acts on the shock absorber block 10i, is greater. The first shock absorber block 10j shown in FIG. 15J is provided with the notches 25 so as to be orthogonal to the fiber direction. By so configuring, a strength and a rigidity of the first shock absorber block 10i can be adjusted. By reducing the rigidity of the overall first shock absorber block 10j, lockup can be delayed. This enables the first shock absorber block 10j to stably absorb the shock energy. FIGS. 15K and 15L are explanatory views of examples of providing wedge notches as spaces provided in the shock absorber block. In FIGS. 15K and 15L, arrows indicate fibers. A first shock absorber block 10k shown in FIG. 15K is configured to form wedge notches 26 on the input side of the shock load P so that tops of the wedges (tops of the notches) are oriented to the acting direction of the shock load P. By so configuring, a strength and a rigidity of the first shock absorber block 10k can be adjusted, an apparent cross-sectional area of the first shock absorber block 10k can be reduced, and an initial stress when the shock load P acts on first shock absorber block 10k can be reduced. Furthermore, cracking and crashing can be generated in sheared parts of the first shock absorber block 10k with the tops of the grooves 26 set as points of origin. In addition, absorption of the shock energy by the first shock absorber block 10k can be accelerated. The grooves 26 can be provided in a part of the shock load input side of the first shock absorber block 10k. Alternatively, the grooves 26 can be provided entirely on the shock load input side as shown in FIG. 15K. A first shock absorber block 10k′ shown in FIG. 15L is configured to form the wedge notches 26 on the input side of the shock load P similarly to the first shock absorber block 10k shown in FIG. 15K. Furthermore, flat portions 26f are formed between adjacent notches 26. By thus providing the flat portions 26f, it is possible to ensure that the first shock absorber block 10k′ absorbs the shock with a narrow crash margin. FIGS. 15M to 15Q are explanatory views of examples of providing first shock absorber block spaces by a combination of different shapes. A first shock absorber block 10l shown in FIG. 15M is provided with the holes 20 and the notches 25 orthogonal to the fibers. By so configuring, a strength and a rigidity of the first shock absorber block 10l can be adjusted. Furthermore, cracking and crashing can be generated in sheared parts of the first shock absorber block 10l with the holes 20 set as points of origin, and absorption of the shock energy by the first shock absorber block 10l can be accelerated. In addition, an apparent cross-sectional area of the first shock absorber block 10l can be reduced, and an initial stress when the shock load P acts on the first shock absorber block 10l can be reduced. A first shock absorber block 10m shown in FIG. 15N is provided with wedge notches 26. The wedge notches 26 are formed on the input side of the shock load P so that the tops of the wedges are oriented to the acting direction of the shock load P. By so configuring, a strength and a rigidity of the first shock absorber block 10m can be adjusted. Furthermore, cracking and crashing can be generated in sheared parts of the first shock absorber block 10m with the holes 20 set as points of origin, and absorption of the shock energy by the first shock absorber block 10m can be accelerated. In addition, an apparent cross-sectional area of the first shock absorber block 10m can be reduced, and an initial stress when the shock load P acts on the first shock absorber block 10m can be reduced. A first shock absorber block 10n shown in FIG. 15O is configured to superimpose a first block 10n1 provided with the notches 25 parallel to the fibers and a second block 10n2 provided with the notches 25 orthogonal to the fibers. By so configuring, a strength and a rigidity of the first shock absorber block 10n can be adjusted. Furthermore, an apparent cross-sectional area of the first shock absorber block 10n can be reduced, and an initial stress when the shock load P acts on the first shock absorber block 10n can be reduced. In addition, by reducing the rigidity of the entire first shock absorber block 10n, lockup can be delayed. In the example of FIG. 15O, at least one of a pitch and a depth of the notches 25, 25 can be changed. For instance, the depth of the notches 25 provided in the second block 10n2 to be orthogonal to the fibers is gradually made smaller toward the acting direction of the shock load P, i.e., toward an outer peripheral side of the first shock absorber B1 (see FIG. 10). The shock absorber block 10n thus configured can be manufactured more easily than the shock absorber block provided with holes cross-sectional areas of which are gradually made smaller toward the acting direction of the shock load P (see FIGS. 15E and 15F). Accordingly, the first shock absorber block 10n can attain the same functions and advantages as those of the shock absorber block provided with holes cross-sectional areas of which are gradually made smaller toward the acting direction of the shock load P more easily. A first shock absorber block 10s shown in FIG. 15P is configured to superimpose a first block 10s1 provided with the holes 20 and a second block 10s2 provided with the notches 25 orthogonal to the fibers. By so configuring, a strength and a rigidity of the first shock absorber block 10s can be adjusted. Furthermore, cracking and crashing can be generated in sheared parts of the first shock absorber block 10s with the holes 20 set as points of origin, and absorption of the shock energy by the first shock absorber block 10s can be accelerated. In addition, by reducing the rigidity of the entire first shock absorber block 10s, lockup can be delayed. Alternatively, the first shock absorber block 10s can be configured by one block without superimposing the two blocks. A first shock absorber block 10o shown in FIG. 15Q is configured to superimpose a first block 10o1 provided with the wedge notches 26 and a second block 10o2 provided with the notches 25 orthogonal to the fibers. By so configuring, a strength and a rigidity of the first shock absorber block 10o can be adjusted. Furthermore, an apparent cross-sectional area of the first shock absorber block 10o can be reduced, and an initial stress when the shock load P acts on the first shock absorber block 10o can be reduced. By reducing the rigidity of the entire first shock absorber block 10o, lockup can be delayed. Alternatively, this first shock absorber block 10o can be configured by one block without superimposing the two blocks. FIGS. 15R to 15T are explanatory views of examples of changing the type, the number or areas of holes provided in the first shock absorber block in the input direction of the shock load. In each of FIGS. 15R to 15T, a fiber direction is parallel to the input direction of the shock load P (arrow X direction). A first shock absorber block 10p shown in FIG. 15R is configured so that the number of holes 20 per unit area is decreased toward the input direction of the shock load P. A first shock absorber block 10q shown in FIG. 15S is configured so that a cross-sectional area of each hole 20a on the input side of the shock load P is smaller than that of each hole 20b on the opposite side to the input side of the shock load P. A first shock absorber block 10r shown in FIG. 15T is configured so that the holes 20 on the input side of the shock load P are through holes and those on the opposite side to the input side of the shock load P are bottomed holes 21. By so configuring, strengths and rigidities of the first shock absorber blocks 10p, 10q, and 10r can be adjusted. In addition, each of the first shock absorber blocks 10p, 10q, and 10r is promptly crashed right after the shock load is input thereto, sufficiently absorbs shock energy, and is difficult to crash as crashing progresses. The motion of the cask 1 can, therefore, effectively stopped. FIGS. 15U and 15V are explanatory views of one example of the first shock absorber block having slots provided in parallel to texture. FIG. 15V shows a state where FIG. 15U is viewed from an arrow D direction. This first shock absorber block 10t is configured so that slots 27 serving as spaces are provided substantially in parallel to a direction of fibers of a wood material that constitutes the first shock absorber block 10t. The fiber direction is a direction indicated by solid lines each having arrows on both ends in FIG. 15U. An input direction of the shock load P is a direction indicated by an arrow X in FIGS. 15U and 15V. In the first shock absorber block 10t, since the slots 27 are provided in parallel to the fiber direction of the wood material, it is possible to make it easier that shear fracture occurs to the first shock absorber block 10t in response to compressive load. As a result, if the shock load P acts as the compressive load and even a material having a high compressive strength is used, it is possible to further ensure absorbing the shock load P. Therefore, during falling or collision of the cask 1, the cask 1 can be ensured to be protected. As shown in FIG. 15V, the slot 27 can either penetrate or not penetrate the first shock absorber block 10t. In addition, a mixture of slots penetrating the first shock absorber block 10t and those which do not penetrate it can be provided. A length of each slot 27 in the fiber direction of the wood material, a width of the slot 27 orthogonal to the fiber direction of the wood material, and the number of slots 27 can be arbitrarily changed according to the material of the first shock absorber block lot, specifications of the cask buffer body, and the like. Besides, the configuration of the first shock absorber block 10a, 10b or the like can be combined with that of the first shock absorber block 10t. As explained so far, according to the first embodiment, the holes, the notches or the other spaces are provided in each of the shock absorbers that constitute the buffer body so as to adjust the shock absorbing characteristics of the shock absorber. By doing so, even if each shock absorber consists of the material or particularly the wood material to which the excessive initial stress occurs at the moment of shock and the crash behavior of which has poor reproducibility, it is possible to make the shock absorbers have uniform shock absorbing characteristics and exhibit a stable shock absorbing performance. Furthermore, by providing the holes, the notches or the other spaces in each of the shock absorbers consisting of the wood materials and constituting the buffer body, the shock absorber can be crashed over the entire region of the wood material. Therefore, the shock absorber can be dealt with as the uniform material. As a result, occurrence of the excessive initial stress right after the shock load acts on the shock absorbers can be suppressed, and the shock absorbers can exhibit the stable shock absorbing performances with good reproducibility. In the first embodiment, the first shock absorber and the first shock absorber block have been mainly explained. The same thing is true for the other shock absorbers and the other shock absorber blocks. In a second embodiment, a block combination structure in which the first, the second and other shock absorbers B1, B2, and the like are constituted by combining the first, second, and other shock absorber blocks 10, 10a, 11, and the like will be explained. The first shock absorber block 10 constituting the first shock absorber B1 will be explained herein by way of example. This can apply to the second and the third shock absorber blocks 11 and 12 constituting the second, the third and other shock absorbers B2, B3, and the like. FIG. 16 is an explanatory view of an example of combining the first shock absorber blocks using antiskid members. In this block combination structure, a groove H is provided in each first shock absorber block 10. Antiskid members 30 are incorporated into grooves H when a plurality of first shock absorbers are combined, thereby preventing the first shock absorbers from being shifted relative to one another. If the same material as that for the first shock absorber block 10 is used for the antiskid member 30, a behavior of the antiskid member 30 during absorption of the shock can be made equal to that of the first shock absorber block 10. FIG. 17A is an explanatory view of another example of combining the first shock absorber blocks using the antiskid members. FIG. 17B is a cross-section taken along a line E-E of FIG. 17A. FIG. 17C is another cross-section taken along the line E-E of FIG. 17A. In this block combination structure, a concave portion H1 is provided in each first shock absorber block 10. Plate-like antiskid members 31 are attached to respective concave portions H1 when a plurality of first shock absorber blocks 10 is combined. In addition, the antiskid members 31 are fixed to the adjacent first shock absorber blocks 10 by screws 32 or bolts serving as fixing units, thereby preventing the first shock absorber blocks 10 from being shifted relative to one another. If a metal plate such as an iron plate or an aluminum plate is used as the antiskid member 31, sufficient rigidity can be ensured even if the antiskid member 31 is made thin. It is thereby possible to reduce a depth of the concave portion H1 formed in each first shock absorber block 10, and minimize an influence of the concave portion H1 on the first shock absorber block 10. As shown in FIG. 17C, the depth of the concave portion H1 is preferably set so that a top of the screw 32 or bolt serving as the fixing unit and the antiskid member 31 do not protrude from an outer peripheral surface of the first shock absorber block 10. By so setting, it is possible to suppress an initial shock generated in the first shock absorber block 10 from being increased by the top of the screw 32 or bolt serving as the fixing means and by the antiskid member 31 during falling or collision of the cask 1. FIGS. 18A to 18C are explanatory views of an example of combining the first shock absorber blocks by forming an antiskid portion in each first shock absorber block itself. This first shock absorber block 10x is configured so that convex portions 33t and concave portions 33v are alternately formed on both sides. The convex portions 33t and the concave portions 33v are formed to be orthogonal to an input direction of the shock load P. At the time of combining a plurality of first shock absorber blocks 10x, the convex portions 33t of one first shock absorber blocks 10x are engaged with the concave portions 33v of the adjacent block, thereby preventing the first shock absorber blocks 10x from being shifted relative to one another. At this time, as shown in FIG. 18C, an antiskid member 30 is incorporated into a groove formed in each first shock absorber block 10x. In addition, it is preferable to suppress the first shock absorber blocks 10x from being shifted relative to a formation direction of the convex portions 33t and the concave portions 33v. FIGS. 19A and 19B are explanatory views of another example of combining the first shock absorber blocks by forming an antiskid portion in each first shock absorber block itself. This first shock absorber block 10y is configured so that a protrusion 34 is formed on one side surface of the first shock absorber block 10y and so that a groove 35 to be engaged with the protrusion 34 is formed on an opposing side surface thereof. The protrusion 34 and the groove 35 are formed in parallel to each other in the input direction of the shock load P. At the time of combining a plurality of first shock absorber blocks 10y, the protrusion 34 of one first shock absorber block 10y is engaged with the groove 35 of the adjacent first shock absorber block 10y, thereby preventing the first shock absorber blocks 10y from being shifted relative to one another. At this time, it is preferable that grooves H1′ and H2′ are formed in a direction crossing a formation direction of the protrusion 34 and the groove 35, and antiskid members 30″ are incorporated into the respective grooves H1′ and H2′ so as to suppress the first shock absorber blocks 10y from being shifted relative to the formation directions of the protrusion 34 and the groove 35. FIGS. 20A to 20C are explanatory views of another example of combining the first shock absorber blocks by forming the antiskid portion in each first shock absorber block itself. This first shock absorber block 10z is configured so that convex portions 36t are formed on one side surface of the first shock absorber block 10z and so that concave portions 36v to be engaged with the convex portions 36t are formed on an opposing side surface thereof. The convex portion 36t and the concave portion 36v are formed in parallel to the input direction of the shock load P. At the time of combining a plurality of first shock absorber blocks 10z, the convex portions 36t of one first shock absorber block 10z are engaged with the concave portions 36v of the adjacent first shock absorber block 10z, thereby preventing the first shock absorber blocks 10z from being shifted. At this time, as shown in FIG. 20C, it is preferable to form grooves in a direction crossing the formation direction of the convex portion 36t and the concave portion 36v, and incorporate the antiskid members 30″ into the grooves so as to suppress the first shock absorber blocks 10z from being shifted relative to the formation direction of the protrusion 34 and the groove 35. FIG. 21 is an explanatory view of another example of combining the first shock absorber blocks using fixing members. In this block combination structure, after the first shock absorber blocks 10 are combined, the blocks 10 are fixed to one another using inverted U-shaped screws 37 serving as the fixing members. This block combination structure can prevent the first shock absorber blocks from being shifted relative to one another by a simple configuration. FIGS. 22A and 22B are explanatory views of a block combination structure using a block fastening unit. In this block combination structure, a through hole h that penetrates a plurality of (three in this example) first shock absorber blocks 10 is provided. After a plurality of first shock absorber blocks 10 are combined, a bolt 38 serving as the fastening unit is inserted into this through hole h, thereby fixing the first shock absorber blocks 10 to one another. This structure can firmly fix the first shock absorber blocks 10 to one another by the fastening unit. If a strength of the bolt 38 serving as the fastening unit is too high, there is a probability that a deformation of the central first shock absorber block 10 is received by the first shock absorber blocks 10 on both sides thereof, and that the central first shock absorber block 10 is insufficiently crashed. For this reason, it is preferable to avoid using an excessively thick bolt or to use a bolt consisting of an easily deformable material if the bolt is used as the fastening unit. In addition, it is preferable to suppress crashing of the central first shock absorber block 10 using, for example, a fastening unit having an adjustable joint structure or a fastening unit, e.g., a wire having a structure in which the fastening unit is bent halfway. FIG. 23A is an explanatory view of a block combination structure using a block binding unit. FIG. 23B is a cross-section taken along a line F-F of FIG. 23A. FIG. 23C is another cross-section taken along the line F-F of FIG. 23A. In this block combination structure, a groove s is formed on an outer periphery of a first shock absorber block 10′. After the first shock absorber B1 is formed by annularly combining a plurality of first shock absorber blocks 10′, a wire 39 serving as the block binding unit is wound entirely around the first shock absorber B1, thereby constraining and fixing the respective first shock absorber blocks 10′. This structure can firmly fix the respective first shock absorber blocks 10′ by a tensile force of the wire 39. The groove s corresponds to the “space” provided in the first shock absorber block 10′ as explained in the first embodiment. As can be seen, in this block combination structure, the space is formed in each first shock absorber block 10′ by the groove s. It is thereby possible to generate cracking and crashing in sheared parts of the first shock absorber block 10′ with the groove s set as a point of origin, and accelerate absorption of the shock energy by the first shock absorber block 10′. At this moment, if a magnitude and a shape of the groove s are changed, a speed of cracking or crashing of the sheared parts of the first shock absorber block 10′ can be adjusted. Furthermore, as seen in a first shock absorber block 10″ shown in FIG. 23C, wedge grooves s′ can be formed by cross sections corresponding to “spaces” and the wire 39 can be wound around the grooves s′. By doing so, it is possible to suppress the wire 39 from being shifted and adjust a strength and a rigidity of the first shock absorber block 10″. Further, an apparent cross-sectional area of the first shock absorber block 10″ can be reduced and an initial stress generated when the shock load P acts on the block 10″ can be reduced. Besides, cracking and crashing can be generated in sheared parts of the first shock absorber block 10″ with tops of the grooves s′ set as points of origin, and absorption of the shock energy by the first shock absorber block 10″ can be accelerated. FIGS. 24A to 24D are explanatory views of an example of a combination structure of the first shock absorber blocks. FIG. 25A is an explanatory view of a stress change when the combination structure, in which first shock absorber blocks each having a larger area on a diametral outside are combined, receives the shock load. FIG. 25B is an explanatory view of a stress change when a combination structure, in which first shock absorber blocks each having a smaller area on a diametral outside and the first shock absorber blocks each having a larger area on a diametral outside are combined, receives the shock load. As explained above, the first shock absorber block constituting the first shock absorber B1 (see FIG. 7) is made of the first material having the highest compressive strength. If the wood material is used, oak, for example, is used as the first material. Since the oak has a high compressive strength, lockup tends to occur when the buffer body is deformed to crash the first shock absorber blocks. For instance, the first shock absorber block 10C shown in FIG. 25A is generally fan-shaped and an area of a diametral outside O is larger than that of a diametral inside I. In the structure in which the first shock absorber blocks 10C thus shaped are made of the material having the high compressive strength such as oak and combined, a reaction force F within the first shock absorber block 10C suddenly rises at a certain strain (εc) by the lockup when the shock load P is applied to the block 10C. As a result, after occurrence of the lockup, the shock load can possibly be absorbed insufficiently. In this combination structure of the first shock absorber blocks, a shock absorber block A (hereinafter, “first shock absorber block 10A”) having a larger area on the diametral inside I than on the diametral outside O is constituted by a material having a high compressive strength such as oak (FIGS. 24A and 24B). A shock absorber block B (hereinafter, “first shock absorber block 10B”) having a larger area on the diametral outside O than on the diametral inside I is constituted by a material lower in compressive strength than the material for the first shock absorber block 10A (FIGS. 24A and 24C). At this time, by making the acting direction of the load substantially parallel to the fiber direction, a rigidity of the first shock absorber block 10A is made high in the acting direction of the load. By making the acting direction of the load substantially orthogonal to the fiber direction, a rigidity of the first shock absorber block 10B is made low in the acting direction of the load and high in the peripheral direction. By so setting, if the shock load P is applied to the first shock absorber blocks 10A, it is possible to suppress the first shock absorber blocks 10A from falling laterally (suppress a motion thereof toward the peripheral direction of the combination structure of the first shock absorber blocks). If oak is used as the material for the first shock absorber block 10A, such a material as oak, red cedar, pine, or spruce is used for the first shock absorber block 10B. The notches shown in FIG. 15I or the like can be provided on a surface of each shock absorber block B (first shock absorber block 10B). In the combination structure of the first shock absorber blocks shown in FIG. 24D, the shock absorber block A (hereinafter, “first shock absorber block 10A′”) having a larger area on the diametral inside I than on the diametral outside O is constituted by a material having a high compressive strength such as oak (FIG. 24D). A shock absorber block B (hereinafter, “first shock absorber block 10B′”) having a larger area on the diametral outside O than on the diametral inside I is constituted by a material lower in compressive strength than the material for the first shock absorber block 10A′ (FIG. 24D). Furthermore, each first shock absorber block 10A′ contacts with the first shock absorber block 10B′ by a predetermined area on the diametral inside I of the first shock absorber B1 (see FIG. 7). As can be seen, the first shock absorber blocks 10A′ and 10B′ can contact with each other by the predetermined area on the diametral inside I of the first shock absorber B1 depending on hardness of the materials and the performances required of the buffer body as shown in FIG. 24D. If the shock load P is applied to such a combination structure of the first shock absorber blocks (see FIG. 25B), the reaction force F within the first shock absorber block 10A gradually increases according to an increase of the strain ε. In addition, occurrence of the lockup can be delayed. As a result, the shock load can be effectively absorbed. During the falling or collision of the cask 1, therefore, the cask 1 can be ensured to be protected. In this combination structure of the first shock absorber blocks, the spaces for dividing the fibers of the wood material, the slots parallel to the fibers of the wood material, or the like can be provided in each of the first shock absorber blocks 10A and 10B as explained in the first embodiment. However, even if the spaces, slots or the like are not provided therein, the stress within the shock absorber block can be gradually increased and the occurrence of the lockup can be delayed. The shock load can be, therefore, effectively absorbed. According to the second embodiment, it is possible to suppress the shock absorbers from being shifted relative to one another by providing the antiskid member or the like on each shock absorber block. This can facilitate operations for assembling the shock absorber blocks and manufacturing the shock absorber. Furthermore, since the shifting of the shock absorber blocks constituting the shock absorber is suppressed, the buffer body can exhibit the required shock absorbing performance when the shock due to the falling or collision acts on the buffer body. As explained so far, the cask buffer body according to the present invention is useful for protection of the cask that stores the recycle fuel and particularly suitable for stably exhibiting the shock absorbing performance.
062298682	summary	TECHNICAL FIELD The present invention relates to a nuclear fuel assembly with a substantially square cross section for a light water reactor comprising fuel rods extending between a top tie plate and a bottom tie plate. BACKGROUND ART In a nuclear reactor, moderated by means of light water, the fuel exists in the form of fuel rods, each of which contains a stack of pellets of a nuclear fuel arranged in a cladding tube, a column of extruded fuel cylinders of an uninterrupted column of vibration-compacted powdered fuel. The cladding tube is normally made of a zirconium-base alloy. A fuel bundle comprises a plurality of fuel rods arranged in parallel with each other in a certain definite, normally symmetrical pattern, a so-called lattice. The fuel rods are retained at the top by a top tie plate and at the bottom by a bottom tie plate. To keep the fuel rods at a distance from each other and prevent them from bending or vibrating when the reactor is in operation, a plurality of spacers are distributed along the fuel bundle in the longitudinal direction. A fuel assembly comprises one or more fuel bundles, each one extending along the main part of the length of the fuel assembly. Together with a plurality of similar fuel assemblies, a fuel assembly is arranged in a core. The core is immersed in water which serves both as coolant and as neutron moderator. During operation, the water flows from below and upwards through the fuel assembly, whereby, in a light-water reactor of boiling water type, part of the water is transformed into steam. The percentage of steam increased towards the top of the fuel assembly. Consequently, the coolant in the lower part of the fuel assembly consists of water whereas the coolant in the upper part of the fuel assembly consists both of steam and of water. This difference between the upper and lower parts gives rise ot special factors which must be taken into consideration when designing the fuel assembly. It is therefore desirable to achieve a flexible fuel assembly for a boiling water reactor which, in a simple manner, may be given a shape in which the upper part of the fuel assembly differs from the lower part thereof. A fuel assembly for a boiling water reactor with these properties is shown in PCT/SE95/01478 (Int. Publ. No. WO 96/20483). This fuel assembly comprises a plurality of fuel units stacked on top of each other, each comprising a plurality of fuel rods extending between a top tie plate and a bottom tie plate. The fuel units are surrounded by a common fuel channel with a substantially square cross section. A fuel assembly of this type may in a simple manner be given a different design in its upper and lower parts. As in a light-water reactor of boiling water type, the water flows during operation from below and up through the fuel assembly of pressurized-water type. The temperature of the water increases the higher it rises in the assembly but it does not boil. A consequence of this is that corrosion of the cladding tubes increase sin the upper part as compared with the lower part of the fuel assembly. This difference between the upper and lower parts gives rise to special factors which must be taken into consideration when designing the fuel assembly. It is therefore desirable, in the same way as has been described for a boiling water reactor, to achieve a flexible fuel assembly for a pressurized-water reactor which, in a simple manner, may be given a design in which the upper part of the fuel assembly differs from the lower part thereof. In UK 1 403 491, a fuel assembly for a pressurized-water reactor is shown, with a possibility of designing the upper part such that is differs from the lower part. This fuel assembly comprises a plurality of fuel units stacked on top of each other, each of which comprises a plurality of fuel rods extending between a top tie plate and a bottom tie plate. The fuel units are fixed to a centrally arranged support tube such that the bottom tie plate of one of the fuel units rests on the top tie plate of an adjacently arranged fuel unit and such that all the fuel rods in the fuel units are parallel to each other. The support tube extends through the whole fuel assembly and the fuel assembly has a substantially circular cross section. This fuel assembly is intended to be used in a nuclear reactor moderated by heavy water where the fuel assemblies are arranged in vertical pressure channels. Another fuel assembly for a pressurized-water reactor with short fuel units with a hexagonal cross section is shown in "Improvements in Water Reactor Fuel Technology, Proceedings of a Symposium Stockholm, 15-19 September 1986, International Atomic Energy Agency, Vienna, 1987". The fuel units are shown vertically arranged and stacked to top of each other. One factor which must be taken into consideration when designing a fuel assembly of light-water type is that the fuel rods grow to varying degrees during operation. The fuel assembly must therefore comprise members so as to allow this growth. Another factor which must be taken into consideration is that the enthalpy across the core becomes as uniform as possible. For achieving enthalpy equalization across the core, it is known to provide fuel assemblies with a burnup-dependent flow resistance. In SE 460 452, elongated elements, such as fuel rods, are provided at their lower ends with restriction bodies which, when the elongated elements during irradiation are gradually, during the burnup of the fuel assembly, adapted to more and more restrict the flow of the coolant through an opening arranged below the restriction body by moving the restriction body closer and closer to this opening. In SE 9003330-9, the spacers are provided with members which are automatically, or manually, gradually activated during the burnup of the fuel assembly such that the coolant flow is deflected to one or more adjoining fuel assemblies. The object of the present invention is to provide a fuel assembly for a light-water reactor comprising a plurality of short fuel units and members for differential growth of the fuel rods in the fuel units. Another object of the invention is to suggest a fuel assembly with a burnup-dependent flow resistance. SUMMARY OF THE INVENTION The present invention relates to a fuel assembly of substantially square cross section for a light-water reactor comprising a fuel assembly with a plurality of fuel rods extending between a top tie plate and a bottom tie plate. The fuel assembly comprises a plurality of fuel units, each comprising a plurality of fuel rods extending between a bottom tie plate and a top tie plate. The bottom and top tie plates may be designed identical. According to one aspect of the invention, the bottom tie plates are fixed to at least one of the water channels (boiling water reactor) or control rod guide tubes (pressurized-water reactor) extending through the fuel assembly, whereas the top tie plates are arranged freely movable in relation thereto. By this embodiment, an axial gap may be obtained between two adjacently located top and bottom tie plates. This axial gap may be utilized as means for accumulation of differential growth of the fuel rods such that the outer dimensions of the fuel assembly are not influenced. Alternately, the top tie plate may be attached to the water channels and the control rod guide tubes, respectively, and the bottom tie plate be freely movable in relation thereto. The fixation of the bottom tie plate and the top tie plate, respectively, to the water channels and the control rod guide tubes, respectively, may be permanent or detachable. A permanent fixation may, for example, be achieved by bulging the respective water channel and control rod guide tube when these are inserted into the bottom tie plate and the top tie plate, respectively. A detachable fixation may, for example, be achieved by arranging a bayonet coupling between the respective water channel and control rod guide tube and the bottom tie plate and the top tie plate, respectively. When designing fuel assemblies for an existing pressurized-water reactor, the length of the fuel units is determined by the distance between the existing spacers. This is because in a pressurized-water reactor the spacers are adapted to rest against the spacers at the same level in adjacently located fuel assemblies. The fuel units are formed with top and bottom tie plates with a supporting function corresponding to that of the replace spacer. To avoid the risk of bending of the rod or vibration, it may be suitable in certain cases to arrange spacers also in a fuel unit. Alternatively, two fuel units may be arranged between two existing spacer levels. This design provides a very stiff construction since a fuel unit will then only have a length of the order of size of typically 250 millimeters. According to one aspect of the invention, a burnup-dependent flow resistance is achieved in the above-mentioned axial gaps between two adjacently located top and bottom tie plates. This is achieved with the aid of flow tongues arranged in flow openings which, in turn, are arranged in the top and bottom tie plates. The flow openings of the top and bottom tie plates are in the fuel assembly arranged one above the other with coinciding centre axes. The flow tongues are arrange such that the flow tongues in a bottom tie plate are arranged across spaces between the flow tongues in a top tie plate arranged adjacent the bottom tie plate. In a new fuel assembly where, for example, the bottom tie plate is arranged fixed to a control rod guide tube such that an axial gap is formed between two adjacently located top and bottom tie plates, this embodiment provides a low flow resistance. After a certain time of burnup of the fuel assembly and a growth of the fuel rods associated therewith, which grow more than the control rod guide tubes, the distance between the adjacently located top and bottom tie plates is reduced, whereby the flow tongues cause the flow resistance to increase to a corresponding extent. In an alternative embodiment of the invention, a burnup-dependent flow resistance is achieved by providing adjacently arranged top and bottom tie plates with flow openings with centre axes displaced in relation to each other. In a partially burnt-up fuel assembly, the flow resistance is greater because the distance between two adjacently arranged top and bottom tie plates has been reduced as a result of the growth of the fuel rods due to neutron irradiation during operation. By displacing the centre axes of the flow openings in a diagonal direction, an advantageous mixing of the coolant may be achieved within and between fuel assemblies. Alternatively, the centre axes of the flow openings may be displaced such that a mixing in the clockwise or counterclockwise direction is obtained substantially within one and the same fuel assembly. An advantage of this mixing is that an equalization of the temperature of the coolant and a reduction of its maximum temperature are obtained in the relevant mixing cross section. Such a mixing is described in greater detail in SE 9402074-0. Especially in fuel assemblies with a burnup-dependent flow limitation it may be advantageous to arrange spacers, according to the above, in the short fuel units. The burnup-dependent flow limitation may give rise to vibration due to transverse flows formed. The advantage of the fuel assembly according to the invention is that axial gaps may be achieved between the fuel units such that a differential growth of the fuel rods may accumulate in this region. Another advantage of the fuel assembly according to the invention is that the flow conditions of the coolant may be influenced by the fuel rods having a greater growth than the control rod guide tubes and the water channels during the life cycle of the fuel assembly. By arranging restriction means in the axial gaps between the fuel units, a burnup-dependent flow resistance may be achieved successively through the fuel assembly gradually in dependence on the burnup. The suspension of the fuel units onto a support structure, such as control rod guide tubes or water channels, implies that the mechanical strength requirements on the fuel units are reduced.
	description	This application is based upon and claims the benefit of U.S. provisional application No. 61/105,339, entitled “HIGH SPATIAL RESOLUTION X-RAY AND GAMMA RAY IMAGING SYSTEM USING DIFFRACTION CRYSTALS”, filed Oct. 14, 2008 by Robert K. Smither, the entire disclosure of which is herein specifically incorporated by reference for all that it discloses and teaches. This application is a continuation in part of U.S. patent application Ser. No. 11/479,797 entitled HIGH RESOLUTION X-RAY AND GAMMA RAY IMAGING USING DIFFRACTION LENSES WITH MECHANICALLY BENT CRYSTALS filed Jun. 30, 2006, Robert K. Smither, all of which is incorporated herein by reference. The United States Government has rights to this invention pursuant to Contract No. DE-AC02-06CH11357 between the U.S. Department of Energy and the University of Chicago, representing Argonne National Laboratory. 1. Field of the Invention This invention relates to a method for improving imaging of a source of radiation and to a device for imaging a source of radiation, and more specifically, this invention relates to a method and device for producing a high spatial resolution three-dimensional image of a source of x-ray and gamma-ray radiation for medical and other application by using a plurality of diffracting crystals, collimators, and detection devices. 2. Background of the Invention Cancer tumor cells have high rates of metabolism and multiply rapidly. Substances injected into the body tend to migrate to locations of such high growth and become incorporated in this new growth. If the injected substance incorporates a short-lived radioactive isotope, the location of a tumor can be detected by locating the region of high radioactivity. Aside from pinpointing tumor location, an image of the tumor is also desirable to ascertain its shape, size, and juxtaposition with adjacent structures. For many medical applications it is imperative that a tumor be detected as early as possible, and early tumors are very small in size. Thus their detection and identification requires the ability to image very small sources. Also, medical research often uses small animals, with very small organs, and the availability of devices with very high spatial resolution is of the utmost importance. One method used to detect tumors is to first inject a body with a biological substance that contains radioactive compounds such as the Technetium isotope 99mTc, which is a 140.5 kiloelectronVolt (keV) gamma emitter having a half-life of 5.9 hours. The gamma rays are detected by a large sodium iodide (Nal) scintillator crystal placed behind a collimator grid yielding at best an 8 millimeter (mm) resolution at the location of the source. The scintillator is viewed by a plurality of photomultiplier tubes and the location of a scintillation event is determined by a computer analysis of the relative intensity of the photomultiplier signals. The collimator/scintillator assembly is placed above and very close to the patient. Aside from this method yielding a low resolution of between approximately 8 mm and 1 centimeter (cm), the image produced is limited to the plane parallel to the surface of the scintillator. As such, the technique provides no depth information about the source. This deficiency can be remedied somewhat by adding another collimator/scintillator assembly below the patient, comparing the counting rate of the two scintillators, and thus estimating the position of the source along the line joining them. In the latest revision of this method the large Nal detector plus collimator is rotated around the patient, taking a plurality of images at different angles. This allows one to generate a three-dimensional image of the radiation emitting area. There are considerable additional costs associated with this method and the fact that this method has been introduced in spite of the additional costs underscores the importance of three-dimensional imaging. Another popular imaging technique is positron emission tomography (PET), used in diagnosis and medical research. In PET, a chemical compound containing a short-lived, positron-emitting radioisotope is injected into the body. The positrons (positively charged beta particles) are emitted as the isotope decays. These particles annihilate with electrons in surrounding tissue. Each annihilation simultaneously produces two 511 (keV) gamma rays traveling in opposite directions. After passing through collimators, these two gamma rays are detected simultaneously by scintillation detectors placed at 180 degrees to each other, and on opposite sides of the patient. The signals from the detectors' photomultiplier tubes are analyzed by a computer to facilitate the production of an image of the radiation-emitting region. Numerous drawbacks exist with scintillation detector tomography. For instance, the typical coarse resolution of no less than 8 mm often results in smaller structures being overlooked. This prevents early detection of cancerous tumors when they are least likely to have metastasized and when treatment is most likely to succeed. This is especially a disadvantage in the detection of breast cancer tumors wherein the tumors often become virulent before growing to a detectable size. Presently, mammography uses x-rays to detect tissue calcification. The assumption is made that this calcification is due to dead cancer cells and that there is a live cancer tumor in the immediate vicinity. Often however, there is no live tumor where calcification has been detected. In fact, the calcification may not have been due to a tumor at all. Unfortunately then, positive mammography results often lead to unnecessary surgical operations. Also, because poor spacial resolution often causes images of actual small tumors to be diffuse, variations in background radiation are often mistaken for actual tumors, leading to unnecessary surgical operations. This inadvertent incorporation of background radiation is an artifact of scintillation detector use wherein the detector must be large enough to cover a given area of the body. Aside from intercepting the radiation emanating from the source under observation, however, the large detectors also detect all ambient background radiation penetrating the scintillating region and this ambient radiation is analyzed as if it had been emitted by the source under observation. Another drawback to using imaging techniques incorporating scintillation detectors is that all of the various radiations emitted by the source are detected by the detectors. As such, a specific radiation having an energy indicative of a specific, injected isotope cannot be easily scrutinized. Recently, efforts have been made to improve scintillation detector tomography. Some PET instruments now achieve a resolution as small as 4 mm. Such improvements entail considerable expenditures and have the additional drawback that the improvement in resolution has come at the cost of a decrease in counting rate. This entails in turn either a longer examination time per patient or the injection of a stronger dose of radiation. Furthermore, the prospects for further improvements in resolution are limited by the fact that such improvements require collimators with ever smaller apertures, and therefore greater mass, together with lower count rates. This increase in collimator mass will increase the number of forward Compton-scattered photons in the collimators and these forward scattered photons are often indistinguishable from those emanating directly from the source. Significant improvements in spatial resolution and in detection efficiency as well as a three dimensional location of the source using a crystal diffraction method for focusing the radiation emanating from the source was disclosed in U.S. Pat. No. 5,869,841 (1999) (granted to the same inventor as the present invention and assigned to the same assignee) and incorporated herein by reference. The enhanced focusing provided in the '841 patent allows the use of much smaller amounts of radioactive substances in order to locate features of interest in the patient. Experiments at the inventor's laboratory have demonstrated the effectiveness of this method and have achieved a spatial resolution of 7 mm. While this is adequate under many circumstances, better spatial resolution would provide significant advantages. Thus a need exists in the art for an improved method and device for imaging x-ray and gamma-ray sources with sufficient spacial resolution to accurately observe structures smaller than 7 mm in size, even down to 1 mm in size or less. The invented method and the resulting device must have sufficient energy resolution to allow the imaging of radiation of a selected energy to the exclusion of others. The method and device also must limit the radiation to which the patient is exposed by incorporating a redirecting mechanism to detect radiation emanating from a tumor while disregarding ambient levels of radiation. It is an object of the present invention to provide a method and a device for high spatial resolution imaging sources of gamma-ray and x-ray radiation that overcome many of the disadvantages of the prior art. Another object of the present invention is to provide a device for improved spatial resolution in the imaging of sources of gamma and x-ray radiation emanating from a subject. A feature of the present invention is the use of an array comprising a plurality of collimator-crystal-collimator-detector elements, each element comprising a source collimator to direct radiation emanating from a narrow region of space onto a diffracting crystal and of a detector collimator to direct onto a detector radiation diffracted by that crystal. An advantage of the invention is the generation of a high-resolution image of the source while the subject is under examination. A further object of the present invention is to provide a device for improved spatial resolution in three-dimensional imaging of sources of gamma and x-ray radiation emanating from a subject. A feature of the present invention is the use of a plurality of non-coplanar assemblies each comprising a plurality of collimator-crystal-collimator-detector elements to record data simultaneously for analysis by a computer. An advantage of the invention is the rapid generation of a two- or three-dimensional image of the source while the subject is under examination. Still another object of the present invention is to provide a method for producing a high-resolution image of a small radiation source which is located in a patient. A feature of the invention is the use of high purity and high quality diffracting crystals 1 mm wide or less. An advantage of the invention is the imaging of millimeter size sources into images of comparable size. Another advantage of the invention is the obviation of unnecessary, invasive surgical procedures to locate a tumor. Yet another object of the present invention is to provide a radiation imaging method having a fast imaging time. A feature of the invention is using scintillation detectors to locate the approximate position of the radiation source and then the use of a high efficiency crystal diffraction system to produce a high-resolution image of the source that can be viewed by a multi-element detector array. An advantage of the invented method compared to typical pure scintillation detector methods is that the amount of radiation necessary to produce a high-resolution image is relatively small, with a one microCurie source producing an image in three minutes. Another object of the present invention is to provide a radiation imaging method wherein there is a sharp one to one correspondence between source location and image location. A feature of the invention is the use of narrow (1 mm or less) collimators between the sources and the diffracting crystals and between these crystals and the detectors. Another feature of the invention is the use of arrays of small size detectors (1 mm or less) to image the radiation. An advantage of the invented method is a sharp image of the radiating source as recorded by the detectors. Yet another object of the present invention is to provide a three-dimensional imaging device. A feature of the invention is the use of a plurality of movable detector arrays. An advantage of the invention is that a detailed three-dimensional image of a source can be obtained quickly, with a small amount of radiation, and at low cost. Another object of the present invention is to provide an economical and manageable imaging device. A feature of the invention is that each of its detector arrays comprises a plurality of thin, long, and arcuately bent individually mounted crystal units. An advantage of the invention is that a detector array can be built rapidly and that a defective crystal in one element of the array can be replaced quickly and at low cost. Still another object of the present invention is to allow the imaging system to observe radiation of a selected energy to the exclusion of other energies. A feature of the present invention is that the angle at which a crystal diffracts radiation depends very sensitively on the energy of the radiation. An advantage of the present invention is that the array can be so constructed as to select only radiation of the desired energy. In brief, the present invention provides a method for high spatial resolution imaging of one or more sources of x-ray and gamma-ray radiation comprising: locating the sources of radiation proximally to a point (this point preselected as a reference point); supplying a plurality of elements, each element comprising a first collimator, defined by a pair of first plates, wherein the first collimator is adapted to direct radiation emanating from locations proximal to the point; a diffracting crystal adapted to receive and diffract the radiation directed by the first collimator; a second collimator defined by a pair of second plates, wherein the second collimator is adapted to direct radiation diffracted by the diffracting crystal; and a detector adapted to detect the radiation directed by the second collimator; supplying a means for analyzing said detected radiation to collect data as to the type and location of the source of the radiation; and supplying a means for converting the data to an image. The present invention also provides a device for high spatial resolution imaging of one or more sources of x-ray and gamma-ray radiation comprising: a means for locating the sources of radiation proximally to a point; a plurality of elements each element comprising a first collimator, defined by a pair of first plates, wherein the first collimator is adapted to direct radiation emanating from locations proximal to the point; a diffracting crystal adapted to receive and diffract the radiation directed by the first collimator; a second collimator defined by a pair of second plates, wherein the second collimator is adapted to direct radiation diffracted by the diffracting crystal; and a detector adapted to detect the radiation directed by the second collimator; the device also comprising a means for analyzing said detected radiation to collect data as to the type and location of the source of the radiation and a means for converting the data to an image. The invented method can be further enhanced by using diffracting crystals bent in an arcuate shape, a multi-element collimator-crystal-collimator-detector array, and a plurality of such multi-element collimator-crystal-collimator-detector arrays. The present invention utilizes the phenomenon of crystal diffraction for gamma- or x-rays, wherein there is a unique correspondence between the energy of the photons incident upon a crystal and the angle at which they are diffracted. In the ‘Bragg diffraction’ case the photons are diffracted near the surface while in the ‘Laue diffraction’ case the photons are diffracted in the interior of the crystal, but in both cases the diffraction phenomenon is a volume effect that depends upon the internal structure of the crystal. Both categories of diffraction can re-focus the diffracted beam. Both Bragg and Laue diffraction are discussed in detail in previously referenced U.S. Pat. No. 5,869,841. For both Laue and Bragg diffraction, diffraction occurs only when the Bragg condition is obeyed, (equation 1):λ=2dhkl sin p (1)where λ is the radiation wavelength, p the angle between the direction of the radiation beam and the atomic layers with which the radiation interacts, and dhkl the spacing between these atomic layers from which are indicated by the Miller indices h, k, l, (one can convert energy E in keV to wavelength λ in Angstrom units by using the relation λ=12.397/E). Performance of the invented device depends on two properties of the diffracting crystals: acceptance angle and diffraction efficiency. With perfectly parallel atomic layers, only rays within a few arc seconds of p will be diffracted (i.e., the “acceptance angle” is only a few seconds of arc), so that one can obtain a large diffraction efficiency only if the rays are nearly parallel, i.e. only if the source is very far away. Acceptance Angle Detail This parallel-ray problem is avoided by increasing the acceptance angle. The acceptance angle controls the solid angle for diffraction at a single wavelength by the crystal and thus the sensitivity of the imaging system. It also controls the range of energies that can be diffracted by the crystal at any one angle. The acceptance angle in bent crystals is controlled by the length of the crystal in the direction of the photon trajectory and by the radius of curvature of the bent crystal. Atomic mass absorption in the crystal limits the size of the crystal one can use. FIG. 1A shows the effect of a mosaic structure for Laue diffraction. If imperfections are either naturally present or else artificially introduced within the crystal so that all the crystal planes are no longer parallel to each other, rays coming at different angles 39 will still find planes 40 for which the Bragg condition is obeyed. As seen in FIG. 1A, the imperfections in the crystal give rise to a three dimensional mosaic structure. The angle 41 between the rays 35 having the lowest angle p and those 39 having the largest p is the acceptance angle, also known as the “rocking angle.” Ordinarily, rocking angles of between 200 and 800 arc seconds are employed. This is adequate for a first scan where a spatial resolution of 4 mm suffices. A rocking angle of between 50 and 150 seconds of arc is required when a 1 mm spatial resolution is required. Thus the degree of imperfection can be tailored to obtain the desired acceptance angle. FIG. 1B shows that for Bragg diffraction the acceptance angle 41 can be increased if the crystal is curved toward the direction “D” of the radiation beam. Rays coming at different angles 47 will still find planes 40 for which the Bragg condition will be obeyed. The angle 41 between the rays 35 with the lowest angle p and the rays 39 with the largest p is the acceptance angle. The curved shape of the crystals produces a significant focusing effect. The highest degree of focusing for Bragg diffraction occurs when the radius of curvature is equal to L/sin p, where L is the distance from the source to the crystal. Performance with Bragg diffraction is improved as one reduces imperfections and impurities in the crystals. Advantages of Bent Crystals Many different kinds of crystals (bent, unbent, mosaic, etc.) can be used in this invention. The examples used in this discussion are bent crystals of nearly perfect silicon or germanium, where ‘nearly perfect’ denotes high-grade commercially available silicon or germanium. They produce high quality imaging in many applications. The acceptance angle of the crystal is controlled by the curvature in the crystal. The diffraction efficiency of the crystal also depends somewhat on the curvature of the crystal, but the fact that the diffraction efficiency remains near 100 percent for a large range of curvatures and energies (80 keV to 200 keV), allows one to vary the acceptance angle of the crystal without changing the diffraction efficiency. The properties and advantages of bent diffraction crystals are discussed in detail in U.S. patent application Ser. No. 11/479,797, “High resolution x-ray and gamma ray imaging using diffraction lenses,” by the present inventor and assigned to the same assignee, incorporated herein by reference. The large acceptance angle in a curved crystal is made possible by the fact that the photon travels through the crystal until it encounters a region where the Bragg condition (λ=2d sin p) is met, where λ is the photon wavelength, d is the spacing between adjacent crystal planes and p is the angle of incidence of the photon on the crystalline planes. At this location the photon is diffracted and, if the crystal curvature is large enough, the photon will not pass through a second region where the Bragg condition is met so it will leave the crystal undeflected. This allows for nearly 100 percent diffraction efficiency. If the curvature is too small, then the photon is partially diffracted back into the original direction. If the curvature is too large then the diffraction efficiency is reduced. In the energy range of 100 keV to 150 keV, the curvature can change by a factor of 20 without affecting the diffraction efficiency by a large mount. The classic Bragg focusing geometry is shown in FIG. 2A. This geometry can give a very sharp line focus at a detector 115 for a point source 111. The radius of curvature, R, needed to obtain a sharp focus is given by equation 2,R=L/sin p, (2)wherein L is the distance from the source to the crystal 89 and sin p is the sine of the Bragg angle. Note the heavily accented volume 93 on the front surface of the crystal where the diffraction takes place. This is the case for low energy photons. With high energy photons where mass absorption is low enough so that the photon can penetrate deep into the crystal, this type of focusing can also take place deep inside the crystal. With bent crystals the photon passes through the crystal until it finds the location where the Bragg condition is satisfied and is diffracted. See FIG. 1B. If the curvature is correct, i.e. if equation 2 is obeyed, then focusing occurs. If mass absorption is low enough for the photon to pass through the crystal in the long direction then this type of focusing can take place even when the photon enters the front end of the crystal and leaves through the back of the crystal. A sharp focus also can be obtained with a Laue diffraction geometry. The Laue focusing geometry is shown in FIG. 2B. The necessary radius of curvature needed for a sharp focus is smaller than in the Bragg geometry (equation 2). It also depends on the set of crystalline planes that has been chosen to provide focusing. Diffraction takes place in a very small volume. Thus the volume 97 depicted in FIG. 2B is a small volume element in a diffracting crystal. This volume is often only a few 10's of microns in all directions. Thus the curvature that counts is only in that volume. The thickness of the bent crystal affects the focusing properties, the diffraction efficiencies, and the area viewed at any one time. Advantages of this invention are that one can vary the thickness, the curvature, the length, the material, the selection of the crystalline planes and both the distance from the source to the crystals and the distance from the crystals to the detector and the size and type of detector (pixel, strip, wide or narrow) with considerable independence, to achieve the best results. The depth of focus for good resolution (1 mm or less) can be quite long, 20 cm to 100 cm. For high resolution one will use thin crystals, 2 mm or less. For low resolution and wide coverage one will use thicker crystals. Crystal Bending Detail The following steps can be used to provide elastically bent crystals: a) selecting a crystalline material and cutting from large single crystals single crystal slabs of desired thickness and with Miller indices orientation determined according to the radiation to be focused; b) forming sets of two or more juxtaposed plates, at least one of which plates is one of said crystal slabs, by contacting said plates with an uniform layer of adhesive placed intermediate the plates, wherein said glue hardens only when it is activated; c) bending to a predetermined curvature one or more of said sets by means of a bending apparatus that allows in-situ measurements of the curvature of the plates; d) activating said glue while the set of plates is in the bending apparatus; and e) releasing said set from the bending apparatus. Typical dimensions for the bent crystal of silicon are: length 1-2 cm, width 5-10 cm (into the plane of the figure and thickness 0.5 mm to 2 mm, bending radius of 20 to 40 meters. Typically, the bent crystal is cemented to a thin backing plate in the curved configuration. The collimators thin plates of lead, tungsten, or other high-atomic-number material. The Imaging System The present x-ray imaging system comprises one or more arrays each comprising a plurality of basic elements. The basic element 110 of the invented imaging system is shown in FIG. 3A. The radioactive source 111 emits gamma- or x-rays. A beam in a first direction 121 passes through a first or “source” collimator 112. X-rays 122 exiting from the first collimator are diffracted by a crystal 113 to a second direction 124 so as to pass through a second or “detector” collimator 114 and be detected by the detector 115 (FIG. 3A depicts Laue diffraction). In the Laue embodiment of the invention shown in FIG. 3A, the crystal is a bent crystal of silicon. Unbent, mosaic, or segmented crystals can also be used. Typical dimensions for a bent crystal of silicon are: length 1-2 cm, width 5-10 cm (into the plane of the figure) and thickness 0.5 to 2 mm, bending radius of 20 to 50 meters. Spatial resolution can be improved by placing a narrow slit 118 in front of the detector. Typically, the bent crystal is cemented to a thin backing of the same curvature as the crystal. The collimators are made of lead, tungsten, or other high-atomic-number material. Typically the collimators comprise parallel plates 123 separated by a distance d so as to form an elongated slit of width d. The orientation of the basic element 110 can be specified by a triad of three vectors of unit length: r, directed from the source 111 to the crystal 113 (the direction of photon travel from the source to the crystal); u, from the crystal 113 to the detector 115; and s, the direction of the elongated slit formed by the collimator plates 123 with, in standard vector notation s=u×r (s projects out of the plane of the paper in FIG. 3A). The detector 115 may comprise an array of photo diodes or a pixel detector similar to that used to record an optical image in a digital camera. A more suitable detector array can be made with a series of cadmium-telluride solid state detectors (CT detectors) or a cadmium-zinc-telluride (CZT detectors). FIG. 3B is a view of FIG. 3A taken along the line 3-3 which depicts three strip CT or CZT detectors 145, 146, 147, each of which comprises a plurality of sections 149. Electric signals from each section 149 is processed by a computer. These detectors made in thin strips and then subdivided in short sections result a long pixel type detector which improves the spatial resolution and reduces background. Also, a substantial fraction of the incident photons lose all their energy in these detectors and the detectors produce voltage pulses proportional to that energy. Then the computer circuitry can select pulses corresponding to photons emitted by the source 111 and exclude those due to ambient background. The width of the strip detector has an affect on the spatial resolution of the system. With a detector strip width of 0.3 mm the measured FWHM (Full Width at Half Maximum) spatial resolution is 0.33 mm. The use of bent crystals allows many different geometries, including focusing ones where the distance from source to crystal is not equal to the distance from the crystal to the detector. Also the lengths of collimators used to focus the radiation do not have to be equal, provided that no photon straight-line path from the source to the detector exists. It is also not necessary that the crystal curvature be uniform. Rather the curvature needs to vary slowly enough so that the diffraction zone can contain enough scattering centers to diffract the photon efficiently. Crystal materials other than silicon and germanium could be quite useful in this type of imaging system when grown as near perfectly as silicon and germanium. For instance quartz can be used for photon energies below 200 keV and tungsten for photon energies above 200 keV. In one embodiment of the invention, the lengths and positions of the collimators are such that no straight-line path exists for photons from the source 111 to the detector 115. The viewing angle and thus the angular resolution of the system using the above parameters can be as low as 50 arcsec. This corresponds to a spatial resolution of 0.25 mm FWHM for a source 50 cm from the crystal and 0.5 mm FWHM for a source at 100 cm from the crystal. Once the curvature is right for Bragg type focusing and the energy of the photon is high enough to penetrate the crystal, then the crystal enables resolutions as low as 0.2 mm FWHM. Bragg type focusing can occur in the Laue geometry as well, resulting in similar high spatial resolution. FIG. 4 is a view of FIG. 3 taken along the line 4-4 and provides a top view of the collimator-crystal-collimator-detector basic element 110 of FIG. 4. The dashed lines 221 and 241 indicate possible trapezoidal shapes for the collimator plates 123. In the configuration shown, the proximal ends 116 of the collimator plates are positioned medially from the distal crystal ends 117 of the first collimator plates allowing an increased viewing area in a direction perpendicular to the plane formed by the photon travel directions 121 and 124 in FIG. 3. The bent crystals can be seen to function as cylindrical mirrors, imaging a strip area of the source, i.e. a strip perpendicular to the plane of curvature onto a strip of detectors perpendicular to the plane of curvature. Imaging Arrays FIG. 5A shows a first embodiment of an imaging array 150 according to the present invention. The array comprises a plurality of the collimator-crystal-collimator-detector combinations 110 described above, with a distance L between adjacent plates 123. A plurality of collimator-crystal-configurations 110 can be combined so that a large source area is covered by the imaging system. If the system uses mainly the Laue mode of diffraction, then the crystal can be stacked close together. If the system uses the Bragg mode, where diffraction occurs at the surface of the crystal, then a small space must be left between the crystal elements for photons to be able to strike the concave surface of the crystal and then diffract from that surface. Typically the source collimators 112 are coplanar with their plates 123 mounted parallel to each other. Specific collimators will be designated as i,j, . . . . The collimators have parallel vectors ri, rj, . . . that are displaced from each other by the plate separation L. A medially located collimator k has its vector rk aligned with a point O around which the sources are located. One can define a viewing direction R for the array, where R is aligned with the vector ri for the array. Similarly, vectors si, sj, . . . designating the direction of the slit openings for the collimators are parallel and from these one can define a slit orientation S for the array, where S is a vector that is parallel with sl. The crystals 113 are long thin strips and their view of the source area is a long thin strip. The multiple crystal array in FIG. 5A allows one to view a large area of the source 111 a at once, depicted as region 111a (‘111 actual’). There are, however, areas in the forms of strips 119 of the source that are not visible at any one time. These are the areas facing the gaps between adjacent crystals. Thus one obtains a striped source image 111i (‘111 imaged’) if the source is viewed from the line 5-5 in FIG. 5A. This is shown in FIG. 5B. To view the whole source area one must move the imaging system or the object being viewed perpendicular to the direction D. This motion is small, just the spacing between crystal strips which can be as small as 1 mm. Thus the full image can be viewed in a short time. To provide a two- or three-dimensional image of a source or sources in a fixed object, a plurality of co-planar non-parallel imaging arrays can be used. In one embodiment shown in FIG. 6, a device having at three co-planar non-parallel imaging arrays A, B, and C, i.e. three co-planar arrays with three non-parallel viewing directions RA, RB, and RC, is used to provide a two-dimensional image of a source 111 (RA being a vector parallel to rAi, rAj, . . . of the collimators in array A, and analogously for RB, and RC. The slit orientation SA, SB, SC is out of the plane of the paper for all three arrays and the same for all three arrays. The use of three non-parallel viewing direction arrays enables the construction of an unambiguous two-dimensional image of a source under most circumstances. The need for three arrays arises from the fact that there can be more than one source point viewed by an array at any one time. Thus if there are sources at points (x1, y1) and (x2, y2), the detector data from two arrays cannot distinguish between this situation and one where the sources are at points (x1, y2) and (x2, y1). The third array removes this ambiguity. In the alternative, just as one can better depict properties of a 3-dimensional object by rotating the object in front of a camera (or revolving the camera around the object), one can remove the above ambiguity in the location of two sources by relative rotation between the sources and a system of one or two imaging arrays. One may construct a three-dimensional imaging system by using six or more fixed arrays. One suitable arrangement is where the six arrays constitute two intersecting orthogonal planes, with each plane containing three arrays. Also, a three-dimensional imaging system can be constructed with three arrays in a plane and with means to impart motion to the source orthogonally to that plane. FIG. 7 shows a preferred embodiment with an assembly of four arrays A, B, C, and D in an arrangement that allows imaging of a three-dimensional source. The directions RA, RB, RC and RD are such that no more than two lie in the same plane. Moreover, none of the vectors SA, SB, SC and SD are parallel to each other. When detectors in specific collimators have detected x-rays is then a simple mathematical procedure to obtain the coordinates of the source points by projecting back from their respective crystals along the r vectors of the collimators. It must be noted that in some specific symmetric cases, some source points may be masked by others. Such masking can be detected by relative translational or rotational motion between the source and the detectors. In an embodiment of the invention, rotation in the plane of the paper of the source with respect to the arrays (or rotation of the arrays with respect to the source) improves the speed with which an image is collected and its quality. In fact, a single array suffices in an imaging system where the source is rotated around a vertical axis and is also moved vertically. Also, it may prove advantageous to provide means to rotate individual arrays around an axis parallel to the source collimators. While in general one can obtain imaging of complex source configurations by using merely one array with a combination of lateral and rotational motions, the larger the number of arrays one uses, the smaller the amounts of radiation that must be injected in a patient. Other Imaging Arrays Using a Similar Approach FIG. 8 shows an alternative embodiment of an imaging array. In this embodiment the source collimators are not parallel but emanate radially from the source so that the crystals view a common source point. The view at that source point is still a strip of the object area and three scanning arrays are still needed to do efficient 2-D imaging but now all the detectors of each array view the same source point. This arrangement enables the detection and imaging of very weak sources, with the imaging being accomplished by relative rectilinear motion of the source with respect to the arrays. It is to be understood that the above description is intended to be illustrative, and not restrictive. For example, the above-described embodiments (and/or aspects thereof) may be used in combination with each other. In addition, many modifications may be made to adapt a particular situation or material to the teachings of the invention without departing from its scope. While the dimensions and types of materials described herein are intended to define the parameters of the invention, they are by no means limiting, but are instead are exemplary embodiments. Many other embodiments will be apparent to those of skill in the art upon reviewing the above description. The scope of the invention should, therefore, be determined with reference to the appended claims, along with the full scope of equivalents to which such claims are entitled. In the appended claims, the terms “including” and “in which” are used as the plain-English equivalents of the terms “comprising” and “wherein.” Moreover, in the following claims, the terms “first,” “second,” and “third,” are used merely as labels, and are not intended to impose numerical requirements on their objects. Further, the limitations of the following claims are not written in means-plus-function format and are not intended to be interpreted based on 35 U.S.C. §112, sixth paragraph, unless and until such claim limitations expressly use the phrase “means for” followed by a statement of function void of further structure.
	abstract	A working apparatus has: a working equipment for doing works on a structure; a folding/unfolding mechanism for conveying the working equipment to the working position in a folded state; a conveyance mechanism (such as a horizontal thruster) for conveying the working equipment and the folding/unfolding mechanism to the working position; a pressing mechanism (such as a ballast tank) for pressing the working equipment against the lower surface of the structure; and a traveling mechanism including a wheel for traveling along the lower surface of the structure and positioning the apparatus.
	abstract	The present invention refers to a radiation system (1) comprising an excentric gantry (100) arranged in connection with multiple treatment rooms (61-68) separated by radiation-shielding separating members (71-78). A movable rotation head (120) is connected to the gantry (100) and is able to move between, and direct a radiation beam (110) into, the treatment rooms (61-68). A simulator head (200-1 to 200-8) is preferably arranged together with the radiation system so it can be used in each respective treatment room (61-68). In such a case, while a first subject (40-1) is being irradiated in a first room (61), a treatment set-up procedure, including correct positioning of subjects (40-2 to 40-8) and irradiation simulation, can simultaneously take place for the other subjects (40-2 to 40-8) in the other treatment rooms (62 to 68).
	summary
	claims	1. A method for vitrifying waste comprising:forming a feed mixture that includes the waste, a source of stable vanadium, and at least one of glass frit or glass forming chemicals;vitrifying the feed mixture in a melter to produce a glass product that includes the waste, wherein the glass product includes no more than 10 wt % vanadium oxide. 2. The method of claim 1 wherein the source of vanadium is an additive that is combined with the waste. 3. The method of claim 1 wherein the source of vanadium is added as a separate component to form the feed mixture. 4. The method of claim 1 wherein the glass frit includes the source of vanadium. 5. The method of claim 1 wherein the source of the vanadium includes a vanadium compound that is capable of reacting and decomposing during vitrification to produce vanadium oxide that is incorporated into the glass product. 6. The method of claim 1 wherein the waste, the source of vanadium, and at least one of glass frit or glass forming chemicals are each fed separately to the melter. 7. The method of claim 1 wherein the waste, the source of vanadium, and at least one of glass frit or glass forming chemicals are combined before being entering the melter. 8. The method of claim 1 wherein at least one of the source of vanadium, glass frit, or glass forming chemicals is combined with the waste before entering the melter and at least one of the source of vanadium, glass frit or glass forming chemicals is fed separately to the melter. 9. The method of claim 1 wherein the feed mixture includes glass frit and the glass frit includes glass beads, cylindrical glass fiber cartridges, glass powder, and/or glass flakes. 10. The method of claim 1 wherein the method reduces the formation of molybdate yellow phases. 11. The method of claim 1 wherein the method reduces the formation of sulfate salt phases. 12. The method of claim 1 wherein the method reduces the formation of salt phases that incorporate molybdate, sulfate, and pertechnetate. 13. The method of claim 1 wherein the method reduces the formation of salt with one or more of chlorine, fluorine, chromium (chromate), and phosphorous (phosphate). 14. The method of claim 1 wherein the melter includes a joule heated ceramic melter or a cold crucible induction melter or a hot wall induction melter. 15. The method of claim 1 comprising calcining the waste in a separate process step prior to vitrification. 16. The method of claim 1 wherein the method increases the waste loading in the glass product. 17. A method for vitrifying high level radioactive waste comprising:forming a feed mixture that includes the high level radioactive waste, a source of stable vanadium, and at least one of glass frit or glass forming chemicals;vitrifying the feed mixture in a melter to produce a glass product that includes the high level radioactive waste, wherein the glass product includes no more than 10 wt % vanadium oxide. 18. The method of claim 17 wherein the glass product includes vanadium oxide.
	abstract	The present invention presents a stencil mask in which various surface patterns can be formed, and in which deformation is suppressed when charged particles are introduced. A stencil mask of the present invention is provided with a semiconductor stack. A first penetrating hole corresponding to an ion introducing area is formed in a first semiconductor layer of the semiconductor stack, and second penetrating holes are formed in a second semiconductor layer, these second penetrating holes having a width greater than the width of the first penetrating hole. The first penetrating hole and the second penetrating holes communicate and pass through the semiconductor stack. Beam members extending between adjacent second penetrating holes connect portions of the first semiconductor layer separated by the first penetrating hole.
	description	This application is a national phase application of PCT Application No. PCT/FR03/002078 filed Jul. 4, 2003, which claims the benefit of French Patent Application No. 02/08537, filed Jul. 8, 2002. which are both hereby incorporated by reference. A subject-matter of the present invention is a decontamination composition, solution and foam. The composition and the solution of the present invention make it possible to obtain an acidic or basic gelled aqueous foam which can be used to decontaminate surfaces. The present invention finds, for example, application in the decontamination of metal surfaces contaminated, for example, by grease, by irradiating inorganic deposits, by a highly adherent oxide layer or in the body of the material. It is very particularly of great advantage in radioactive decontamination, for example of large-scale nuclear plants of complex design or which are inaccessible, for which economy with regard to chemical reactants and liquid effluents used is necessary. For example, it is difficult to decontaminate the inside of large-scale tanks, for example of 20 to 100 m3, or plants for the reprocessing of spent fuel comprising solutions of fission products as the medium is a highly irradiating one. Specifically, the dose rate can reach up to 40 GyH at the bottom of the tank, at a depth of 7.5 m. This level of irradiation prohibits virtually any modification of the existing fittings of the tank. In addition, the presence of numerous cooling coils in the vessels does not allow the introduction of devices for the application of the decontamination treatments. Finally, the contaminated fluid cannot be extracted from the vessels with a view to recycling the foam without very expensive additional facilities. The existing transfer means and discharge lines for the fluids have therefore to be used. Numerous compositions and foams intended for surface treatments, in particular for cleaning, degreasing and/or radioactive decontamination treatments of surfaces, have been developed. Unfortunately, they all exhibit the same disadvantages: they have lifetimes which are too short and difficult to control. This is because the foams of the prior art rapidly drain, in a few minutes, and exhibit a lifetime, defined as the time necessary for the complete conversion of a given volume of foam to liquid, generally ranging from 1 to 10 minutes. This often means that, to ensure the effectiveness of the treatment, it is necessary to apply the foam repeatedly to the surface to be treated. The amount of cleaning effluents and the difficulty of the treatment are therefore enhanced. In addition, as the duration of contact of the foam with the surface is limited as a result of the short lifetime of the foams, the cleaning and treatment agents used often have to be chosen so as to be highly active over a very reduced time. Only high concentrations of products, or more corrosive products, can therefore be used. This limits the type of surface which it is possible to treat and results in greater pollution, increased difficulties in rinsing the surfaces and an increase in the cost of the treatment. There thus exists a real need for a foaming composition which makes it possible to overcome the disadvantages of the compositions of the prior art, that is to say which makes it possible in particular to prolong and to control the lifetime of the foam, to reduce the amount of effluents, to use less corrosive cleaning agents, to use these agents at a reduced concentration, and to reduce the difficulty, the pollution and the cost of the treatment. The purpose of the present invention is specifically to solve the numerous problems of the prior art by providing a composition intended to prepare an aqueous foaming solution which makes it possible to generate a foam which does not exhibit the disadvantages of the prior art. The composition of the present invention comprises: a foaming organic surface-active agent or a mixture of foaming surface-active agents, a gelling agent and, optionally, a decontaminating agent. The foams generated from the composition of the present invention thus comprise a gelling agent. This is because, unexpectedly, the lifetime of this foam is greatly increased in comparison with the foams of the prior art and the foam thus prepared exhibits a markedly improved ability, in comparison with the foams of the prior art, to remain in contact with a surface, even a vertical surface, for several hours, thus making it possible to provide for the decontamination of the said surface statically or in spray mode. This unexpected result leads to greater effectiveness of the treatment of the surface, if appropriate with reduced concentrations of decontaminating agents, for example cleaning, degreasing or decontaminating agents, and a reduction in the amount of effluents produced. In addition, it is possible to use active decontaminating agents which are less corrosive than those of the prior art owing to the fact that the contact of the foam of the present invention with the surface to be treated is prolonged. Advantageously, the composition of the present invention is an aqueous solution which comprises, per liter of solution: 0.2 to 2% by weight of a foaming organic surface-active agent or of a mixture of foaming surface-active agents, from 0.1 to 1.5% by weight of a gelling agent and, optionally, 0.2 to 7 mol of a decontaminating agent. This solution can be prepared very easily, for example at ambient temperature, by adding the surfactant or surfactants, the gelling agent and, if it is of use, the decontaminating agent of the composition of the present invention to an aqueous solution, for example water, with simple mixing. According to the present invention, the gelling agent is preferably biodegradable. It is advantageously an organic thickening agent exhibiting a Theological behaviour of pseudoplastic type. According to the invention, the gelling agent can be chosen, for example, from the group consisting of a water-soluble polymer, a hydrocolloid and a heteropolysaccharide chosen, for example, from the family of the polyglucoside polymers comprising trisaccharide branched chains, such as xanthan gum, for example, Rhodopol 23 (trade mark) sold by Rhodia. It can also be chosen from the group consisting of cellulose derivatives, such as carboxymethylcellulose or a polysaccharide comprising glucose as sole monomer, for example Amigel (trade mark) sold by Alban Muller International. According to the invention, the surface-active agent can be a foaming nonionic surfactant chosen, for example, from the family of the alkylpolyglucosides or alkylpolyetherglucosides. These surfactants are derivatives of natural glucose and exhibit the advantage of being biodegradable. Mention may in particular be made, by way of example, of the surfactants “Oramix CG-110” (trade mark) sold by Seppic or “Glucopon 215” (trade mark) sold by AMI. According to the invention, the surface-active agent can be an amphoteric surfactant chosen, for example, from the family of the sulphobetaines, from the family of the alkyl amidopropyl hydroxysulphobetaines, for example Amonyl 675 SB (trade mark) sold by Seppic, or from the family of the amine oxides, for example Aromox MCD-W (trade mark) or the cocodimethylamine oxide sold by Akzo Nobel. The composition of the present invention can comprise a single surfactant or a mixture of surfactants chosen, for example, from the abovementioned families. The composition of the present invention is presented mainly as a composition which makes it possible to generate a foam for the decontamination of a surface. Of course, the present invention also covers any composition which makes it possible to generate a foam, whatever its use, provided that it comprises a surface-active agent and a gelling agent. For example, the composition of the present invention can also be a composition comprising only these last two components and intended to prepare a rinsing foam or alternatively a composition additionally comprising a surface-treatment agent and intended to prepare a surface-treatment foam. The surface-treatment agent can, for example, be an antioxidant, an antiseptic, and the like. The decontaminating agent, when it is present, is chosen according to the use for which the composition is intended. When the composition is intended to generate a decontamination foam, the active agent is chosen in particular as a function of the nature of the contamination and of the surface to be decontaminated, for example an acid or a mixture of acids, a base or a mixture of bases, an oxidizing agent, for example H2O2, a reducing agent, a disinfectant, and the like. A person skilled in the art will know how to choose the decontaminating agent according to his requirements. According to the invention, the active decontaminating agent can be an acid or a mixture of acids, for example inorganic, advantageously chosen from the group consisting of hydrochloric acid, nitric acid, sulphuric acid, phosphoric acid and oxalic acid. According to the invention, the acid is advantageously present at a concentration of 0.2 to 7 mol, preferably of 0.3 to 7 mol, more preferably of 1 to 4 mol. These concentration ranges relate, of course, to the concentration of H+ ions. In addition, they are given for the preparation of 1 liter of foaming solution. They thus represent the concentration in mol/l in 1 liter of foaming solution prepared from this composition. According to the invention, the active decontaminating agent can be a base or a mixture of bases, for example inorganic, advantageously chosen from the group comprising sodium hydroxide, potassium hydroxide, sodium carbonate, and the like. According to the invention, the base is advantageously present at a concentration of less than 2 mol.l−1, preferably ranging from 0.5 to 1.5 mol.l−1. These concentration ranges relate, of course, to the concentration of OH− ions. In addition, they are given for the preparation of 1 liter of foaming solution. They thus represent the concentration in mol/l in 1 liter of foaming solution prepared from this composition. Thus, depending on the abovementioned composition chosen in accordance with the present invention, an acidic or alkaline foam may exhibit either properties of dissolution of irradiating radioactive deposits, for example for the removal of contaminating materials not attached to the surface, or properties of controlled corrosion of the surface for a contaminating material fixed to the latter. Advantageously, the composition of the present invention exhibits a viscosity at 0.3 rpm (Brookfield LVT, module x) of between 100 and 50 000 cP. This is because this viscosity makes it possible for the foam to have a prolonged lifetime and also makes possible the possibility of spraying this solution using a nozzle or of passing it through a porous packing to generate a foam. The foam can be generated from this foaming solution by any system for the generation of foam of the prior art: mechanical stirring, sparging, bead static mixer or any other device which provides gas-liquid mixing, such as the devices disclosed in FR-A-2 817 170, or then a device using a spray nozzle, and the like. The foam generated can act statically, it has a long lifetime, generally of between 1 and 10 hours, and makes possible a controlled duration of action on the surface as a result of the control of the drainage time by virtue of the gelling agent. The present invention also relates to a process for the decontamination of a surface comprising a stage consisting in bringing the surface to be decontaminated into contact with a foam prepared from the composition of the present invention, that is to say with a foaming solution in accordance with the present invention. The invention relates generally to the treatment, in particular to the decontamination, of surfaces of any type, for example of glass, plastic, metals, and the like, which may be large and which are not necessarily horizontal but which can be inclined or even vertical. It can be used, for example, to decontaminate tanks, ventilation conduits, storage pools, glove boxes, steam generators, pipes, floors, and the like. The decontamination foams can be used both in the context of the periodic maintenance of existing industrial plants and in the context of the dismantling of such plants. These plants can, for example, be nuclear plants or chemical industry plants in general. The foam can be brought into contact with the surface to be treated by conventional processes for filling, for example, a tank, a vessel or a pipe, the walls of which are to be decontaminated; for spraying onto the surface to be decontaminated; for circulating the foam in a plant, the surfaces of which are to be decontaminated; and the like. For example, the foam can be applied to the surface to be decontaminated by any conventional process for spraying by means of a pump and of a nozzle. For the spraying, the break-up of the jet of foam over the surface to be decontaminated can be obtained, for example, with a flat jet or round jet nozzle. The short time for recovery of the viscosity of the composition of the present invention allows the sprayed foam to adhere for a sufficiently long time to the surface onto which the foam is sprayed. For example, to decontaminate a tank, the process of the present invention can consist simply in filling the tank with the foam of the present invention in order for its surfaces to be in contact with the foam. The foam then naturally decomposes “statically” under the effect of its gravitational drainage. The term “static” is then in contrast to the dynamic application of the foams, consisting of circulating or spraying. The foam can also be applied solely to the surfaces of the tank without necessarily filling it. Consequently, another subject-matter of the invention is a process for the decontamination of a plant which comprises the simple introduction of the foam by simple filling inside the plant, the “static” maintenance of this foam inside the space, for example at a temperature of between 20° C. and 50° C., during the drainage time of the foam, generally between 1 and 10 hours and sufficient to guarantee the decontamination, and then, finally, the removal of the drained liquid simply by emptying. The decontamination treatment of the surface can consist of several applications of the same foam or with foams of different natures applied successively. Each of these treatments can comprise filling the space to be decontaminated or spraying the foam over a surface, statically maintaining the foam for several hours during its draining and removing the drained liquid simply by emptying. However, the inventors have noted that, as a result of the longer lifetime of the foam of the present invention than that of the foams of the prior art, a reduced number of applications, indeed even a single application, is sufficient to obtain effective treatment of a surface where several applications were necessary with the foams of the prior art. The duration of the contacting operation will depend essentially on the nature of the decontamination, on the composition and on the nature of the foam, and on the nature of the surface. Generally, a duration of the contacting operation which can range from 15 minutes to 10 hours is sufficient for an effective treatment. This duration will be adjusted according to requirements in the application made of the present invention. The present invention guarantees an effective treatment, in particular an effective decontamination, owing to the fact that the lifetime of the foam, and thus the contact time of the foam with the wall, is increased and adjusted by the addition of the gelling agent, which slows down the draining. In addition, on vertical surfaces or even roofs, the foams of the present invention, as a result of the presence of the gelling agent, adhere better than the foams of the prior art, which further increases the treatment or decontamination effectiveness on these surfaces. The drained liquid obtained at the end of the life of the foam of the present invention can be easily discharged by emptying and can be treated by conventional procedures for decontaminating liquid effluents. It can also be regenerated, for example in the way disclosed in the document FR-A-2 817 170, to reconstitute a foam. The process of the present invention can additionally comprise, after the stage of bringing the surface to be decontaminated into contact with the foam, a stage of rinsing the said surface using a rinsing foam or solution. The rinsing foam or solution can be any appropriate foam or solution depending on the nature of the decontamination foam and/or of the surface to be rinsed. It can be simply a conventional rinsing foam or a rinsing foam in accordance with the present invention, that is to say comprising simply a surfactant and a gelling agent and, optionally, a conventional buffer compound which makes it possible to neutralize the acidic or basic decontaminating agent used above or a compound for the treatment of the surface. It can also be an aqueous solution, for example water. Such a treatment, by “gelled foam”, in accordance with the present invention has many advantages in comparison with the existing treatments. First, the conventional advantages of foam treatment apply, that is to say in particular the reduction in the volume of effluents produced. This is because the foam is composed of a dispersion of air bubbles in liquid and can be characterized by its expansion “F” defined under standard temperature and pressure conditions by the following relationship (1):F=(Vgas+Vliquid)/Vliquid=Vfoam/Vliquidin which V represents a volume of liquid, of gas or of foam as is indicated. The decontamination foams prepared from the composition of the present invention advantageously exhibit an expansion of the order of 10 to 15. They thus make it possible to decontaminate a large volume, for example of 100 m3, with less than 10 m3 of liquid. Another advantage, in particular in the case of decontamination by spraying gelled foam over surfaces of radioactive plants, is that the gelled foam of the present invention produces smaller amounts of radioactive effluents as a result of its long lifetime, whereas spraying foams or aqueous solutions of the prior art produces large amounts of radioactive effluents for a limited effectiveness due to the low contact time with the surfaces treated. Another advantage of the present invention lies in the fact that, following the natural draining of the foam of the present invention, the contaminated drained liquid is recovered and the surface only has to be rinsed with a very small amount of water, that is to say approximately 1 liter/m2. Thus, less liquid effluent to be treated is generated subsequently. This results in a simplification in terms of overall procedure for the treatment of the contamination and a decrease in the pollution. Other characteristics and advantages of the present invention will become more apparent on reading the following examples, given, of course, by way of illustration and without limitation, with reference to the appended figures. The draining and effectiveness properties of foams prepared from five foaming solutions each comprising a reference mixture of two surfactants: Oramix (trade mark) at 8 g/l and Amoyl (trade mark) at 3 g/l, were studied. One formulation, referred to as reference formulation (allowing the generation of a reference foam), did not comprise decontaminating agent. The other four formulations differ in the nature of the decontaminating agent: 1st formulation: sodium carbonate at a concentration of 1 mol.l−1, 2nd formulation: a mixture of hydrofluoric acid at a concentration of 0.05 mol.l−1 and of nitric acid at a concentration of 2 mol.l−1, 3rd formulation: a mixture of oxalic acid at a concentration of 0.6 mol.l−1 and of nitric acid at a concentration of 0.5 mol.l−1, 4th formulation: a mixture of hydrogen peroxide at a concentration of 1 mol.l−1 and of nitric acid at a concentration of 3 mol.l−1. No cloud point was observed between 20 and 50° C. These foaming solutions were subsequently used to generate foams of controlled expansion using a glass bend static generator (Ql=foaming solution flow rate, Qg=air flow rate, F=(Qg+Ql)/Ql). An experimental protocol was developed to plot the drainage kinetics of each of the foams under conditions close to industrial reality by virtue of the device (I) represented diagrammatically in FIG. 1. In this figure, the following references indicate the following components of the device (I): (3): vessel for preparing the foaming solution; (5): foaming solution; (7): mechanical stirrer; (9) pump; (11): system for supplying compressed air; (13): flow rate controller; (15): foam generator; (17) pipes; (19): vessel for receiving the foam; (21): foam; (23): manual valve; and (25): tank for recovering the drained liquid. Each of the five formulations exhibits excellent foamability since foams with an expansion of greater than 10 were prepared. It emerges from the kinetic studies that the presence of the decontaminating agents does not modify or only very slightly modifies the drainage kinetics in comparison with the reference foam without decontaminating agent, as is shown in FIG. 2. Over all the formulations prepared, more than half the liquid drains in less than 8 minutes and the lifetimes of each of the formulations remain short (15 to 25 minutes). The addition of a small amount, that is to say of 0.1% by weight or 1 g/l, of xanthan gum, used as gelling agent within the meaning of the present invention, to the various foaming solution formulations of Example 1 stabilizes all the foams, as is shown in FIG. 3. The addition of 1 g/l of xanthan gum has the effect of considerably slowing down the draining of each of the foam formulations and of thus increasing the lifetime of the foam. The time t1/2 necessary for half the liquid present in the foam to drain and the lifetime t1, time for all the liquid of the foam to drain, are collated in Table 1 below for the various foams studied. TABLE 1Lifetime t1 and time t1/2 for various foam formulations1 g/l ofxanthanWithout xanthan gumgumLifetime t and half life t1/2 (in mint1/2t1t1/2t1and s)Nitric acid + hydrogen peroxide4′30″15′18′50′foamNitric acid + hydrofluoric acid4′30″15′24′60′foamNitric acid + oxalic acid foam6′20′36′80′Alkaline foam with sodium7′30″25′90′>120′carbonate With an amount of xanthan gum of 1 g/l, the time t1/2 is approximately 20 minutes for the two acidic formulations comprising hydrogen peroxide and hydrofluoric acid. The foam comprising oxalic acid is the most stable of the acidic foams, with a time t1/2 of close to 40 minutes. Finally, the alkaline foam drains very slowly since close to one and a half hours are required to recover half the liquid. These results show that the addition of a small amount of xanthan gum, of 0.1% by weight or 1 g/l, stabilizes all the formulations of Example 1. This is because very substantial gains with regard to the stability of the formulations are obtained since lifetimes of between 50 and 120 minutes could be observed from the simple addition of a small amount of xanthan gum. Tests have been carried out to demonstrate the connection between the amount of gelling agent and the lifetime of the foam. FIG. 5 is a graph illustrating the influence of the amount of xanthan gum on the delay in the draining (stability of the foam). As long as a liquid does not drain, the foam is stable. The foam does not drain for 20 minutes for 1 g/l of xanthan gum, 60 minutes for 2 g/l of xanthan gum and 120 minutes for 3 g/l of xanthan gum. It is also apparent, on this graph, that the foaming solutions without gelling agent drain immediately (t=0 minute). The effectiveness of the foams of Example 2 was furthermore tested for the decontamination of surfaces. This is because the object is to demonstrate that the foams prepared with the foaming solutions of the present invention can, for example, dissolve a reconstituted deposit of insoluble materials simulating a true irradiating deposit adhering to a wall. Stainless steel plates covered with adherent deposits are suspended in a 30 liter column made of plexiglass in the device (II) represented diagrammatically in FIG. 4. In this figure, the following references indicate the following components of the device (II): (40): plexiglass column; (42): suspended steel plate; (44): valve; (46) glass bead bed foam generator; (48): system for introducing compressed air; (50): pipe for conveying the foam generated to the column (40); (52): pipe for recovering the draining liquid; (54) valve; (56): pump; (58): pulsation damper; (60): filter; (62): withdrawal pipe; (64): reactor for the preparation of the foaming solution; (66): foaming solution; (68): mechanical stirrer; (70): thermometer; (72): water feed pipe; (74): pipe for feeding the foaming solution with compound; (76): pipe for feeding the foam generator (46) with foaming solution; (78): alcohol reservoir; (80): alcohol metering pump; (82): pipe for recovering the foam. The two plates (42) covered with the deposit to be dissolved are deliberately placed at the centre of the column. The column is filled until the samples are completely immersed and generation is halted when the topmost edge of each of the two plates is at a depth of 10 cm in the foam. This level of foam corresponds to 20 liters of foam and is intentionally limited in order to quantify the effectiveness of the top part of the foam. The slight immersion of the plates is disadvantageous since the foam dries up from the top under the effect of gravitational draining. The foam/deposit contact times are then shorter and may prove to be inadequate to provide effective dissolution. However, if the dissolution proves to be significant in the top part of the foam, it will be even more significant within the foam. The stopwatch is started when the column has been filled with 20 liters of foam and the foam is allowed to act statically. The sample is withdrawn after a given time in order to evaluate, by weighing, the dissolution of the deposit. If two samples have been positioned, one can be withdrawn after immersing for one hour, for example, and the other after two hours. To carry out these deposit dissolution experiments, the foams are obtained successively in the following way. 4 liters of a solution comprising one of the three reactants, the surfactants and the xanthan gum are prepared. The solution is placed with stirring in the reactor (64) thermostatically controlled between 20 and 50° C. A gas-liquid mixture in a known proportion is then subsequently generated through a bed of glass beads: approximately 12 liters per hour of acidic solution are mixed with a controlled gas flow of 180 liters per hour of air, to generate a relatively wet foam with a known expansion in the region of 14. Tests in the foam phase were carried out, for example with the carbonate-comprising foaming formulation comprising 1.5 g/l of xanthan gum with the Rhodopol 23 trade mark. The lifetime of the foam is then of the order of 2 to 3 hours. A first adherent deposit, sample 1, with a thickness of 0.5 mm, i.e. 1 g over 25 cm2, is placed within the column. The object of the test is to allow a carbonate-comprising foam to act statically and to recover the sample once the foam has returned to the liquid state. The test is carried out by preheating the foaming solution to 50° C., which makes it possible to obtain a temperature within the foam of 33° C. After one hour, the temperature of the foam is 30° C. and, after two hours, 28° C. After 3 hours, the temperature is that of the laboratory (27° C.) and the carbonate-comprising foam, obtained from a 1M solution, has completely drained. The sample, invisible at the beginning as immersed in the foam, appears completely freed from the deposit. This very positive qualitative result justified the introduction of two thicker MoZr deposits with a thickness of approximately 1.2 mm, so as to quantify the rate of dissolution over the first two hours of the treatment. One deposit is withdrawn after 1 hour of contact with the foam and the other after 2 hours. The results are collated in Table 2 below. It is apparent that the mass dissolved in two hours, that is to say 0.74 g or 0.71 g, is virtually double that dissolved over the first hour, that is to say 0.42 g (slightly less since the mean temperature over the second hour is lower by 3° C.). At a mean temperature of the foam of 30° C., the rate of dissolution of a deposit of 25 cm2 in contact with the foam is of the order of 0.4 g/h, or alternatively 0.2 mm/h, to be compared with the 0.8 g/h obtained in the liquid phase at 30° C. This rate of dissolution, virtually constant over the first two hours, shows, as in the case of the liquid phase, that the dissolution is uniform and homogeneous over the surface. It advantageously makes it possible to completely dissolve an irradiating deposit of 0.5 mm over 3 hours at 30° C. This result makes it possible to envisage using, instead of a rinsing operation with sodium carbonate in the liquid phase, a rinsing operation in the foam phase according to the invention, which makes it possible to reduce the amount of sodium used, which is disadvantageous for the subsequent formation of the containment glasses. TABLE 2Losses in weight of an MoZr deposit in contact with a static foamcomprising sodium carbonate (1 M)Loss inweightImmersionWeight ofduring thetime in theinitial depositimmersionSamplefoamTemperature(25 cm2) in gtime (g)21 hour1 h from 33° C.2.40.42to 30° C.32 hours1 h from 33° C.2.490.74to 30° C.1 h from 30° C.to 28° C.31 additional1 h from 36° C.1.750.49hour (48 hto 30° C.afterwards)32 additional2 h from 35° C.1.260.71hoursto 28° C.
	claims	1. A glass composition for vitrifying a mixed waste product, the glass composition comprising:40 to 50 wt % of SiO2, 11 to 16 wt % of Al2O3, 8 to 15 wt % of B2O3, 3 to 6 wt % of CaO, 1 to 3 wt % of K2O, 1 to 3 wt % of Li2O, 1 to 3 wt % of MgO, 15 to 19 wt % of Na2O, 0.5 to 3 wt % of TiO2, 0.5 to 3 wt % of VO2, 0.5 to 3 wt % of As2O5, 0.5 to 3 wt % of CeO2, 0.1 to 2 wt % of CoO, 1 to 3 wt % of Fe2O3, 0.01 to 0.1 wt % of MnO2, 0.1 to 1.0 wt % of P2O5, and 0.5 to 3 wt % of ZrO2. 2. The glass composition of claim 1, wherein the glass composition includes 0.5 to 1.5 wt % of As2O5.
043366147	abstract	A tube-in-shell heat exchanger wherein the tube bundle has a central spine which carries a series of bracing grids for the tubes. The grids are resiliently mounted on the spine so that differential thermal expansion of one group of tubes relative to other groups of tubes and structure can be accommodated without inducing severe thermal stress.
	claims	1. A combination of a reactor vessel, a reactor core arranged in the reactor vessel, and a decay heat removal system for removing decay heat of the reactor core,wherein the reactor core includes a control rod insertion hole into which a control rod is allowed to be inserted,wherein the reactor vessel includes an insertion hole formed on a top thereof, andwherein the decay heat removal system comprises:a first heat pipe which is placed in an upper plenum of the reactor vessel and arranged vertically corresponding to a position of the insertion hole formed on the top of the reactor vessel and corresponding to a position of the control rod insertion hole;a control rod drive mechanism which is connected to an upper plenum of the first heat pipe and drives the first heat pipe to move up and down so that the first heat pipe is selectively inserted in the control rod insertion hole of the reactor core; anda second heat pipe which is coupled to and in close contact with a bottom surface of the reactor vessel and removes the decay heat generated in the reactor core,wherein the first heat pipe is a hybrid heat pipe which is internally provided with both a neutron absorber and a coolant, the first heat pipe not only absorbing a neutron generated in the reactor core by the neutron absorber but also removing the decay heat generated in the reactor core by the coolant. 2. The combination according to claim 1, wherein the first heat pipe receives the coolant from an upper plenum of the reactor vessel as a heat sink, and performs cooling by absorbing the decay heat generated in the reactor core, and transferring the absorbed heat to the coolant. 3. The combination according to claim 1, wherein a fin is provided in a condenser of the first heat pipe, which employs air, water, nanofluid, seawater, nitrogen or liquid metal as a coolant for exchanging heat with the first heat pipe. 4. The combination according to claim 1, wherein the second heat pipe is a hybrid heat pipe which is internally provided with both a neutron absorber and a coolant, the second heat pipe not only absorbing the neutron generated in the reactor core by the neutron absorber but also removing the decay heat by the coolant. 5. The combination according to claim 1, wherein the second heat pipe is made of a flexible material, is curved corresponding to the shape of the bottom surface of the reactor vessel, and closely contacts the bottom surface. 6. The combination according to claim 1, wherein the second heat pipe is curved corresponding to the shape of the bottom surface of the reactor vessel, and closely contacts the bottom surface. 7. The combination according to claim 4, wherein the coolant in the second heat pipe comprises one of water (H2O), a nanofluid, a refrigerant, mercury (Hg), lithium (Li) and FLiBe (LiF-BeF2). 8. The combination according to claim 1, wherein a wick placed in the first heat pipe comprises one of carbon fiber, copper, stainless steel, zirconium alloy, silicon carbide (SiC), and boron carbide (B4C). 9. The combination according to claim 1, wherein a case material forming an outer appearance of the first heat pipe comprises one of stainless steel, zirconium alloy, inconel alloy and molybdenum alloy. 10. The combination according to claim 1, wherein the coolant in the first heat pipe comprises one of water (H2O), a nanofluid, a refrigerant, mercury (Hg), lithium (Li) and FLiBe (LiF-BeF2). 11. The combination according to claim 1, wherein a wick placed in the second heat pipe comprises one of carbon fiber, copper, stainless steel, zirconium alloy, silicon carbide (SiC), and boron carbide (B4C). 12. The combination according to claim 1, wherein a case material forming an outer appearance of the second heat pipe comprises one of stainless steel, zirconium alloy, inconel alloy and molybdenum alloy.
053316797	abstract	A fuel spacer for a fuel assembly comprises a plurality of tubular ferrules each forming a fuel rod insertion passage, a belt-like support member for supporting the tubular ferrules bundled in a lattice arrangement and a spring member for axially supporting the fuel rods disposed in the ferrules. The adjoining ferrules are joined together horizontally, each of the ferrules has at least one end to which a plurality of cutout portions are formed circumferentially of the end portion and at least one flat portion is formed between adjoining petal portions at which the adjoining ferrules are spot welded. Each of the cutout portions and petal portions of the tubular ferrule has various shapes such as trapezoidal, rectangular, triangular, V or M shape. The cutout portions and petal portions may be formed at both axial ends of the tubular ferrule. The adjoining ferrules are joined together such that an end portion of one ferrule to which the cutout portions are formed is spot welded to an end portion of another ferrule to which any cutout portion is not formed. The petal portions formed on a downstream side with respect to a coolant flow in a fuel assembly is twisted outward to provide revolutional flow and an opened end of the ferrule.
051184664	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates generally to nuclear reactor coolant system pumps and, more particularly, is concerned with an improved reactor coolant pump having an internal self-cooling arrangement. 2. Description of the Prior Art In pressurized water nuclear power plants, a reactor coolant system is used to transport heat from the reactor core to steam generators for the production of steam. The steam is then used to drive a turbine generator. The reactor coolant system includes a plurality of separate cooling loops, each connected to the reactor core and containing a steam generator and reactor coolant pumps. In one version of the reactor coolant system used in a nuclear power plant, the reactor coolant pumps are high inertia pumps hermetically sealed and mounted to the one steam generator in the respective coolant loop. Each pump has an outer casing, a central axially extending rotor rotatably mounted at its opposite ends by upper and lower bearings, and a canned motor located about the pump rotor between the upper and lower bearings. The motor includes a rotor section mounted for rotation on the pump rotor and a stator stationarily mounted to the casing about the rotor section. An impeller mounted at one end of the pump rotor rotates therewith and draws reactor coolant water axially through a central inlet nozzle in the pump casing and discharges the water tangentially through an outlet nozzle in the pump casing. The temperature of the reactor coolant water is typically in the range of from approximately 500.degree. to 600.degree. F. which is too hot to also use to cool the motor and bearings of the pump. Thus, a heat removal arrangement separate from, and which does not employ, the reactor coolant water has been utilized in the prior art. One heat removal arrangement includes an annular hollow jacket surrounding the motor, a set of coils contained in the jacket and surrounding the motor, and other sets of coils located adjacent the upper and lower bearings. The multiple sets of coils are connected in flow communication so as to define a closed path for circulation of an internal coolant fluid therein for cooling the bearings and motor. The annular jacket of the heat removal arrangement has an inlet and outlet connected in flow communication with an external source of a secondary coolant fluid which can then flow through the jacket over the set of coils contained therein. The secondary coolant fluid is typically at a temperature much lower than the temperature of the internal coolant fluid circulating about the closed path such that the heat carried by the internal coolant fluid gained from cooling the motor and bearings is readily transferred to the secondary coolant fluid through the one set of coils in the jacket. Use of the above-described heat removal arrangement of the prior art is necessary in reactor cooling systems where the temperature of the reactor coolant water is too high to also be useful in cooling the pump motor and bearings. A drawback of this prior art heat removal arrangement, however, is that it does increase the complexity of the pump. SUMMARY OF THE INVENTION The present invention provides an improved reactor coolant pump having an internal self-cooling arrangement designed to avoid the aforementioned drawback. The self-cooling arrangement of the present invention employs reactor coolant water from the main flowstream to cool the pump motor and bearings. The reactor coolant water from the main flowstream, and thus the self-cooling arrangement of the invention, can be used in those situations where the temperature of the reactor coolant water entering the pump is below approximately 200.degree. F. Reactor coolant water at that temperature circulated by the self-cooling arrangement of the improved pump can readily remove motor heat generated by electrical losses and bearing heat generated by friction, eliminating the need for use of an external secondary coolant fluid and a separate internal closed path coolant fluid. Accordingly, the present invention is directed to a pump for pumping a fluid. The pump comprises: (a) a casing defining an inlet for receiving a fluid, an outlet for discharging the fluid, and a passage interconnecting the inlet and the outlet through which the fluid can flow in a main stream from the inlet to the outlet; (b) a central rotatable rotor having an end disposed adjacent the annular passage of the casing; (c) at least one bearing rotatably mounting the rotor adjacent to the end thereof to the casing; (d) a motor disposed about the rotor and adjacent the bearing and being operable for rotatably driving the central rotor; (e) means mounted to the end of the rotor in communication with the annular passage and the flow of fluid therethrough and being rotatable with the rotor for creating a lower pressure at the inlet of the casing than at the outlet thereof for drawing fluid into the casing through the inlet thereof and discharging fluid from the casing through the outlet thereof after flow of the fluid in the main stream through the annular passage; and (f) a self-cooling arrangement defining a fluid flow loop in flow communication with the annular passage and in heat transfer relationship with the bearing and motor and being operable for diverting only a fraction of the fluid from and back to the main stream through the annular passage to cool the bearings and motor. More particularly, the fluid flow loop is composed of outer and inner annular loop portions. The outer loop portion extends generally coaxial with, but is located farther radially outwardly from, the central rotor than is the inner loop portion. The fluid flow loop also includes a plurality of entry and exit ports which open respectively into and from the outer and inner loop portions. The entry and exit ports are defined in flow communication with the annular passage. Particularly, the entry ports are located downstream of the exit ports and thus at points of greater pressure in the main stream of the fluid through the annular passage. Further, the self-cooling arrangement includes foreign particle deflectors provided with respect to the fluid flow loop so as to minimize passage of particles into the fluid flow loop and to collect those particles which do enter the loop at a desired location along the loop. These and other features and advantages of the present invention will become apparent to those skilled in the art upon a reading of the following detailed description when taken in conjunction with the drawings wherein there is shown and described an illustrative embodiment of the invention.
	description	The present disclosure claims priority from U.S. provisional patent application No. 61/805,626, filed Mar. 27, 2013, the entirety of which is hereby incorporated by reference. The present disclosure relates generally to methods for preparation of alpha sources of polonium. In particular, the present disclosure relates to methods using sulfide micro-precipitation. Polonium-210 (210Po) is naturally present at trace levels in the environment as a part of the uranium-238 (238U) decay chain. It is typically considered as one of the most radiotoxic nuclides: only one microgram of this alpha emitter (t1/2=138 d) may be sufficient to be fatal to an average adult, making it around 250000 times more toxic than hydrogen cyanide1,2. Due to its toxicological properties, studies have been done to determine 210Po in a variety of samples such as soils, sediments, water, food, tobacco leaves, cigarettes, urine, and biological materials3-12. Polonium (Po) samples for alpha counting are typically prepared by spontaneous plating on metallic discs. Although silver discs have typically been used for Po plating13, nickel, copper, and stainless steel discs may also be employed due to their lower costs14,45. Prior to being used, the metallic discs are typically polished and cleaned to remove the dust and the oxide layer at the surface16. They are then typically brought in contact with the sample in a minimum volume of diluted HCl solution (typically about 0.1 to 1 M) and agitated for about 3-6 hours at a higher temperature (e.g., 80-95° C.) to obtain the highest yields possible (typically about 90%)8,13-16. The metallic discs are typically subsequently rinsed with water17 and heated at relatively high temperatures (typically about 300° C.) for few minutes to oxidize the polonium and reduce the risk of contamination to the alpha detector15. Although this sample preparation technique is widely performed, this technique, in particular the heating step, may be inconvenient and time consuming. In addition, the plating is typically performed using in-house assemblies resulting in a low analysis throughput. In some example aspects, the present disclosure provides a method for preparing alpha sources of polonium, which may include: providing a sample of polonium in a solution; introducing a controlled amount of sulfide and a controlled amount of a metal capable of forming an insoluble sulfide salt in the solution, in order to co-precipitate polonium from the solution; and filtering out the precipitates. It will be noted that throughout the appended drawings, like features are identified by like reference numerals. In the presence of sulfide, Po2+ is expected to be insoluble in 1 M HCl with a solubility product constant of about 5×10−29 (see FIG. 8)18. This low solubility has been applied to separate polonium sulfide (PoS) from tellurium and bismuth in 1 M HCl19. As shown in FIG. 8, mercury, silver and copper sulfides are also insoluble sulfide salts20, which may enable their use as co-precipitating agent of PoS for the preparation of thin-layer counting sources by alpha spectrometry. The present disclosure considers the use of certain sulfide salts (such as certain transition metal salts), in particular CuS. However, other sulfide salts may be suitable, such as sulfide salts with low solubility (including those that may not be listed in FIG. 8, such as Cr and Co sulfide salts). For example, sulfide salts of Fe, Pb, Ni, Cd, Co, Cr, Mn, Tl and Zn may be suitable. Investigations similar to those described herein may be carried out to determine the suitability of using such other sulfide salts, as well as the suitable reaction conditions when using such other sulfide salts. Although the present disclosure describes investigation of reaction conditions for micro-precipitation of polonium from a solution containing HCl, other solutions (e.g., acidic or non-acidic solutions) may be suitable. Although the described example investigations consider HCl being added to provide an acidic solution, other acids (e.g., hydrofluoric acid, phosphoric acid or sulfuric acid) may be used in order to achieve an acidic solution. Investigations similar to those described herein may be carried out to determine the suitability of a given solution, as well as the suitability of other reaction conditions. Based on the present disclosure, relatively large batches of Po samples may be relatively rapidly processed to increase sample analysis throughput. A vacuum box system may be suitable for such an application of the present disclosure. Micro-precipitation methodologies using lanthanide fluoride for actinides21-23 and barium sulfate for radium-226 (226Ra) have been employed for the preparation of thin-layer counting sources by alpha spectrometry24,25, however techniques used in micro-precipitation of actinides and radium-226 typically are not expected to work for micro-precipitation of polonium. In various examples and embodiments, the present disclosure may provide a relatively robust, simple and/or fast method for the preparation of polonium counting sources for alpha spectrometry using sulfide micro-precipitation, for example using copper sulfide micro-precipitation. Although copper sulfide is discussed herein as an example, other sulfide salts may be suitable. Copper may be practically useful, for example compared to silver and mercury, as silver is typically light sensitive and less stable in solution (and may result in poor spectral resolution in alpha spectrometry); while mercury may be undesirable due to its toxicity. However, use of silver sulfide and/or mercury sulfide for co-precipitation of polonium may be appropriate in some applications, and is within the scope of the present disclosure. The present disclosure discusses suitable conditions for the co-precipitation of PoS, and examples of suitable ranges are described. Other ranges and sub-ranges may be possible. In various example studies, potential radionuclide and chemical interferences were also investigated. The possibility of using 209Po as a yield tracer to determine 210Po was also investigated. In the disclosed examples, described further below, Po was co-precipitated with CuS, filtered onto Eichrom Resolve™ filters and counted. The disclosed method may be faster, cheaper, and/or more convenient than conventional spontaneous plating on metallic discs and example studies disclosed herein found that similar yields may be obtained (about 80-90%). In example studies, described below, suitable conditions for the micro-precipitation method using CuS as co-precipitate were found (e.g., about 50 μg of Cu2+ in about 10 mL of about 1 M HCl). These reaction conditions may be compatible with conventional preparation and purification procedures for polonium samples (typically using about 0.1 to 1 M HCl). The example results showed that most susceptible radionuclide interferences (e.g., Ra, Th, U, Np, Pu and Am) for polonium isotopes (namely, 208Po, 209Po and 210Po) may be effectively removed. The effects of several transition metals (namely, Cu2+, Ag+, Fe3+, Fe2+, Pb2+ and Ni2+) on the yield and the resolution of alpha peaks were also assessed. The example results demonstrated the versatility of the presently disclosed method for environmental and/or biological matrix. In various example studies, the disclosed method has been applied to determine various amounts of 210Po using 209Po as a yield tracer. Development of an example method for micro-precipitation of Po using a sulfide salt as co-precipitate is now described. This example is provided for the purpose of illustration only and is not intended to be limiting. For example, although CuS is described as an example co-precipitate, other sulfide salts may be used, and may be appropriately selected (e.g., based on solubility product constants). Similarly, although HCl is described as being added to achieve an acidic solution, other acids may be added, or the solution need not be acidic. Certain example reaction conditions are also described as being suitable. These are also provided for illustration only and may be varied as appropriate, for example using appropriate investigation to determine suitability. Suitable Reaction Conditions Example studies were carried out to determine suitable reaction conditions for obtaining polonium using micro-precipitation, with CuS as co-precipitate. Conditions that were investigated included: amount of Cu2+ added to the solution, reaction time before filtering out precipitates, concentration of HCl in the solution, and total volume of solution used in the reaction. In these investigations, temperature and pressure were kept at ambient levels, as this may be more practical to implement. The disclosed investigations also arrived at a set of reaction conditions that were found to be particularly useful. However, other reaction conditions may be suitable. Similar investigations may be readily carried out to determine suitable reaction conditions using other sulfide salts as co-precipitate. For the solutions investigated, about 50 mBq of 209Po was added in disposable 50 mL conical polypropylene tubes. Suitable amounts of Cu2+ for the reaction was first investigated by adding known quantities of Cu2+ from a copper solution (about 500 mg/L in 1% v/v HCl). The co-precipitation was carried out in about 10 mL of about 1 M HCl by adding about 1 mL of 0.3% m/v Na2S solution. Then, the influence of the precipitation time was investigated. Using selected suitable conditions (in these examples, about 50 μg of Cu2+ and a reaction time of about 10 minutes), the influence of the HCl molarity and volume of the solution was investigated. FIG. 1 shows example results of micro-precipitation yields of 209Po as a function of Cu2+ amount added in about 10 mL of about 1 M HCl. Without the addition of Cu2+, a yield of 25±2% was obtained. By adding controlled amounts of Cu2+, the yield was found to improve. As shown in the example results, with the addition of about 1 μg of Cu2+, a yield of around 80% was reached. The yield was substantially the same when controlled amounts of about 5 μg, about 25 μg, about 50 μg and about 100 μg of Cu2+ were added. The yield obtained was close to the conventional spontaneous plating technique (typically about 90% in optimal conditions). The low yield observed when Cu2+ was not added suggests that part of divalent or tetravalent Po was precipitated as PoS, which may be consistent with its low solubility (see FIG. 8). This may also be consistent with the observed improved yield when a co-precipitating agent (in this case, Cu2+) is added. Note that Po4+ is a relatively strong oxidant (E°=1.03 V) compared to S2− (E°=0.14 V)27 and is expected to be reduced to Po2+ in those conditions, suggesting that no valence adjustment is required. Polonium yield was also investigated as a function of different interfering transition metals added, at different amounts (see FIG. 2). The spectral quality of alpha spectrometry of the resulting alpha counting source was also investigated (see FIG. 3). From these example results, it was determined that a controlled amount of as little as about 1 μg of Cu2+ added to about 10 mL solution of polonium in HCl may be sufficient to obtain an acceptably high yield of polonium. It may be useful to introduce more Cu2+, in order to ensure that a sufficiently high yield is obtained, and to insure against the possibility that Cu2+ is caught up by impurities. However, as shown in FIGS. 2 and 3, using too large an amount of Cu2+ (e.g., much more than about 100 μg in a 10 mL solution) may be undesirable, as the co-precipitate obtained may have unacceptably low polonium yield (see FIG. 2) and/or may have poor energy resolution for alpha spectrometry (see FIG. 3). This drop in performance may be due to self-absorption of alpha particles in the thicker counting sources. In the example investigations, about 50 μg of Cu2+ added to about 10 mL of solution was found to be suitable. A lower quantity of added Cu2+ may also be suitable, for example depending on the specific sample matrix. Using more than a minimum amount of Cu2+ may also have a practical merit, since the formation of the brown colloidal CuS precipitate in the solution and on the filters may be observed, which may be convenient for routine laboratory work. FIG. 4 shows Po yield as a function of time. These example results were obtained using about 50 μg of Cu2+ in about 10 mL of about 1 M HCl, and measuring Po yield after a reaction time of about 10 min, about 20 min, about 30 min, about 1 hr, about 2 hr, about 3 hr and about 4 hr. The example results show that a Po yield around 80% was achieved after about 10 minutes. This yield was substantially the same up to at least 3 hours and slowly decreased afterwards. It was observed that, beyond 3 hours, the brown colloidal precipitate slowly coagulated with time and adhered onto the surface of the plastic tubes. After about 24 hours, the precipitate was completely adsorbed on the surface of the tubes leading to a clear solution. Thus, filtration may be conducted within about 3 hours after the addition of the sulfide, in order to avoid the loss of the precipitate due to adsorption. These example results also indicate that, using the disclosed method, a sufficient polonium yield may be obtained in as little as about 10 min, which may be advantageously faster than the conventional spontaneous plating methods, which typically takes at least 3 hours to reach an equivalent yield. FIG. 5 shows example results from an investigation of the influence of HCl molarity on the Po yield. This example study was carried out using about 10 mL of the solution, about 50 μg of Cu2+ over a reaction time of about 10 minutes. For relatively low HCl molarities (e.g., about 0.01 to about 1 M), the yield remained about the same (at about 80%), but decreased for higher HCl concentrations (e.g., above about 1 M). Solutions with concentrations higher than about 1 M were almost colorless and little precipitate was formed. The precipitates found on the filters obtained from reactions using about 0.01 M solutions were darker and less granular than those from reactions using about 1 M HCl. The FWHM (full width half maximum) of the alpha peak for 209Po was wider for the precipitates obtained using about 0.01 M HCl (150 keV) in comparison to those obtained using about 0.1 M HCl (55 keV) and about 1 M HCl (32 keV) (example results not shown). At lower HCl molarities, the crystallinity of the precipitates might be different, resulting in the observed variation in color and poor alpha energy resolution. For higher HCl molarities (e.g., above about 1 M), the loss in Po yield may be due to PoS and CuS salts being more soluble at lower pH and practically not precipitated and/or PoS and CuS not being precipitated because H2S was formed too fast and immediately vaporized, which may have prevented the micro-precipitation from occurring. To better understand the observed behavior, tests were carried out. To a solution of about 3 M HCl, 7 times more sulfide than normal was added, which formed the brown CuS precipitate with a yield of 93±4%. For a higher concentration of HCl (about 10 M), no precipitate was observed even though excessive amount of solid sodium sulfide was added to the solution and a significant amount of gaseous H2S was produced. Another test was performed to first form a CuS precipitate in about 1 mL of about 1 M HCl; about 20 mL of concentrated HCl was then added, which brought the concentration to approximately 10 M; but the brown CuS precipitate remained undissolved with a yield of 77±4% for 209Po. The results of these tests suggested that the low Po yield in low pH conditions likely was not caused by the dissolution of the precipitate, but rather by the fast vaporization of H2S that made the precipitate more difficult to form. Based on the results of these example studies, it was determined that a solution using a HCl concentration between about 0.1 and about 1 M may be suitable for co-precipitation of Po and CuS. Higher concentrations of HCl may help to reduce some potential interferences, thus HCl at a concentration of about 1 M HCl may be more useful. FIG. 6 shows example results from an investigation of the effect of solution volume on the micro-precipitation yield. The example investigation was carried out using about 1 M HCl and about 50 μg of Cu2+ in solution volumes of about 5 mL, about 10 mL, about 20 mL, about 30 mL, about 40 mL and about 50 mL. The Po yields were found to be 83±3% for reactions carried out using solution volumes of about 10 mL or less, and slightly decreased for reactions carried out using solution volumes of about 20 mL (75±3%). However, for reactions carried out using solution volumes of about 30 mL or more, the Po yields dropped to almost zero and the solutions became colorless. It was found that it was possible to achieve an acceptable yield (90±4%) by adding 7 times more sulfide to about 40 mL of about 1 M HCl solution. Since a larger amount of HCl may facilitate the formation of H2S and prevent the precipitation, more sodium sulfide may need to be added to maintain a sufficiently high concentration of S2− in the solution to initiate the micro-precipitation. As described above, Po micro-precipitation with CuS as a co-precipitate may be achieved relatively quickly (e.g., in about 10 min, and up to about 3 hr) with sufficiently high yields (e.g., about 80% or greater) in about 5 to about 10 mL of about 0.01 to about 1 M HCl using about 1 to about 100 μg of Cu2+. In particular, a sufficiently high yield of Po was found to be obtained with CuS as a co-precipitate, using a reaction time of about 10 min, a solution volume of about 10 mL, HCl concentration of about 1 M and about 50 μg of Cu2+. An example suitable method for co-precipitation of polonium with CuS is described below. Suitable conditions for this example method were determined based on the investigations described above. Variations on this example method may be possible. In 50 mL polypropylene conical tubes, about 50 mBq of 209Po may be added and mixed into about 10 mL of 1 M HCl. For each sample, 7.87×10−7 moles of Cu2+ (about 50 μg) followed by 4.17×10−5 moles of S2− may be added and the sample vigorously shaken. After sitting for about 10 minutes, the sample may be filtered using a suitable filter, for example through a 0.1 μm Resolve™ filter (Eichrom Technologies Inc., Lisle, Ill.). Prior to the filtration, the hydrophobic filter should be wetted, such as with 1-2 mL of 80% ethanol, followed by 1-2 mL of UPW. The sample may be then filtered, for example at a low flow (approximately 3-4 mL/min) using a vacuum box (Eichrom Technologies Inc., Lisle, Ill.). After the final rinse (e.g., using 1-2 mL of 80% ethanol), the precipitate is dried (e.g., air dried for few minutes) and subsequently mounted on a stainless steel disc, for example using double-sided adhesive tape, for counting by alpha spectrometry. The present disclosure may provide a useful alternative to the conventional spontaneous plating methodology for the preparation of Po alpha counting sources. The disclosed micro-precipitation method may be faster and easier to operate. It has been found that, for example, using a 12-holes vacuum box for filtration, it may be possible to perform all the preparation steps and process 12 samples within about one hour. Since current conventional spontaneous plating methods reported for the determination of Po are typically performed in a relatively small volume of 0.1 to 1 M HCl, similar to the reaction conditions of the present disclosure, the disclosed micro-precipitation technique may be readily implemented into current practice. Interference Assessment Further example studies were carried out to investigate the possibility of interference by radionuclide and other chemicals in examples of the disclosed method. Radionuclide Interferences Other alpha emitters, including Ra and actinide nuclides (e.g., Th, U, Np, Pu and Am), may interfere with counting of Po isotopes of interest (208Po, 209Po or 210Po) if they were to co-precipitate with the sulfide salt. Decontamination factors for Ra and example actinides were determined and are shown in FIG. 9, along with their alpha energies and emission intensities of the most susceptible interfering isotopes. An investigation of this possible interference is described below. For each sample, approximately 50 mBq of Ra and actinide standards were added in about 10 mL of about 1 M HCl. The example CuS micro-precipitation procedure described above (using about 50 μg of Cu2+ in about 10 mL of about 1 M HCl, over a reaction time of about 10 min) was followed and the filtrate solution was collected. Radium was determined in the filtrate using a barium sulfate micro-precipitation procedure as previously published by Maxwell25. Actinides were measured in the filtrate using cerium fluoride (CeF3) micro-precipitation as previously described by Dai24,26. The CuS filter and the filtrate samples were both counted by alpha spectrometry, and the decontamination factor was calculated as the ratio of the Ra or actinide activity in the filtrate to that on CuS filter. A moderate decontamination factor (134) was obtained for Ra; whereas higher decontamination factors (greater than 400) were obtained for actinides. These example results demonstrate that Ra and actinides are not expected to form insoluble sulfides in acidic solutions. Therefore, similar to the conventional spontaneous plating technique, no purification may be required to remove potential radionuclide interferences for Po samples obtained using the disclosed method. Chemical Interferences The effects of some example transition metals (in this example study, Ag+, Cu2+, Fe3+, Fe2+, Pb2+ and Ni2+ were considered) that could co-precipitate with sulfide were also evaluated. After the addition of known quantities of those elements and 50 mBq of 209Po, the example CuS micro-precipitation procedure described above (using about 50 μg of Cu2+ in about 10 mL of about 1 M HCl, over a reaction time of about 10 min) was applied to prepare the counting sources. FIG. 2 shows example results illustrating the influence of the example transition metals on the Po yield. FIG. 3 shows example results illustrating the influence of the example transition metals on alpha energy resolution. For Cu2+, which is also the co-precipitating agent, the Po yield remained substantially constant (about 80%) up to about 1000 μg of added Cu2+ and then decreased to 25±3% for about 10000 μg of added Cu2+ (see FIG. 2). The FWHM increased as the amount of added Cu2+ increased. For the Po samples obtained from solutions containing more than 100 μg of added Cu2+, FWHM of about 328 keV was reached with about 10000 μg of added Cu2+ (see FIG. 3). The precipitates on the filters were observed to be darker as the amount of added Cu2+ increased. Similar results were obtained for Ag+. The Po yield was relatively stable up to about 1000 μg of Ag+ added to the solution and decreased afterwards (see FIG. 2). The energy resolution was more affected for the Po samples obtained from solutions with more than 100 μg of added Ag+ (see FIG. 3), and the filters were observed to be darker as the amount of added Ag+ increased. Because of the relatively low solubility of silver sulfide (see FIG. 8), silver was expected to completely precipitate in the presence of S2-, resulting in a lower yield and poor energy resolution due to the self-absorption of alpha particles by the thicker precipitate. Although not tested, it is expected that Hg2+ would behave similarly to Ag+ due to its low solubility (see FIG. 8). For biological and environmental samples with a high concentration of Cu2+ or Ag+, additional purification (e.g., using an Eichrom Sr Resin)28 may be required to reduce these chemical interferences before the micro-precipitation. The example results also show that the micro-precipitation Po yield decreased as the amount of Fe3+ added to the solution surpassed about 100 μg and a minimal yield was found at about 1000 μg of added Fe3+ (40±3%). The Po yield increased to 62±3% at about 10000 μg of added Fe3+ (see FIG. 2). However, the alpha energy resolution was found to be not affected as the amount of added Fe3+ increased (see FIG. 3). The brown color characteristic of CuS was observed on the filters for the precipitate obtained from solutions containing about 100 μg of added Fe3+ or less; but the filters for the precipitate obtained from solutions containing about 500 and about 1000 μg of added Fe3+ were white. For the precipitate obtained from solutions containing about 10000 μg of added Fe3+, the filter was pale yellow. Also, the CuS precipitate formed very slowly for the solutions with about 100 μg of added Fe3+ and no visible coloration in the solutions was observed for higher added Fe3+ quantities. These results suggest that the precipitation of CuS was hampered as excessive ferric ion might compete with Cu2+ and form a complex with S2− in the solution. It may be that, as the ferric sulfide complex was quickly formed, fast consumption of the S2− in the solution and high solubility of ferric sulfide prevented the micro-precipitation of CuS from occurring. To verify this hypothesis, the solubility of ferric sulfide was examined. In about 1 M HCl solution containing about 1000 μg of Fe3+, 10 times more sulfide was added, which changed the yellow complex of FeCl2+ to colorless with no precipitate formed. Furthermore, the black Fe2S3 precipitate prepared in water was found to be soluble in about 1 M HCl and the H2S gas was produced. These tests indicate solubility of Fe2S3 in 1 M HCl. In another test, 4 times more sulfide was added to a solution of about 1000 μg of added Fe3+, and an improved yield of 93±4% was achieved. For the solution containing about 10000 μg of added Fe3+, a pale yellow precipitate was observed, possibly due to the formation of trace Fe2S3 that adsorbed FeCl2− in the presence of high concentration of Fe(III) in the solution. For verification, a test was performed by filtering a mixture of S2− and about 10000 μg of added Fe3+ with no Cu2+ added, and a yellow precipitate was observed. In addition, the alpha energy resolution was found to be not affected by the amount of Fe3+ added, confirming that only low quantity of the precipitate was produced. For Fe2+, the Po yield was found to be consistently high except for about 10000 μg of added Fe2+ (46±3%, see FIG. 2). The alpha energy resolution was found to be not affected (see FIG. 3). The filters showed the characteristic brown color of CuS precipitate, except that the precipitates obtained from solutions with about 10000 μg of added Fe2+ were white. A slower CuS precipitation was observed for the solutions of about 1000 μg added Fe2+. Adding more sulfide to a solution of about 10000 μg added Fe2+ increased the yield to 87±4%. Since ferrous ion is more soluble in 1 M HCl with sulfide than ferric ion (see FIG. 8), a higher quantity of ferrous iron (about 10000 μg) may be needed to compete for S2− and interfere with the CuS precipitation. These results may be useful since Fe(OH)3 pre-concentration procedures are typically used for the determination of Po in environmental and biological samples13,16. A reduction of Fe3+ to Fe2+ may be helpful to alleviate the influence of Fe3+ on the recovery. For Pb2+, the Po yields were relatively constant at 85±5% for the different amounts of added Pb2+ (see FIG. 2), but the FWHM value was found to increase considerably for the solution of about 10000 μg added Pb2+ (about 308 keV) (see FIG. 3). The filters had the characteristic brown color of CuS except that the precipitate obtained from the solution with about 10000 μg of added Pb2+ was dark black. The solubility of PbS is slightly higher than Fe2S3 (see FIG. 8), which may lead to less interference of Pb2+ on the micro-precipitation than Fe3+. For solutions with added Ni2+, the Po yield and the FWHM were not affected in the quantity range studied (see FIGS. 2 and 3). All the filters had the brown color of CuS precipitate. This may be expected since NiS is expected to be completely soluble in 1 M HCl. For the solutions with less than about 100 μg of transition metal impurities, no additional purification step may be needed. Similar to the conventional spontaneous plating technique, the interfering transition metals for the CuS micro-precipitation technique may be removed using suitable additional sample pre-treatment steps such as extraction chromatography13,28. The addition of such a purification step may be dependent on the sample matrix used. Determination of 210Po in Spike Samples To evaluate the performance of the disclosed micro-precipitation method, replicate samples spiked with known amounts of 210Po were analyzed using 209Po as the tracer for yield correction. A set of samples spiked with known amounts of 210Po were prepared by adding 5-100 mBq of 210Pb standard (in secular equilibrium with its daughter 210Po) to about 10 mL of about 1 M HCl. Then 50 mBq of 209Po tracer was added to the spike and blank samples for yield monitoring and correction. All the samples were then processed through the micro-precipitation procedure described above (using about 50 μg of Cu2+ in about 10 mL of about 1 M HCl, over a reaction time of about 10 min), and the counting sources were prepared for the determination of 210Po by alpha spectrometry. Example results are shown in FIG. 7. As shown, the measured activities of 210Po in the spiked samples ranging from about 5 to about 100 mBq agreed with the expected values. Good linearity (slope=1.0141) and correlation (R2=0.9999) were observed, demonstrating acceptable accuracy and precision of the disclosed method for the determination of 210Po in environmental and biological samples. The present disclosure may provide a relatively fast method for the preparation of alpha counting sources of polonium using sulfide micro-precipitation. In particular, CuS co-precipitation was investigated. The disclosed method may be relatively robust, rapid and simple, and may be easier and faster than conventional spontaneous plating methods for the measurement of 210Po by alpha spectrometry. Since the disclosed method may not require the use of a relatively expensive silver disc, the disclosed method may help to reduce the cost of Po analysis. The disclosed micro-precipitation technique may help to increase the sample analysis throughput and/or reduce the analysis cost. Thus, the disclosed method may be useful for 210Po radioassays in emergency samples, among other applications. In some examples (e.g., using about 0.1 to about 1 M HCl), the disclosed method may be compatible with typical conventional sample preparation procedures for 210Po using ion exchange or extraction chromatography purification techniques. Potential interferences of alpha emitting radionuclides and transition metals on the micro-precipitation yield and alpha energy resolution were also examined in example studies. Example results suggest that the disclosed method may be suitable to be adapted for the determination of Po in a variety of sample matrices by alpha spectrometry. The disclosed method may be applicable to routine and/or emergency radioanalytical procedures for the measurement of Po in environmental and/or biological samples. The embodiments of the present disclosure described above are intended to be examples only. Alterations, modifications and variations to the disclosure may be made without departing from the intended scope of the present disclosure. While the systems, devices and processes disclosed and shown herein may comprise a specific number of elements/components, the systems, devices and assemblies could be modified to include additional or fewer of such elements/components. For example, while any of the elements/components disclosed may be referenced as being singular, the embodiments disclosed herein could be modified to include a plurality of such elements/components. Selected features from one or more of the above-described embodiments may be combined to create alternative embodiments not explicitly described. All values and sub-ranges within disclosed ranges are also disclosed. The subject matter described herein intends to cover and embrace all suitable changes in technology. All references mentioned are hereby incorporated by reference in their entirety. 1. Cornett, J.; Tracy, B.; Kramer, G.; Whyte, J.; Moodie, G.; Auclair, J. P.; Thomson, D. Radiat. Prot. Dosim. 2009, 134, 164-166. 2. Harrison, J.; Leggett, R.; Lloyd, D.; Phipps, A.; Scott, B. J. Radiol. Prot. 2007, 27, 17-40. 3. Aoun, M.; El Samrani, A. G.; Lartiges, B. S.; Kazpard, V.; Saad, Z. J. Environ. Sci. 2010, 22, 1387-1397. 4. Gans, I. Sci. Total Environ. 1985, 45, 93-99. 5. Carvalho, F. P. Radiat. Prot. Dosim. 1988, 24, 113-117. 6. Savidou, A. K. K.; Eleftheriadis, K. J. Environ. Radioact. 2006, 85, 94-102. 7. Tso, T. C.; Fisenne, I. Radiat. Bot. 1968, 8, 457-462. 8. Khater, A. E. M. J. Environ. Radioact. 2004, 71, 33-41. 9. Meli, M. A.; Desideri, D.; Roselli, C.; Feduzi, L. J. Environ. Radioact. 2009, 100, 84-88. 10. Hill, C. R. Nature 1960, 187, 211-212. 11. Takizawa, Y.; Zhao, L.; Yamamoto, M.; Abe, T.; Ueno, K. J. Radioanal. Nucl. Chem. 1990, 138, 145-152. 12. Shabana, E. I.; Elaziz, M. A.; Al-Arifi, M. N.; Al-Dhawailie, A. A.; Al-Bokari, M. A. Appl. Radiat. Isot. 2000, 52, 23-26. 13. Matthews, K. M.; Kim, C. K.; Martin, P. Appl. Radiat. Isot. 2007, 65, 267-279. 14. Kelecom, A.; Gouvea, R. C. S. J. Environ. Radioact. 2011, 102, 443-447. 15. Karali, T.; Ölmez, S.; G. Yener. Appl. Radiat. Isot. 1996, 47, 409-411. 16. Eichrom Thecnologies, LLC., Analytical procedures, Lead-210 and Polonium-210 in Water, 2009. 17. Benedik, L.; Vasile, M.; Spasova, Y.; Wätjen, U. Appl. Radiati. Isot. 2009, 69, 770-775. 18. Bagnall, W.; Robertson, D. S. J. Chem. Soc. 1957, 1044-1046. 19. Figgins, P. E. The radiochemistry of polonium; N.A.S.-N.R.C., 1961. 20. Sillen, L. G.; Martell, A. E. Stability Constants of Metal Ligand Complexes; The chemical Society: London, 1964. 21. Vajda, N.; Törvényi, A.; Kis-Benedek, G.; Kim, C. K.; Bene, B.; Macsik, Z. Radiochim. Acta 2009, 97, 395-401. 22. Dai, X.; Kramer-Tremblay, S. Health Phys. 2011, 101, 144-147. 23. Maxwell, S. L.; Culligan, B. K.; Noyes, G. W. J. Radioanal. Nucl. Chem. 2010, 286, 273-282. 24. Dai, X.; Kramer-Tremblay, S.; Li, C. Radiat. Prot. Dosim. 2012, 151, 30-35. 25. Maxwell, S. L.; Culligan, B. K. (2012). J. Radioanal. Nucl. Chem. 2012, 293, 149-156. 26. Dai, X. J. Radioanal. Nucl. Chem. 2011, 289, 595-600. 27. Schweitzer, G. K.; Pesterfield, L. L. The Aqueous Chemistry of the Elements; Oxford University Press: New York, 2010. 28. Vajda, N.; LaRosa, J.; Zeisler, R.; Danesi, P.; Kis-Benedek, G. J. Environ. Radioact. 1997, 37, 355-372.
	description	This is a divisional application of application No. 12/205,513, filed Sep. 5, 2008; which was a continuation, under 35 U.S.C. §120, of International application PCT/EP2007/000572, filed Jan. 24, 2007; the application also claims the priority, under 35 U.S.C. §119, of German patent application No. DE 10 2006 010 826.4, filed Mar. 7, 2006; the prior applications are herewith incorporated by reference in their entirety. The invention relates to a nuclear engineering plant, in particular to a pressurized-water reactor, with a containment. The interior of which is divided by a gas-tight intermediate wall into a plant space containing a reactor pressure vessel and a primary cooling circuit and into an operating space which is walkable during normal operation. In addition, a plurality of overflow openings are formed in the intermediate wall. The invention further relates to a closure apparatus for an overflow opening disposed in the intermediate wall. Within the context of the safety-related configuration of a nuclear power plant, the reactor pressure vessel enclosing the reactor core is usually arranged in a containment. The containment contains a large number of redundant and diversified safety apparatuses which ensure reliable cooling of the reactor core in cases of disruptions to the normal operation of the nuclear power plant, in particular if the reactor core experiences a loss in coolant. The containment additionally ensures, as a gas-tight mantle which is often kept at a negative pressure with respect to the surrounding area, that even in case of an incident no radioactivity can leak into the surrounding area. The term containment also encompasses the atmosphere it encloses. Most of the plants built to date have so-called single-space containments without clear separation of the internal volumes which are not walkable during normal operation. This makes control and maintenance work in the containment difficult. If the interior of the containment is to be accessed, the reactor usually needs to be shutdown beforehand in a timely manner and the containment atmosphere be decontaminated, which can entail relatively long downtimes. In order to circumvent such difficulties, so-called two-space containments have already been proposed, in which the interior of the containment is divided by a gas-tight intermediate wall, generally a concrete intermediate wall, into a plant space containing the reactor pressure vessel and the primary cooling circuit and into an operating space which is screened off therefrom in terms of radiation and ventilation and can also be accessed during operation. In this concept, during normal operation the plant space with the active and high energy primary circuit components which are contaminated to a greater degree is therefore completely separated and screened off from the remaining containment areas which are walkable and thus accessible for maintenance purposes, the so-called operating space. In the case of incidents and especially accidents with increased core temperatures, which can be caused for example by a leak in the primary cooling circuit, a massive release of steam and explosive gases, primarily hydrogen, into the internal separated-off space areas of the containment can occur under certain circumstances. Especially in the case of a two-space containment of the type described above, a relatively rapid rise in pressure and a strong concentration of ignitable gases can occur within the comparatively small volume of the plant space. Even with smaller leak cross sections which lead only to a comparatively slow build-up in pressure in the plant space, critical concentrations of ignitable gases or explosive gas mixtures can arise in localized fashion due to the limited expansion volume. In such cases of incidents or accidents, the aim is therefore for an effective distribution of the incident atmosphere in the entire containment volume which restricts the local and global maximum concentrations of ignitable gases. To this end, the intermediate wall contains, between the plant space and the operating space, overflow openings which have hitherto been closed primarily with bursting screens and which open when the bursting pressure is reached. Since certain fluctuations in the bursting tolerance are unavoidable in the configuration and manufacture of the bursting elements, the solution described has the disadvantage, however, that, especially in the more likely incident situations with smaller leaks and slower build-up in pressure, typically only that bursting element with the lowest individual bursting pressure opens. The pressure equalization thus induced generally prevents the opening of the other bursting elements. As a result of the only partial space opening, the convective distribution of the gases which are capable of detonating is severely restricted, with the result that the risk of explosion must be reduced by way of increased use of inertization apparatuses and recombiners or the like. Such measures are generally comparatively complex and expensive and cannot be regarded as optimum in view of the level of safety which can be achieved. Other safety concepts which have to date been conceived and, in part, realized envisage, additionally or alternatively to the bursting screens, hand-operated flaps which can be operated or actuated by way of cable pulls or else electromotorically by the operators stationed outside the containment in order to release the overflow cross sections in case of emergency. However, hand-operated emergency devices have proven too slow and too unreliable in many incident scenarios, especially in the case of those which involve a very early hydrogen release, and are also severely limited in terms of number and opening cross section due to the high outlay necessary. In general, any erroneous operations which are based on human failure or on situational misjudgement should be precluded from happening in advance in a safety-critical plant such as a nuclear power plant. Alternative solutions with motor-driven closure flaps are problematic due to their dependence on an intact energy supply. Furthermore, these motor-driven closure flaps are not effective in the case of large fracture cross sections of high energy pipelines due to the delayed opening to relieve pressure and the significant space requirement (drive, transmission, etc.) results in that no relevant installation spaces for pressure-relief elements are available. Detailed H2 distribution and pressure build-up analyses carried out in accident situations involving massive H2 releases utilizing the above devices generally confirm this problem. It is accordingly an object of the invention to provide a nuclear engineering plant and a closure apparatus for its containment which overcome the above-mentioned disadvantages of the prior art devices of this general type, which ensures particularly high operational safety, in particular also in incident situations with relevant hydrogen release in the core area or primary cooling circuit, while keeping complexity in terms of manufacture and operation low. The invention furthermore specifies a closure apparatus, which is particularly suitable for use in such a plant, for a pressure relief and overflow opening disposed in the intermediate wall between plant space and operating space. With the foregoing and other objects in view there is provided, in accordance with the invention, a nuclear engineering plant. The plant contains a containment including an interior and a gas-tight intermediate wall dividing the interior into a plant space having a reactor pressure vessel and a primary cooling circuit and into an operating space being walkable during normal operation. The gas-tight intermediate wall has a plurality of overflow openings formed therein. The containment further has closure apparatuses with closure elements each closing a respective one of the overflow openings. The closure elements open automatically when a trigger condition associated with the respective overflow opening has been reached. The closure apparatuses opens in dependence on a pressure and opens independently of the pressure. The object is achieved according to the invention with respect to the nuclear engineering plant by virtue of the fact that the respective relief and overflow opening is closed by a closure element of a closure apparatus which opens automatically when a trigger condition associated with the respective overflow opening has been reached, wherein closure apparatuses which open as a function of pressure and which open independently of pressure are provided. The invention proceeds from the consideration that in a nuclear engineering plant of the type mentioned in the introduction, even in the case of an incident with possibly massive release of steam and flammable or explosive gases, the local and global maximum concentrations of these gases should be kept to a minimum already for type and design reasons. Mixtures which are capable of detonating and could endanger the integrity of the containment should not even be allowed to occur in the first place. In case of an incident occurring inside the plant space, rapid opening of the space should therefore follow for the purposes of the effective distribution of the incident atmosphere and limitation of the gas concentrations in the entire containment volume. The closure apparatuses, which are arranged in the overflow openings of the intermediate wall and which, in the closed state, ensure the space separation and the separation of plant space and operating space in terms of ventilation and radiation protection during normal operation, should additionally be constructed in accordance with the design principles diversity, passivity, redundancy and failure safety such that, if an incident occurs, an automatic and self-supporting opening or release of the overflow cross sections preferably without the need for outside energy follows. For a particularly effective distribution of the gases which are released due to the incident and their mixing with the entire remaining containment atmosphere and for effective pressure relief in the plant spaces in the case of large pipeline fractures, not merely individual but a number of, ideally many or even all of the closure elements should release the flow passages they block during normal operation at the same time or at least nearly at the same time. This should also apply in particular in the case of comparatively small leaks in the primary circuit and the associated slow build-up of pressure. To this end, according to the present concept, provision is made first for in each case one incident-related trigger condition to be associated with the closure apparatuses which are independent of one another, with the trigger condition taking into consideration the specific ambient conditions, operational parameters and influence factors at the respective site of use, i.e. at the site of the respective overflow opening. Second, in addition to a purely pressure-dependent trigger mechanism which can be realized for example simply by way of a conventional bursting film or the like, or as an alternative thereto, at least for some of the closure apparatuses at least one further trigger principle which is not dependent on pressure is provided. Such a diversified design of the closure apparatuses or of the associated trigger apparatuses and the selection, which is matched to the respective overflow opening, of the trigger parameters, the threshold values, the sensitivities etc. result in an early and nearly simultaneous response to even comparatively harmless incident situations in not just a single one, but in a large number of decentrally triggered closure apparatuses which are independent of one another. In order to achieve the intended diversity, the nuclear engineering plant can advantageously have at least two types of closure apparatuses, where the operating principles thereof which form the basis of the trigger procedure and/or the operation procedure differ. Alternatively or additionally thereto, however, at least one closure apparatus may also be provided, in which a plurality of trigger apparatuses which are based on differing operating principles are combined. A nuclear engineering plant, where closure apparatuses which open as a function of the temperature are provided in addition to the closure apparatuses which open as a function of pressure, is particularly advantageous. This is because, in the case of a slow build-up of pressure in the plant space, generally only one or few of the pressure-sensitive closure apparatuses or bursting elements open when the individual trigger pressure or bursting pressure is reached, which in itself could entail an only inadequate space opening and mixing of the containment atmosphere. Since during an incident usually also the temperature in the plant space increases at the same time, for example due to hot steam escaping from a leak in the primary circuit, further overflow cross sections are released by the temperature-sensitive closure apparatuses which result in an effective distribution of the incident atmosphere. Advantageously, at least one closure apparatus is configured such that it opens automatically as soon as the atmosphere pressure in the plant space exceeds a predetermined trigger pressure. Rather than a trigger condition relating to an absolute value of the pressure, it is also possible for a relative criterion to be used such that at least one closure apparatus opens automatically as soon as the pressure difference between the plant space and the operating space exceeds a predetermined trigger value. The trigger value is preferably approximately 20 mbar to 300 mbar. Furthermore, at least one closure apparatus is configured such that it opens automatically as soon as the local atmosphere temperature at a measurement location in the plant space exceeds a predetermined trigger temperature. Advantageously, the associated temperature-dependent trigger or unlocking apparatus is integrated in the closure apparatus, i.e. the measurement location is situated directly where the closure apparatus is mounted or at or in the overflow opening. Alternatively or additionally to this, it is also possible, however, for at least one closure apparatus to be provided, in which a closure element is coupled, via a remote-controlled apparatus, to a temperature-dependent trigger apparatus which is positioned near the ceiling of the plant space. It is expedient here if a closure element arranged in the lower region of the plant space can be actuated or unlocked by way of a fusible solder device arranged at a higher level or the like, with the result that the higher temperatures occurring in the upper regions in case of a temperature stratification (temperature layering) are utilized for the unambiguous triggering and reliable opening of the lower overflow devices, too. The trigger temperature is expediently matched to the room temperature in the plant space which is provided for normal operation and is advantageously kept below 60° C. by an air recirculation and cooling system. It is preferably selected from the interval from approximately 65° C. to 90° C. In a particularly expedient refinement, the trigger temperatures of the closure apparatuses which open as a function of temperature are selected in a staggered manner or such that they increase as the installation height of the apparatuses increases, e.g. from 65° C. to 90° C., which takes the temperature stratification in the plant space into account. Thus, a rapid and approximately simultaneous opening of all the closure apparatuses is favored in this case, too. In an advantageous development, at least one closure apparatus is configured such that it opens automatically as soon as the concentration of a gas which is flammable or capable of exploding and is present in the atmosphere of the plant space, in particular the concentration of hydrogen, exceeds a predetermined trigger concentration. Therefore, another diversification of the trigger criteria in addition to a pressure-sensitive and temperature-sensitive triggering is achieved by the concentration of the ignitable gases being monitored, wherein exceeding a predetermined limit value leads to the automatic opening of the closure apparatuses which close during normal operation the overflow cross sections. Advantageously, the trigger concentration is approximately 1 to 4 percent by volume of H2. In a particularly advantageous refinement, such a concentration-sensitive triggering can be achieved by a catalytic element, which releases heat in the presence of explosive or ignitable gases, or a H2 recombiner being arranged near a closure apparatus which opens as a function of temperature or near a temperature-sensitive trigger apparatus which is associated therewith such that the heat released thereby triggers the opening of the closure apparatus when a threshold value is exceeded. By way of example, the respective recombiner is arranged just underneath a fusible solder opening device which initiates the opening of the closure element, with the result that the increased operating temperature in the recombiner in case H2 is present causes reliable opening of the overflow cross section in a diversified manner—i.e. even in case of otherwise low room temperatures and independently of any release of steam. Instead of a recombiner, or in addition thereto, a catalytic element on the basis of a metallic carrier with washcoat coating and the catalytically active materials platinum and/or palladium can, for example, also be provided. If hydrogen is present, the exothermic heat of reaction ensures even at a concentration of from 1 to 4 percent by volume reliable triggering of the temperature-sensitive closure apparatus, even independently of the other room temperature. Furthermore, a first subset of the overflow openings, which are in each case provided with one of the closure apparatuses, is advantageously arranged in the lower third, in particular near the floor of the plant space, and a second subset is advantageously arranged in the higher regions, in particular in the sections of the intermediate wall which form the ceiling of the plant space. The difference in height between the high and low overflow openings is here preferably more than 5 m, in particular more than 20 m, in order to effectively drive in a passive manner the convection rollers by way of the difference in density of the atmosphere columns between the plant and the operating spaces. Here, there is expediently a number of recombiners mounted on the inside wall of the plant space between the lower and the higher overflow openings. Due to the reduced density resulting in the case of an incident from the failure of the space cooling or due to the release of steam, the chimney effect in the plant spaces which is possibly even increased by the heat of reaction of the catalytic recombiners is used to drive large-area convection rollers in the containment. In this manner, an increase of the overflow speeds to 0.5 m/s to 2 m/s or more is achieved. The manner in which the recombiners are expediently arranged here is selected such that an increased release of heat of reaction in the so-called chimney region of the plant spaces, i.e. in the upper regions of the steam generator towers and above the low overflow openings is achieved. In this manner, the recombiners are impacted by flow with increased entry speed after the space opening, which favors a particularly effective hydrogen reduction. Particularly preferably, the recombiners are arranged inside the plant space such that the H2 reduction power in the plant spaces is more than 20% of the overall H2 reduction power of for example more than 50 kg/h. Large pressure-relief areas become possible through the combination of these relief and overflow elements which can now all largely open as a function of pressure. Preferably, the area covered overall by the closure elements is approximately 0.1 to 0.5 times the horizontal cross-sectional area of the plant space. This also leads to a serious limitation in the difference pressure loading in the plant spaces even in the case of serious pipeline fractures, with the result that a planar steel beam holding and sealing construction is made possible even in the case of only limited load reduction (due to the technically possible section moduli). In this manner, the pressure loads occurring in the case of serious pipeline fractures and large leaks are limited to less than 0.5 times, in particular to less than 0.2 times, the design pressure for the containment. In order to ensure a particularly effective convection even in the case of small leaks, the total area of the overflow openings per steam generator tower is moreover advantageously more than 1 m2, in particular more than 5 m2. Advantageously, the nuclear engineering plant has a cooling apparatus, in particular in the form of an air recirculation and cooling system, for cooling the air in the plant space, wherein the cooling power of the cooling apparatus is preferably such that the room temperature in the plant space can be reduced during normal operation permanently to less than 60° C. The cooling minimizes in particular the chimney effect, which is established during normal operation, in the plant space with respect to the operating space and thus the pressure difference influencing the seals of the closure elements. Furthermore, the nuclear engineering plant contains a suction apparatus for the air which is in the plant space and a purifying apparatus, which is connected upstream or downstream on the flow side of the suction apparatus, for purifying and decontaminating the sucked-off air. During normal operation, a slight negative pressure of approximately 10 Pascal or more with respect to the operating space is thus produced in the plant space due to the air being sucked off from the plant space, wherein the pressure difference must not exceed the trigger value which leads to bursting of the bursting elements or else to pressure-dependent opening of the closure apparatuses. The sucked-off space air is purified as far as possible in aerosol and iodine filters before it is given off to the surrounding area. A particularly expedient closure apparatus for one of the overflow openings of the nuclear engineering plant advantageously has a closure element containing a bursting film or a bursting screen, wherein the closure apparatus is configured such that it automatically releases the flow cross section when a predetermined trigger temperature on the side of the surrounding area is reached. In other words, the two functions “pressure-dependent opening” and “temperature-dependent opening” are combined in a single closure apparatus, wherein the closure apparatus which is configured as a passive element opens automatically, autonomously, without delay and preferably without the need for outside energy (as per the failsafe principle) when one of the two trigger conditions (trigger pressure or trigger temperature) is reached, and releases the overflow cross section. This minimizes the number of closure apparatuses necessary for a diversified design of the opening mechanisms and enables the accommodation of correspondingly large opening areas of closure elements within planar ceiling constructions. Such a construction furthermore limits the difference pressure acting on the respective apparatus during maximum pressure increase, so that a particularly cost-effective construction of the respective opening element and of the structures of the plant spaces loaded by the difference pressure becomes possible. Here, the pressure-dependent opening in proven fashion by way of a bursting film or bursting screen, which forms the actual closure element or is integrated in the closure element, is realized. Advantageously, the respective closure apparatus contains an actuating apparatus which acts, in the case of a temperature-dependent triggering, directly on the bursting screen or on the bursting film and results in the destruction or tearing of the latter. This is preferably a spring-driven actuating apparatus. The closure apparatus furthermore preferably contains a locking apparatus which blocks the actuating apparatus before the trigger temperature is reached or compensates for it in terms of its effect. By way of example, a tension spring, which is pre-tensioned but blocked during normal operation by a fusible solder device, can be fixed directly and approximately centrally on the bursting film such that the tension spring is released when the melting temperature is reached, with the tension spring in that case tearing the bursting film following its release. According to an advantageous alternative embodiment, the bursting screen or the bursting film can also be mounted, or clamped in, on a frame element which is mounted such that it can rotate or pivot on the wall. The frame element can be fixed in the closed position by a locking element and the locking element is configured, or coupled to a temperature-dependent trigger apparatus, such that it is unlocked when the trigger temperature is reached. By way of example, the bursting element can be held in a frame element which is pressed against a planar supporting and sealing construction via a fusible solder apparatus. When the melting temperature is reached, the entire frame element is released and opens due to gravity and/or the spring force. If temperature-dependent triggering has not yet taken place, alternatively the bursting film or the bursting screen in the closure element opens when the bursting pressure is reached. The fusible solder apparatuses are expediently concentrated on a holding element or a few holding elements on the frame and are provided with tensioning elements. On account of the construction, the contact-pressure force is distributed over the sealing elements such that a sufficient tightness is achieved and a simple functional examination or a simple exchange is made possible. In another advantageous variant, the triggering and opening of the closure element is effected using a trigger apparatus which is disposed remote from the closure element and acts in a trigger event via a mechanical or pneumatic/hydraulic remote-controlled apparatus on the closure element or the associated locking element. For example, a trigger apparatus containing a fusible solder or a fusible bead can be mounted in the higher regions of the plant space, wherein the triggering is transmitted via a cable pull or a tensioned spring element or the like to a lower closure element. In an alternative system, in a trigger event, a pipeline which is connected to a pressure accumulator or to one or more hydraulic reservoirs and is closed during normal operation by a fusible bead or a fusible solder is released, so that as a consequence of the starting application of pressure, a closure element which is arranged at a distance is unlocked or opened. The closure element is preferably configured and mounted on the wall such that the process of opening is driven or assisted by the inherent weight of the closure element. Suitable for this purpose is in particular horizontal fitting in the ceiling, wherein for example a closure element, which is in the form of a type of closure flap, is mounted with one or more hinges in the ceiling wall or another supporting construction and flaps downward to open during the process of opening. In particular if the closure element is fitted vertically, it is expedient to provide a spring element or a compression leg which assists in the process of opening. Alternatively or in addition thereto, overflow cross sections can also be opened by way of motor-actuated overflow flaps. These flaps are in this case preferably kept permanently closed in a motor-driven way and open due to spring force and/or their inherent weight. Here, opening can be triggered via a suitable measurement and control apparatus when a predetermined absolute or difference pressure or a trigger temperature is reached. If there is no voltage at the drive motor or the associated process control technology or when the trigger criterion (e.g. pressure, temperature or gas concentration) is reached, the closure element then opens reliably and without the need for outside energy as per the failsafe principle. In order to limit the difference pressure loading, closure elements of this type can also be fitted with bursting films. The use of these overflow flaps, which are motor-actuated or held closed using motor force and can be especially in the form of louver flaps or of rotary pendulum flaps with spring return actuator, is expedient particularly in the low, temporarily cold inflow region of the plant space or of the intermediate wall delimiting it from the operating space, especially in combination with the temperature-sensitive and pressure-sensitive, large-area closure apparatuses, described above, in the high overflow regions. In order to ensure sufficient screening in the inflow opening region of the plant space, screening walls are provided upstream of the respective inflow device. The screening devices can also be arranged such that they are aligned radially in the lower third of the intermediate wall. The free convection cross section should be at least twice that of the inflow cross section. The respective bursting film or bursting screen of the closure element is advantageously configured such that it tears or breaks, when a predetermined trigger pressure is applied to it, in the direction of both sides depending on the direction of attack of the pressure force. Thus, an opening of the overflow cross sections is also triggered in the case of an external positive pressure in the operating space, caused for example by a fracture of a secondary-side live-steam line. Moreover, the bursting elements are expediently configured such that, when radiation forces impact, no relevant fractured pieces can occur which could possibly cause secondary damage. Therefore, the bursting films are expediently configured such that they break in the one direction by tearing of the provided bursting material webs and in the opposite direction primarily by kinking. For sealing purposes, easily tearable sealing films with a thickness of less than 0.05 mm can additionally be used. In the event of negative or positive pressures of 20 mbar to 300 mbar, the bursting films thus expediently open in both directions. The advantages attained by the invention are in particular that first the active and high energy plant components in a nuclear engineering plant, in particular in a pressurized-water reactor, are hermetically screened during normal operation against the surrounding operating spaces which thus remain walkable for any occurring control and maintenance work, and second, in incident and accident situations, with the danger of the release of ignitable gases, rapid, reliable and large-area space opening using passive, failure-safe elements on the basis of diversified operating principles is effected. The convective distribution of the released gases, optimized by a manner of arranging the overflow cross sections such that they are staggered in height, particularly reliably and simply effects a limitation of the maximum local concentrations, so that mixtures which are capable of detonating and could endanger the integrity of the containment are reliably avoided. Thus, the invention serves to increase the safety reserves in a nuclear reactor with simultaneous significant reduction in complexity and costs. Other features which are considered as characteristic for the invention are set forth in the appended claims. Although the invention is illustrated and described herein as embodied in a nuclear engineering plant and a closure apparatus for its containment, 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. Components having the same design or the same function are indicated by the same reference numbers in each of the figures. Referring now to the figures of the drawing in detail and first, particularly, to FIG. 1 thereof, there is shown in a longitudinal section a containment 2, which extends largely symmetrically with respect to a longitudinal axis 1, of a nuclear engineering plant 4 with a pressurized-water reactor. The containment 2 has inside nuclear components for generating hot and pressurized water steam, in particular a reactor pressure vessel 8 and a primary cooling circuit 10 connected to the former with a number of steam generators 12, and other system components. The shell forming the containment 2 also prevents activity transfers to the surrounding area in the case of a comparatively serious operational incident by closing in and keeping back coolant from escaping for example from a leak in the primary cooling circuit 10. Usually, the shell of the containment 2 is made of steel and is additionally surrounded by a non-illustrated concrete cladding. The nuclear engineering plant 4 is configured for a particularly high operational reliability, wherein at the same time an extremely economical mode of operation is made possible. For this purpose, an interior 14 of the containment 2 is divided by a gas-tight intermediate wall 16 into a plant space 18 containing the reactor pressure vessel 8 and the primary cooling circuit 10, i.e. containing the active and high energy components, and into an operating space 20 which contains the other system components which are only comparatively weakly loaded with radioactivity. The plant space 18 and the operational space 20 are possibly in turn divided into further subspaces, but this is of no fundamental importance for the safety-related concept which will be explained below and for the associated design principles. The intermediate wall 16 is generally configured in a comparatively massive steel concrete construction, so that in addition to the ventilation-side separation of operational space 20 and plant space 18, a not insignificant screening effect against the radiation which is released inside the plant space 18 is also still realized. Thus, the operational space 20 remains walkable even during the operation of the nuclear engineering plant 4, and is thus easily accessible for any necessary control and maintenance work. The operation of the reactor generally only needs to be interrupted for maintenance work on the high energy nuclear components which are contaminated to a greater degree and are situated within the plant space 18. Within the context of an operational incident, caused or accompanied for example by a leak in the primary cooling circuit 10, relevant amounts of ignitable and possibly explosive gases, in particular hydrogen, could also be released into the plant space 18 in addition to a release of hot steam. Since the volume of the plant space 18 is kept comparatively small, after a certain time critical concentrations would under certain circumstances be exceeded, at which an increased risk of explosion would have to be expected. In order to reliably preclude such critical states in advance, arranged in the intermediate wall 16 separating the plant space 18 from the operational space 20 is a number of overflow openings 22a, 22b, 22c which are closed in a gas-tight manner during normal operation by way of respectively assigned closure elements 24, but which release in the event of an incident or of a disaster state in the plant space 18 the respective overflow cross section early and, for reasons of an effective distribution of the incident atmosphere, in the entire containment 2. The respective closure element 24 is a constituent part of a passive closure apparatus 26 which effects, when a trigger condition which is considered to be an incident indicator is reached or exceeded, the opening of the overflow cross section automatically and without the need for outside energy. During normal operation, a slight negative pressure with respect to the operational space 20 is generated in the plant space 18 by sucking off the atmosphere therein; the pressure difference here is approximately 10 Pa to at most 1,000 Pa. For reasons of sucking off the air, a suction apparatus denoted 28 in FIG. 1 is provided, which contains a suction line 30, which is guided through the intermediate wall 16 and through the containment 2 from the plant space 18 to the outside, and a suction pump 32 connected thereto. Before it escapes into the surrounding area, the air sucked off from the plant space 18 is purified and decontaminated with the aid of a multistage purifying apparatus 34 which is connected into the suction line 30 and contains a number of aerosol and iodine filters. Furthermore, the heat arising in the plant space 18 is continuously removed by an air recirculation and cooling system (not illustrated further here) such that the room temperature in the plant space 18 is, during normal operation, below a maximum value of approximately 60° C. In the event of an incident which is due to a leak in the primary cooling circuit 10, the release of hot steam and gases produces an increase in pressure and also an increase in temperature inside the plant space 18, wherein furthermore the hydrogen content in the air or the concentration of other ignitable gases can rise. The pressure in the plant space 18 or the difference pressure with respect to the operational space 20 and the atmospheric temperature are therefore particularly relevant and suitable parameters for monitoring the nuclear engineering plant 4 for the occurrence of incidents. The closure apparatuses 26 are configured and constructed such that they automatically respond and effect the opening of the overflow cross sections as soon as one or more of the operational parameters (pressure, temperature, H2 concentration) exceeds a threshold value which is to be considered to be an incident indicator. Due to the diversified configuration of the closure apparatuses 26 with respect to their trigger and actuation mechanisms, an extremely reliable and comparatively large-area space opening is achieved here even in the case of the more likely incident situations with comparatively slow build-up of pressure. Moreover, the specific manner in which the overflow openings 22a, 22b, 22c are arranged in the intermediate wall 16 separating the plant space 18 from the operating space 20 ensures in the trigger event optimized flow conditions which promote rapid and uniform distribution of the released ignitable gases in the entire containment volume. For one, a number of comparatively low overflow openings 22a with pressure-sensitive and temperature-sensitive closure apparatuses 26 which open automatically in the trigger event are provided in vertical installation in the lower third of the plant space 18. Secondly, a plurality of such closure apparatuses are arranged in the overflow openings 22b of a ceiling 36 above the steam generator 12 (horizontal installation) and in the process form, as it were, an overflow ceiling. In order to adequately take into account the possible temperature stratification in the plant space 18, the trigger temperatures of the high closure apparatuses 26 with approximately 90° C. are selected to be higher than those of the low closure apparatuses 26 at approximately 65° C. Furthermore, in the area near the floor, motor-actuated overflow flaps 38 are installed in the overflow openings 22c located there, which are normally kept closed by an electromotor 40 and which in the trigger event, i.e. if the electromotor 40 is switched off or without voltage, open by way of spring force or gravity as per the failsafe principle. As already explained, the diversification of the trigger mechanisms ensures that in the event of an incident in the core region of the reactor or in the primary cooling circuit 10, the predominant or at least a relevant number of closure apparatuses 26 or closure elements 24 open. This results in the flow course illustrated by flow arrows 42 in the right-hand part in FIG. 1, where the hot gases and steam escape the high overflow openings 22b in the ceiling 36 of the plant space 18 upward, cool off on the above ceiling of the containment 2 and subsequently drop downward in the ring-shaped intermediate space 44 between the intermediate wall 16 and the inside wall of the containment 2, in order to finally flow in again through the opened overflow cross sections 22a, 22c into the plant space 18. Circulating the flow causes an extremely effective mixing of the previously separated “internal” and “external” containment atmosphere, which reliably limits the maximum concentrations of gases or gas mixtures which are capable of detonating and could endanger the integrity or stability of the containment 2. The flow cross section in the intermediate space 44 between the intermediate wall 16 and the inside wall of the containment 2 is narrowed by screening elements 114, which results in a particularly advantageous flow guidance. In the exemplary embodiment, the screening elements 114 form altogether a type of ring stop which is installed in the intermediate wall 16, approximately at a third of the height of the plant space 18. The remaining free convection cross section or inflow cross section 115 is about three times as large as that of the overflow openings 22a, 22c which are located underneath the screening elements 114 just as the inflow cross section. Moreover, a plurality of catalytic recombiners 46 are used to reduce the released hydrogen and are disposed preferably on the inside wall of the plant space 18, in particular around the steam generators 12, such that they are impacted by the flow of the rising atmosphere at comparatively high entry speed when the overflow ceiling 36 is opened. The reduction in hydrogen is particularly effective in that case. Due to the heat of reaction, the overflow speeds in the high, ceiling-side overflow cross sections are increased to up to 2 m/s or more. Finally, at least some of the recombiners 46 are disposed near the temperature-sensitive closure apparatuses 26 such that the heat of reaction thereby produced with the occurrence of H2 results in the trigger temperature being exceeded, i.e. in the opening of the respective overflow opening, even if the room temperature is otherwise rather relatively low. Due to this multiple use of the recombiners 46 used in the nuclear engineering plant 4 it is possible therefore to implement in a simple and cost-effective manner an additional trigger mechanism for the closure apparatuses 26 which is dependent on the hydrogen concentration. FIG. 2 shows by way of example the closure apparatus 26 which combines the two functions of pressure-dependent opening and temperature-dependent opening. In the embodiment shown here, the closure apparatus 26 is particularly suitable for horizontal installation in the ceiling 36 or the ceiling wall of the plant space 18 of the nuclear engineering plant 4; however, modifications with inclined or vertical installation positions, e.g. in a side wall, are also conceivable. The closure apparatus 26 contains the closure element 24 which is provided for covering an overflow opening 22 and has a bursting film 50 which is clamped in a first frame element 48. The first frame element 48 has, in plan view from the direction denoted by directional arrow 52, a rectangular basic area. The frame element 48 itself is made of hollow beams which have a square cross section or with the use of other, e.g. open, steel profiles (L, U profile combinations). The first frame element 48 supporting the bursting film 50 is fixed using hinges 56 such that it can pivot to a second frame element 58 such that the complete closure element 24 can be opened in the case of need in the manner of a wing or a door or a pull-open window with respect to the second frame element 58 which is fixedly connected to a sealing and holding construction 60 and serves at the same time as a bearing and stop for the first frame element 48. The sealing and holding construction 60 for its part is fixedly anchored in a non-illustrated concrete ceiling or intermediate wall of the nuclear engineering plant. The two frame elements 48, 58 are arranged underneath the holding construction 60 so that the closure element 24 can freely flap downward to open in a trigger event. During normal operation of the nuclear engineering plant 4, the closure element 24 is kept closed in a gas-tight manner, that is to say it is in the uppermost of the three positions which are illustrated in FIG. 2 and indicate the movement profile during opening. The two frame elements 48, 58 then congruently overlap each other, wherein a high gas-tightness is achieved by way of providing sufficient contact pressure and, if appropriate, by way of a sealing device (not illustrated further here). This closed state is maintained during normal operation with the aid of a locking apparatus 64 containing fusible solder 62, which locking apparatus 64 is arranged on the two frame elements 48, 58 opposite that side of the frame elements which has the hinges. The rod-shaped or belt-shaped fusible solder 62 is, as can be easily seen in the detail illustration at the bottom right in FIG. 2, fixedly connected at its two ends in each case with the aid of a fastening screw 66 to the first frame element 48 which is mounted such that it can move on the one hand, and to the immovable second frame element 58 or the positionally fixed sealing and holding construction 60 on the other hand and in this manner prevents in the normal case the closure element 24 from opening. In addition, the contact pressure prevailing between the two frame elements 48, 58 can be adjusted by a tensioning apparatus 70 which can be adjusted using an adjustment screw 68, and therefore the tightness of the arrangement can be adjusted. The fusible solder device can expediently contain an arrangement of a plurality of parallel fusible solders which (in the manner of a comb) are mounted with a spacing of a few millimeters and thus enable the application of larger closure forces during normal operation of the plant. Such a design is furthermore particularly suitable to accommodate catalytic elements in the gap region between the individual fusible solder strips. The melting point of the fusible solder 62 is selected such that the closure element 24 is released as soon as the room temperature at the site of the fusible solder 62 exceeds a specifically predetermined trigger value which is matched to the overflow opening. Therefore the fusible solder 62 is separated in its middle region between the two fixedly clamped or screwed-in ends by the starting melting process, whereupon the closure element 24 flaps downward to open as a result of its inherent weight and due to its “suspended” installation and releases the overflow opening 22 for through-flow purposes. After it has been triggered, the original state can easily be restored again by inserting a new fusible solder 62. Additionally, the exchange of the fusible solder device 62 can be used to easily carry out a repeating functional examination of the elements, and the trigger temperature can be changed and, for example, be matched to altered operational conditions of the nuclear engineering plant 4 or to altered safety regulations etc. Alternatively to the temperature-dependent unlocking and opening of the closure element 24, the overflow cross section can also be opened by bursting of the bursting film 50 when the bursting pressure is reached. FIG. 3 shows a combination of two closure apparatuses 26 which are arranged directly next to one another in the overflow ceiling 36, of which the left one corresponds to the closure apparatus 26 known from FIG. 2 in terms of its configuration and its function, but the right one is configured somewhat simpler than a purely pressure-sensitive closure apparatus 26 with a bursting film 50 clamped in at a fixed frame 72. In general it is sufficient for many areas of use and applications to provide diversified trigger and opening mechanisms for only some of the closure apparatuses 26, as a result of which the overall complexity for conception, manufacture and maintenance of the nuclear engineering plant 4 can be kept comparatively low. FIG. 4 illustrates an alternative embodiment of the closure apparatus 26 which is particularly suitable for use in the region of lower incident and accident temperatures, e.g. for vertical installation in a side wall 74 or in a section of the intermediate wall 16. Similarly to the closure apparatus 26 known from FIG. 3, this variant contains the bursting film 50 fixed to the frame element 48. The frame element 48 is mounted by a number of hinges 56 (which are not shown here) at its bottom longitudinal side 76 such that it can rotate on a frame-type supporting construction 78 which is inserted into the steel concrete wall such that it can flap downward to open into the opening position shown here largely due to its inherent weight when the locking is released. The process of opening is assisted, in particular in its initial phase, by the two telescope spring elements 80 which are arranged laterally and are fastened in each case at the ends of the movable frame element 48 and on the wall 74 in an articulated manner. In the normal case, the frame element 48 is kept closed by a locking element 83 which is mounted on that side of the supporting and holding construction 78 which lies opposite the hinges 56 and in the process engages in a corresponding fitting part 82 of the frame element 48. In the present case, this is a pneumatic locking element 83 which unlocks when pressurized air is applied to it. The locking element 83 is additionally coupled via a pipeline 84, which is during normal operation of the nuclear engineering plant 4 without pressure, to a temperature-dependent trigger apparatus 86 which, in the trigger event, makes available the operating pressure which is necessary to unlock the locking element 83. The trigger apparatus 86 can be arranged relative to the locking element 83 expediently in a region of relatively high incident temperatures and can be conceived, depending on the length of the pipeline 84, also as a remote-controlled apparatus. The trigger apparatus 86, which is illustrated in FIG. 5 before and in FIG. 4 after the trigger process, contains a pressurized-gas-filled pressurized-gas vessel 88 which can be closed on the outlet side by a stop valve 90. In the normal case, i.e. before the trigger condition is reached, a spring-loaded valve tappet 92 of the stop valve 90 is held in the closed state by a glass fuse 94. The fuse 94 can be in the form of a glass jacket, for example, and additionally contains a glass jacket 100 which is arranged outside the valve housing 96 and is fixed in position by a stop 98, to which glass jacket 100 an actuation rod 102 which presses against the valve tappet 92 and in the process fixes it in the closure position is attached. The glass jacket 100 is made of a glass material which, when it reaches a predetermined trigger temperature, typically 65° C. to 90° C., melts or breaks such that the previously blocked valve tappet 92 is made accessible. As a consequence of the positive pressure in the pressurized-gas vessel 88, and also assisted by the spring force of the compression spring 104, the valve tappet 92 opens so that the gas located in the pressurized-gas vessel 88 can escape there from and can flow first into an intermediate space 108 surrounded by a housing 106 and then into the pipeline 84 which is connected to the pneumatic locking element 83, as a result of which the locking element 83 is unlocked and the closure element 24 arranged in the wall 74 opens. In the present embodiment, the glass jacket 100 is located outside the intermediate space 108, wherein the actuation rod 102 is guided through a corresponding cutout 110 in the housing wall. The actuation rod 102 can be displaced in the cutout 110 in its longitudinal direction. The gap between the actuation rod 102 and the housing 106 is kept small and is sealed off with the aid of a sealing ring or the like (not illustrated further here), which is fixed on the edge of the cutout 110, such that no leak-related loss of pressure occurs here. The previously mentioned function of remote triggering can also be achieved by hydraulic devices or cable pull systems. FIG. 6 finally shows again the closure apparatus 26 known from FIG. 4, but here after a pressure-related trigger event which has led to tearing or bursting of the bursting film 50 with the frame element 48 kept closed. The arrows 112 indicate the starting flow.
	claims	1. A support cage for receiving a cylindrical-shaped expansion tank having a selected radius and a selected length comprising: a lower support ring having an outside radius smaller than said selected radius of said cylindrical tank; a plurality of radial members having a first end secured to said lower support ring, said radial member comprising a first portion extending radially outward from the circumference of said lower support ring for a distance greater than said radius of said cylindrical tank and a second portion extending perpendicular to said lower support ring for a distance less than said length of said cylindrical tank to a top end; an upper support member secured to said top end of said plurality of radial member such that said lower support ring, said plurality of radial members, and said upper support member form said support cage; and a pair of attaching members, each one of said pair joined to said upper support member and each one defining an attaching portion lying in the same plane for rigidly attaching said support cage to a support member. 2. The support cage of claim 1 further comprising a bottom attaching member extending downward and radially from said lower support ring to said same plane for producing a bottom attachment to a support member. claim 1 3. The support cage of claim 1 wherein said support cage is formed from the group of materials consisting of plastic, stainless steel and aluminum. claim 1 4. The support cage of claim 1 wherein said lower support ring and said upper support member are secured to said radial member by welding. claim 1 5. The support cage of claim 1 wherein said upper support member and said pair of connecting members are formed as a single member. claim 1 6. The support cage of claim 1 further comprising a threaded member attached to said upper support member for receiving a set screw for tightening against said tank. claim 1 7. The support cage of claim 1 wherein said upper support member is a band. claim 1 8. The support cage of claim 7 wherein said band comprises a first flexible portion member suitable for coupling with a second flexible portion, said first and second flexible portion located opposite said attaching member to allow easy placement of said tank. claim 7 9. The support cage of claim 1 wherein said lower support ring allows access to a bottom connection on said expansion tank. claim 1
058964312	claims	1. A condensing system for a vacuum breaker of a nuclear reactor, said system configured to couple to the vacuum breaker, said system comprising: a steam inlet pipe, said steam inlet pipe comprising a loop seal between a first end and a second end; and a condenser, said condenser positioned proximate said steam inlet pipe and configured to substantially condense steam flowing through said steam inlet pipe. a wetwell; a drywell; a vacuum breaker coupling said wetwell to said drywell; and a condensing system coupled to said vacuum breaker, said condensing system comprising: coupling the steam inlet pipe to the vacuum breaker; positioning the condenser proximate the steam inlet pipe; and coupling the condenser to the pool of water. 2. A condensing system in accordance with claim 1 wherein the nuclear reactor includes a wetwell, and wherein said condenser is positioned in the wetwell. 3. A condensing system in accordance with claim 1 wherein the nuclear reactor includes a drywell, and wherein said condenser is positioned in the drywell. 4. A condensing system in accordance with claim 1 wherein the reactor includes a pool of water, and wherein said condenser is coupled to the pool of water. 5. A nuclear reactor comprising: 6. A nuclear reactor in accordance with claim 5 wherein said condenser is positioned in said wetwell. 7. A nuclear reactor in accordance with claim 5 wherein said condenser is positioned in said drywell. 8. A nuclear reactor in accordance with claim 5 further comprising a pool of water, wherein said condenser is coupled to said pool of water. 9. A nuclear reactor in accordance with claim 8 wherein said pool of water comprises a Gravity Driven Cooling System pool. 10. A nuclear reactor in accordance with claim 8 wherein said pool of water comprises a passive cooling containment system pool. 11. A nuclear reactor in accordance with claim 5 wherein said steam inlet pipe comprises a first end, a second end, and a loop seal between said first end and said second end, said second end coupled to said vacuum breaker, said first end positioned in said drywell. 12. A nuclear reactor in accordance with claim 5 further comprising a wall separating said wetwell from said drywell, said wall having a spill-over-hole therein, said spill-over-hole spaced from a floor of said drywell, said steam inlet pipe first end positioned above said spill-over-hole. 13. A method for preventing steam from flowing from a drywell to a wetwell in a nuclear reactor utilizing a condensing system, the nuclear reactor including a vacuum breaker and a pool of water, the vacuum breaker coupling the drywell to the wetwell, the condensing system including a steam inlet pipe and a condenser, said method comprising the steps of: 14. A method in accordance with claim 13 wherein the steam inlet pipe includes a first end, a second end, and a loop seal between the first end and the second end, and wherein coupling the steam inlet pipe to the vacuum breaker comprises the step of coupling the second end of the steam inlet pipe to the vacuum breaker so that the first end is positioned adjacent the drywell. 15. A method in accordance with claim 13 further comprising the step of positioning the condenser in the drywell. 16. A method in accordance with claim 13 further comprising the step of positioning the condenser in the wetwell.
053234316	description	DESCRIPTION OF THE PREFERRED EMBODIMENTS In the following description, like reference characters designate like or corresponding parts throughout the several views of the drawings. Also in the following description, it is to be understood that such terms as "forward", "rearward", "left", "right", "upwardly", "downwardly", and the like are words of convenience and are not to be construed as limiting terms. IN GENERAL Referring to the drawings wherein like reference numerals refer to like elements, FIG. 1 depicts a portion of a nuclear plant facility, such as a pressurized water nuclear reactor 1 containing a reactor vessel, generally referred to as 10, for generating heat by nuclear reactions. Typically, a primary coolant such as borated water (not shown) inside the reactor vessel 10 functions to control the reaction and convey heat away from the reactor vessel 10. The primary coolant which may become radioactive in the reactor vessel 10 flows through a portion of a closed loop, generally referred to as a primary loop 20. Referring to such primary loop 20, the primary coolant flows out of the reactor vessel 10 via a pipe 25, referred to in the art as the hot leg, and transfers its heat to a water-filled secondary system, generally referred to as 30. Steam is created in a steam generator 40 from water in the secondary system 30 and is conveyed to a turbine 50-generator (not shown) set. In the primary loop 20, the primary coolant exits the steam generator 40 via a pipe 60, referred to in the art as the cold leg, and returns to the reactor vessel 10 to repeat the above described cycle. With respect to the secondary system 30, the steam of the secondary system 30 is isolated from the primary coolant and exits the steam generator 40 via a steam line 70 and is conveyed to an energy utilization device such as the steam turbine 50 which in conjunction with an electrical generator (not shown) produces electricity. Once entering the turbine 50, the steam turns turbine blades (not shown) and then exits the turbine 50 via a pipe 75 and flows into a condenser 80. Within the condenser 80, the steam is condensed back to water and is conveyed through a pipe 90 to a pump 100 which pumps the water through a pipe 110 back to the steam generator 40. Now referring to FIG. 2, the reactor vessel 10, which is typically made of carbon steel, is shown having an interior wherein components such as fuel assemblies 120 for generating heat by nuclear reactions are disposed. The reactor vessel includes a semicircular dome 130 mounted atop a vessel body 140. The dome 130 has an exterior surface 150 and an interior surface 160 spaced from each other to define a wall 170 extending between the two. The dome includes an upper portion 180 and a lower portion 190. A flange 200 having a plurality of holes 205a (one shown) is disposed circumferentially around the dome lower portion 190 and extends radially from the lower portion 190. Two lifting lugs 210 are attached to the dome upper portion 180 and are configured to receive a crane hook (not shown) for moving the dome 130. The vessel body 140 has a generally cylindrical side wall 220 having an upper portion 230 and a lower portion 240. The side wall 220 has, in this particular embodiment, two outwardly extending nozzles 245a and 245b disposed at opposite sides of the side wall 220. One nozzle 245a conveys the coolant away from the reactor vessel 10, and the other nozzle 245b allows the coolant to re-enter the reactor vessel 10. A semicircular domed bottom 250 is integrally attached to the lower portion 240 of the side wall 220. The body 140, likewise, includes an exterior surface 270 and an interior surface 280 spaced from each other to define a wall 290 extending between the two. A circumferentially extending flange 260 is attached to the side wall 220 upper portion 230 and extends radially therefrom. The flange 260 has a plurality of holes 205b (one shown) disposed circumferentially around the flange 260. When the dome 130 is placed on the vessel body 140, the body flange holes 205b and the dome flange holes 205a are positioned in registry with each other forming a mated flange. Each dome flange hole 205a is, therefore, aligned with a corresponding body flange hole 205b forming a pair of matched holes 205a and 205b. A plurality of studs 300 having two ends 305a and 305b are positioned in a pair of matched holes 205a and 205b. One end 305b of the stud is positioned in the matched holes 205a and 205b, and the other end 305a, which receives the nut 320, extends upwardly from the flanges 200 and 260. The stud 300 includes a cylindrical threaded stud shaft. An annular washer 310 is disposed on the stud 300 and abuts an outer surface 340 of the dome flange 200. A nut 320 is threaded onto one end 305a of the stud 300 for securing the dome 130 to the body 140. Referring to FIG. 3, a detailed view of the stud 300 positioned in a pair of matched holes 205a and 205b is illustrated. The head flange 200 includes a top surface 340 and a bottom surface 341 spaced from each other to define a wall thickness 342 extending between the two. The body flange 260 includes a top surface 343 abutting the bottom surface 341 of the flange 200. The annular shaped washer 310 includes a top surface 350 and a bottom surface 360 spaced apart from each other to define a wall thickness 370 between the two, and the bottom surface 360 is disposed abuttingly atop the top surface 340 of the head flange 200. As best seen in FIG. 3A, the wall thickness 370 of the washer 310 is bored to define an opening having an arcuate shaped inner peripheral surface 380 disposed adjacent the stud 300 upon assembly and an arcuate shaped outer peripheral surface 390. Referring back to FIG. 3, the nut 320 also includes a top surface 400 and a bottom surface 410 spaced apart from each other to define a wall thickness 420 between the two. Referring to FIG. 3B, the wall thickness 420 of the nut 320, likewise, is bored to define an opening having an arcuate shaped inner peripheral surface 430 disposed adjacent the stud 300 upon assembly and an arcuate shaped outer peripheral surface 440. The inner peripheral surface 430 is threaded to mate with the threaded shaft of the stud 300. Referring back to FIG. 3, the nut bottom surface 410 abuts the washer top surface 350. The stud 300 includes a cylindrical threaded shaft, and a top portion 460 includes a L-shaped gripper 470 allowing the stud 300 to be readily attached to a mechanical threading lifting device. Referring to FIG. 4, a device 480 for removing and installing the reactor vessel studs 300 is shown and is generally referred to in the art as an Automated Stud Handling and Transport System as is well known in the art. To remove each of the studs 300, the device 480 is positioned on the dome flange 200 to engage a stud 300 and then operated to unscrew the stud 300. After all the studs 300 are removed, a crane (not shown) is attached to each lifting lug 210 for removing the dome 130 providing access into the interior of the reactor vessel 10. To install the dome 130, the process is reversed. RETAINER OF THE PRESENT INVENTION Referring to FIG. 5, there is illustrated a retaining device 620 of the present invention operable to mate the washer to the stud to maintain the positional relationship of the stud and washer. The retaining device 620 includes a resilient, generally "U" shaped strap 630. The strap 630 has two arcuate shaped sides 650 connected respectively at two bend portions 660 to a substantially straight middle portion 670. The strap 630 has two ends 640a and 640b each having a pair of holes (not shown). A wedge 680, typically made of nylon, is attached to each of the ends 640a and 640b of the strap 630. Each wedge 680 includes a thickened, rectangular shaped lower body 690 having two threaded holes 700 (one pair shown) therethrough. The lower body 690 is attached to each strap side 650 preferably by a pair of bolts 710 each having a threaded shaft (not shown) and a bolt head 720. The bolt shaft is disposed through each strap hole and is threaded into the wedge hole 700 until the bolt head 720 abuts the strap side 650. The lower body 690 is thickened to increase the wedge 680 strength in this area. The wedge 680 includes an upper body 730 which extends outwardly from the lower body 690. The upper body 730 is tapered and terminates at a generally pointed lip portion 740. The wedge 680 includes an outer surface 750 and an inner surface 760 each having a generally arcuate shape and spaced apart from each other defining a wall thickness 770 between the two. The inner surface 760 is arcuate shaped to conform to the shape of the stud 300 (see FIG. 3). Referring to FIGS. 6 and 7, to install a retaining device 620 on a stud 300, the ends 640a and 640b of the strap 630 are slide over the stud 300 so that the wedges 680 are spaced generally one hundred and eighty degrees from each other on opposite sides of the stud 300. The strap 630 is made from a resilient material to allow it to expand to receive the stud and then return to its original shape to fit generally concentrically around the stud 300. The lip portion 740 of each wedge 680 is slidably inserted between the stud 300 and the washer inner peripheral surface 380, and the washer 310 and retaining device 620 are urged towards each other to allow the wedges 680 to fit firmly between the washer 310 and the stud 300. In this manner, the washer 310 is retained on the stud 300 until the retaining device 620 is removed. To remove the retaining device 620, the above described process is reversed. Referring to FIGS. 8 and 9, an alternative embodiment of the retaining device of the present invention is illustrated. A pair of semi-circular shaped clamps 780a and 780b having two ends 790a and 790b are respectively disposed abutting the top and bottom surfaces 350, 360 of the washer 310. Each clamp 780a and 780b is disposed circumferentially around the stud 300 so that the ends 790a and 790b generally face each other. The clamps 780a and 780b, likewise, are made of a resilient material allowing it to pass around the stud 300 without losing its original shape. Each clamp 780a and 780b includes a top surface 800, and a bottom surface 810 spaced apart from each other defining a wall 820 between the two. The wall 820 includes an arcuate shaped inner peripheral surface 830 adjacent the stud 300, and an arcuate shaped outer peripheral surface 840 outwardly spaced from the inner peripheral surface. One clamp 780a bottom surface 810 abuts the washer top surface 350, and the other clamp 780b top surface 800 abuts the washer bottom surface 360. Each clamp 780a and 780b frictionally engages the shaft of the stud 300 and, therefore, is not movable on the stud 300. The pair of clamps 780 in combination with each other keep the washer 310 from sliding. Therefore, what is provided is a device for removably securing a reactor vessel washer to a vessel stud allowing the stud and washer to be maintained in a mated pair during the removal and installation of the stud. Although the invention is fully described herein, it is not intended that the invention as illustrated and described be limited to the details shown, because various modifications may be obtained with respect to the invention without departing from the spirit of the invention or the scope of equivalents thereof. For example, two wedges, without a connecting strap, may be disposed between the stud and washer for retaining the washer and stud as a mated pair.
050013513	claims	1. An object holder for positioning an object in a radiation beam comprising: a base; means for securing a plate member to the base, said plate member having a planar face surface; an X-guide movably secured to the base and translatable in X-directions relative to said plate member, said guide including first slide means for sliding in abutting relation on said face surface in said X-directions; and a Y-guide secured to the X-guide and translatable in Y-directions, said Y-guide including second slide means for sliding in abutting relation on said face surface in said Y-directions; one of said guides including means to receive said object to be positioned in said radiation beam. a base; means for securing a plate member to the base, said plate member having a planar face surface; an X-guide movably secured to the base and translatable in X-directions relative to said plate member; an X-transporter secured to said guide; first slide means secured to the X-transporter for sliding in abutting relation on said face surface in said X-directions; a Y-guide secured to the X-transporter and translatable in Y-directions; a Y-transporter secured to said Y-guide; second slide means secured to the Y-transporter for sliding in abutting relation on said face surface in said Y-directions; and means for rotatably securing an object support table to said Y-guide. 2. The object holder of claim 1 wherein said X-guide includes an X-transporter and means for securing the first slide means to the X-transporter and the Y-guide includes a Y-transporter and means for securing the second slide means to the Y-transporter. 3. An object holder as claimed in claim 2, including permanent magnets which are secured to the X-transporter and to the Y-transporter and which magnets bear on the supporting face of the supporting plate via intermediate sliding pieces. 4. An object holder as claimed in claim 1, further including a rotation mechanism for rotating the supporting plate about an axis extending transversely of the supporting face thereof. 5. An object holder as claimed in claim 4, characterized in that rotation has an extent of at least 360.degree.. 6. An object holder as claimed in claim 1, further including a tilt mechanism for tilting the supporting plate about an axis extending parallel to the supporting face thereof. 7. An object holder as claimed in claim 6, characterized in that said tilt mechanism includes a plurality of bushings for providing a tilting movement, said bushings having cut-outs which act as passages for a radiation beam. 8. The object holder of claim 1 wherein said first and second slide means each include magnet means for magnetically forcing the X and Y slide means in said abutting relations. 9. The object holder of claim 8 wherein said magnet means are permanent magnets, said sliding means each including sliding members intermediate said magnets and said face surface. 10. The object holder of claim 1 wherein said Y-guide is to receive said means adapted to receive said object. 11. An object holder for positioning an object in a radiation beam comprising: 12. The object holder of claim 11 wherein said first and second slide means each include magnet means and ductile interface means, said magnet means being secured to said transporters and said ductile means being secured to said magnet means intermediate said magnet means and said face surface, said ductile means for sliding on said face surface.
056617666	abstract	An apparatus for measuring fuel assembly bow and twist comprising a fixturing apparatus, a reference device and an ultrasonic measuring device.
	summary
	summary
	claims	1. A process for synthesizing a compound C1 selected from the group consisting of mixed peroxides and hydroxo-peroxides of an actinyl and of at least one cation X1, wherein:the actinyl has formula AnO2q+ where An is an actinide selected from the group consisting of uranium and neptunium, and q equals 1 or 2;the cation X1 is a doubly, triply or quadruply charged metal cation, wherein the metal differs from An;which process comprises a reaction in a solvent of a salt of the cation X1 with a compound C2 selected from the group consisting of mixed peroxides and hydroxo-peroxides of the actinyl and of at least one singly charged cation X2, whereby compound C2 is converted to compound C1 by a replacement of the cation X2 by the cation X1. 2. The process of claim 1, wherein the cation X1 is a cation of an alkaline-earth metal, a cation of a post-transition metal, a cation of a transition metal, a cation of a lanthanide or a cation of an actinide. 3. The process of claim 2, wherein the cation X2 is a cation of an alkaline metal, a cation of a transition metal or a polyatomic cation. 4. The process according to claim 1, wherein the reaction of the cation X1 with compound C2 comprises adding a solution of the salt of the cation X1 to compound C2 to obtain a reaction medium, and leaving the reaction medium to stand for sufficient time to obtain the replacement of the cation X2 by the cation X1. 5. The process of claim 4, wherein the solution of the salt of the cation X1 is an aqueous solution. 6. The process of claim 1, further comprising a synthesis of compound C2. 7. The process of claim 6, wherein the synthesis of compound C2 comprises a reaction of a first aqueous solution comprising a salt of the actinide An with an alkaline second aqueous solution comprising a salt or hydroxide of the cation X2 and hydrogen peroxide. 8. The process of claim 7, wherein the synthesis of compound C2 comprises adding the first solution to the second solution under agitation to obtain a reaction medium and leaving the reaction medium to stand for sufficient time to obtain the formation of compound C2. 9. The process of claim 1, wherein compound C1 has general formula (I):(X1m+)r1[(AnO2q+)n(O22−)p−x(OH−)2x](2p−qn)− (I)where:m equals 2, 3 or 4;n is an even integer of 2 or higher;x is an integer of 0 or higher;p is an integer higher than x; andn, p and r1 are such that 1.5≤p/n≤2 and 0<r1=(2p−qn)/m. 10. The process of claim 9, wherein compound C2 has general formula (II):(X2+)r2[(AnO2q+)n(O22−)p−x(OH−)2x](2p−qn)− (II)where:0<r2=2p−qn. 11. The process of claim 9, wherein n is an even integer ranging from 2 to 60. 12. The process of claim 1, wherein the cation X1 is a cation of an actinide or lanthanide. 13. The process of claim 1, wherein the cation X2 is an ammonium cation. 14. A process for synthesizing a mixed oxide of an actinide An selected from the group consisting of uranium and neptunium, and of at least one metal able to form a doubly, triply or quadruply charged metal cation X1, the metal differing from An, which process comprises:synthesizing a mixed peroxide or hydroxo-peroxide of an actinyl of formula AnO2q+ where q equals 1 or 2, and of at least the cation X1, the synthesis comprising a reaction in a solvent of a salt of the cation X1 with a compound C2 selected from the group consisting of mixed peroxides and hydroxo-peroxides of the actinyl and of at least one singly charged cation X2, whereby compound C2 is converted to compound C1 by a replacement of the cation X2 by the cation X1; andcalcining the mixed peroxide or hydroxo-peroxide thus synthesized. 15. A mixed peroxide or hydroxo-peroxide of an actinyl and of at least one cation X1, wherein:the actinyl has formula AnO2q+ where An is an actinide selected from the group consisting of uranium and neptunium, and q equals 1 or 2;the cation X1 is a doubly, triply or quadruply charged metal cation, wherein the metal differs from An;the peroxide or hydroxo-peroxide has following general formula (I):(X1m+)r1[(AnO2q+)n(O22−)p−x(OH−)2x](2p−qn)− (I)where:m equals 2, 3 or 4;n is an even integer of 2 or higher;x is an integer of 0 or higher;p is an integer higher than x; andn, p and r1 are such that 1.5≤p/n≤2 and 0<r1=(2p−qn)/m. 16. The peroxide or hydroxo-peroxide of claim 15, wherein n is an even integer ranging from 2 to 60. 17. The peroxide or hydroxo-peroxide of claim 15, wherein the cation X1 is a cation of an actinide or lanthanide. 18. The process of claim 10, wherein n is an even integer ranging from 2 to 60.
	claims	1. Neutron adsorbent material, characterized in that said material is a composite material comprising hafnium diboride and hafnium dioxide, in which the hafnium diboride represents at least 80% by volume of the material. 2. Material according to claim 1 , in which the hafnium diboride represents about 90% by volume of the material. claim 1 3. Neutron adsorbent material, characterized in that said material is a composite material consisting of at least 80% by volume hafnium diboride and up to 20% by volume hafnium dioxide. 4. Material according to claim 1 , in which the hafnium dioxide represents up to 20% by volume of the material. claim 1 5. Neutron adsorbent material, characterized in that said material is a composite material comprising hafnium diboride and hafnium dioxide, in which the hafnium dioxide represents about 10% by volume of the material. 6. Material according to claim 1 , in which the hafnium dioxide represents about 10% by volume of the material. claim 1 7. Material according to claim 2 , in which the hafnium dioxide represents about 10% by volume of the material. claim 2 8. Method of manufacturing a neutron absorbent material, said neutron absorbent material being a composite material hafnium diboride, said method comprising steps that consist of, in this order: adding hafnium dioxide powder to hafnium diboride powder, mixing the hafnium diboride powder and the hafnium dioxide powder in a way that produces a homogeneous mixture, wherein when up to 20% by volume of hafnium dioxide is added, the homogeneous mixture of the hafnium diboride and hafnium dioxide represent 100% by volume and sintering the homogeneous mixture in a way that produces the composite material. 9. Method according to claim 8 , in which, about 10% by volume of hafnium dioxide is added, the homogeneous mixture of the hafnium diboride and hafnium dioxide representing 100% by volume. claim 8 10. Method according to claim 8 , in which the hafnium diboride powder has a particle size ranging up to about 50 xcexcm. claim 8 11. Method according to claim 9 , in which the hafnium diboride powder has a particle size ranging up to about 50 xcexcm. claim 9 12. Method according to claim 10 , in which the hafnium diboride powder has a particle size ranging up to about 20 xcexcm. claim 10 13. Method according to claim 11 , in which the hafnium diboride powder has a particle size ranging up to about 20 xcexcm. claim 11 14. Method according to claim 10 , in which the hafnium diboride powder has a particle size ranging up to about 10 xcexcm. claim 10 15. Method according to claim 11 , in which the hafnium diboride powder has a particle size ranging up to about 10 xcexcm. claim 11 16. Method according to claim 8 , in which the mixture of the hafnium diboride powder and hafnium dioxide powder is produced by the application of ultrasound to a slip comprising said powders dispersed in a dispersion liquid. claim 8 17. Method according to claim 8 , in which the homogeneous mixture is sintered under vacuum. claim 8 18. Method according to claim 8 , in which the homogeneous mixture is sintered in a graphite mold lined with a sheet of graphite. claim 8 19. Method according to claim 17 , in which the homogeneous mixture is sintered in a graphite mold lined with a sheet of graphite. claim 17 20. Method according to claim 8 , in which the mixture is sintered at a temperature of from about 1600 to 2100xc2x0 C., under a pressure of from 15 to 100 MPa for a period of about 15 to 90 minutes. claim 8 21. Method according to claim 8 , in which the mixture is sintered at a temperature of about 1900xc2x0 C., under a pressure of about 83 MPa for a period of about 1 hour. claim 8 22. Method according to claim 8 , comprising additionally a step of machining the composite material over a thickness of about 500 to 1000 xcexcm. claim 8 23. Material according to claim 7 , in which the hafhium dioxide is in the form of particles having a diameter ranging up to about 50 xcexcm. claim 7 24. Material according to claim 7 , in which the hafnium dioxide is in the form of particles having a diameter ranging up to about 20 xcexcm. claim 7 25. Material according to claim 7 , in which the hafnium dioxide is in the form of particles having a diameter ranging up to about 10 xcexcm. claim 7 26. Method according to claim 17 , in which the mixture is sintered at a temperature of from about 1600 to 2100xc2x0 C., under a pressure of from 15 to 100 MPa for a period of about 15 to 90 minutes. claim 17 27. Method according to claim 17 , in which the mixture is sintered at a temperature of about 1900xc2x0 C., under a pressure of about 83 MPa for a period of about 1 hour. claim 17 28. Material according to claim 3 , in which the hafnium dioxide represents about 10% by volume of the material. claim 3 29. Material according to claim 1 , in which the hafnium dioxide is in the form of particles having a diameter ranging up to about 50 xcexcm. claim 1 30. Material according to claim 29 , in which the hafnium dioxide is in the form of particles having a diameter ranging up to about 20 xcexcm. claim 29 31. Material according to claim 29 , in which the hafnium dioxide is in the form of particles having a diameter ranging up to about 10 xcexcm. claim 29 32. Material according to claim 3 , in which the hafnium dioxide is in the form of particles having a diameter ranging up to about 50 xcexcm. claim 3 33. Material according to claim 32 , in which the hafnium dioxide is in the form of particles having a diameter ranging up to about 20 xcexcm. claim 32 34. Material according to claim 32 , in which the hafnium dioxide is in the form of particles having a diameter ranging up to about 10 xcexcm. claim 32 35. Material according to claim 1 , having a density of about 10000 to11000 kg/m 3 . claim 1 36. Material according to claim 35 , having a density of about 10550 to10630 kg/m 3 . claim 35
059011920	summary	FIELD OF THE INVENTION This invention relates generally to nuclear reactors and, more particularly, to an apparatus and methods for securing piping within reactor pressure vessels of such reactors. BACKGROUND OF THE INVENTION A reactor pressure vessel (RPV) of a boiling water reactor (BWR) typically has a generally cylindrical shape and is closed at both ends, e.g., by a bottom head and a removable top head. A core shroud, or shroud, typically surrounds the core and is supported by a shroud support structure. Boiling water reactors have numerous piping systems, and such piping systems are utilized, for example, to transport water throughout the RPV. For example, core spray piping is used to deliver water from outside the RPV to core spray spargers inside the RPV. The core spray piping and spargers deliver water flow to the reactor core. Stress corrosion cracking (SCC) is a known phenomenon occurring in reactor components, such as structural members, piping, fasteners, and welds, exposed to high temperature water. The reactor components are subject to a variety of stresses associated with, for example, differences in thermal expansion, the operating pressure needed for the containment of the reactor cooling water, and other sources such as residual stresses from welding, cold working and other inhomogeneous metal treatments. In addition, water chemistry, welding, heat treatment and radiation can increase the susceptibility of metal in a component to SCC. Reactor internal piping, such as T-boxes and core spray line risers, occasionally require replacement as a result of SCC. Replacing the core spray piping often requires removing and replacing the core spray line riser. The core spray line attachment to the shroud, however, typically is installed during original reactor construction and is difficult to access. In addition, replacing the core spray line riser, particularly the attachment to the shroud, is complicated by the limited available working space. The core spray line riser includes a lower elbow secured to a core spray sparger T-box. There are a number of welds between the lower elbow and T-box, and the integrity of these welds must be maintained to ensure proper operation of the core spray piping. As explained above, it is difficult to access these welds. It would be desirable to provide an apparatus which facilitates providing additional support for the welds between the core spray sparger T-box and the lower weld of the lower elbow of the core spray line riser. It would also be desirable to provide such apparatus which is easy to assemble and install in a limited working space. SUMMARY OF THE INVENTION These and other objects may be attained by an apparatus which, in one embodiment, includes a sleeve sized to be partially inserted within a T-box pipe, and a draw bolt which extends through the sleeve and engages a block so that the draw bolt draws the block into tight engagement with the lower elbow of the riser. More particularly, the sleeve is substantially cylindrically shaped and has a main cylindrical body and a flange at one end. An axial bore extends through the main body, and two opposing water flow openings are located in the sleeve main body. The main body is sized to be inserted into the T-box pipe so that the sleeve flange engages one end of the T-box pipe. The sleeve flange includes an opening sized to receive the draw bolt. The draw bolt is threaded and is of sufficient length to extend through the main body bore and an opening machined in the core spray line riser lower elbow. The block is substantially wedge shaped and includes a threaded opening to engage the draw bolt. To install the repair apparatus, after removing a core spray line riser, a replacement core spray line riser is coupled to the T-box, for example by welding the lower elbow to the core spray sparger T-box. The T-box end plate is then removed and the repair apparatus sleeve is inserted into the open end of the T-box so that the sleeve flange is in contact with and closes off the open end of the T-box. The sleeve is rotated to ensure that the water flow openings are substantially aligned with the core spray spargers attached to the T-box. The core spray line repair apparatus draw bolt is then inserted through the sleeve flange opening, through the main body bore, and through the opening machined in the replacement riser lower elbow. The repair apparatus block is then engaged to the draw bolt. The draw bolt is then torqued and locked in place. As the bolt is tightened into engagement with the block, the apparatus provides axial restraint. The above-described apparatus provides support for the welds between the core spray sparger T-box and the lower weld of the core spray line riser lower elbow. In addition, such apparatus is easy to assemble and can be installed in a limited work space.
042082471	claims	1. A thermal nuclear reactor including a core, neutron detecting means disposed externally of said core, and a control element selectively positionable within said core, said control element comprising a plurality of rods of a material substantially black to thermal neutrons encased in a sealed cladding, one of said rods also including a spontaneous fast neutron emitting source encapsulated in said said material substantially black to thermal neutrons, said source being of sufficient strength to emit neutrons detectable by said detection means upon shutdown of said reactor and location of said control element within said core. 2. A thermal nuclear reactor comprising: a. a reactor vessel; b. a neutron detector disposed externally of said vessel; c. a spontaneous fast neutron emitting source and control element rectilinearly movable within said vessel, said element including a neutron source which spontaneously emits fast neutrons through (.alpha.,n) reactions of sufficient strength to be detected by said detector, said source being encapsulated within a sealed enclosure, said enclosure also bounding a plenum in fluid communication with said source, and a sealed cladding encapsulating said enclosure, said cladding also bounding a neutron absorbing control material substantially black to thermal neutrons, said neutron absorbing control material encapsulating said neutron source and said enclosure; and d. a plurality of vertical coextending fuel assemblies arranged within said vessel to form a nuclear core, one of said assemblies having a guide tube for slidingly receiving said neutron source and control element. an elongated structure of neutron (n) poison material along a substantial portion of its length, said poison material being substantially black to thermal neutrons, a minor portion of said elongated structure including a spontaneous fast neutron emitting source encased in a material substantially black to thermal neutrons and substantially transparent to neutrons in other energy ranges, said source emitting neutrons of sufficient strength to be detected by said detector while said reactor is in a shutdown condition, said neutron emitting source comprising plutonium-238 and beryllium, whereby plutonium-238 spontaneously emits alpha (.alpha.) particles which react with the beryllium in an (.alpha.,n) reaction to spontaneously produce said fast neutrons. 3. The reactor of claim 2 including means for controlling the position of said neutron emitting source and control element such that said source is selectively positionable within said core at power levels below a preselected value and positionable out of said core at power levels above a preselected value, and at least a portion of said substantially black material is positionable within said core when said source is either within or out of said core. 4. A control element adapted for a thermal nuclear reactor having a core, an ex-core neutron detector, and means for reciprocatingly inserting said element into said core comprising:
	abstract	A neutron-optical component array in which the beam paths of the individual moderators are combined in a concerted manner so as to create a superimposed neutron beam with an effective mean beam direction. The superimposed neutron beam has a multi spectrum composed of the single spectrums of several moderators, whereby a larger spectral width is obtained, making various applications in different neutron energy fields possible. The multi spectrum can be further improved in terms of the intensity thereof and the beam quality by adding further neutron-optical components, particularly in the form of an energy-depending switching super reflector, and by switching between moderators.
	abstract	A canister apparatus, basket apparatus and combinations thereof for transporting and/or storing high level radioactive waste, such as spent nuclear fuel. The canister apparatus comprises a cavity for receiving the spent nuclear fuel that is surrounded by two independent gas-tight containment boundaries. The structures that form the two independent gas-tight containment boundaries are in substantially continuous surface contact with one another, thereby facilitating sufficient heat removal from the cavity. In another aspect, the invention is a basket apparatus having a plurality of disk-like grates arranged in a stacked and spaced arrangement so that the cells of the disk-like grates are aligned. In still another aspect, the invention can be a basket apparatus having a disk-like grate having a ring-like structure encompassing a gridwork of beams specially arranged to achieve a unique cell configuration.
054460751	summary	TECHNICAL FIELD OF THE INVENTION The present invention relates in general to therapeutic exercise putties and more particularly to a method and apparatus for monitoring a patient's progress in manipulative therapy using exercise putty. BACKGROUND OF THE INVENTION Borosiloxanes exhibit peculiar physical characteristics which make them suitable for therapeutic use. Borosiloxane bouncing putties are shown, for example, in U.S. Pat. No. 2,541,851, issued to Wright, and U.S. Pat. No. 3,677,997, issued to Kaiser et al. Both of these patents are fully incorporated by reference herein. These bouncing putties have the peculiar characteristic of being able to be kneaded and worked as a putty-like material, while at the same time exhibiting elastic properties under a greater degree of force. A borosiloxane bouncing putty has long been commercially available as a toy under the trademark "SILLY PUTTY". Because of their characteristics, bouncing putties have found application in physical therapy to strengthen muscular control and performance of, e.g., patients'hands. Because a lump of exercise putty is by its nature a shapeless mass, there is no easy way for a physical therapist (or the patient) to monitor the progress being made in manipulating the putty. Different amounts of kneading or manipulation will, in conventional exercise putties, create the same result--the same shapeless mass. There is no good way to measure the amount of manipulation which has been done. A need therefore exists for a method and apparatus to monitor the progress made by a patient in exercising his or her hands. SUMMARY OF THE INVENTION According to one aspect of the present invention, a method of manipulative therapy is disclosed by which a first malleable mass having a first color is supplied to a patient. A second malleable mass having a second color different from the first color is also supplied to the patient, and the patient is directed to knead or manipulate the two putties together. The putty is kneaded or manipulated until such time as the two colors are completely blended together to yield a uniform color which is a result of blending the first color with the second color. According to a further aspect of the invention, the malleable masses provided to the patient are based on borosiloxane and are pigmented to different colors. According to another aspect of the invention, the second malleable mass is significantly smaller and has a more concentrated color than the first malleable mass. Once this second malleable mass has been completely combined with a first malleable mass, further highly colored malleable masses are provided to the patient for combination with the product of the last combination. This process is repeated a number of times suitable for the patient'therapy or until such time as the combined mass has achieved a color which can no longer be altered by additional amounts of putty, such as a uniform dark brown color. It is preferred that apparatus according to the invention be provided to physical therapists and the alike in a kit including a relatively large mass of relatively colorless exercise putty, and a plurality of relatively small masses of putty having distinctive colors, and each being chromatically distinct from at least one other small mass. In this way, the combined putty mass will change color as additional amounts of small, highly colored putty of varying hues are added to it. The present invention confers a technical advantage in that the progress of the patient's manipulative therapy can be closely monitored. If a large, relatively uncolored mass of putty is poorly combined with a smaller, highly colored mass, the two masses will simply be lumped together and will be visually distinct from one another. At an intermediate stage, the masses will have been blended together such that stripes of color will appear in the combined mass. Only after a considerable amount of kneading and manipulating will the mass attain a uniform color which is the product of combining the colors in the original two constituents. This provides the physiotherapist or physician some indication of the amount of working being done by the patient, allowing the therapist or physician to monitor the patient'progress.
	summary
	claims	1. A brachytherapy device comprising: a) a body comprising a top surface, a bottom surface, and a staging area surface, wherein said top surface comprises an insert receiving well capable of holding an insert device, and wherein said staging area surface slopes away from said top surface and toward said bottom surface, and wherein said staging area surface comprises a loading area having one or more seed holder supports capable of holding a seed holder; and b) a lid attached to said body, wherein said lid is capable of opening to expose said insert receiving well and is capable of closing over said top surface to cover said insert receiving well. 2. The brachytherapy device of claim 1 , wherein said body comprises a vent that creates a line of sight through said body. claim 1 3. The brachytherapy device of claim 2 , wherein said vent extends from said top surface to said bottom surface. claim 2 4. The brachytherapy device of claim 1 , wherein said body comprises a vent that does not create a line of sight through said body. claim 1 5. The brachytherapy device of claim 4 , wherein said vent extends from said insert receiving well to said bottom surface. claim 4 6. The brachytherapy device of claim 1 , wherein said top surface is substantially parallel with a flat surface when said bottom surface rests on said flat surface. claim 1 7. The brachytherapy device of claim 1 , wherein said top surface comprises multiple insert receiving wells. claim 1 8. The brachytherapy device of claim 1 , wherein said top surface comprises a transfer device well capable of holding a transfer device. claim 1 9. The brachytherapy device of claim 1 , wherein said top surface comprises multiple transfer device wells. claim 1 10. The brachytherapy device of claim 1 , wherein a portion of said top surface is raised. claim 1 11. The brachytherapy device of claim 1 , wherein said insert receiving well is cylindrically shaped. claim 1 12. The brachytherapy device of claim 1 , wherein the bottom of said insert receiving well comprises a pin capable of aligning said insert device. claim 1 13. The brachytherapy device of claim 12 , wherein said pin is conically shaped. claim 12 14. The brachytherapy device of claim 1 , wherein the bottom of said insert receiving well comprises multiple pins capable of aligning said insert device. claim 1 15. The brachytherapy device of claim 1 , wherein said bottom surface comprises a hollow bottom. claim 1 16. The brachytherapy device of claim 1 , wherein at least a portion of said hollow bottom is positioned underneath said staging area surface. claim 1 17. The brachytherapy device of claim 1 , wherein the angle of said staging area with respect to a flat surface is less than 60 degrees when said bottom surface rests on said flat surface. claim 1 18. The brachytherapy device of claim 1 , wherein the angle is of said staging area with respect to a flat surface is between 80 and 10 degrees when said bottom surface rests on said flat surface. claim 1 19. The brachytherapy device of claim 1 , wherein the angle of said staging area with respect to a flat surface is between 60 and 30 degrees when said bottom surface rests on said flat surface. claim 1 20. The brachytherapy device of claim 1 , wherein said one or more seed holder supports are capable of restricting longitudinal movement of said seed holder. claim 1 21. The brachytherapy device of claim 1 , wherein said one or more seed holder supports are capable of restricting latitudinal movement of said seed holder. claim 1 22. The brachytherapy device of claim 1 , wherein said loading area comprises a restraining wire attached to said one or more seed holder supports, wherein said restraining wire is capable of restricting axial movement of said seed holder. claim 1 23. The brachytherapy device of claim 1 , wherein said staging area comprises a top portion and a bottom portion, wherein said bottom portion comprises a trough capable of collecting a radioactive seed that rolls down said top portion. claim 1 24. The brachytherapy device of claim 1 , wherein said lid is attached to said body via an integral hinge. claim 1 25. The brachytherapy device of claim 1 , wherein said lid defines an interior region, wherein a portion of said interior region comprises a protrusion capable of restraining movement of a seed holder within said insert device when said lid is closed. claim 1 26. The brachytherapy device of claim 25 , wherein said protrusion extends between the pick-up handles of said insert device when said lid is closed. claim 25 27. The brachytherapy device of claim 25 , wherein the clearance between said protrusion and said seed holder is less than 0.1 inches. claim 25 28. The brachytherapy device of claim 25 , wherein the clearance between said protrusion and said seed holder is less than 0.01 inches. claim 25 29. The brachytherapy device of claim 25 , wherein the clearance between said protrusion and said seed holder is less than 0.005 inches. claim 25 30. The brachytherapy device of claim 1 , wherein said brachytherapy device comprises a handle attached to said body. claim 1 31. The brachytherapy device of claim 30 , wherein said handle comprises a lock that locks said lid in the closed position. claim 30 32. The brachytherapy device of claim 31 , wherein said lock locks said lid in the closed position when said handle is in a vertical position. claim 31 33. The brachytherapy device of claim 30 , wherein movement of said handle from a vertical position toward a horizontal position moves said lid from a closed position to an opened position. claim 30 34. The brachytherapy device of claim 30 , wherein the center of gravity of said brachytherapy device is such that said lid closes from an open position when said brachytherapy device is lifted by said handle. claim 30 35. The brachytherapy device of claim 1 , wherein said body and lid are aluminum. claim 1 36. A brachytherapy device comprising: a) an insert device comprising: i) multiple seed holder pockets, wherein each of said multiple seed holder pockets is capable of holding a seed holder, and ii) a pick-up handle; b) a body comprising a top surface, a bottom surface, and a staging area surface, wherein said top surface comprises an insert receiving well capable of holding said insert device, and wherein said staging area surface slopes away from said top surface and toward said bottom surface, and wherein said staging area surface comprises a loading area having one or more seed holder supports capable of holding a seed holder; and c) a lid attached to said body, wherein said lid is capable of opening to expose said insert receiving well and is capable of closing over said top surface to cover said insert receiving well. 37. The brachytherapy device of claim 36 , wherein a connector connects at least two of said multiple seed holder pockets. claim 36 38. The brachytherapy device of claim 36 , wherein each of said multiple seed holder pockets comprises a top portion and a bottom portion, wherein said top portion comprises a cylindrical inner space, and wherein said bottom portion comprises a rectangular inner space. claim 36 39. The brachytherapy device of claim 36 , wherein said insert device comprises at least two pick-up handles. claim 36 40. The brachytherapy device of claim 39 , wherein each of said at least two pick-up handles comprises an aperture. claim 39 41. The brachytherapy device of claim 36 , wherein said insert device is plastic. claim 36 42. The brachytherapy device of claim 36 , wherein each of said multiple seed holder pockets is capable of holding a seed holder having a rectangular shaped body. claim 36 43. The brachytherapy device of claim 30 , wherein at least one of said multiple seed holder pockets comprises a vent. claim 30
044850689	summary	BACKGROUND OF THE INVENTION The present invention relates to an installation for the storage and/or transfer of dangerous products such as those presenting an irradiation hazard (radioactive waste, feed materials, sources, etc), or chemical contamination (plutonium, liquid effluents, etc) or such as explosives. More specifically, the invention relates in preferred manner an installation for the storage and/or transfer of irradiated fuel assemblies in the core of a fast neutron nuclear reactor. It is known that irradiated fuel assemblies in the core of a fast neutron nuclear reactor are located in a tightly sealed sheath containing liquid sodium as soon as they are removed from the reactor cylinder, in order that the calories which the fuel continues to dissipate are removed under favourable conditions, despite the absence of storage in the reactor vessel. Another solution consists of transforming the assemblies under water or some other liquid (e.g. sodium.) after washing on leaving the reactor cylinder. It is also known that the irradiated fuel is reprocessed with minimum delay. However, it may be necessary to provide for a provisional storage of the fuel assemblies, particularly when there is a difference between the time in which the reactor is put into operation and that in which the reprocessing plant is put into operation. Moreover, even when the reactor and reprocessing plant are both operating, the storage of the irradiated assemblies awaiting reprocessing is virtually indispensible due to the significant difference between the removal speeds from the reactor cylinder during an irradiated fuel replacement phase and the reprocessing possibilities of the reprocessing plant. As the reprocessing plant is not in principle installed on the site on the reactor, it is possible to store the irradiated fuel on the same site as the reactor or on the site of the reprocessing plant. Technical problems are simplified by the first solution which makes it possible to defer the assembly transportation operations, thus reducing costs. For this reason, the storage installation according to the invention is preferably located on the site of the reactor. However, this location is not limitative and it is to be understood that the invention also covers the case when the storage installation is located on the site of the reprocessing plant. BRIEF SUMMARY OF THE INVENTION The present invention therefore relates to an installation for the storage and/or transfer of dangerous products, wherein it comprises a storage enclosure defining at least one loading and unloading station and a plurality of storage stations, modules each of which receives the dangerous products such as fuel assemblies (exposed or in sheets) each of these modules being located in one of the stations defined in the enclosure, the number of modules being less than that of the stations, lifting means for creating a fluid cushion below at least part of the modules in order to raise the latter within the enclosure and means for moving the modules between the different stations when they are raised by the lifting means. According to a secondary feature of the invention, the loading and storage stations are aligned in rows in accordance with two different directions, the lifting means acting independently in each of the said rows, and the means for moving the modules acting simultaneously on all the modules of the same row. According to a particular embodiment of the invention, the loading and storage stations are aligned in rows in two orthogonal directions, the storage enclosure then being rectangular and the module square. Preferably, means are provided for guiding the modules during their movements between the different stations. When these stations are aligned in rows, these module guidance means can comprise guide rails located on the periphery of the storage enclosure and between certain of the rows, except at each of the ends of the rows, and lateral rollers located on each module to cooperate with the guide rails and with adjacent modules. Preferably, the means for moving the modules are constituted by jacks located outside the enclosure at least at one end of at least part of said rows. According to another secondary feature of the invention, the lifting means comprise parallel pipes arranged level with the base of the enclosure and incorporating uniformly distributed support nozzles, the pipes being supplied with a pressurized fluid such as gas or liquid. When the installation is a fuel assembly storage installation located on the site of the reactor, it is important to provide a circuit making it possible to remove the calories diffused by the fuel assemblies. This circuit can be supplied in preferred manner by a cooled fluid such as gas or liquid. According to another secondary feature of the invention, each module then comprises an upper plate, means for supporting the fuel assemblies, a lower plate whose lower face is flat so as to permit the lifting of the module by the lifting means and an intermediate ferrule surrounding the fuel assemblies and defining in the vicinity of each of the plates a passage enabling the cooled fluid to enter and leave the module.
	description	The present invention relates to the field of X-ray imaging, wherein an object under examination is illuminated by X-radiation and the X-radiation, which has penetrated the object, is detected in order to acquire a two-dimensional image of the object. In particular, the present invention relates to a portable X-ray detection device comprising a two-dimensional X-ray detector unit for detecting the X-radiation having penetrated the object under examination. The present invention further relates to an X-ray imaging system, in particular a medical X-ray imaging system, the X-ray imaging system comprising the described portable X-ray detector. Further, the present invention relates to a method for acquiring X-ray image data by means of the described portable detector. Furthermore, the present invention relates to a computer-readable medium and to a program element having instructions for executing the above-mentioned method for acquiring X-ray image data by means of the described portable detector. In medical X-ray imaging patients are X-ray examined either by means of a stationary X-ray imaging system being located typically in special designed X-ray laboratory rooms or by means of a movable X-ray imaging system. Movable X-ray imaging systems are frequently used if a patient is not transportable. A stationary X-ray imaging system typically comprises a so-called bucky unit. A bucky unit is a box, which comprises a tray for an X-ray cassette, additionally an anti scatter grid and a so-called Automatic Exposure Control (AEC) unit. An anti scatter grid is used for instance for chest exposures in the intensive care department, especially for heavy patients. The grid improves the image quality significantly in particular for thick objects. The anti scatter grid can be optionally removed from the bucky unit. The AEC may be used for controlling an X-ray source in order to allow for optimally exposed images with and without an anti scatter grid. A movable X-ray imaging system is usually operated by using a free cassette for detecting X-rays, which have traversed a non transportable patient under examination. A free cassette is typically positioned just below the patient. U.S. Pat. No. 4,205,233 discloses an X-ray radiographic table, which comprises a bucky frame having a front opening through which a bucky unit respectively a bucky tray may be inserted into the frame. The bucky tray supports a cassette, which carries an X-ray sensitive film. The cassette is centered on the bucky tray between adjustable clamps and is adapted to be aligned with an X-ray source. JP 2004-073356 A discloses radiographic equipment comprising (a) a grid detection means for detecting the presence and absence of the grid, (b) a fixing means for fixing the grid, (c) a fixing detection means for detecting the fixing state of the grid and (d) a posture change restriction means for restricting the operation of the posture change of the photographic equipment. The operation of the posture change restriction means may be controlled according to the detecting result of the grid detection means. U.S. Pat. No. 6,850,597 B2 discloses an X-ray image photographing apparatus including an X-ray source and an X-ray detector. The X-ray image photographing apparatus further comprises a grid detecting means having a construction for detecting at least (a) the presence or absence of a grid, (b) the kind of the grid and (c) the presence or absence of the replacement of the grid. Further, the X-ray image photographing apparatus comprises an image processing system for image processing and outputting image data collected by the X-ray detector and a memory having stored a plurality of sets of image processing parameters for controlling the image processing system based on an output of the grid detecting means. The image processing starts from selecting a gain image. EP 1 420 618 A2 discloses an X-ray imaging apparatus including an X-ray source and an X-ray detector on which a plurality of different types of grids are detachably mountable. The X-ray detector includes an automatic exposure control (AEC) unit, which detects the quantity of X-rays emitted from the X-ray generation means and outputs a signal based on the detected quantity. The X-ray imaging apparatus also includes control means controlling the X-ray generation means based on the signal output from the AEC detector, and correcting the AEC detector output according to the type of grid used. There may be a need for providing X-ray equipment for increasing the reliability of the emitted radiation dose for X-ray imaging in particular for patients being not transportable. This need may be met by the subject matter according to the independent claims. Advantageous embodiments of the present invention are described by the dependent claims. According to a first aspect of the invention there is provided a portable X-ray detection device. The portable X-ray detection device comprises (a) a two-dimensional X-ray detector unit and (b) a sensing unit, which is adapted to recognize whether an anti scatter grid is attached to the X-ray detector unit. This first aspect of the invention is based on the idea that an automatic sensing of the presence of an optional anti scatter grid provides a reliable and useful information for a radiographer in order to adapt proper default exposure settings of an X-ray source in the moment an anti scatter grid is attached. Therefore, the risk of over- or under-exposures of X-ray images may be reduced. In particular, the risk to apply a too high radiation dose to a patient in case that an anti scatter grid is erroneously not used can be minimized effectively. Therefore, the workflow of acquiring X-ray images is simplified such that also a less experienced user can operate an X-ray imaging system comprising the described portable detector. In this context it is mentioned that an anti scatter grid may be any channel type X-ray absorption device providing for an X-ray attenuation, which compared to direct X-rays is different for scattered X-rays impinging onto the detector under at least a slightly slanted angle. Therefore, scattered radiation having a different angle of incidence may be suppressed such the contrast of the resulting X-ray images may be increased significantly. This means that the anti scatter grid removes radiation being scattered predominately in thick objects and thus improves the image contrast and the signal to noise ratio. For thin objects an anti scatter grid is typically not used because compared to the effect of a contrast enhancement the effect of deteriorating the signal to noise ratio is too big. The X-ray detector unit may be any type of X-ray sensitive element such as a radiographic film. However, since modern X-ray imaging systems usually rely on a digital image acquisition the X-ray detector unit might be a digital detector, which comprises an array of X-ray sensor elements. Thereby, each sensor element is capable of receiving an individual radiation dose, wherein the height of an output signal of that sensor element depends or is proportional to the individual radiation dose impinging onto the corresponding sensor element. Preferably, the portable X-ray detection device comprises a holder or a fixations means, which allows for a precise spatial positioning of an anti scatter grid relative to the X-ray detector unit. A detachably fixation of an anti scatter grid may be realized for instance by means of a clip mechanism, which allows for an easy handling of the anti scatter grid. According to an embodiment of the invention the portable X-ray detection device comprises a handle, which is adapted to facilitate a manual transportation of the portable X-ray detection device. The provision of a handle has the advantage that the portable X-ray detection device may be transported in an ergonomic advantageous manner to almost any location where an X-ray examination of a non-transportable patient is required. According to a further embodiment of the invention the portable X-ray detection device has a weight of less than 10 kg, preferably less than 8 kg and more preferably less than 6 kg. By contrast to so-called stationary X-ray detectors, which are used for stationary X-ray imaging systems and which typically have a weight of approximately 20 kg or even more, for the benefit of a comparatively easy transportation a portable X-ray detector comprises less lead shielding. According to a further embodiment of the invention the portable X-ray detection device has a flat structure with a height of less than 5 cm, preferably less than 3 cm. This may provide the advantage that the portable X-ray detection device may be positioned even within narrow or small regions, which are available close to a patient. For instance the portable X-ray detection device may be positioned directly under a patient's mattress. Therefore, the flat structure of the portable X-ray detection device allows for an X-ray image data acquisition without amending significantly the posture of a patient under examination. As a consequence, the described portable X-ray detection device is useful for a variety of different medical X-ray imaging applications. According to a further embodiment of the invention the sensing unit is adapted to recognize the type of an anti scatter grid being attached to the X-ray detector unit. The capability of identifying the type of an anti scatter grid may allow for an even more precise adaptation of default exposure settings of an X-ray source. According to a further embodiment of the invention the portable X-ray detection device further comprises an automatic exposure control unit (AEC), which is adapted to measure the radiation dose impinging onto the X-ray detector unit in real time. This may provide the advantage that the radiation dose can be monitored and, in case a sufficient radiation dose has been received by the X-ray detector unit, an X-ray source generating the X-radiation can be switched of and/or a shutter can be closed such that no more X-rays impinge onto the X-ray detector unit. This may allow for a reliable operation of the X-ray source even if the X-rays are detected by means of a portable and not by means of a stationary detector. Preferably, the AEC unit comprises an X-radiation dose measurement device such as an ionization chamber. According to a further embodiment of the invention the portable X-ray detection device is adapted to be inserted into a bucky unit. In this respect a bucky unit may be any box which is adapted to receive an anti scatter grid and an X-ray detection device in a precise spatial orientation with respect to each other. Optionally, a bucky unit may also be configured in order to receive an ACE unit. A portable X-ray detection device, which may be inserted into a bucky unit, has the advantage that the portable X-ray detection device may also be used for stationary X-ray systems. Therefore, the portable X-ray detection device may be used for a variety of different purposes. According to a further aspect of the invention there is provided an X-ray imaging system, in particular a medical X-ray imaging system. The X-ray imaging system comprises (a) an X-ray source, which is adapted to generate X-rays penetrating an object under examination, and (b) a portable X-ray detection device as described above, wherein the portable X-ray detection device is adapted to receive X-rays, which have been penetrated the object under examination. This aspect of the invention is based on the idea that the portable X-ray detection device might be employed in a beneficial manner for X-ray imaging, in particular for medical X-ray imaging. Thereby, the information being provided by the sensing unit with respect to the presence or absence of an anti scatter grid might be used for operating the X-ray source. In particular, when an anti scatter grid is detected, the X-ray source may be operated in such a manner, that an increased radiation dose is emanated from the X-ray source. By contrast thereto, when there is no anti scatter grid placed in front of the two-dimensional X-ray detector unit, the X-ray source may controlled such that only a reduced X-radiation dose is emanated from the X-ray source. This allows for an automatic exposure control such that an object under examination is only subjected to a radiation dose, which is sufficient in order to acquire high contrast X-ray images. The X-ray source typically is an X-ray tube. However, the provided portable X-ray detection device may also be used in connection with other X-ray sources such as e.g. a synchrotron radiation source providing for a quasi-focused X-radiation. According to an embodiment of the invention the X-ray imaging system further comprises an X-ray generator device for providing electric energy to the X-ray source, which X-ray generator device is coupled to the sensing unit. This may provide the advantage that the radiation dose an object under examination is exposed to may be effective controlled in such a manner, that the presence or the absence of an anti scatter grid has a strong influence on the electrical control of the X-ray source. Of course, the X-ray generator device may be coupled directly or indirectly to the sensing unit. According to a further embodiment of the invention the X-ray imaging system further comprises a control unit, which is coupled both to the portable X-ray detection device and to the X-ray generator device. This may provide the advantage that the operation of the X-ray generator device can be controlled by means of an appropriate software, wherein depending on the presence or absence of an anti scatter grid different parameter datasets are used for the operation of the X-ray generator device. Preferably, a data connection in between the sensing unit of the portable X-ray detection device and the X-ray generator device is provided, which allows for a real time communication. This means that the information regarding the presence or absence of an anti scatter grid is available immediately after the anti scatter grid has been attached to or removed from the two-dimensional X-ray detector unit. According to a further embodiment of the invention the control unit is adapted to select one of at least two predefined parameter datasets for operating the X-ray generator device, whereby the selection depends on the presence or absence of an anti scatter grid. Such parameter datasets may be for instance so called Automatic Programmed Radiography (APR) parameter datasets, which may be stored in a memory of the control unit. This means that information from sensing an anti scatter grid is used to control the selection of APR-settings, whereby different APR-settings are available for X-ray examination with and without grid. Preferably, the APR parameter datasets are selected based on an examination code, which may be available in the control unit. Thereby, an examination code non-ambiguously refers to a selected body part of a patient under examination. Therefore, the optimal radiation dose, which depends on both the respective body part of the patient under examination and the presence or absence of an anti scatter grid may be automatically adjusted. Therefore, an over- or an under-exposure of images can be avoided. As a consequence, the reliability of the X-ray imaging system is increased significantly such that the X-ray imaging system may also be operated by a comparatively poor skilled radiographer without increasing the risk for a wrong X-ray exposure. However, in this respect it has to pointed out that the parameter datasets as well as the examination codes may also be available from a remote computer or a remote memory by means of a network, such as the WorldWideWeb, from which the parameter datasets may be downloaded. Of course, also the examination codes may also by acquired via a computer network. According to a further embodiment of the invention the predefined parameter dataset is designed to be used both with and without an anti scatter grid by attaching a grid parameter representing the presence or the absence of the grid. This means that a grid correction factor is not included explicitly within the APR parameter sets. The grid correction factor is rather stored outside the APR parameter sets, wherein one parameter of the APR parameter sets refers to the grid correction factor. This may provide the advantage that the total number of parameter datasets can be reduced effectively by a factor of 2. Thereby, an additional grid correction factor is used for adapting the respective APR parameter set on the presence or absence of an anti scatter grid. The grid correction factor may include two values. For instance a first value may describe a change of the acceleration voltage of the X-ray source by shifting the acceleration voltage by a predetermined voltage difference. Another second value may describe a multiplication factor for the electron beam current with which the X-ray source is operated. Preferably, the additional grid correction factor reflects a “rule of thumb” for modifying the high voltage and/or the current of X-ray generator device, which is fed to the X-ray source. The “rule of thumb” knowledge of radiographers how to take an X-ray image with and without an anti scatter grid may by stored within a whole list of APR parameter set. The application of the “rule of thumb” may be triggered by an anti scatter grid detection prior to exposure of the X-ray image. In order to be sure that an over-exposure of a patient under examination can be excluded each predefined parameter dataset contains a programmed reference for selecting an X-ray exposure corresponding the absence of an anti scatter grid. Therefore, only if an anti scatter grid is detected, the radiation dose originating from the X-ray source will be increased. Notably too, selection of the APR data set corresponding to the body part under examination, in combination with grid correction factor to be applied, constitute selection of an active mode of the X-ray source. The X-ray source is operated in the selected mode to deliver the corresponding radiation dose. In the following there will be described exemplary embodiments of the present invention with reference to a method for acquiring X-ray image data. It has to be pointed out that of course any combination of features relating to different subject matters is also possible. According to a further aspect of the invention there is provided a method for acquiring X-ray image data, in particular for acquiring medical X-ray imaging data of a patient under examination. The provided method comprises the steps of (a) recording X-ray attenuation data by means of a portable X-ray detection device as described above, (b) determining the presence or the absence of an anti scatter grid in front of the two-dimensional X-ray detector, and (c) operating an X-ray source based on an output signal of the sensing unit indicating the presence or the absence of an anti scatter grid. This aspect of the invention is based on the idea that a determination of the presence or the absence of an anti scatter grid may automatically trigger a predetermined operating mode of the X-ray source. Thereby, the radiation dose originating form the X-ray source may be automatically adapted in order to compensate for the X-ray attenuation caused by the anti scatter grid. This may allow for a reliable exposure setting such that an erroneously under- and in particular an erroneously over-exposure of a patient under examination can be effectively avoided. According to an embodiment of the invention the step of operating an X-ray source is further based on an output signal of an automatic exposure control unit, which is associated with X-ray detector unit. As has already been described above such an online monitoring of the radiation dose may provide the advantage that the X-ray exposure of an object under examination may be immediately stopped if a sufficient radiation dose has been received by the X-ray detector unit in order to allow a high quality X-ray image without having the risk of an over-exposure of a patient under examination. According to a further embodiment of the invention the step of operating an X-ray source comprises selecting one of at least two predefined parameter datasets for operating the X-ray generator device. This may provide the advantage that the described method may be carried out with taking benefit of so called Automatic Programmed Radiography (APR) parameter datasets. Such APR datasets may include pre-programmed values for e.g. the acceleration voltage of an X-ray tube and the electron beam current of an X-ray tube. Of course, the APR datasets may also depend on the body part of a patient under examination. According to a further aspect of the invention there is provided a computer-readable medium on which there is stored a computer program for acquiring X-ray image data, in particular for acquiring medical X-ray imaging data of a patient under examination. The computer program, when being executed by a control unit, is adapted for performing embodiments of the above-described method for acquiring X-ray image data. According to a further aspect of the invention there is provided a program element for acquiring X-ray image data, in particular for acquiring medical X-ray imaging data of a patient under examination. The program element, when being executed by a control unit, is adapted for performing embodiments of the above-described method for acquiring X-ray image data. The computer program element may be implemented as computer readable instruction code in any suitable programming language, such as, for example, JAVA, C++, and may be stored on a computer-readable medium (removable disk, volatile or non-volatile memory, embedded memory/processor, etc.). The instruction code is operable to program a computer or other programmable device to carry out the intended functions. The computer program may be available from a network, such as the WorldWideWeb, from which it may be downloaded. It has to be noted that embodiments of the invention have been described with reference to different subject matters. In particular, some embodiments have been described with reference to apparatus type claims whereas other embodiments have been described with reference to method type claims. However, a person skilled in the art will gather from the above and the following description that, unless other notified, in addition to any combination of features belonging to one type of subject matter also any combination between features relating to different subject matters, in particular between features of the apparatus type claims and features of the method type claims is considered to be disclosed with this application. The aspects defined above and further aspects of the present invention are apparent from the examples of embodiment to be described hereinafter and are explained with reference to the examples of embodiment. The invention will be described in more detail hereinafter with reference to examples of embodiment but to which the invention is not limited. FIG. 1 shows a flow chart how a method for acquiring X-ray image data with a portable X-ray detector may be carried out. The method starts with a step S1. In step S2 there is recorded a two-dimensional X-ray attenuation data set. Thereby, a portable X-ray detection device is employed, which portable X-ray detection device comprises a two-dimensional X-ray detector unit and a sensing unit, which is adapted to recognize whether an anti scatter grid is attached to the X-ray detector unit. In step S3 there is determined the presence or the absence of an anti scatter grid in front of the two-dimensional X-ray detector. Thereby, an output signal provided from the sensing unit is evaluated. In step S4 there is selected a predefined application programmed radiography (APR) parameter dataset from a plurality of different APR parameter datasets, which are stored in a memory associated to a control unit for carrying of the described method for acquiring X-ray image data. The selection depends on the presence or absence of an anti scatter grid. This means that information from sensing an anti scatter grid is used to control the selection of APR-settings, whereby different APR-settings are available for X-ray examination with and without grid. According to the embodiment described here the APR parameter datasets are selected based on an examination code, which may be available in the control unit. Thereby, an examination code non-ambiguously refers to a defined body part of a patient under examination. Therefore, the optimal radiation dose, which depends on both the respective body part of the patient under examination and the presence or absence of an anti scatter grid may be automatically adjusted. In step S5 an X-ray tube is operated based on the selected APR dataset. Since the information whether an anti scatter grid is positioned within the X-radiation beam has already been used for selecting an appropriate APR parameter dataset, the X-ray tube is operated under conditions which take into account the presence or absence of an anti scatter grid. The APR parameter dataset in particular includes a first value for the acceleration voltage for electrons impinging onto the anode of the X-ray tube and a second value for the current of the electron beam hitting the anode. Both values have a strong influence on the radiation dose. The first value determines the spectral distribution of the X-radiation wherein the second value determines the intensity of the X-radiation. In step S6 there is carried out an automatic exposure control of the X-radiation being detected by the X-ray detector unit. Thereby, if an accumulated radiation dose has been reached, which is sufficient for a good quality X-ray image, the X-radiation impinging onto the object under examination is switched off or blocked. This allows for an online monitoring and for an online controlling of the radiation dose a patient is exposed to. It has to be pointed out that the step S6 is optionally. This means, that the method for acquiring X-ray image data with a portable X-ray detector may be carried out also without this step. However, according to the embodiment described here, carrying out step 6 makes the method more reliable with respect to an erroneous radiation dose. FIG. 2 shows in a schematic representation a block diagram of a medical X-ray imaging system 200. The X-ray imaging system 200 comprises an X-ray tube 210, which is adapted to generate X-rays 212 originating from a not depicted anode of the X-ray tube 210. The X-rays 212 penetrate at least partially a patient 220 under examination such that attenuated X-ray 222 impinge onto an anti scatter grid 230, which is fixed directly in front of a portable X-ray detection device 240. The device 240 is operable in connection with a configuration 205, e.g., the rest of FIG. 2 except for the patient 220. The X-ray tube 210 is controlled respectively driven by an X-ray generator 290 for providing electric energy to the X-ray source 210. The anti scatter grid 230 may be any channel type X-ray absorption device providing for an X-ray attenuation, which compared to direct X-rays is different for scattered X-rays impinging onto the grid 230 under at least a slightly slanted angle. Therefore, the anti scatter grid 230 removes X-radiation being scattered within nuclei of the patient 220. Predominately direct X-rays 232, which have not been scattered within the patient 220, penetrate the anti scatter grid 230 and impinge onto a two-dimensional detector array 241 of the portable detector 240. However, independent from the ratio of scattered X-rays the grid 230 definitely reduces the intensity of the X-radiation, which intensity can be detected by the detector array 241. Therefore, according to the embodiment described here, it is ensured that the X-ray intensity being generated by the X-ray source 210 is automatically adapted to the presence or the absence of the anti scatter grid 230. In order to provide for such an automatic adaptation the portable X-ray detector 240 comprises an active grid sensing device 245. The grid sensing device 245 is capable of detecting an anti scatter grid 230, which is attached in a predefined position in front of the portable X-ray detector 240. The grid detection is realized by means of a passive grid sensing device 235, which is provided at the anti scatter grid 230. The interaction between the passive grid detection device 235 and the active grid sensing device 245 can for instance comprise a mechanical engagement of a nose or any other element projecting from the inner surface of the anti scatter grid 230 into a recess of the active grid sensing device 245. However, the presence of the anti scatter grid 230 can also be detected be closing or opening an electrical contact the passive grid detection device 235 and the active grid sensing device 245. Of course also other interactions such as a magnetic interaction by means of e.g. a reed relay between the passive grid detection device 235 and the active grid sensing device 245 can be used for reliably and effectively sensing the presence of the anti scatter grid 230. Further, the operation of the grid sensing device 245 may also take benefit from a transponder unit and/or any other device using RFID technology. Anyway, according to the embodiment described here, the active grid sensing device 245 generates an output signal, which output signal depends on the presence respectively the absence of the anti scatter grid 230. In order to effectively avoid an over-exposure of the patient under examination 220 the portable X-ray detector 240 is provided with an automatic exposure control unit 247. The automatic exposure control unit 247 is a measurement device comprising an ionization chamber, which outputs a signal as soon as an accumulated X-ray radiation dose has been reached by means of a single X-ray exposure or a plurality of X-ray exposures. Therefore, the automatic exposure control unit 247 may be useful in order to effectively prevent an erroneously over-exposure of a patient under examination 220. As can be seen from FIG. 2, the portable X-ray detector 240 is equipped with a handle 249: The handle 249 is adapted to facilitate a manual transportation of the portable X-ray detector 240. In order to contribute to a comfortable transportation the portable X-ray detector 240 has a weight of less than 8 kg and more preferably less than 6 kg. Therefore, by contrast to so-called stationary X-ray detectors, which are used for stationary X-ray imaging systems and which typically have a weight of approximately 20 kg or even more, the described portable X-ray detector 240 allows for a comparatively easy transportation. The portable X-ray detector 240 further has a flat structure with a height of less than 5 cm, preferably less than 3 cm. This may provide the advantage that the portable X-ray detector 240 may be positioned for instance directly under a patient's mattress without disturbing significantly the posture of a patient under examination. The output signal generated by the active grid sensing device 245 is applied to a control unit 250 of the X-ray imaging device 200. The control unit 250 may be realized by means of a host computer such as a PC or a workstation. Further, a stop signal, which may be generated by the automatic exposure control unit 247 is also transferred to the control unit 250. The control unit 250 comprises, realized either by software, by hardware of by a combination of software and hardware, an application programmed radiography (APR) selection means 261. Since the signal provided by the active grid sensing device 245 includes to information whether a grid is used or not, the APR selection means 261 can select a proper APR parameter dataset from a series of different APR parameter datasets 265a, 265b, 265c and 265d. Each parameter dataset 265a, 265b, 265c or 265d includes values for driving the X-ray source 210 with a predetermined acceleration voltage and a predetermined electron beam current, wherein apart from being adapted to the presence of the anti scatter grid 230 these values are optimized for a specific body part of the patient 220. The portable detector 240 is adapted to be used both in a so-called “free cassette operation” as well as in a bucky unit of a stationary medical X-ray imaging system. i.e., in either configuration 205. Therefore, according to the embodiment described here, there are four different APR parameter dataset available for controlling the X-ray generator 290 respectively the X-ray tube 220 in order to expose a particular body part of the patient 220 with the proper X-radiation dose. The first APR parameter dataset (i) is used for an X-ray imaging in the “free cassette operation”, wherein neither the anti scatter grid 230 nor the AEC unit 247 is used. This APR parameter dataset (i) is typically used for thin objects like hands or for median thick objects like knees. The second APR parameter dataset (j) is used for an X-ray imaging in the “free cassette operation”, wherein the anti scatter grid 230 is used, but the AEC unit 247 is not used. This APR parameter dataset (i) is typically used for thick objects like abdomen or median thick objects like knees. The third APR parameter dataset (k) is used for an X-ray imaging, wherein the portable detector 240 is inserted in to a bucky tray of a stationary medical X-ray imaging system. Both the anti scatter grid 230 and the AEC unit 247 are used. This APR parameter dataset (k) is typically used for thick objects like abdomen or chests. The fourth APR parameter dataset (l) is used for an X-ray imaging, wherein the portable detector 240 is inserted in to a bucky tray. The anti scatter grid 230 is not used whereas the AEC unit 247 is used. This APR parameter dataset (l) is typically used for thin objects under examination. In order to effectively prevent an over-exposure of the patient 220 the control unit 250 further comprises, realized either by software, by hardware of by a combination of software and hardware, an AEC signal detection means 262, which is coupled to both the AEC unit 247 and the X-ray generator 290. However, it is also possible that the AEC unit 247 is coupled directly to the X-ray generator 290. The control unit 250 further comprises a memory 270 comprising data representing a patient list with information about the scheduled X-ray imaging examinations. The memory 270 is coupled to the data processor 260 in order to allow for a quick and secure selection of proper APR programs 265a, 265b, 265c and 265d. Thereby, the proper APR program will be selected automatically following an examination code that is available in the memory 270. According to the embodiment described here, the examination code is transferred from a Radiology Information System RIS 280 via a network 281 or a bus system. Based on the information given with the examination code the radiographer may be able to decide about the use of a grid during the preparation of the X-ray examination. This means that an initially pre-selected APR program will not always match with the current need. If initially the pre-selected APR program (i) has been selected for a “no grid operation” but now a grid has to be used, the APR setting has to be overridden to APR program (j). According to the embodiment described herewith, the proper APR program selection of (i) to (j) can be done automatically by the medical X-ray imaging system 200. Therefore, the probability of generating a proper X-ray radiation dose is increased significantly such that a potential source of error is eliminated, which source of error is existent when APR programs are selected manually. The APR program may include a so-called grid correction factor, which includes two values. A first value describes a change of the acceleration voltage of the X-ray source by shifting the acceleration voltage by a predetermined voltage difference. A second value describes a multiplication factor for the electron beam current with which the X-ray source is operated. For example, the APR program for a knee may have an acceleration voltage of 60 kV and an electron beam current of 5 mAs when no anti scatter grid is used. Using an anti scatter grid would increase these values by a voltage difference of 3 kV and by a factor of 2. This would lead to an acceleration voltage of 63 kV and an electron beam current of 10 mAs. Of course, such a rule of thumb can be applied to any APR programs, which do not need to be modified at all. Such rules would be very similar to the rules applied by the radiographer today. This rule of thumb can be for instance implemented in a program code by the following instructions: IF grid=yes and APR-default=grid_no THEN increase kV and mAs using delta_kV and mAs-factor IF grid=no and APR-default=grid_yes THEN decrease kV and mAs using delta kV and mAs-factor. It should be noted that the term “comprising” does not exclude other elements or steps and the “a” or “an” does not exclude a plurality. Also elements described in association with different embodiments may be combined. It should also be noted that reference signs in the claims should not be construed as limiting the scope of the claims. In order to recapitulate the above described embodiments of the present invention one can state: It is described a portable X-ray system 200, which has sensing means for detecting whether an anti scatter grid 230 is attached to a portable detector 240 or not. The system is able to automatically change the default exposure settings 265a, 265b, 265c, 265d, when a grid 230 is removed or attached to the portable detector 240. Thus, the risk of an under- or an over-exposure of the image will be reduced. List of reference signs:S1step 1S2step 2S3step 3S4step 4S5step 5S6step 6S7step 7200medical X-ray imaging system205configuration210X-ray source/X-ray tube212X-radiation originating from X-ray source220object/patient under examination222X-radiation penetrating patient230anti scatter grid232X-radiation leaving anti scatter grid/X-radiation impinging ontodetector235grid sensing device (passive)240portable X-ray detection device/portable X-ray detector241two dimensional X-ray detector unit/detector array245grid sensing device (active)247automatic exposure control unit249handle250control unit/host computer260data processor261APR selection means262AEC signal detection means265afirst APR program265bsecond APR program265cthird APR program265dfourth APR program270memory280Radiology Information System281bus system/network290X-ray generator device
	abstract	In a radiographic apparatus, positions of a radiation grid and a radiation detector are determined such that, when a radiation source and the radiation detector are in a standard position, an arrangement pitch of shadows of absorbing foil strips appearing on a detecting plane of the radiation detector as a result of a radiation beam being emitted from the radiation source and blocked by the radiation grid is an integral multiple of an arrangement pitch in a transverse direction of radiation detecting elements. Further, the shadows of the absorbing foil strips appear without covering transversely adjacent pairs of the detecting elements.
	summary
	claims	1. A windshield repair device comprising a bridge, an injector attached to the bridge for injecting resin into a damaged area of a windshield and a plurality of LED light sources attached to the bridge to provide light around the injector, wherein the LED light sources are attached to the bridge by being held in openings and extend downwardly from the openings closely spaced around the injector so as to allow the light from the LED light sources to cure resin injected by the injector into the damaged area of a windshield. 2. The windshield repair device according to claim 1, wherein an LED light source is provided within the injector. 3. The windshield repair device according to claim 1, wherein the light sources are UV LED light sources. 4. The windshield repair device according to claim 1, wherein the plurality of LED light sources are attached directly to the bridge closely spaced around the injector. 5. The windshield repair device according to claim 4, wherein the LED light sources are held in the openings in the bridge by a socket arrangement. 6. The windshield repair device according to claim 5, further comprising a battery compartment provided in the bridge. 7. The windshield repair device according to claim 5, wherein the injector comprises an injector seal configured to be pressed against outer glass of the windshield, the injector seal being made of a clear material. 8. The windshield repair device according to claim 1, further comprising a battery compartment provided in the bridge. 9. The windshield repair device according to claim 1, wherein the bridge comprises a main support portion and an extension arm to which the injector attaches, a suction cup attached to an underside of the main support portion, and adjusting screws attached to the main support portion adapted to level the bridge on the windshield. 10. The windshield repair device according to claim 9, wherein the plurality of LED light sources are attached directly to the bridge closely spaced around the injector. 11. The windshield repair device according to claim 10, wherein the LED light sources are held in the openings in the bridge by a socket arrangement. 12. The windshield repair device according to claim 11, further comprising a battery compartment provided in the bridge. 13. The windshield repair device according to claim 1, wherein no reflector is provided in the openings in which the LED light sources are held. 14. A light source for a windshield repair device, comprising holding device having a shape adapted to be connected to a bridge of a windshield repair device and fit around an injector of the windshield repair device for injecting resin into a damaged area of a windshield, and a plurality of LED light sources provided on the holding device, wherein the LED light sources are attached to the holding device by being held in openings and extend downwardly from the openings, the holding device and openings being configured such that the LED light sources are closely spaced around the injector so as to allow the light from the LED light sources to cure resin injected by the injector into the damaged area of a windshield. 15. The UV light source for a windshield repair device according to claim 14, wherein LED light sources are UV LED light sources. 16. The UV light source for a windshield repair device according to claim 14, wherein no reflector is provided in the openings in which the LED light sources are held. 17. A method for curing resin provided in a crack in a windshield, comprising:providing a windshield repair device having comprising a bridge, an injector attached to the bridge for injecting resin into a damaged area of a windshield and at least one a plurality of UV LED light source sources attached to the bridge or the injector to provide light within or around the injector, wherein the UV LED light sources are attached to the bridge by being held in openings and extend downwardly from the openings closely spaced around the injector so as to allow the light from the UV LED light sources to cure resin injected by the injector into the damaged area of a windshield;injecting resin into the damaged area of a windshield; andexposing the resin to UV light from the plurality of UV LED light sources of the windshield repair device according to claim 1.
	description	This application is a National Stage of International Application No. PCT/AU2016/050282 filed with the Australian Patent Office on Apr. 15, 2016, which claims priority to Australian patent application No. 2015/901,349 filed Apr. 15, 2015, wherein the entirety of each of the aforementioned applications is hereby incorporated herein by reference. The present disclosure relates to selectively separating elements or commodities from aqueous streams. In particular, the disclosure relates to selectively separating a soluble metallic constituent from other soluble metals and other materials accompanying the metallic constituent in a mixture. Extraction of metallic constituents such as uranium from its ores is commonly carried out by processes which include leaching the ore or a concentrate thereof. The process of leaching yields a solution which contains both uranium and other commodities and impurities. It is often difficult to separate uranium from a mixture that contains other commodities and impurities such as rare earth elements. Aqueous streams such as mineral processing/metallurgical streams that result from one or more processes including leaching or chemical extraction in acidic, neutral or alkali reagents may often contain a range of elements (as ions, molecules, complexes etc.) that may be considered both as commodities and contaminants in the form of metals and metalloids (e.g. As, Cd, Cr, Se, Tl, radionuclides). In addition, where seawater, saline surface or groundwaters are used as part of the extraction or treatment process, they may contain a range of other cations (e.g. Na+, Ca2+) or anions (e.g. Cl−, SO42−) or one or more cationic, anionic or neutral additives introduced during the process (e.g. surfactants, complexing moieties, other organic or inorganic compounds) that may be considered problematic or deleterious in terms of optimisation of element extraction, recovery or purification in subsequent processes (e.g. ion exchange, dissolution, precipitation). Therefore, it is desirable to provide a process for selective separation of one or more constituents of the aqueous streams such as the streams described above. Layered double hydroxides (LDH) are a class of both naturally occurring and synthetically produced materials characterised by a positively-charged mixed metal hydroxide layers separated by interlayers that contain water molecules and a variety of exchangeable anions. LDH is most commonly formed by the co-precipitation of divalent (e.g. Mg2+, Fe2+) and trivalent (e.g. Al3+, Fe3+) metal cation solutions at moderate to high pH. An LDH compound may be represented by the general formula (1):M(1-x)2+Mx3+(OH)2An−yH2O (1)where M2+ and M3+ are divalent and trivalent metal ions, respectively and An− is the interlayer ion of valence n. The x value represents the proportion of trivalent metal ion to the proportion of total amount metal ion and y denotes variable amounts of interlayer water. Common forms of LDH comprise Mg2+ and Al3+ (commonly known as hydrotalcites [HT]) and Mg2+ and Fe3+ (known as pyroaurites), but other cations, including Ni, Zn, Mn, Ca, Cr and La, are known. The amount of surface positive charge generated is dependent upon the mole ratio of the metal ions in the lattice structure and the conditions of preparation as they affect crystal formation. The formation of HT (the most commonly synthesised LDH frequently with carbonate as the principal “exchangeable” anion) may be most simply described by the following reaction:6MgCl2+2AlCl3+16NaOH+H2CO3→Mg6Al2(OH)16CO3·nH2O+2HCl Typically, ratios of divalent to trivalent cations in Hydrotalcites vary from 2:1 to 4:1. Other synthetic pathways to form HT (and other LDH) include synthesis from Mg(OH)2 (brucite) and MgO (calcined magnesia) via neutralisation of acidic solutions. This can be described by the following reaction:6Mg(OH)2+2Al(OH)3+2H2SO4→Mg6Al2(OH)16SO4·nH2O+2H2O A range of metals of widely varying concentrations may also be simultaneously co-precipitated, hence forming a polymetallic LDH. HT or LDH were first described over 60 years ago. Sometimes, they can also occur in nature as accessory minerals in soils and sediments. Layered double hydroxides may also be synthesised from industrial waste materials by the reaction of bauxite residue derived from alumina extraction (red mud) with seawater, as described by the following reaction:6Mg(OH)2+2Al(OH)3+2Na2CO3→Mg6Al2(OH)16CO3·nH2O+2NaOHor by the reaction of lime with fly ash derived from fossil fuel (e.g. coal fired power stations). Within the LDH or HT structure there are octahedral metal hydroxide sheets that carry a net positive charge due to limited substitution of trivalent for divalent cations as described above. As a consequence, it is possible to substitute a wide range of inorganic or organic anions into the LDH or HT structure. These anions are often referred to as “interlayer anions” as they fit between the layers of hydroxide material. Layered double hydroxides are generally unstable below a pH of approximately 5 but may act as buffers over a wide range of solution pH. Layered double hydroxides or HT, and in particular those that contain carbonate as the predominant anion, have also been demonstrated to have a considerable capacity to neutralise a range of mineral acids via consumption of both the hydroxyl and carbonate anions contained within the LDH structure. A number of studies have been conducted to investigate ways to exploit the anion exchange properties of LDH. These studies have focused on the removal of phosphate and other oxyanions and humic substances from natural and wastewater(s). Phosphate is one of the many anions that may be exchanged into the interlayer space in LDH. Laboratory studies of phosphate uptake using synthetically prepared Mg—Al HT and a range of initial dissolved phosphate concentrations indicate an uptake capacity of from ca. 25-30 mg P/g to ca. 60 mg P/g with uptake also influenced by initial phosphate concentration, pH (with maximum phosphate absorption near pH 7), degree of crystallinity and the HT chemistry. A major obstacle to the use of HT for phosphate removal in natural and/or wastewaters is the selectivity for carbonate over phosphate, with a selectivity series in the approximate order CO32−>HPO42−>>SO42−, OH−>F−>Cl−>NO3−. Many HT are also synthesised with carbonate as the predominant anion and thus require anion exchange before they are exposed to phosphate. When carbonate is also combined with sulphate, nitrate and chloride (as might commonly occur in natural or wastewaters) the reduction of phosphate absorption to the HT is further decreased. A number of recent studies have focused on the formation and study of synthetic LDH or specifically HT or similar and their subsequent reactivity to a range of anions, particularly silicate with a view to forming polymetallic aluminosilicates, which as potential precursors to clay materials, are thought to limit metal mobility and bioavailability. A potential also exists for the co-precipitation of silicate and aluminate anions as another precursor of analogue of clay minerals. Thus, other structural elements or interlayer ions may be incorporated (both inorganic and organic) to assist in both substitution and/or incorporation of ions from solution and/or increased stability. Subsequent formation of chlorite- or phyllosilicate-like minerals from pure Mg—Al or predominantly Mg—Al HT which may be similar to the HT, or iso-chemical in composition when compared to the HT, or may possess a similar chemistry as the HT with substitution of some ions as determined by the nature of Mg and/or Al added or the nature and chemical composition of the natural or wastewater which may influence the final geochemical composition, crystallinity or mineralogy. This increased stability of LDH or HT or chlorite-like minerals or other LDH or HT derivatives may also be achieved possibly in combination with chemical methods described above by partial or complete evaporation, calcination or vitrification leading to part or complete dehydration and partial/total recrystallisation. The use of co-amendments with, or encapsulation of, the LDH or HT may also be an option to further increase physical or chemical stability. The International Atomic Energy Agency (which is the international centre of cooperation in the nuclear field working with member states and multiple partners worldwide to promote safe, secure and peaceful nuclear technologies) published a report in 2004 summarising the state of the art in the field of treatment of effluents from uranium mines and mills. Importantly, the report omits any reference to one or more processes whereby the addition of chemical compounds to modify solution chemistry to form LDH or HT for the treatment of effluents from uranium mines. Throughout this specification, the word “comprising” and its grammatical equivalents is to be taken to have an inclusive meaning unless the context of use indicates otherwise. It will be clearly understood that, if a prior art publication is referred to herein, this reference does not constitute an admission that the publication forms part of the common general knowledge in the art in Australia or in any other country. In a first aspect, the disclosure provides a process for selectively separating a soluble metallic constituent from other soluble metals and other materials accompanying the metallic constituent in a mixture, the process comprising: (a) providing the mixture in an aqueous solution such that the metallic constituent forms a complex anion in the solution and wherein one or more of the other metals forms a cation or a complex cation in the solution; (b) contacting the solution with one or more additives to form layered double hydroxide (LDH) material in situ such that the complex anion is intercalated within interlayers of the LDH material and wherein one or more of the other metals are incorporated into the LDH material's crystal structure or matrix; and (c) selectively recovering the metallic constituent from the interlayer of the LDH by subjecting the LDH from step (b) to a recovery treatment step. In a second aspect, the disclosure provides a process for selectively separating a metallic constituent from one or more of other metals and other materials accompanying the metallic constituent in a mixture, the process comprising: (a) providing the mixture in an aqueous solution such that the metallic constituent forms a complex anion in the solution and wherein one or more of the other metals forms a cation or a complex cation in the solution; (b) contacting the solution with a layered double hydroxide (LDH) material such that the complex anion is intercalated within interlayers of the LDH material; and (c) selectively recovering the metallic constituent from the interlayer of the LDH by subjecting the LDH from step (b) to a recovery treatment step. In a third aspect, the disclosure provides a process for selectively separating uranium from rare earth elements and other materials in a mixture, the process comprising: (a) providing the mixture in an aqueous solution such that uranyl complex anions are present in the solution; (b) contacting the solution with one or more additives to form layered double hydroxide (LDH) material in situ; (c) such that the uranyl complex anion is intercalated within interlayers of the LDH material and wherein one or more of the rare earth elements is incorporated into the LDH material's crystal structure or matrix; and (d) selectively recovering uranium from the interlayer of the LDH by subjecting the LDH from step (b) to a recovery treatment step. In a fourth aspect, the disclosure provides a process for selectively separating uranium from rare earth elements and other materials in a mixture, the process comprising: (a) providing the mixture in an aqueous solution such that uranyl complex anions are present in the solution; (b) contacting the solution with LDH material such that the uranyl complex anion is intercalated within interlayers of the LDH material and wherein one or more of the rare earth elements is incorporated into the LDH material's crystal structure or matrix; and (c) selectively recovering uranium from the interlayer of the LDH by subjecting the LDH from step (b) to a recovery treatment step. Throughout the specification, a skilled person may understand that incorporation of the one or more metals into the crystal structure or matrix of the LDH material non-exclusively refers to incorporating the said metal(s) as a building block of the crystal structure of the LDH material and does not refer to mere adsorption of materials on the surface of the LDH material. The term “aqueous solution” as used herein refers to all of the waters, other liquids or solutes or solvents or mixtures whether miscible or immiscible and solids such as but not limited to mineral processing/metallurgical streams and electronic waste (e-waste) streams that result from one or more processes including leaching or chemical extraction in acidic, neutral or alkali reagents. The aqueous solution may contain a range of chemical species (as ions, molecules, complexes, micelles, aggregates, particulates or colloids and so forth). The aqueous solution may also contain a plurality of metals, metalloids, lanthanide or rare earth elements (REE), actinides, transuranic metals and radionuclides, any one of which may be regarded as a commodity or a contaminant. The term “metallic constituent” particularly but not exclusively encompasses metallic species capable of forming complex anions in aqueous solutions described herein. Suitable examples of metallic constituents include, but are not limited to, uranium, vanadium, thorium, chromium, and some transuranic metals. capable of forming complex anions in aqueous solutions. The metals referred to herein also include rare earth elements (REE) (15 metallic elements with atomic numbers ranging from 57 to 71). The REE are often described as part of the Lanthanide series and, for convenience, from time to time are represented as Ln3+. Other examples of suitable metals include but are not limited to radionuclides or transuranics capable of being present in aqueous solution in the 3+ oxidation state. The term “transuranic” will be used herein to refer to chemical elements with atomic numbers greater than 92 (the atomic number of uranium). All of these elements are unstable and decay radioactively into other elements. The term “transuranic” may be taken as a reference to transuranic metal species or transuranyl metal species. It will be appreciated by persons skilled in the art that although some transuranic metals may be present in aqueous solution in the 3+ oxidation state (e.g. Am3+ and Cm3+), these and other transuranic metals may alternatively be present in solution as transuranic cations (e.g. AmO22+ and PuO22+), depending on one or more characteristics of the aqueous solution such as pH, ionic strength, presence and concentration of one or more ligands, oxidation state of the transuranic metals and so forth. Transuranic cations are capable of forming transuranic complex anions with one or more ligands such as CO32− or SO42− in a similar manner to uranyl cations. The embodiments described herein provide an improved process for precipitating a range of elements from aqueous solutions such as process waters/metallurgical solutions which is particularly advantageous in separating or differentiating a mixture of elements or contaminants of interest within the LDH material. The embodiments utilise the different uptake mechanisms and behaviours of LDH materials (irrespective of whether the LDH material is formed in situ or is added to the aqueous solution) to efficiently separate or differentiate or partition a mixture of elements and contaminants. It is recognised that cations such as metal cations may be incorporated into the metal oxide layers of the LDH that forms the crystal structure or matrix. The applicant has also recognised that some metallic constituents, particularly metals such as uranium (vanadium or chromium or some transuranic metals in further embodiments) may exist as large sized oxy-cations such as UO22+ that cannot be accommodated into the crystal structure or matrix. The applicant has discovered that by tailoring or controlling reaction conditions such as pH conditions and/or addition of reaction agents etc., such large size oxy-cations such as UO22+ may be utilised for preferentially forming one or more complex anion species such as any of UO2(CO3)22−, UO2(CO3)34−, CaUO2(CO3)32−, UO22+—SO4. For example, at lower pH that UO22+—SO4 complexes (e.g. UO2(SO4)34−) may predominate while at intermediate to higher pH UO22+—CO32− anionic complexes (e.g. UO2(CO3)22−, UO2(CO3)34−, CaUO2(CO3)32−) may predominate. Given this speciation of the uranyl ion (UO22+) as anionic complexes, these anionic complexes preferentially partition into the anionic interlayers of the LDH material. Therefore, if the aqueous solution comprises a mixture of metals such as Cu, Mn, Ni, Pb, Zn and rare earth elements (REE; 15 metallic elements with atomic numbers ranging from 57 to 71; The REE are often described as part of the Lanthanide series and, for convenience, from time to time be represented as Ln3+) and uranium, the process would separate the uranium from the remaining metals, metalloids and rare earth elements by preferentially forming uranyl complex anionic complexes which would be intercalated into the interlayer of the LDH whereas at least some of the metals and rare earth elements would be incorporated into the crystal structure of the LDH. For example, REE as predominantly Ln3+ cations (Ce as +3 and +4 and Eu and +2 and +3 oxidation states) are strongly partitioned into the primary metal hydroxide layer of LDH materials substituting for other +3 cations such as Al and Fe. As a result of the process described herein, REE are for example contained within the metal hydroxide layers of the LDH and valuable uranium is contained as anionic complexes within the LDH interlayers. Some embodiments also result in the formation of an LDH material that typically may contain in excess of 30% U and 0-50% REE. Such resulting quantities of uranium and rare earth metals is typically 100-300 times higher than typical ore grades of these elements thus allowing substantial enrichment of the commodities of value. Another significant benefit afforded by some embodiments is that the process results in effective separation of potentially problematic ions such as Na, Cl and SO4 or other additives or components from the mineral processing or aqueous stream, thus potentially facilitating simpler processing, further enrichment, recovery or purification). Yet another advantage is the production of a cleaner effluent that may potentially be reused in mineral processing or other site applications or other operations without (or minimal) additional treatment. Certain embodiments may be employed to selectively separate transuranic metallic constituents and/or transuranic metals from complex mixtures comprising one or more of Pu, Np, Am and Cm in addition to U and daughter radionuclides. Advantageously, one or more characteristics of the aqueous solution or reaction conditions may be varied so that one or more transuranic elements of interest may be present in solution as a respective transuranic complex anion or a transuranic (III) cation. For example, it will be appreciated by those skilled in the art that any suitable redox reaction or disproportionation reaction may be employed to yield the desired transuranic and non-transuranic ion of appropriate oxidation state in solution, and that the resulting metal cations or anions may then be reacted with one or more ligands to produce a desired metal cation or metal complex anion in solution capable of being incorporated or intercalated into the LDH or excluded from the LDH. Accordingly, selective separation of one or more transuranic elements from said mixture may comprise contacting the solution with one or more additives to form layered double hydroxide (LDH) material in situ such that a transuranic complex anion formed in step a) may be intercalated within interlayers of the LDH material. Alternatively, selective separation of one or more transuranic elements from said mixture may comprise contacting the solution with one or more additives to form layered double hydroxide (LDH) material in situ such that a transuranic (III) cation formed in step a) may be incorporated into the LDH material's crystal structure or matrix. In alternative embodiments, selective separation of one or more transuranic elements from said mixture may comprise contacting the solution with a layered double hydroxide (LDH) material such that a transuranic complex anion formed in step a) may be intercalated within interlayers of the LDH material. In still further embodiments, selective separation of a transuranic element from said mixture may comprise contacting the solution with one or more additives to form layered double hydroxide (LDH) material in situ or by contacting the solution with a layered double hydroxide (LDH) material such that one or more complex anions other than the transuranic element of interest formed in step a) and, optionally, one or more metal cations other than the transuranic element of interest formed in step a) is incorporated into the LDH material's crystal structure or matrix. In this particular embodiment, the transuranic element of interest may be selectively separated from the mixture by excluding it from incorporation or intercalation with the LDH material, thereby leaving it in solution for subsequent recovery. Selective recovery steps include, but are not limited to, ion-exchange of the LDH interlayers or (partial) dissolution of the LDH to isolate specific transuranic, radionuclides or other metals. The inventor opines that the embodiments of the process as described herein may prove an effective alternative means of selective separation in comparison to existing techniques that may involve one or more sequential co-precipitation steps, ligand formation and ion-exchange. Some embodiments also provide a method of recovering the selectively separated constituents of the aqueous solution wherein the different constituents have been taken up by the LDH by differing uptake mechanisms (for example uranium in the interlayer of the LDH, whereas REE in the crystal structure or metallic oxide layers of the LDH) by subjecting the LDH from step (b) of the first, second or third aspects to a further recovery treatment step. The various recovery treatment steps that may be utilised have been detailed in the foregoing passages of the present specification. In some embodiments, the process comprises the recovery of the LDH from the aqueous stream before subjecting the separated LDH from step (b) to step (c) of the process described herein. Such recovery of the LDH may be carried out by recovery means such as sedimentation, flocculation or filtration. It may be appreciated that in some further aspects, the aqueous stream may also be contacted with pre-formed LDH material wherein at least a part of the total quantity of LDH material in the process is formed in-situ. In at least some embodiments, the step of contacting the pre-formed LDH material with the aqueous solution (as described in the second aspect and fourth aspect) comprises dissolving at least a part of the pre-formed LDH material into the solution thereby obtaining dissolved LDH in the solution. [Such dissolving of the LDH material typically results in the divalent and trivalent cations (forming the metal oxide layers of the pre-formed LDH) going into the solution as ionic species] This dissolution step is followed by controlling the reaction conditions in the aqueous solution for promoting in situ precipitation of LDH material from the dissolved LDH material such that the complex anion is intercalated within interlayers of the LDH material formed in situ and wherein one or more of the other metals are incorporated into the crystal structure or matrix of the LDH material formed in situ. Preferably, the step of dissolving the LDH in the aqueous solution comprises controlling the pH of the aqueous solution, preferably at pH levels of less than 7 and more preferably less than 5 and even more preferably less than 3. Furthermore, the subsequent step of controlling the reaction conditions in the aqueous solution may also comprise controlling the pH of the aqueous solution preferably at a pH level of greater than 8 to promote the precipitation of the LDH material from the dissolved LDH material. Such a process comprising initially lowering the pH of the solution for dissolving the LDH material followed by an increase in the pH for promoting reformation of LDH from the dissolved LDH may be referred to as pH cycling step. Thus in at least some embodiments, the uptake mechanism of the LDH materials relies on the dissolution and reformation of the added LDH material. A skilled person would appreciate that dissolving the pre-formed LDH material in an aqueous solution results in dissolution of at least a part of the metal oxide layer of the LDH material into the solution which results in divalent and trivalent cations being dissolved in the solution. The subsequent step of controlling the reaction conditions results in the reformation of the LDH material by precipitation and the dissolved divalent and trivalent ions along with some of the other metals form the crystal structure or matrix and interlayer anions of the LDH material reformed in situ. In an embodiment, the process further comprises the step of controlling the pH levels of the aqueous solution thereby controlling speciation of the complex anion. The pH conditions speciation of the anionic complexes may be suitably tailored or controlled. For example, at lower pH that UO22+−SO4 complexes (e.g. UO2(SO4)34−) may be preferentially formed whereas under intermediate to higher pH conditions UO22+−CO32− anionic complexes (e.g. UO2(CO3)22−, UO2(CO3)34−, CaUO2(CO3)32−) may predominately formed. In some embodiments, the LDH from step (b) may be separated from the aqueous solution by conventional separation methods such as sedimentation and/or filtration and/or cyclonic separation or other suitable solid-liquid separation means. The separated LDH may be treated to recover the constituents therefrom. In some further embodiments, the recovery treatment step may be carried by introducing the separated LDH to an ion-exchanging solution to cause ion-exchange to occur whereby the complex anion of the metallic constituent in the interlayer of the LDH is ion-exchanged with an anion in the ion-exchanging solution. In this manner the complex anion of the metallic solution goes into solution by carrying out the ion exchanging step. It will be appreciated that this ion-exchanging step involves the ion-exchange solution having at least one substituent anion such that the substituent anion displaces at least some of the intercalated anion or complex anion by an ion exchange mechanism thereby resulting in the anion or complex anion being released from the LDH interlayer into the ion-exchange solution. Whilst, the intercalated anion or complex anion of the metallic constituent is released from the interlayer of the LDH, the other metals (such as the REE/metals) that are present in the crystal structure or matrix of the LDH remain incorporated in the crystal structure or matrix of the LDH material. Recovering the LDH from step (b) and then conducting an ion exchange process is particularly beneficial when the initial aqueous solution in step (a) comprises leaching solutions having high salt concentrations (such as those used in leaching processes for recovering uranium) because achieving optimal ion exchanging efficiencies in such leaching solutions such as mining waste solutions was found to be difficult. Adopting the process described in some embodiments that involves the step of separating the LDH from step (b) and as described above results in higher ion exchanging efficiencies thereby producing better separation of the intercalated metallic constituent from the LDH. In a further embodiment, the ion exchanging step also comprises controlling pH conditions to promote displacement of the anion or complex anion from the interlayer and/or to promote speciation of a preferred type of anion or complex anion over other anions or complex anions. For example, a strong alkali may be added to displace UO22+−SO4 or UO22+−CO3 complexes by OH− anions by increasing the pH. It would be appreciated that such a recovery treatment step involves the recovery of the uranium (in the form of the uranyl complex anion) back into aqueous solution even though the REE remain incorporated in the LDH crystal structure. Alternatively, addition of a strong acid thereby reducing the pH such that charged or neutral UO22+ complexes are displaced from the interlayers of the LDH is also possible. In this, sustained acid addition may also sufficiently decompose the LDH crystal structure to liberate the REE or other metals or elements. In some embodiments, the substituent agent may comprise one or more of the following nitrilotriacetic acid (NTA), ethylenediaminetetraacetic acid (EDTA) or a range of other complexing agents such as crown ethers or other organic or (complex) inorganic ligands and/or wherein the substituent agent is substantially more electronegative relative to the intercalated complex anion in the LDH material thereby resulting in the substituent agent such as EDTA and/or NTA displacing the complex anion from the interlayer. In a further embodiment, the recovery step further comprises separating the LDH material after the ion exchanging step is completed. It may be appreciated that separating the LDH material after the ion-exchanging step is complete results in obtaining a separated LDH material which comprises the incorporated metallic cations or REE present in the crystal structure of the separated LDH material. The incorporated metallic cations or REE from the separated LDH may be recovered by methods such as heat treatment or thermal decomposition of the separated LDH material thereby resulting in the formation of a collapsed or metastable material. In a further embodiment, the process may include the addition of a further additive (such as silica) to the LDH material prior to or during the heat treatment or thermal decomposition. Preferably, the process may involve controlling the ratio of the further additive to the LDH material for selectively controlling formation of oxide materials upon the heat treatment or thermal decomposition. Additives such as silica may also be added to the LDH prior to or during the heat treatment or thermal decomposition step in a range of forms including crystalline silica (e.g. quartz), amorphous or chemically-precipitated silica, silicic acid, organic forms including tetra-ethylsilica(te) or silica added to the LDH interlayers. Controlling the ratio of the silica to the LDH and/or controlling the temperature of heat treatment may result in a series of reactions between the added forms of silica and the LDH resulting in the formation of a range of materials like minerals such as (or in addition to spinel and periclase) pyroxenes such as enstatite, olivines including forsterite and other minerals including silica transformed into high temperature forms including cristobalite. It will be appreciated that by varying the amount of silica or other elements in one or more forms relative to the LDH (by controlling the ratio of the silica to the LDH material) that different proportions or suites of minerals may form as a result of the heat treatment step. This process of including an additive as described above is particularly advantageous for two reasons. The first advantage is that secondary mineral oxides such as metallic silicates or pyroxenes may be formed that may constitute a suitable long-term repository for a range of contaminants including radionuclides. The second advantage is that given selected elements may be partitioned into materials formed as a result of the heat treatment, the materials formed (as determined by the composition of the LDH and the type and proportion of the additive) this may assist in the selective recovery of particular elements contained within selected minerals. Silica may be replaced by other additives in further embodiments and the embodiment described above is in no way limited by the addition of silica. In some alternative embodiments, the separated LDH may be subjected to a dissolution step wherein the separated LDH is dissolved in a dissolving solvent such as an acid that results in the release of the intercalated complex anion and the metal cations from the crystal structure of the separated LDH into the dissolving solvent. Capturing the metallic species from the initial aqueous solution containing low concentrations of metallic species in the LDH material in accordance with step (b), separating the LDH material and then redissolving the separated LDH in a dissolving solvent results in obtaining a solvent with relatively higher concentration levels of the metallic species (in comparison with the low concentration levels of the initial aqueous solution). It would be appreciated that recovery of metallic species from a solution containing relatively higher concentration levels of the metallic species is more desirable and cost-effective and therefore offers the skilled person an opportunity to use conventional metallurgical recovery methods which would otherwise be ineffective for capturing trace amounts of metallic species as present in at least some embodiments of the aqueous solution utilised in step (a) of the present process. Therefore, this embodiment present a significant commercial advantage by offering a viable method or process for recovering metallic species from aqueous solutions having low concentration of the metallic species. In an alternative embodiment of the process, ion exchange of the intercalated complex anion may not be carried out. Instead, the LDH comprising the intercalated complex anion and the incorporated one or more metals and other materials obtained from step (b) may be separated and subsequently subjected to a heat treatment process as described above. Such a heat treatment process initially results in the collapsing of the LDH material resulting in the loss of the layered structural characteristics of the LDH material and subsequently results in recrystallisation of the LDH material. Specifically, the recrystallisation of the heat treated and collapsed LDH results in formation of a first oxide material comprising the metallic constituent and a second oxide material comprising one or more of the other metals. For example, the inventor has surprisingly discovered that calcination of LDH comprising intercalated uranyl complex cations and rare earth metals incorporated in the crystal structure produce a first crystalline oxide material in the form of periclase, a spinel and a third material that incorporates a proportion of uranium and other commodities such as the REE. In further embodiments, the heat treatment may be carried out under substantially reducing conditions for reducing the intercalated complex anion present within the interlayers of the LDH material obtained in step (b). For example, during heat treatment of LDH comprising intercalated uranyl complex anions, the heat treatment may be carried out under anoxic conditions (e.g. N2) or reducing (e.g. CO or C) conditions to form reduced U mineralogy, for example to produce uraninite (UO2). In some alternative embodiments, other agents may be added to form UF6 as a gas phase to assist in separation and recovery of U or specific U isotopes. In some further embodiments, the process may comprise optimisation of the crystal structure or matrix of the LDH material for selectively incorporating one or more of the other metals into the crystal structure or matrix of the LDH. For example, the optimisation may be carried out by introducing additives to the aqueous solution such as carbonates in an alkaline liquor for tuning uptake of selected or specific rare earth elements in the crystal structure or matrix of the LDH. Without being bound by theory, it is theorised that amount/speciation of bicarbonate/carbonate in the aqueous solution may potentially endow the LDH material (such as hydrotalcite) with some selectivity given the increasing affinity of mid to heavy REE for carbonates or bicarbonates. In an embodiment, the pre-determined metallic constituent comprises uranium or vanadium and wherein the one or more of the other metals comprises REE. As described earlier, the complex anion [anion] may comprise a uranyl complex anion such as but not limited to: UO2(CO3)22−, UO2(CO3)34−, CaUO2(CO3)32− UO2(SO4)34−, a vanadyl complex anion including but not limited to VO2(OH)2−, VO3OH2−, V10O286−, a chromium complex anion including but not limited to Cr2O72−, or a transuranic complex anion In at least some embodiments the pH of the solution determines speciation of the complex anion. In one embodiment the intercalated complex anion may be displaced from the interlayer of the LDH by the ion exchanging step described in the earlier section by the addition of a substituent agent such as EDTA, NTA, crown ethers, etc. In alternative embodiments, the LDH material may be subjected to the heat treatment step in accordance such that the heat treatment results in the thermal decomposition of the LDH material to recrystallise as a first crystalline oxide and a second crystalline oxide such that the uranium is incorporated in the first metal oxide and one or more of the REE is incorporated in the second crystalline oxide. Preferably the heat treatment may be carried out under substantially reducing conditions for reducing the uranyl ion from a +6 to +4 oxidation state or a mixture thereof. In some further embodiments, the LDH material may be provided in the form of a hybrid material comprising magnetic materials and LDH. Providing magnetic LDH hybrids assists in recovering the LDH material by methods such as magnetic separation. In some further embodiments, the LDH material may be provided in the form of a hybrid material comprising carbonaceous materials and LDH. Providing carbonaceous LDH hybrids assists in the co-recovery of precious metals by methods such as physi- or chemisorption/absorption. In at least some embodiments, the LDH material may be separated as a solid or the dissolved components recovered and recycled for executing the process described herein. The process described herein may be utilised as a pre-concentration step in mineral processing for elements/commodities of interest. The process may also be utilised as a method of separating elements/commodities of interest, in particular separation of a pre-determined metallic constituents as described in earlier sections from other metals and materials that may be contained in an aqueous solution or stream such as process water, leach solution (e.g. in situ or heap) or pregnant liquor. In an embodiment of at least the first aspect the step of contacting the solution with one or more additives to form layered double hydroxide (LDH) material in situ further comprises: (a) adding a magnesium and/or aluminium containing silicate material in the aqueous solution and dissolving at least a part of the silicate material in the solution thereby leaching at least a part of the magnesium and/or aluminium from the silicate material into the water; and (b) controlling reaction conditions for achieving an appropriate Mg:Al ratio in the solution for formation of the layered double hydroxide (LDH) in situ. The applicants have realised that addition of magnesium and/or aluminium containing silicates such as, for example Mg-bearing sepiolite, vermiculite, attapulgite or talc or kaolinite or natural or synthetic minerals such as zeolite that may yield magnesium and/or aluminium ions into the water when at least a part of the silicate material is subjected to the dissolving step (step [c]) of the process. Without being bound by theory, it is theorised that leaching of at least some of the magnesium and aluminium ions into the water results in the leached magnesium and/or aluminium ions being taken up for formation of the LDH material in situ. The utilisation of magnesium containing and/or aluminium containing silicate materials as a source of magnesium and/or aluminium ions for the formation of the LDH material presents several advantages. Addition of the said silicate material to the water followed by dissolving at least a part of the silicate results in leaching of at least some of the magnesium and/or aluminium ions initially present in the silicate material. Surprisingly, however, the remaining undissolved silicate material provides nucleation sites for facilitating the formation or precipitation of the LDH material in situ thereby improving yield of the LDH formed in the water. The undissolved silicate material also functions as an agent for increasing density and/or aggregate particle size of the LDH formed in situ thereby assisting in settling of the LDH in the water and/or dewatering or physical separation and recovery. Another advantage presented by the process of the present embodiment is that the undissolved silicate material may also function as an additional cation or anion exchange agent in the water. This implies that in addition to the LDH material, the undissolved silicate material may contribute towards functioning as an adsorbent for ionic species dissolved in the water. The silicate materials mentioned above may also include, but are not limited to one or more of the following: Attapulgite; Clinoptilolite; Sepiolite; Talc; Vermiculite, mineral aggregates or associations in the form of rocks (e.g. ground granite, greenstone or serpentinite), overburden, soils, sediments or waste materials, for instance from alumina refining (red mud) or coal combustion (fly ash). In at least some embodiments, undissolved silicate material from step (a) and the LDH formed in situ in step (b) form an insoluble clay material mixture wherein the clay material mixture incorporates said at least one or more dissolved cation species and/or the one or more dissolved anion species. This mixture may be also referred to a hybrid clay mixture. In one embodiment, the step of dissolving the magnesium and aluminium containing silicate material comprises leaching the magnesium and aluminium from the silicate material under acidic pH conditions. For example, the silicate material may be dissolved by way of introducing an acidic solution such as hydrochloric acid solution and/or sulphuric acid solution. The applicants have realised that acid treatment or acid leaching of silicate materials containing both aluminium and magnesium may result in the leaching or release of aluminium ions and magnesium ions from the silicate material into the water. Therefore, conducting the dissolving step under acidic conditions in at least some embodiments can result in leaching of magnesium and aluminium ions into the water. The leached ions may be utilised for tailoring the Mg:Al ratio and using the leached magnesium and aluminium ions as building blocks for the LDH formed in situ. In an alternative embodiment, the step of dissolving the magnesium and/or aluminium containing silicate material comprises leaching the magnesium and/or aluminium from the silicate material under alkaline conditions. Conducting the dissolving step for silicate materials (containing both magnesium and aluminium) under alkaline conditions results in low dissolution of magnesium and comparatively higher dissolution of aluminium ions into the water. Conducting the dissolving step under alkaline conditions in at least some embodiments can result in leaching of at least the aluminium ions into the water which may be utilised for tailoring the Mg:Al ratio and using the leached aluminium ions as building blocks for the LDH formed in situ. Furthermore, in at least some embodiments, Si may also be leached into the water as a result of the dissolving step of the process. Excessive leaching of silica can potentially occupy the interlayer anion exchange site within the LDH during formation or may combine with the leached aluminium ions to form other compounds during the LDH formation. In some embodiments, the process also comprises controlling the leaching of silica into the water. In at least some embodiments, the step of dissolving involves agitating the silicate material in the water for leaching said at least part of the magnesium and/or aluminium from the silicate material. The agitation may be carried out by way of one or more methods such as stirring and/or ultrasonication and/or any other desirable agitation means. It is also envisaged that a series of agitation steps may be utilised for agitating the silicate materials. Agitating the silicate material results in increased leaching of the magnesium and/or aluminium ions from the silicate material into the water. In at least some embodiments, the adding step comprises adding a mixture comprising the said magnesium and/or aluminium containing silicate material and an additional silicate material. It is important to appreciate that by carefully combining one or more magnesium and/or aluminium containing silicate materials in desired proportions, a desired divalent to trivalent ratio (Mg:Al) as required to form the LDH may be achieved. As a result, the required amount of additional Mg or Al to be added is reduced which can result in significant benefits. In at least some embodiments, step (c) comprises adding a mixture to the solution, the mixture comprising the silicate material and an additional material such as barren overburden or rocks or mineral processing wastes or slags In further embodiments the step of controlling the reaction conditions comprises adding at least one Mg-containing compound and/or at least one Al-containing compound for achieving the appropriate Mg:Al ratio in the water for formation of the LDH in situ. The Mg or Al dissolved in the water may comprise the leached magnesium and/or aluminium ions derived from the dissolved silicate materials and in at least some embodiments may also comprise magnesium and or aluminium ions forming a part of the dissolved cations in the water being subjected to the process described herein. This recognises that many natural or in particular wastewaters may include dissolved magnesium and or aluminium ions. Mg ions and Al ions present in the water are taken up by the formation of LDH (containing Mg and Al as the predominant metal species in the lattice structure of the LDH). Advantageously, the LDH also can take up and largely immobilise other ions into the interlayer spaces between the lattice. Thus, other ions can also be removed from the water and largely immobilised. For example, the said at least one aluminium containing compound may comprise aluminate (Al(OH)4′ or AlO2—.2H2O) or aluminium sulphate, aluminium hydroxide or organometallic compounds containing aluminium. Other inorganic compounds such as aluminium sulphate (e.g. Al2(SO4)S.18H2O), aluminium hydroxide (Al(OH)3) or organometallic compounds (e.g. aluminium acetylacetonate Cl5H21 AlO6) may also be used where a source of Al is required. Preferably these sources of Al will be alkaline to raise solution pH to an appropriate level for LDH or HT formation, but also may be used where the final solution pH or the combination of these or other compounds is alkaline. In some embodiments it may also be necessary to add additional Mg to the water in order to adjust the ratio of Al to Mg in the water to the desired level to obtain LDH or HT containing Mg and Al as predominant metal species in a lattice. This may be achieved, for example, by adding MgO or Mg(OH)2 to the water. Advantageously, MgO or Mg(OH)2 also assist in obtaining desirable pH characteristics that are suitable for the formation of LDH, such as HT. It will be appreciated that although LDH material is generally predominantly composed of Mg2+ and Al3+ cations, other divalent and trivalent metal cations, in particular Fe2+ and Fe3+ may be substituted for Mg2+ and Al3+, respectively in the LDH material. In embodiments where the aqueous solution is derived from leaching an ore, a concentrate, or a metallic constituent-bearing material or an alternative metallurgical process, the LDH material formed in accordance with the process described herein may also comprise Fe2+ and Fe3+. Accordingly, it is envisaged that in some embodiments the ratio of Al to Mg in the aqueous solution, when adjusted to the desired level to obtain LDH or HT containing Mg and Al as predominant metal species in the lattice, also takes into account the relative concentrations of Fe(II) and Fe(III) in the aqueous solution. In some embodiments it may be necessary or desirable to add additional alkaline or acid-neutralising material in addition to the at least one Mg-containing compound or the at least one Al-containing compound to the natural or wastewater. The additional alkaline or acid neutralising material may be selected from one or more of alkaline or acid-neutralising solutes, slurries or solid materials or mixtures thereof, such as lime, slaked lime, calcined magnesia, sodium hydroxide, sodium carbonate, sodium bicarbonate or sodium silicate. This list is not exhaustive and other alkaline or acid-neutralising materials may also be added. The additional alkaline or acid-neutralising material may be added before the addition of the at least one Mg-containing compound or the at least one Al-containing compound to the natural or wastewater, together with the addition of the at least one Mg-containing compound or the at least one Al-containing compound to the natural or wastewater, or after the addition of the at least one Mg-containing compound or the at least one Al-containing compound to the natural or wastewater. In some embodiments the order or sequence of addition of various alkalis or acid-neutralising materials to acid waters, wastewaters, slurries or process waters as described elsewhere in this specification may confer certain benefits. For example, the order of addition may confer geochemical and/or operational advantages to the neutralisation process and the formation of Layered Double Hydroxides (LDH) and other mineral precipitates. Selective, partial or total removal of Layered Double Hydroxide (LDH) and/or other undissolved silicate materials and/or mineral precipitates or slurry components at various stages of the reactions whether via addition of various alkalis or acid-neutralising materials to acid waters, wastewaters or process waters or via addition of acid waters, wastewaters or slurries to various alkalis or acid-neutralising materials as described elsewhere in this specification may also be considered advantageous. Such an example involves the removal of precipitates or existing solids or aggregates, mixtures or co-precipitates thereof prior to the introduction of reverse osmosis to remove some or all of remaining solutes or evaporation. This removal of Layered Double Hydroxide (LDH) and/or other mineral precipitates including the undissolved silicates at various stages of the reactions whether via addition of various alkalis or acid-neutralising materials to acid waters, wastewaters or process waters or via addition of acid waters, wastewaters or process waters to various alkalis or acid-neutralising materials as described elsewhere in this specification may be facilitated or enhanced by mechanical (e.g. centrifugation) or chemical (e.g. via addition of flocculants) means or a combination thereof In some embodiments partial or total removal of water or other solvents or miscible or immiscible solutes, such as by partial or total evaporation or distillation, may be used to increase the concentrations of one or more of dissolved, colloidal or particulate constituents or additional added constituents such as Mg and/or Al, (e.g. to tailor the appropriate Al to Mg ratio) to increase the concentration by a sufficient degree to induce the formation of LDH. At least some embodiments are also directed to water and water streams including process waters that may contain little or no Mg and/or Al or be dominated by other dissolved cations and/or anions. (e.g. such as those derived from some acid sulphate soils, industrial processes or nuclear power plants, weapons or research facilities). It is noted that not all waters (e.g. processing or wastewaters) have a major ion chemistry suitable for the formation of LDH or specific types of LDH such as Mg—Al HT or similar compositions. Thus, it may be necessary to tailor this chemistry for the formation of LDH or more specifically Mg—Al HT. The tailoring of the solution chemistry includes the step of adding the silicate material in a manner as set out in step (a) and may also addition of one or more reagents such as those containing Mg and/or Al to achieve a suitable Mg:Al ratio for promoting formation of the LDH in situ. In some embodiments, at least one of the dissolved anions in the water from a stream such as process stream may comprise a complex anion such that at least one of the complex anions is intercalated into an interlayer of the LDH formed in situ and wherein one or more dissolved cations are incorporated into the LDH material's crystal structure or matrix. Preferably, the process may further comprise the steps of controlling pH levels in the water thereby controlling speciation of the complex anion. It will be appreciated that the LDH formed in situ or added to the stream or derivatives of these LDH may also provide a substrate for a range of chemisorption or physisorption reactions that may also be used to recover one or more commodities or contaminants. Any of the features described herein can be combined in any combination with any one or more of the other features described herein are within the scope of the disclosure. Referring to the flow diagram illustrated in FIG. 1, an ore body containing a plurality of metallic constituents such as uranium and REE may be introduced to an aqueous leaching solution to obtain a pregnant leaching solution or aqueous stream. Some of the metallic constituents such as uranium may form complex anions in the solution, such as uranyl anionic complexes, as described in previous sections. Some of the other metallic constituents particularly constituents such as the REE may typically form cations in the aqueous solutions. The liquid phase of the pregnant leach solution containing the dissolved anions and cations may be separated from the undissolved solids and directed to a reaction step. The reaction step may comprise steps such as controlling the pH to determine the speciation of the uranyl complexes as illustrated in FIG. 2. The reaction step may form complex anions containing the metallic constituent (e.g. uranium). The reaction step may be followed by an LDH formation step or alternatively an LDH addition step or LDH addition and cycling of the pH to induce partial dissolution and then reformation of the LDH. In the LDH formation step, additives such as divalent additives such as MgO may be added in combination with trivalent additives such as soluble alumina salts in specific ratios and under suitable pH (alkaline pH) to promote the in-situ formation of LDH material in the solution. Such an LDH formation step also results in intercalation of the complex anion (such as the uranyl complex anion) interlayers of the LDH material formed in-situ. The metallic cations are also incorporated into the metal oxide layers of the LDH material formed in situ thereby forming a part of the crystal structure or matrix of the LDH. Such separation of the metallic species is based upon the differing uptake mechanisms for different ions provided by the LDH material formed in situ. As described earlier, the LDH formation step may be substituted or complemented by an LDH addition step in which pre-formed LDH material may be added to the solution containing the complex anions (uranyl complex anion) and the metallic cations. The step of adding pre-formed LDH material also results in intercalation of the complex anion (such as the uranyl complex anion) interlayers of the LDH material. This step may also include controlling pH so that part of the LDH may be initially dissolved at a pH of less than 9 and as low as pH 1, for specified time intervals as required to yield a sufficient degree of LDH dissolution, followed by an increase in the pH to promote reformation of the LDH material in situ. During this reformation step, other cations such as REE cations may be incorporated into the metal hydroxide layer substituting for the original cations in the initially added LDH. During this process, some anions, in particular those comprised of uranyl anionic complexes may also be substituted into the interlayer of the LDH material. Note that other techniques may also be used in the dissolution or reformation steps including (ultra) sonication or the addition of other solvents or reagents as required. The LDH material containing the intercalated complex anion and obtained from the LDH formation step or the LDH addition step may be separated by processes such as sedimentation, flocculation, filtration, cyclonic separation or other known separation methods. The separated LDH material may then be subjected to a further process for recovering the intercalated complex anion (e.g. the uranyl complex anion) such as an ion exchange process in accordance with the steps described in the preceding sections of the specification. Alternative methods of recovering the intercalated metallic constituent may also be employed in accordance with the process step detailed in the preceding sections. As discussed earlier, the recovery treatment step may not be limited to recovery of the intercalated metallic constituent such as the uranyl complex anion but may further include recovery of the metallic cations such as REE incorporated in the LDH matrix in the LDH formation step. The process described herein utilises the differing uptake mechanisms for different metallic ionic species as a way of separating the metallic species. In some embodiments, desirable separation and recovery is achieved by intercalating at least one metallic constituent in the interlayer (such as the uranyl complex anion) of the LDH (formed in-situ or added to the solution) and subsequently recovering the metallic constituent from the LDH by a further recovery step. In a first exemplary embodiment (example 1) the process may be utilised for the processing of uranium-bearing ores. It is common in uranium bearing ores that a range of other elements are present in addition to uranium. The other elements may include elements such as As, Se, Cu and the rare earth elements (REE— Ln3+ comprising La—Lu+Sc+Y). The inventor has found that REE predominantly exist as Ln3+ cations in a +3 oxidation state. Cerium exists in +3 and +4 oxidation states. Europium exists in +2 and +3 oxidation states. In the exemplary process, a uranium bearing solution derived from leaching of a uranium ore was contacted with LDH material. There are two different ways in which intercalation of the uranyl complex anion may be achieved. In a first possible way, the uranyl complex anion would readily intercalate into the interlayer of the LDH material added to the solution. However, utilising such a method does not result in uptake of the REE into the matrix or crystal structure of the LDH material added to the uranium. In a more preferred way, the LDH material added to the uranium bearing solution was dissolved in the uranium bearing solution by reducing the pH of the solution to less than 3. Reducing the pH level resulted in dissolution of the LDH material thereby resulting in the release of divalent and trivalent cations (that form the metal oxide layers of the LDH material) into the solution. After dissolving the LDH material, the pH was increased to provide alkaline reaction conditions in the solution. Providing such alkaline conditions resulted in reformation of the LDH material as a result of precipitation of the LDH material in the solution. During the reformation of the LDH material the divalent and trivalent cations that were dissolved into the solution (as a result of the initial dissolving step) precipitated to form the metal oxide layer of the reformed LDH material. During the reformation step at least some of the REE cations were also incorporated into the crystal structure of the reformed LDH material. Anionic uranyl complexes were also intercalated into the interlayer of the reformed LDH material. Importantly it has been recognised that as the divalent to trivalent ratio of metals in the primary metal hydroxide layer of the LDH may typically vary between 2:1 and 4:1, changes in this ratio may occur in the reformed LDH due to incorporation of other cations from solution that still allow a stable LDH to form. During the course of the process, the REE were shown to be strongly partitioned into the primary metal hydroxide layer of the reformed LDH material substituting for other +3 cations such as Al and Fe that were present in the initially added LDH material. Unlike the REE cations, the uranyl ion (uranium is known to exist as a UO22+ oxy-cation in solution) is considered too large to substitute for the +2 cations such as Mg2+ Alkaline earth and transition metals generally present in the metal hydroxide layers of LDH material. As shown in FIG. 2, under low pH conditions, anionic uranyl complexes are formed especially UO22+−SO4 complexes (e.g. UO2(SO4)34−). Under intermediate to higher pH UO22+−CO32− anionic complexes (e.g. UO2(CO3)22−, UO2(CO3)34−, CaUO2(CO3)32−) may predominate. Given this speciation of the UO22+ as anionic complexes, these uranyl anionic complexes preferentially partition into the anionic interlayers of LDH. As a result, the process of example 1 provides the following advantages: Valuable REE are contained within the metal hydroxide layers of the LDH Valuable U is contained as anionic complexes within the LDH interlayers. Separation of these two valuable commodities U and REE, not only from each other in terms of the way they are bound in the initial solution, but also from other components including some contaminants, salts or ions etc. that may otherwise interfere in the U or REE recovery process is highly beneficial for later separation, recovery and purification. A solid LDH is produced that typically may contain in excess of 30% U and 0-50% REE, typically 100-300 times typical ore grades of these elements thus allowing substantial enrichment of the commodities of value. Effective separation of potentially problematic ions such as Na+, Cl− and SO42− or other additives from the mineral processing stream (with the potential to make for simpler processing, further enrichment or recovery). Production of a cleaner effluent that may potentially be reused in mineral processing or other site or other operations without (or minimal) additional treatment. In addition to the above, given the different partitioning or separation of U from REE, several methods may be utilised for recovering a commodity of interest based upon the separation of commodities achieved, as elucidated above. Recovery of one or more commodities may be carried out effectively by one or more of the following further steps: the addition of a strong alkali to displace UO22+−SO4 complexes by OH− anions, or reducing the pH such that less charged or neutral UO2 complexes are displaced from the LDH interlayers. other complexing ligands or other anions (e.g. NTA, EDTA) may be added to the LDH to displace the UO2-complexes and form new NTA, EDTA complexes. addition of other chemical reagents such as phosphates, vanadates or inorganic or organic peroxides, or combinations thereof, to induce uranium precipitation. partial or complete dissolution of a U-, REE-metal-containing LDH by the addition of acid and recovery of the constituents by conventional means. addition of reducing agents, anoxia or gases (e.g. CO) to reduce uranyl complexes (U +6 oxidation state) to U (+4 oxidation state) for example as UO2 to eliminate the uranyl complexation with carbonate on the basis of charge and allow recovery of U in the +4 oxidation state. Such recovery methods may include physical (e.g. ultrasonication) or otherwise chemical (solvent-based) delamination of the LDH to recover the reduced U or the application of other physicochemical methods as required. other methods of separation that may include calcination such that with heating, typically in the range 100-1200° C., there will be layer collapse and re-crystallisation of the LDH leading to the formation of discrete or intimately associated mineral phases such as spinel and periclase. These phases, by virtue of their chemistry and crystal structure, may accommodate one of more elements of interest or may provide enhanced opportunities for recovery of particular elements given the different physicochemical properties of the mineral phases formed from calcination. The methods of stabilisation described here may also find applications in the nuclear energy or weapons industries to assist in the containment of Uranium bearing materials or wastes including transuranics or daughter radionuclides. In a second exemplary embodiment (example 2) the process may be utilised for the processing of uranium-bearing ores, in which LDH can be formed in situ within a mineral processing or metallurgical stream that includes the uranium bearing ores. The uranium ore containing stream was dosed, typically with one of or both of Mg and Al containing compounds, to achieve a desired ratio of Mg/Al in the stream which results in precipitation of LDH such as hydrotalcites. As explained in example 1, uranium bearing ores include a range of other elements that are present in addition to uranium which includes heavy metals, metalloids and/or REE. Forming the LDH material in situ also results in incorporation of the cations such as Ln3+ cations and/or Ce3+ and Ce4+ and/or Eu2+ or Eu3+ oxidation states. In situ formation of the LDH also results in REE cations being shown to be strongly partitioned into the primary metal hydroxide layer of LDH. As discussed earlier, since uranium exists as an oxy-cation commonly known as a uranyl (UO22+) cation, the uranyl ion is too large to be substituted for +2 cations such as Mg2+ into the LDH. Alkaline earth and transition metals generally present in the metal hydroxide layers of the LDH. Once again, under low pH conditions, anionic uranyl complexes are formed, especially UO22+−SO4 complexes (e.g. UO2(SO4)34−). Under intermediate to higher pH UO22+−CO32− anionic complexes (e.g. UO2(CO3)22−, UO2(CO3)34−, CaUO2(CO3)32−) may predominate. Given this speciation of the UO22+ as anionic complexes, these uranyl anionic complexes preferentially partition into the anionic interlayers of LDH formed in situ. The process described in example 2 also provides one or more of the several advantages of the process of Example 1 as summarised above. The commodities of interest may also be recovered by one or more of the further recovery steps listed under Example 1. In a third exemplary embodiment (Example 3) the process may be utilised for the processing of uranium-bearing ores, in which LDH can be formed in situ within an alkaline mineral processing or metallurgical stream that includes the uranium bearing ores. The uranium ore containing stream was dosed, typically with one of both of Mg and Al containing compounds, to achieve a desired ratio of Mg/Al in the stream which results in precipitation of LDH such as hydrotalcites. Due to the pre-existing alkaline conditions (pH of at least greater than 7 and preferably greater than 8) of the alkaline mineral processing or metallurgical stream, in situ formation of LDH is favourable when the desired ratio of Mg/Al is achieved. As explained in Example 1, uranium bearing ores include a range of other elements that are present in addition to uranium which includes heavy metals, metalloids and/or REE. Forming the LDH material in situ also results in incorporation of the cations such as Ln3+ cations and/or Ce3+ and Ce4+ and/or Eu2+ or Eu3+ oxidation states and a range of anions including oxo-metallic anions or oxyanions. Laboratory trials have demonstrated that the Al containing compound is preferably to be added first or in conjunction with any Mg containing compound to prevent the precipitation of the Mg as Mg carbonate compounds such as MgCO3 rather than it being utilised in the formation of the LDH In situ formation of the LDH also results in REE cations being shown to be strongly partitioned into the primary metal hydroxide layer of LDH. As discussed earlier, since uranium exists as an oxy-cation commonly known as a uranyl (UO22+) cation, the uranyl ion is too large to be substituted for +2 cations such as Mg2+ into the LDH. Alkaline earth and transition metals generally present in the metal hydroxide layers of the LDH. Once again, under the alkaline conditions of the stream, anionic uranyl complexes are formed. Under the intermediate to higher pH conditions of the stream, UO22+−CO32− anionic complexes (e.g. UO2(CO3)22−, UO2(CO3)34−, CaUO2(CO3)32−) may predominate. Given this selective speciation of the UO22+ as anionic complexes, these uranyl anionic complexes preferentially partition into the anionic interlayers of LDH formed in situ. It is important to appreciate that under the reaction conditions of example 3, as explained above only carbonate complexes will predominate and some REE, particularly the mid (MREE) to heavy REE (HREE) may be preferentially retained in the solution due to the known preferential complexation of MREE and HREE by carbonate ligands. This preferential speciation under alkaline conditions may be used advantageously given that the MREE and HREE are generally considered the most valuable components of the REE due to their often low abundance. In another exemplary embodiment, the step of contacting the solution with one or more additives to form layered double hydroxide (LDH) material was carried out by adding a magnesium and aluminium containing silicate material in the aqueous solution and dissolving at least a part of the silicate material in the solution thereby leaching at least a part of the magnesium and/or aluminium from the silicate material into the water; and controlling reaction conditions for achieving an appropriate Mg:Al ratio in the solution for formation of the layered double hydroxide (LDH) in situ. Raw materials, primarily Mg—Al or Al-bearing aluminosilicate clays (vermiculite, attapulgite, sepiolite, talc kaolinite) and zeolites (white and pink clinoptilolite), were procured from industrial and commercial sources. These clays and zeolites were used as sources of raw materials, principally Al and Mg, during acid and alkali dissolution experiments enhanced by the use of ultrasonication. Initial batch decomposition reactions of the aluminosilicates in both acid and alkali and with the additional use of agitation including ultrasonication were completed. Results of ICP analyses to quantify the extent of dissolution due to acid or alkali in combination with stirring (1-4 hours) or ultrasonication+stirring (1 hour) are presented in Table 1 and FIG. 3. These results indicate that substantial Mg and Al release (preferably >3:1 Mg/Al molar ratio) as required for hydrotalcite synthesis can be achieved from clays or zeolites during acid extraction. In addition, some clays such as sepiolite (Table 1) yielded both high concentrations of Mg and Al and high Mg/Al molar ratios. Under acidic conditions, all clays and zeolites demonstrated incongruent dissolution with Mg/Al and Al/Si ratios higher in the solute than the solid. In contrast, under alkali conditions any incongruent dissolution was obscured by secondary precipitation reactions. TABLE 1Geochemistry of filtered solutions produced by 1M HCl or 1M NaOHdigestion after stirring (1-4 hours) and ultrasonication + stirring (1 hour) of clay and zeolite suspensions.MgAlSiMg/Clay/zeoliteAlSiMgCaFeKNamMmMmMAlAl/SiAttapulgite sonicated8817218320652611438362.30.51 hr in 1M HClAttapulgite sonicated3117603034242020160.00.21 hr in 1M NaOHAttapulgite stirred 1272511317210283845114.61.1hr in 1M HClAttapulgite stirred 177006031225910030.00.1hr in 1M NaOHAttapulgite stirred 229341171731228825114.50.9hr in 1M HClAttapulgite stirred 288106032226200030.00.1hr in 1M NaOHAttapulgite stirred 438631251701731905123.60.6hr in 1M HClAttapulgite stirred 41010305032215190040.00.1hr in 1M NaOHClinoptilolite (pink)192743513932171141730.22.7sonicated 1 hr in 1MHClClinoptilolite (pink)5415802113254190260.00.4sonicated 1 hr in 1MNaOHClinoptilolite (pink)85161890119491310.25.4stirred 1 hr in 1MHClClinoptilolite (pink)103201907230710010.00.3stirred 1 hr in 1MNaOHClinoptilolite (pink)85181789118511310.24.9stirred 2 hr in 1MHClClinoptilolite (pink)1035017017232110010.00.3stirred 2 hr in 1MNaOHClinoptilolite (pink)107282199149631410.24.0stirred 4 hr in 1MHClClinoptilolite (pink)1341011010227190010.00.3stirred 4 hr in 1MNaOHClinoptilolite (white)3476745146131987621320.15.4sonicated 1 hr in 1MHClClinoptilolite (white)1607770101602448606280.00.2sonicated 1 hr in 1MNaOHClinoptilolite (white)1511821754157491610.29.0stirred 1 hr in 1MHClClinoptilolite (white)1398030120220800030.00.1stirred 1 hr in 1MNaOHClinoptilolite (white)1522022784162491610.27.9stirred 2 hr in 1MHClClinoptilolite (white)22130020119224960150.00.2stirred 2 hr in 1MNaOHClinoptilolite (white)1782824855163491710.26.5stirred 4 hr in 1MHClClinoptilolite (white)31182020120230540160.00.2stirred 4 hr in 1MNaOHSepiolite sonicated 1187532622675131319.80.3hr in 1M HClSepiolite stirred 1 h20105357227105151419.90.2in 1M HClSepiolite stirred 2 h231253742216134151418.40.2in 1M HClSepiolite stirred 4 h76406131226501985431419.10.2in 1M HClTalc sonicated 1 hr in1053711146573028.10.21M HClTalc stirred 1 h in 1M1516913710017.30.2HClTalc stirred 2 h in 1M1818913610016.90.2HClTalc stirred 4 h in 1M43040922720111.40.1HClVermiculite sonicated538784160519570532286620283.30.71 hr in 1M HClVermiculite sonicated21501159257330010.10.11 hr in 1M NaOHVermiculite stirred 1281371473013973102.92.1hr in 1M HClVermiculite stirred 11313137224420001.30.3hr in 1M NaOHVermiculite stirred 2848523448901701010333.11.0hr in 1M HClVermiculite stirred 21303046226360000.10.2hr in 1M NaOHVermiculite stirred 412716935949136193615563.10.8hr in 1M HClVermiculite stirred 40404055223350001.10.1hr in 1M NaOH Importantly, the dissolution of Mg and Al is substantially enhanced using the combination of ultrasonication+stirring relative to stirring alone. During an acid digest, substantial Si and other elements such as Fe and Ca may also be released depending on the chemistry and purity of the clay or zeolite. This is undesirable as excess silica can potentially occupy the interlayer anion exchange site within the LDH or HT during formation or may combine with Al to form other compounds during LDH or HT synthesis. In particular, it is desirable that the Al/Sl molar ratio is <0.5 as depicted in FIG. 3. In addition, abundant Fe may result in substitution for one or both of Mg and Al in the LDH or HT structure. If Fe is present in sufficient quantities this may lead to the formation of unstable green rusts. Alkali dissolution using either stirring or ultrasonication+stirring, as expected, yielded a substantially different solution composition with enhanced dissolution of Si over that of Al, while Mg was low as it is likely to have precipitated as brucite —Mg(OH)2. Whilst excess silica is generally undesirable in the formation of LDH or HT as described above, potential exists to use the remnant clay or zeolite after dissolution as substrates for LDH or HT nucleation. In cases where high Si is present, this may occupy at least part of the anionic interlayers of the LDH or HT structure. This property may be exploited if calcination is required to form other high temperature phases as described elsewhere. Further clay dissolution experiments were undertaken with H2SO4 in place of HCl to investigate the effects, if any, of using a different acid. These results are presented in Table 2 and illustrate that relatively less dissolved Si is produced in the presence of H2SO4 yielding lower Al/Si ratios. As outlined above, this is considered important in the synthesis of LDH or HT from solutions produced by clay or zeolite dissolution. In addition, Mg/Al ratios generally increased using H2SO4 in place of HCl. TABLE 2Ratios of concentrations of Al, Si and Mg and Mg/Al and Al/Si insolutions produced by 1M H2SO4 and 1M HCl digestion usingultrasonication + stirring (1 hour) of clay and zeolite suspensions.Clay/AlSiMgMg/AlAl/Sizeolite(H2SO4/HCl)(H2SO4/HCl)(H2SO4/HCl)(H2SO4/HCl)(H2SO4/HCl)Vermiculite sonicated 1 hr in 1M0.70.50.71.01.4HCl or H2SO4Sepiolite sonicated 1 hr in 1M2.70.73.11.23.8HCl or H2SO4Attapulgite sonicated 1 hr in 1M0.60.60.81.31.0HCl or H2SO4Kaolinite sonicated 1 hr in 1M1.51.45.03.31.1HCl or H2SO4Pink clinopt sonicated 1 hr in 1M0.40.20.41.11.9HCl or H2SO4White clinop sonicated 1 hr in0.70.60.71.01.11M HCl or H2SO4 On the basis of the above dissolution experiments and supplementary experiments using H2SO4 in place of HCl, synthesis of the nano-hybrid materials was undertaken using a range of clay and zeolite. In addition, aluminate was also used as both a source of additional Al and as a neutralising agent. A list of the nano-hybrid material produced and their P-uptake capacity is given in Table 3. TABLE 3Phosphorus uptake capacity of a range of clay/zeolite nano-hybridmaterials synthesised in this study.P-uptakeClay/zeolitemg/gUnground vermiculite + white clinoptilolite4.6ALL SOLIDSUnground Vermiculite + aluminate ALL5.4SOLIDSSepiolite + aluminate ALL SOLIDS9.0Unground vermiculite/white clinoptilolite11.7NO SOLIDSSepiolite/white clinoptilolite ALL SOLIDS13.2Sepiolite + white clinoptilolite ALL13.5SOLIDSVermiculite + white clinoptilolite ALL13.7SOLIDSVermiculite + aluminate ALL SOLIDS14.1Unground vermiculite/white clinoptilolite14.3ALL SOLIDSVermiculite/white clinoptilolite ALL14.5SOLIDSUnground vermiculite + aluminate NO15.4SOLIDSVermiculite/white clinoptilolite NO17.5SOLIDSVermiculite + aluminate NO SOLIDS19.8Sepiolite/white clinoptilolite NO SOLIDS28.4Sepiolite + aluminate NO SOLIDS41.7 Mixing ratios of solutions both with and without residual clay or zeolite solids present were determined using the equation:v1/v2=(r[Mg]2−[Al]2)/([Al]1−r[Mg]1)where v1 and v2 are the volume ratio of the two clay or zeolite solutions required to give r which is the required Mg:Al ratio in the final solution (in this case 3), and [Mg]1, [Mg]2 and [Al]1 and [Al]2 are the concentrations of Mg and Al in solutions 1 and 2, respectively. Where aluminate was added, target Mg/Al molar ratios of 3 were calculated. Mineralogical (XRD) analysis of the nano-hybrid materials indicated the presence of hydrotalcite in addition to the residual clay or zeolite mineral which acted as a scaffold for hydrotalcite nucleation and precipitation is depicted in FIG. 4. The significance of the examples presented here is that a new class of material has been synthesised using a novel preparation method utilising elements contained within commercial clays to produce nano-hybrids which contain LDH in the form of HT grafted onto the original clay or zeolite substrate. The beneficiation process adds significant utility and value to commercially-mined clays and zeolites as demonstrated by the high P-uptake (as phosphate) achieved as depicted in FIG. 5. The high P-uptake demonstrates that other simple or complex anions, for instance uranyl-carbonate complexes, may also be removed from solution using these materials. Four samples of hydrotalcite were calcined by heating the sample up to 1350° C. with a Pt strip heater. FIGS. 6-7 depict the decomposition of the hydrotalcite samples as they undergo calcination alone or in the presence of crystalline silica (quartz), amorphous silica, and with interlayer silica, and the progressive formation of spinel (Al silicate) and periclase (Mg silicate) phases with increasing temperature. FIG. 6 shows that hydrotalcite decomposes to a dehydrated hydrotalcite form between 330-350° C. A periclase phase begins to form between about 450-550° C. with spinel forming at around 850° C. FIG. 7 shows that hydrotalcite in the presence of quartz decomposes to a dehydrated hydrotalcite form between 310-355° C. Quartz alpha to beta phase transformation is indicated as forming at about 550° C.; a periclase phase begins to form between about 750-800° C. with spinel forming at around 1200° C. Forsterite forms at about 1300° C. corresponding to the disappearance of quartz. FIG. 8 shows that hydrotalcite in the presence of amorphous silica decomposes to a dehydrated hydrotalcite form between 375-425° C. A periclase phase begins to form between about 800-850° C. with spinel forming at around 1100° C. Forsterite forms at about 1210° C. FIG. 9 shows that hydrotalcite in the presence of interlayer silica decomposes to a dehydrated hydrotalcite form between 310-340° C. A periclase phase begins to form between about 400° C. with spinel and forsterite forming at around 495° C. The quantitative extent of decomposition of hydrotalcite and formation of other minerals as a function of temperature in the presence of quartz for different quartz: hydrotalcite ratios were also examined by heating the sample up to 1350° C. with a Pt strip heater. FIG. 10 shows the decomposition of hydrotalcite with quartz:hydrotalcite ratio of 1:1 and FIG. 11 shows the decomposition of hydrotalcite with quartz:hydrotalcite ratio of 3:1. Calcination of hydrotalcite results in the mineral formation of spinel and periclase as well as element segregation. The back scatter SEM image of the bright (high atomic mass) discrete U-bearing grain is shown in FIG. 12 indicating that there is migration of U and some other elements into discrete phases during calcination. In view of the mineral/elemental segregation, it may be possible to selectively leach the calcined hydrotalcite to remove U or, alternatively, to crush the calcined hydrotalcite and employ flotation or heavy mineral separation techniques to remove and recover U. The type of ambient atmosphere used to form the hydrotalcite also has an effect on the elemental uptake of U and REE in the hydrotalcite, as well as mineral segregation post-calcination. In Table 4 below is an example of the elemental uptake when the hydrotalcite was formed under an inert (e.g. nitrogen (N2)) or a reducing (e.g. carbon dioxide (CO2)) atmosphere. TABLE 4ElementMean N2Mean CO2Al1.6810.10Mg7.5620.34U18.287.73Fe2.318.07Th16.851.09Si0.513.22Y2.750.46Ca0.299.49Ce32.755.42O18.7634.39Total101.56100.31 Table 5 below shows the selective separation of U from a synthetic raffinate containing uranium, and two rare earth elements, lanthanide and yttrium by formation of a Fe(II)/Fe(III) LDH material. The concentration of Al, Mg, Fe, U, La and Y in the synthetic raffinate is shown in the second column of Table 5. The concentration of Al, Mg, Fe, U, La and Y remaining in the solution following formation of the Fe(II)/Fe(III) LDH material is shown in the third column of Table 5. Additional Mg(II) in the form of MgCl2.6H2O was added to adjust the M2+:M3+ cation ratio to about 2.5 so as to cause formation of Fe(II)/Fe(III) LDH material. The precipitate so formed was a blue-green colour which is characteristic of Fe(II)/Fe(III) LDH material containing mixed Fe valency. Formation of the Fe(II)/Fe(III) LDH material results in uptake of substantially all of the uranium, lanthanide and yttrium from the synthetic raffinate. TABLE 5SampleSynthetic raffinateFe(II)/Fe(III) LDHAl3183<0.1Mg9191.7Fe154991.7U183<0.5La73.70.1Y25.9<0.1 In the present specification and claims (if any), the word ‘comprising’ and its derivatives including ‘comprises’ and ‘comprise’ include each of the stated integers but does not exclude the inclusion of one or more further integers. Reference throughout this specification to ‘one embodiment’ or ‘an embodiment’ means that a particular feature, structure, or characteristic described in connection with the embodiment is included in at least one embodiment. Thus, the appearance of the phrases ‘in one embodiment’ or ‘in an embodiment’ in various places throughout this specification are not necessarily all referring to the same embodiment. Furthermore, the particular features, structures, or characteristics may be combined in any suitable manner in one or more combinations. It is to be understood that the embodiments are not limited to specific features shown or described since the means herein described comprises preferred forms of putting the embodiments into effect. The embodiments are, therefore, claimed in any of its forms or modifications within the proper scope of the appended claims (if any) appropriately interpreted by those skilled in the art.
050842315	abstract	A refueling mast for a reactor complex includes four generally cylindrical tubes. Each inner tube has vertical grooved tracks formed therein. Each outer tube has a guide roller mounted thereon with grooves which mate with track grooves of a respective track on the adjacent inner tube. Track grooves are cold formed moving a roller die tool up and down each inner tube, while increasing pressure on the incorporated die rollers. This process flattens the inner tube where the tracks are being formed. The grooved tracks and the associated flattening provide the torsional rigidity required on the mast tubes for precise positioning and orientation of fuel elements.
	abstract	Provided is an X-ray analysis apparatus including: a goniometer including an incident-side arm extending in a first direction, a fixing portion, and a receiving-side arm; an X-ray source portion, which is arranged on the incident-side arm and generates an X-ray source extending in a second direction, which crosses the first direction; a support base, which is arranged on the fixing portion, and is configured to support a sample; a parallel slit, which is arranged on the fixing portion, and is configured to limit a line width along the second direction of the X-ray source generated by the X-ray source portion; and a detector, which is arranged on the receiving-side arm, and is configured to detect a scattered X-ray generated by the sample.
	abstract	Irradiation device for proton and/or ion beam therapy, said device comprising a radiation source, a beam guiding device, and a therapy room comprising a treatment site and an access, wherein the therapy room is arranged in a first plane, and the treatment beam is directed into the therapy room from a second plane above or below the first plane, and oriented towards the treatment site so that the treatment beam is directed away from the access. In the therapy room, a shielding is provided, which is open towards the treatment site and associated with the entrance region of the treatment beam into the therapy room so that the access is arranged on the side of the shielding opposing the treatment site, and a labyrinth leading from the access to the treatment site is provided laterally offset to the treatment beam proceeding in the therapy room and to the shielding.
039719553	claims	1. A shielding container for radiopharmaceuticals comprising a cover and a cylindrical body wherein said cover includes a disk shaped lid having a depending member and cap cover, said cylindrical body is formed with a squared off well and an outer recess communicating with said well, a sealed bottle containing radiopharmaceutical material having a retaining ring located around its neck is located in said well, said cap cover contacting the bottle cap, and a pad of absorbent, flexible, resilient material located around said depending member and filling said outer recess. 2. The container of claim 1 wherein said outer recess is cylindrical and said pad is cylindrical with an inner hole and split so as to enable the pad to be placed around the depending member. 3. The container of claim 2 wherein said disk shaped lid and said cylindrical body both have an outwardly extending flange and said container is sealed in the area where said flanges meet. 4. The container of claim 3 wherein said cover includes a second inner disk concentric but of smaller diameter than the first dimensioned to sit within the container body and wherein said depending member is attached to said second disk. 5. The container of claim 4 wherein both of said lid disks and said cover flange are formed as a single unit and said depending member and cap cover are formed as a second unit and the smaller disk includes a centrally located hole into which the top of the depending member is force fit. 6. The container of claim 5 wherein said lid disks and cover flange and said cylindrical container body and body flange are formed of lead, said depending member and cap cover are formed of rigid polymeric material, and said retaining ring is formed of semi-rigid deformable polymeric material. 7. The container of claim 6 wherein said absorbent pad is formed of polyurethane foam. 8. The container of claim 7 wherein said bottle is sealed by a screw type cap, said bottle cap and said cap cover are both of the same polygonal shape, and said cap cover is dimensioned to engage said bottle cap so that rotation of the container cover will in turn rotate and unlock the cap and the cap will be lifted and removed by lifting the container cover. 9. The container of claim 8 wherein said screw type cap and said cap cover are both octagonal. 10. The container of claim 7 wherein said bottle is sealed by a non-rotatable pierceable cap, grooves are located in the side walls of said well and the retaining ring locks into said grooves, and said cap cover is dimensioned to fit loosely over said pierceable cap.
	claims	1. A method for imaging a tumor of a patient, comprising the steps of:delivering a set of n protons from a synchrotron: through a beam transport system exit nozzle, through a proton radial cross-section beam expander, through a first prior imaging sheet, through a second prior imaging sheet, through a patient position, through at least one posterior imaging sheet, and into a scintillation material of a beam energy scintillation detector system, said first prior imaging sheet positioned between said proton radial cross-section beam expander and the patient position, said second prior imaging sheet positioned between said proton radial cross-section beam expander and the patient position, wherein the patient occupies the patient position during use, wherein n comprises a positive integer of at least two;simultaneously detecting spatially resolved prior position photon emissions, resultant from passage of the set of n protons;determining a set of prior proton vectors entering the patient using the spatially resolved prior position photon emissions;simultaneously detecting spatially resolved posterior position photon emissions, resultant from passage of the set of protons;determining a set of posterior proton vectors exiting the patient using the spatially resolved posterior position photon emissions;determining a set of n probable proton paths through the patient using spatial correlations of entry points of the set of prior proton vectors into the patient and exit points of the set of posterior proton vectors from the patient; andgenerating an image of the patient using the n probable proton paths. 2. The method of claim 1, said step of simultaneously detecting spatially resolved prior position photon emissions further comprising the steps of:detecting first simultaneous photon emission positions resultant from: (1) a first proton, of said set of n protons, transmitting through said first sheet and (2) a second proton, of said set of n protons, transmitting through said first sheet;detecting second simultaneous photon emission positions resultant from: (1) the first proton transmitting through said second sheet and (2) the second proton transmitting through said second sheet,wherein a first proton vector, of said set of prior proton vectors, passes through a first member of said first simultaneous photon emission positions and a first member of said second simultaneous photon emission positions,wherein a second proton vector, of said set of prior proton vectors, passes through a second member of said first simultaneous photon emission positions and a second member of said second simultaneous photon emission positions. 3. The method of claim 1, said step of simultaneously detecting spatially resolved prior position photon emissions further comprising the steps of:simultaneously detecting a first set of n photon emission positions, resultant from passage of the set of n protons, from said first sheet;simultaneously detecting a second set of n photon emission positions, resultant from passage of the set of n protons, from said second sheet,wherein individual prior vectors of said set of prior proton vectors pass through respective members of: (1) the first set of n photon emission positions on said first sheet and (2) the second set of n photon emission positions on said second sheet, wherein n comprises a positive integer of at least five. 4. The method of claim 3, said step of simultaneously detecting spatially resolved posterior position photon emissions further comprising the steps of:simultaneously detecting a third set of n photon emission positions, resultant from passage of the set of n protons, from said third sheet;simultaneously detecting a fourth set of n photon emission positions, resultant from passage of the set of n protons, from said fourth sheet,wherein individual posterior vectors of said set of posterior proton vectors pass through respective members of: (1) the third set of n photon emission positions on said third sheet and (2) the fourth set of n photon emission positions on said fourth sheet. 5. The method of claim 3, said proton radial cross-section beam expander comprising a set of atoms, wherein at least twenty percent of said set of atoms comprise a form of hydrogen. 6. The method of claim 5, said step of simultaneously detecting spatially resolved prior position photon emissions occurring on a time scale less than a fifty percent decay in flux of the spatially resolved posterior position photon emissions. 7. The method of claim 5, said step of generating an image of the patient further comprising the step of:using a residual energy, of each of the set of n protons, determined using spatially resolved photon output positions from said scintillation material of said beam energy scintillation detector system associated with each of said n protons. 8. The method of claim 1, said step of simultaneously detecting spatially resolved prior position photon emissions occurring on a time scale less than a fifty percent decay in flux of the spatially resolved posterior position photon emissions. 9. The method of claim 8, further comprising the step of:moving said beam transport system exit nozzle and said scintillation material relative to the patient;repeating said steps of: (1) simultaneously detecting spatially resolved prior position photon emissions and (2) simultaneously detecting spatially resolved posterior position photon emissions; andgenerating a three-dimensional tomographic image of the patient through multiple uses of said step of moving and said step of repeating. 10. The method of claim 8, further comprising the step of:determining relative positions of said beam transport system exit nozzle, the patient, and said scintillation material using a set of fiducial indicators, said set of fiducial indicators comprising: (1) at least two fiducial markers and (2) at least one fiducial detector configured to detect photons from positions of said at least two fiducial markers. 11. The method of claim 8, further comprising the step of:using fiducial indicators positioned on at least said beam transport system exit nozzle in targeting the patient with said set of n protons. 12. The method of claim 1, wherein said proton radial cross-section beam expander comprises a diffusing element. 13. The method of claim 1, wherein said proton radial cross-section beam expander comprises a plastic. 14. An apparatus for imaging a tumor of a patient, comprising:an imaging system, comprising:a beam transport system exit nozzle configured to simultaneously deliver a set of n protons from a synchrotron, wherein n comprises a positive integer of at least two;a proton beam radial cross-section expander configured to radially expand a beam of the set of n protons;a patient position, the patient occupying the patient position during use;a first prior imaging sheet, said first prior imaging sheet positioned between said proton beam radial cross-section expander and the patient position;a second prior imaging sheet, said second prior imaging sheet positioned between said proton beam radial cross-section beam expander and the patient position;at least one posterior imaging sheet; anda scintillation material of a scintillation detector system,said imaging system configured to:simultaneously detect spatially resolved prior position photon emissions, resultant from passage of the set of n protons;determine a set of prior proton vectors entering the patient using the spatially resolved prior position photon emissions;simultaneously detect spatially resolved posterior position photon emissions, resultant from passage of the set of protons;determine a set of posterior proton vectors exiting the patient using the spatially resolved posterior position photon emissions;determine a set of n probable proton paths through the patient using spatial correlations of entry points of the set of prior proton vectors into the patient and exit points of the set of posterior proton vectors from the patient; andgenerate an image of the patient using the n probable proton paths and output of said scintillation detector system. 15. The apparatus of claim 14, further comprising:a set of fiducial indicators, said set of fiducial indicators comprising: (1) at least two fiducial markers and (2) at least one fiducial detector configured to detect photons from positions of said at least two fiducial markers, said imaging system configured to determine relative positions of said beam transport system exit nozzle, the patient, and said scintillation material during use. 16. The apparatus of claim 14, said proton beam radial cross-section expander further comprising:a pathway for the set of n protons comprising a set of atoms, at least thirty percent of said set of atoms, by number, comprising a form of hydrogen.
	claims	1. A collector comprising:a first mirror shell positioned inside a second mirror shell that has a chamfered end. 2. The collector of claim 1, wherein said collector receives light from a light source and reflects said light to illuminate an area in a plane. 3. The collector of claim 2, wherein said chamfered end receives said light from said light source. 4. The collector of claim 1,wherein said second mirror shell includes a portion that receives light and reflects said light, andwherein said chamfered end receives said light from said portion, and further reflects said light. 5. The collector of claim 1, wherein said second mirror shell has a thickness that decreases in a direction of said chamfered end. 6. The collector of claim 1,wherein said first mirror shell receives light from a light source at an angle of incidence of less than 20° to a surface tangent of said first mirror shell, andwherein said second mirror shell receives light from said light source at an angle of incidence of less than 20° to a surface tangent of said second mirror shell. 7. The collector of claim 1, wherein at least one of said first and second mirror shells includes a first segment having a first optical surface and second segment having a second optical surface. 8. The collector of claim 7,wherein said first segment is annular in a section of a hyperboloid, andwherein said second segment is annular in a section of an ellipsoid. 9. The collector of claim 7,wherein said first segment is annular in a section of a hyperboloid, andwherein said second segment is annular in a section of a paraboloid. 10. An illumination system comprising:a light source;a plane; anda collector that receives light from said light source, and illuminates an area in said plane,wherein said collector has a first mirror shell positioned inside a second mirror shell that has a chamfered end. 11. The illumination system of claim 10, further comprising an optical element having a plurality of raster elements in a light path from said light source to said plane. 12. The illumination system of claim 11,wherein said collector illuminates an annular region in said plane, andwherein said plurality of raster elements are in said annular region. 13. An EUV-projection exposure facility comprising:the illumination system of claim 10 for illuminating a mask; anda projection objective for imaging said mask onto a light sensitive object. 14. A method of manufacturing a micro-electronic component, comprising using the EUV-projection exposure facility of claim 13.
058928065	claims	1. A spacer for maintaining an inner tube in spaced relation within an outer tube in a nuclear reactor, said spacer comprising: a split ring adapted to be disposed about the outer surface of said inner tube, said ring having a central annular body portion with a raised annular bearing surface thereon adapted to contact the inner surface of said outer tube and prevent contact between said outer surface and said inner surface, an annular land projecting from each side of said central body portion, and a transverse split across said central annular body portion and said lands ;and a collar adapted to be received on one of said lands and effective to create an interference fit between said ring and said inner tube and thereby constrain axial movement of said spacer on said inner tube. a split ring adapted to be disposed about the outer surface of said pressure tube, said ring having a central annular body portion with a raised annular bearing surface thereon adapted to contact the inner surface of said calandria tube and prevent contact between said outer surface and said inner surface, an annular land projecting from each side of said central body portion, and a transverse split across said central annular body portion and said lands; and a collar adapted to be received on one of said lands and effective to create an interference fit between said ring and said pressure tube and thereby constrain axial movement of said spacer on said pressure tube. 2. The spacer of claim 1 wherein each said land has a groove or ridge to retain said collar thereon. 3. The spacer of claim 1 wherein said split ring has a concavity formed in the central annular body portion, said concavity forming an annular void space between said raised bearing surface and said inner tube. 4. The spacer of claim 1 wherein said bearing surface has a coating to reduce heat transfer between said ring and said outer tube. 5. The spacer of claim 1 wherein said bearing surface has a coating to reduce wear between said ring and said outer tube. 6. A spacer for maintaining a pressure tube in spaced relation with a calandria tube of a nuclear reactor, said spacer comprising: 7. The spacer of claim 6 wherein each of said lands has a groove or ridge to retain said collar thereon. 8. The spacer of claim 6 wherein said split ring has a concavity formed in the central annular body portion, said concavity forming an annular void space between said raised bearing surface and said pressure tube. 9. The spacer of claim 6, 7 or 8 wherein said collar is formed of material having substantially the same coefficient of diametrical creep as said pressure tube. 10. The spacer of claim 6, 7 or 8 wherein said collar is formed of the same material as said pressure tube. 11. The spacer of claim 10 wherein said collar is formed of zirconium alloy. 12. The spacer of claim 6, 7 or 8 wherein said collar and said split ring are formed of the same material as said pressure tube. 13. The spacer of claim 12 wherein said collar and said split ring are formed of zirconium alloy. 14. The spacer of claim 6, 7 or 8 where said bearing surface has a coating to reduce heat transfer between said ring and said calandria tube. 15. The spacer of claim 6, 7 or 8 where said bearing surface has a coating to reduce wear between said ring and said calandria tube. 16. The spacer of claim 14 wherein said coating is zirconium oxide.
	description	This application is a continuation of the Paris Convention Treaty (PCT) Application PCT/DE2004/000955 filed on May 7, 2004, designating the United States and published in German, which is hereby incorporated by reference. The present embodiments relate, generally, to radiation grids, and particularly, to a scattered radiation grid or a collimator that absorbs secondary radiation scattered by an object, including a support or substrate with a plurality of spaced-apart absorbing elements. In radiology, stringent demands are made or imposed on the quality of images. For radiology images made for radiological medical diagnosis for example, X-radiation from a virtually punctate X-ray source is passed through an object to be examined, and a distribution of an attenuation of the X-radiation is detected two-dimensionally on a side of the object diametrically opposite the X-ray source. In computed tomography, line-by-line detection of the X-radiation attenuated by the object is made. Solid-state detectors are increasingly used as radiation detectors. These solid-state detectors have a matrix like array of semiconductor elements that act or operate as receivers. The X-ray image or projection achieved or made is composed of a plurality of individual pixels, and ideally, the attenuation of the radiation through the object along a straight axis or path from the X-ray source to a location on the detector surface corresponds to each of the plurality of pixels. The radiation that strikes the detector along this straight axis is referred to as a primary radiation. However, during the passage of the X radiation through the object, interactions necessarily occur between the X-ray beams and the object, which leads to scattering effects. That is, besides the primary beams, which pass un-scattered through the object, secondary beams also occur, which strike the detector having deviated from their respective rectilinear axis or path. These secondary beams, which can make up a substantially high proportion of an entire signal modulation of the detector, are an additional source of noise and reduce a capability of detecting finely contrasting image distinctions. For reducing the scattered radiation striking the detector, it is known to employ scattered radiation grids. Known scattered radiation grids comprise regularly arranged structures which absorb X-radiation, and between which through conduits (channels or ducts), or the like, for primary radiation are provided. A distinction is made between focused grids and unfocused grids. In focused grids, the through conduits and thus the absorption structures that determine them are aligned with the focus of the X-ray source, in contrast to unfocused grids, in which the conduits are perpendicular to the detector surface. A mode of operation of a scattered radiation grid is such that primarily the secondary radiation, and in unfocused grids also part of the primary radiation, are absorbed via the absorbing structures, and thus do not strike the detector and do not contribute to the proportion of radiation that generates the X-ray image. On one hand, the scattered beams should be maximally absorbed, yet on the other hand, a maximal proportion of primary radiation should pass un-attenuated through the grid. Reducing the proportion of scattered radiation can be achieved via a substantially high shaft ratio of the conduits. This high ratio is between a height of the grid and a thickness or diameter of the through conduits. However, due to the thickness of the absorbing elements located between the conduits, image distortion can occur from absorption of part of the primary radiation. When the grid is used in conjunction with a matrix detector, a discontinuity in the grid causes image distortion because of the projection of the grid in the X-ray image. A potential risk is that the projection of the detector element structures and the scattered radiation grid may interfere with one another, which may lead to an occurrence of interfering moiré effects. The above discussed grid problems or issues were also described in German Patent Application DE 102 41 424.6, which was published after the priority date of the present application. In this German patent application document, a novel type of grid is described in comparison with the conventional lead lamination grids. Conventional lead lamination grids are referred to as “placed grids.” Thin lead laminations and elements, which are usually made of radio-transparent paper to form the through slits between the laminations, are placed alternatingly. However, these placed grids are limited in terms of production and manufacture and may lead to problems, such as in solid-state detectors. The grid of DE 102 41 426 is different, since being produced via a rapid prototyping technique or method using a layer-wise solidification of a buildup material. With this technique, substantially fine and exact structures can be built up, which are used for the configuration of the absorption structure. The absorption structure thus manufactured is then coated, both on the inside faces of the through conduits provided in the structure and on the diametrically opposite surfaces, with a substantially high absorbent material, and the surface coating is either reduced substantially or removed entirely in a post-treatment step or act. Although with this grid, the detectability of grid projections can be reduced and shifted into a substantially high location frequency range so that they cannot be sharply projected by the imaging systems. These grids may be expensive to manufacture, and may make stringent technical demands in terms of the course or process of manufacture. This is applicable when removing the coating from the face ends of the structure produced by stereo-lithography, which during the removal process itself may not be affected. However, a homogeneous reduction in the layer thickness or a substantially homogeneously complete removal may be necessary, so that a locally varying absorption behavior may not occur. Moreover, the coating of the insides of the through conduits needs to be or remain unaffected. Similar problems to those in radiological diagnosis also occur in nuclear medicine, when gamma scanners or cameras are used for example. There again, care is taken such that a minimal amount of scattered gamma quanta may reach the detector. In this type of examination, the X-ray source for the gamma quanta is located in the interior of the object being examined. After an unstable nuclide has been injected, an image of an organ is generated by the detection of the quanta emitted from the body because of the decomposition of the nuclide. The course of the activity or decomposition in the organ over time allows conclusions to be drawn about a function of that organ. In this technique, as in a scattered radiation grid, a collimator is placed in front of the gamma detector and the collimator determines the projection direction of the image. In operation and construction, this collimator may be similar to the scattered radiation grid described at the outset. The present embodiments are defined by the appended claims. This summary describes some aspects of the present embodiments and should not be used to limit the claims. A scattered radiation grid or collimator may be relatively simple to produce or manufacture. In the provided scattered radiation grid or collimator, absorbing elements are embodied in the form of small tubes or pins and are fixed on plug-in fixtures or clamping fixtures that are provided on a support. In one embodiment of the scattered radiation grid or collimator, a mechanical fixation of the absorbing elements on the plug-in fixtures or clamping fixtures provided on the support is achieved. That is, the absorbing elements are mounted firmly or clampingly fixed on these plug-in fixtures or clamping fixtures. The absorbing elements are embodied in the form of small tubes or as pins, and the plug-in fixtures or clamping fixtures are configured to suit the embodiment of the absorbing elements. Since the absorbing elements are prefabricated parts, which may need not be further machined and which may intrinsically have their own absorption properties, the manufacture of the scattered radiation grid or collimator can proceed markedly more simply, as the support, after the mechanical fixation of the absorbing elements, may not need to be post-machined for the sake of the absorption properties. Various types of absorbing elements can be used. In one embodiment, all of the absorbing elements may comprise an absorbent material. The absorbing elements are shaped in the form of small metal tubes or metal pins. In an alternate embodiment, each of the absorbing elements has a support element, which comprises a radio-transparent material and is coated on at least one side face with a coating of an absorbent material. These absorbing elements accordingly comprise different materials, namely first a material of the support and second the coating material. However, since the absorbing elements are prefabricated parts, no further provisions may need to be made after the absorbing elements are placed on the support; that is, the absorbing elements are used in their prefabricated form. When the absorbing element has a tubular form or shape, a support element that is likewise in the form of a small tube and thus hollow on the inside can be coated on the inner and/or outer side face; that is, either one or two coating faces may be provided. The face ends, however, are not coated. Alternatively to the use of the tubular absorbing elements, as described pin-like absorbing elements may be used, which for a two-component structure have the support element that is coated only on its outside. The tubular absorbing elements may be embodied with or have various cross sections. These cross sections may have hollow cylindrical, or hollow polygonal outer and/or inner shape. Numerous potential shapes are conceivable, even in mixed forms; that is, the outer shape may be cylindrical while the inner shape may be polygonal, and vice versa. Correspondingly, pin-like absorbing elements may also have cylindrical and/or polygonal cross sectional shapes. The absorbing elements are appropriately made from elongated prefabricated wires or small tubes from which they are suitably cut to length. In absorbing elements that entirely comprise absorbent material, the long wire or small tube is a metal wire or small metal tube. While in the multi-component absorbing elements, a corresponding nonmetallic wire or a corresponding small tube is provided with the absorbent coating on the inside and/or outside of the absorbing elements. An absorbing element has a length of 1 mm to 10 mm, 2 mm to 6 mm, or 2 mm to 3 mm. This is applicable for both the small tubes and the pin-like absorbing elements. The outer diameter is between 0.3 mm to 2 mm, or between 0.5 mm to 1 mm, which likewise applies to both types of absorbing elements. For tubular absorbing elements, the wall thickness is between 20 μm to 50 μm, and for two-component elements, this figure describes the entire wall thickness comprising both the support element and the inner and/or outer coating. Other larger or smaller dimensions may be provided. As described, the absorbing elements are mechanically fixed via the plug-in fixtures or clamping fixtures provided on the support. The plug-in fixtures or clamping fixtures may either protrude from the plane of the support or alternatively be molded into the plane of the support. In terms of the embodiment of the plug-in fixtures or clamping fixtures and the fixation of absorbing elements, different designs may be conceivable—depending on the type of absorbing element used. The tubular absorbing elements can be mounted on the plug-in fixtures or clamping fixtures that engage the interior of an absorbing element. The diameter and shape of a plug-in fixture or clamping fixture may correspond to the diameter and shape of the through conduit of an absorbing element, so that the absorbing element can be mounted or clamped onto the plug-in fixture or clamping fixture. In other words, the plug-in fixture or clamping fixture may engage the interior of the absorbing element, and the diameter or shape of the fixture is selected such that a secure mechanical hold is assured, and yet the mounting or assembling process can be effortless. Alternatively to the placement on the fixture, tubular or pin-like absorbing elements can be received between at least two, or four, plug-in fixtures or clamping fixtures that engage the outside. That is, the absorbing elements are clamped in place between the fixtures. In a further alternative, the fixation of the tubular or pin-like absorbing elements in plug-in fixtures or clamping fixtures are embodied as indentations or as through holes that correspond to the outer shape of the absorbing elements. That is, the absorbing elements are placed in pre-shaped recesses or holes in the support and are retained therein. Because of the radio-transparency of the support and hence also of the plug-in fixtures or clamping fixtures integrally formed onto it, the plug-in fixtures or clamping fixtures may correspond to the length of the absorbing elements, so that the absorbing elements—in whatever way—may be received quasi-entirely on the support. Alternatively, the plug-in fixtures or clamping fixtures may be shorter than the absorbing elements, or at most half as long as the absorbing elements, which economizes on support material. In an advantageous aspect, the plug-in fixtures or clamping fixtures may be disposed such that the absorbing elements are received while aligned with a focus of the radiation source. By suitable disposition or embodiment of the plug-in fixtures or clamping fixtures, focusing may be achieved even in this “plug-in or clamping grid or clamping collimator.” The support may be radio-transparent plastic and may be produced by stereo-lithography by the substantially rapid prototyping technique. In this regard, see DE 102 41 424.6 or related U.S. Pat. No. 6,968,041 (Publication No. 2004/0156479 Ser. No. 10/772,471)), the disclosure of which is incorporated herein by reference, already mentioned, in which the production of a support by this technique is described. In such a method, whatever previously described structure of the individual layers of a three-dimensional volumetric model of the support has been made is “inscribed” in a liquid polymer resin using a UV laser beam under computer control. By the action of the laser, the polymer resin hardens at the points or surfaces exposed to light. Once the first structure plane is “inscribed”, the construction platform on which the structure is built up is sensibly lowered, after which a new resin layer is applied, and the second structure plane is “inscribed.” This process is repeated until the desired structure is achieved. By using this technique, arbitrarily configured support structures can be generated. The use of a structure produced by stereo-lithography by the rapid prototyping technique has manifold advantages. First, by this technique, the support in terms of its surface structure can be produced substantially exactly and with a substantially precise shape along with the plug-in fixtures or clamping fixtures embodied there, which is practical with respect to the mechanical mounting of the absorbing elements. In another advantageous aspect, the location or disposition of the plug-in fixtures or clamping fixtures can be relatively simply varied with respect to the targeted focusing of the absorbing elements over the support plane. The plug-in fixtures or clamping fixtures themselves—since the absorbing elements are seated vertically on or in them or are parallel to them—are necessarily also focused. This “focusing” can be done substantially exactly, as described for the stereo-lithography method. Moreover, the absorbing elements may be potted with a radio-transparent potting composition, such as an X-ray transparent plastic or a casting resin, in order to lend the structure improved stability. The absorbing elements, and their coatings, may be of various absorption materials. For instance, W, Ta, Mo, Cu, Ni, Co, Fe, Mn, Cr, and V can be named, along with all the absorbent alloys that can be made from them, among others. In another aspect, a method for producing a scattered radiation grid or a collimator, including a support having a plurality of spaced apart absorbing elements, is provided. In this method, via an automatic positioning mechanism, the tubular or pin-like absorbing elements are secured to plug-in fixtures or clamping fixtures provided on the support. Due to the fact that the absorbing elements are substantially thin, and given the surface area of a scattered radiation grid or collimator, which is for example 40×40 cm, up to several hundred thousand absorbing elements may be put in place, and an automatic positioning mechanism is therefore expediently used that mounts the absorbing elements on the fixtures or clamps them between them. Via the positioning mechanism, the absorbing elements can be placed individually, or secured as a plurality of absorbing elements simultaneously. After the positioning of the absorbing elements, they are embedded in a position-fixing way via a potting composition. In the following detailed description of the drawings, illustrative and exemplary embodiments that are not to be understood as limiting are described and discussed along with their characteristics in further detail below with reference to, and in conjunction with the drawings. FIG. 1 schematically shows a mode of operation of a scattered radiation grid in radiological diagnostics. The X-ray beams 2 originating at a focus of an X-ray source 1 propagate in a straight line in the direction of an object 3. The X-ray beams pass through the object 3, and in the form of rectilinear primary radiation 2a strike a radiation detector 4 downstream of the object 3. As such, the primary radiation beams 2a produce a location-resolved distribution of attenuation for the object 3. A portion of the radiation 2 passing through the object 3 is scattered in the object 3, thereby creating secondary radiation or scattered radiation 2b, which does not contribute to the targeted image information. When the secondary radiation or scattered radiation 2b strikes the detector, the actual image information may be adulterated and the signal-to-noise ratio may be worsened or reduced. To minimize the adverse effects of the secondary radiation 2b on the image taken at the detector, a scattered radiation grid 5 is provided, which is located between the object 3 and the detector 4. The scattered radiation grid 5 has beam channels 6, which are determined by a basic structure 7. The basic structure 7 in turn forms an absorption structure, which may absorb the striking secondary radiation 2b. As FIG. 1 shows, the beam channels 6 are focused or in other words aligned in the direction of the X-ray source 1. Arriving primary radiation 2a, as shown in FIG. 1, points along rectilinear paths through the scattered radiation grid 5 at the detector 4; and substantially all of the other radiation, forming an angle with the paths, is absorbed or substantially attenuated by the scattered radiation grid 5. The set-up conditions are similar when making images in nuclear medicine. A radiation generating vehicle, not identified by a reference numeral in FIG. 2, that emits gamma rays is placed in an organ 3a of the object 3 being examined and is enriched therein and upon its decomposition emits gamma quanta 8a and—because of scattering in the organ 3a or object 3—also emits gamma quanta 8b as scattered radiation. Via a collimator 5, the primary radiation, in the form of the quanta 8a, reaches the detector 4 directly, while the secondary radiation at an angle with the detector, in the form of the gamma quanta 8b, is absorbed by the collimator 5. FIG. 3, schematically, shows the production of a scattered radiation grid, using a substantially rapid prototyping technique, primarily based on stereo-lithography. A laser beam 9 is aimed at a surface of a UV-cross-linkable polymer 11 located in a container 10. The laser beam 9 is moved over the surface, as indicated by the double arrow A, and the motion control, which is performed via a suitable controlling computer, is based on a three-dimensional volumetric model of the basic structure 7 to be set up or produced. Via the moving laser beam 9, the pattern of the basic structure 7 to be created is quasi-inscribed into the polymer resin 11, causing a corresponding resin layer to solidify in accordance with the inscribed pattern. This resin layer is built up on a platform 12, which once the first structural layer has been “inscribed,” is lowered, as represented by the double arrow B, so that the second structural layer is inscribed. Substantially fine filigreed structures 7 can be inscribed by the laser 9 due to the good focusing ability of the laser 9, so that even substantially thin-walled structures can be made with an arbitrary configuration. The basic structure 7 can be built up either directly on the platform 12 or on a support plate, not identified by a reference numeral. Regarding the description of the other drawing figures, it will first be pointed out that each figure describes a corresponding scattered radiation grid, while keeping similar layouts for the collimator. FIG. 4 schematically shows a cross section of an embodiment of a scattered radiation grid 13. A support 14 of radio-transparent material is plastic. This support 14 may be produced by stereo-lithography in a substantially rapid prototyping process. Near a top of the scattered radiation grid 13, a plurality of plug-in or clamping receptacles 15 distributed in a matrix-like fashion is provided, and one absorbing element 16 is mounted on each of them. The absorbing elements 16 include radio-absorbent material, such as W or Ta. The absorbing elements 16 are shaped in the form of small tubes, or in other words are hollow on the inside. The shape and diameter of the plug-in or clamping receptacles 15 corresponds to the inner shape or inner diameter of such a tubular absorbing element 16, which may be hollow with cylindrical, oval or polygonal cross sections. The absorbing elements 16 are mechanically retained on the plug-in or clamping receptacles 15 to be fixed in a stable position. Once the absorbing elements have been placed (the number of absorbing elements to be placed can amount to several hundred thousand), the entire absorption structure is potted with a potting composition 17, such as a casting resin. FIG. 5 schematically shows a top view on the scattered radiation grid 13 of FIG. 4. The absorbing elements 16 can be seen arranged in rows above and below each other. They are placed substantially close together. The spacing of the plug-in fixtures or clamping fixtures 15 is selected in accordance with the wall thickness and the diameter of the absorbing elements 16. Incident X-radiation can pass through the through conduits or channels 18 formed in the absorbing elements 16, as well as through the voids located between each two absorbing elements 16. FIG. 6 by comparison schematically shows, via a top view, a different or alternate arrangement pattern. In order to increase a packing density, the absorbing elements 16 are located in rows offset from one another. An exemplary layout, however, is similar to that described for FIG. 4. FIG. 7 schematically shows a cross section of another embodiment of the scattered radiation grid 19, in which tubular absorbing elements 16 used. The elements 16 are mounted on plug-in or clamping receptacles 15 that project from the support surface. However, with respect to the surface area of the support 14, the plug-in or clamping receptacles 15 are each at different angles, which also enables positioning the absorbing elements 16 at a corresponding angle relative to one another. The absorbing elements 16 may be focused by aligning them with respect to a fictive focus. This fictive focus may be a radiation source that generates the primary radiation that fans out toward the scattered radiation grid 19. Because of the focusing, the primary radiation, in accordance with its alignment, that passes un-scattered through the object to be examined reaches a region of the scattered radiation grid 19 in which the absorbing elements 16 are aligned and focused in accordance with the primary radiation. This primary radiation can pass un-attenuated through the focused absorbing elements 16. However, at least some secondary radiation or scattered radiation that is scattered by the object is absorbed via the absorbing elements 16. While FIG. 7 describes the embodiment of the scattered radiation grid 19 in which plug-in or clamping receptacles 15 rise from the surface of the support 14, FIG. 8 schematically shows an alternate scattered radiation grid 20 in which the plug-in or clamping receptacles are embodied as indentations 21 that extend into the plane of the support. The indentations 21 are also embodied or positioned at an angle, so that the absorbing elements 16—in the exemplary embodiment shown as tubular absorbing elements—are aligned at an angle relative to a focus. The absorbing elements 16 are inserted into the indentations, which in their shape or diameter correspond to the outer diameter or outer shape of the absorbing elements 16, and are mechanically fixed therein. FIG. 9 schematically shows a cross section of another embodiment of a scattered radiation grid 22. On the support 14, a plurality of plug-in or clamping receptacles protruding from the top surface are embodied, in the form of pegs 23, between which the absorbing elements 24 are placed and are retained in clamping fashion. This arrangement of the scattered radiation grid 22 is shown from the top view of FIG. 10. A width or shape of the pegs 23 is dimensioned such that the absorbing elements 24 can be located substantially close to each other. Unlike the embodiments described above, an absorbing element 24 comprises a support element 25, made primarily of radio-transparent plastic, which forms the through conduit for the radiation. On an outside surface, the support element 25 has a coating 26 made of absorbent material. In this arrangement, the absorbing elements 24 are fixed securely via the plug-in fixtures or clamping fixtures embodied as pegs 23. The pegs 23 may also be longer than shown, optionally as long as an absorbing element. FIG. 11 schematically shows a cross section of another embodiment of a scattered radiation grid 27. The support 14 is quasi-perforated with a plurality of holes 28 that form the plug-in or clamping receptacles. One absorbing element—shown as a tubular absorbing element 16—is inserted in clamping fashion into each hole 28. It is understood that—as in the pattern of FIG. 9—focusing of the radiation may be achieved by suitable alignment of the holes 28. A top view of the scattered radiation grid 27 is shown in FIG. 12. FIG. 13 schematically shows a method for positioning the absorbing elements on the corresponding support. As an example, a support 29 is shown from which relatively long peg-like plug-in fixtures or clamping fixtures 30 protrude. One absorbing element 31—a microtubule as an example—is to be placed between a plurality of these plug-in fixtures or clamping fixtures 30. This absorbing element placement is performed via a tool 32, which is supplied with the absorbing elements 31 to be placed from a reservoir 33 that is shown as an example. These absorbing elements 31 reach a conduit 34 in the tool that is positioned substantially exactly vertically above a position between a plurality of plug-in or clamping receptacles 30 at which an absorbing element 31 is to be placed. From within this tool conduit 34, the absorbing element 31 slides into the receiving position between the plug-in or clamping receptacles 30. To improve the insertion, a slight overpressure may be applied in the tool conduit 34 via the reservoir 33, so that the absorbing element 31 is pressed in the plug-in or clamping receptacles 30. This overpressure may also be applied intermittently, whenever the tool 32 is to be positioned exactly and the absorbing element 31 is to be put in place. Alternatively or in addition, an under pressure may be applied to the support 29, via an opening 35 and may be optionally provided at each absorbing element position. This opening 35, like the arrow that symbolizes the underpressure, is shown in dashed lines. Moreover, rinsing may be performed to an absorbing element 31, for instance with a suitable liquid from the reservoir 33 or from the tool conduit 34, for example via water, with which the aforementioned pressure can be built up. Moreover, the inner walls of the tool conduit 34 may be provided with a coating that improves sliding along the inner walls. To improve the sliding into the position toward the support, the plug-in or clamping receptacles 30 are provided with a narrowing or pointing edge 36 at their respective tops. Such edges 36 may furthermore be provided on each plug-in or clamping holder of the type described above. Since as described, depending on the size of the scattered radiation grid, up to several hundred thousand absorbing elements are placed (in a mammography scattered radiation grid, this may be from 100,000 to 500,000 absorbing elements), it is advantageous if a plurality of absorbing elements 31 can be placed simultaneously. As such, the tool 32 may be embodied accordingly. FIG. 13 schematically shows one method. The tool 32 may be configured such that every other absorbing element receptacle can be mounted simultaneously, so that after one assembling step or act, the tool is moved onward by only a single position, and the yet unoccupied positions located in between are then filled. The tool 32 may be produced or manufactured from plastic in the course of the stereo-lithography by the substantially rapid prototyping technique, to achieve the targeted precision, even when implementing a simultaneous mounting of a plurality of absorbing elements. It should be understood that the proposed assembling method or process described is merely one example, and other types of assembly method are conceivable. As already described, the absorbing elements, or the coatings, may be of any potential materials that absorb radiation, such as X-radiation. The length of the absorbing elements may be between 1 mm and 10 mm, or between 2 mm and 6 mm. The outer diameter—whether for a tubular absorbing element or a pin-like absorbing element—may be between 0.3 mm and 2 mm, or between 0.5 mm and 1 mm. The wall thickness of tubular absorbing elements may be between 20 μm and 50 μm. While only FIG. 4 shows the embedding of the absorbing elements in the potting composition, it is understood that all the structures shown may be embedded in a suitable potting composition.
	summary
048881503	claims	1. A control rod for nuclear reactors, comprising a number of absorber plates (13-16) which are connected to each other along a centre line on the rod and which are each provided with a plurality of bored channels (18), which extend at least substantially perpendicularly to the centre line of the rod, contain boron carbide or other absorber material (20) which swells upon irradiation and are sealed off from communication with the surroundings of the control rod, characterized in that within at least one region of an absorber plate each channel (18b) is arranged at a smaller distance (a) to an adjacent channel than (c) to the surface of the absorber plate. 2. A control rod according to claim 1, characterized in that within at least the stated region of the absorber plate, each channel (18b) is arranged at a smaller distance (a) to the adjacent channel on one of its sides than (b) to the adjacent channel on its other side. 3. A control rod according to claim 1, characterized in that within at least the stated region of the absorber plate, each channel (18b) is arranged at a distance (a) to an adjacent channel which is smaller than half of the distance (c) to the surface of the absorber plate.
045445206	claims	1. A neutron producing laser target comprising: a pair of crossed glass fibers disposed across an orifice on a target supporting plate; a microsphere disposed in a quadrant formed by said crossed fibers glued at an equator tangentially to each of said fibers; and at least one shell enclosing said microsphere comprising two hemispheres glued to said fibers with said microsphere disposed essentially at the center thereof. 2. The invention of claim 1 wherein said fibers are at most about 5 to about 10 .mu.m in diameter. 3. The invention of claim 1 wherein said glass fibers are juxtapositioned essentially at right angles to one another. 4. The invention of claim 1 wherein said shell is about 100 .mu.m in outside diameter. 5. The invention of claim 1 further comprising at least a second shell enclosing said first shell comprising two hemispheres glued to said fibers. 6. The invention of claim 5 wherein said second shell is essentially 500 .mu.m in outside diameter.
	claims	1. An EMI shield for mounting in a rigid housing for a circuit board, the housing having peripheral sidewalls, comprises a thermoform formed by heating thermoformable sheet and drawing it into a mold or onto a die, the thermoform having a vacuum deposited metal coating thereon of a thickness of at least one micron, said thermoform conforming to said rigid housing and fitting conformingly between the sidewalls thereof, said thermoform having a peripheral, outwardly extending lip thereon, said lip having a first surface and an opposing second surface, the circuit board having a ground trace fixed to an outer surface thereof, said first surface of said lip abuttable to the ground trace of the circuit board, a gasket of elastomeric material disposed between said sidewall and said second surface, whereby said gasket urges said first surface of said lip into touching engagement with said ground trace. 2. A system for containment of EMI and RFI in an electronic device having a generally rigid housing and having a circuit board mountable within the housing comprises a polymeric thermoform having a peripheral sidewall, said sidewall having an outwardly extending lip thereon, said peripheral sidewall defining at least one polygonal compartment on said thermoform, said compartment having an open side, said thermoform having a first face and a second face, said thermoform having a conductive metal coating on at least the first face thereof, the open side of said at least one compartment coincident with said first face of said thermoform, said circuit board having a first side populated with at least one emitting component and having a ground trace fixed thereto, said polygonal compartment overlying said at least one emitting component, said ground trace in registry with said lip and touchingly engaged therewith, said housing having at least one opentopped enclosure formed therein, said at least one enclosure defined by upstanding ribs on said housing, said at least one enclosure receiving said compartment of said thermoform, said lip in registry with said upstanding ribs, an elastomeric gasket interposed between said ribs and said second face of said thermoform, whereby said elastomeric gasket urges said lip into touching engagement with said ground trace. 3. A system for containment of EMI and RFI in an electronic device having a generally rigid housing and having a circuit board mountable within the housing comprises a polymeric form having multiple compartments defined by hollow walls integrally formed in said form, each of said compartments having an open side, said form being a thermoform, said form having a first face and a second face, each of said open sides of said compartments coincide with said first face of said form, said form having a conductive metal coating on all of at least said first face thereof, said circuit board having a first side populated with a plurality of electronic components and having a ground trace fixed thereto, said compartments overlying at least some of said electronic components, said ground trace in registry with said hollow walls and touchingly engaged therewith, said housing having at least multiple opentopped enclosures formed therein, said enclosures defined by upstanding ribs on said housing, said enclosures receiving said compartments of said form, said hollow walls in registry with said upstanding ribs, an elastomeric gasket interposed between said ribs and said second face of said form, whereby said elastomeric gasket urges said hollow walls into touching engagement with said ground trace. 4. The system of claim 3 wherein claim 3 said metal coating is continuous and smooth, comprising a vacuum deposited layer at least one micron in thickness. 5. The system of claim 3 wherein claim 3 said form has a peripheral sidewall having an outwardly extending lip thereon, said enclosure having an outer wall, said lip in registry with said outer wall of said enclosure, said lip in registry with said ground trace, said gasket disposed upon said outer wall of said enclosure and under said lip, said ground trace further in registry with said lip, whereby said gasket further urges said lip into touching engagement with said ground trace. 6. A system for containment of EMI and RFI in an electronic device having a generally rigid housing and having a circuit board mountable within the housing comprises a polymeric form having a peripheral sidewall, said sidewall having an outwardly extending lip thereon, said form having a first face and a second face, said form being a thermoform, said peripheral sidewall defining at least one polygonal compartment on said form, said compartment having an open side, said form having a conductive metal coating on at least the first face thereof, the open side of said at least one compartment coincident with said first face of said form, said circuit board having a first side populated with at least one emitting component and having a ground trace fixed thereto, said polygonal compartment overlying said at least one emitting component, said ground trace in registry with said lip, said housing having at least one opentopped enclosure formed therein, said at least one enclosure defined by upstanding ribs on said housing, said at least one enclosure receiving said compartment of said form, said lip in registry with said upstanding ribs, said lip having a multiplicity of spaced apart protrusions formed therein. 7. The EMI containment system of claim 6 wherein claim 6 said protrusions comprise dimples pressed into said lip of said form. 8. The EMI containment system of claim 6 wherein claim 6 said protrusions comprise die cut tabs formed on and extending from said lip of said form. 9. The EMI containment system of claim 6 wherein claim 6 said protrusions comprise die cut punctures formed on and extending from said lip of said form. 10. The containment system of claim 6 wherein claim 6 said protrusions comprise dimples pressed into said lip of said form. 11. The containment system of claim 6 wherein claim 6 said protrusions extend from said first face of said form, whereby said protrusions are urged by said ribs into touching engagement with said ground trace. 12. The containment system of claim 6 wherein claim 6 said protrusions extend from said second face of said form, whereby said protrusions urge said lip into touching engagement with said ground trace. 13. A method of shielding EMI/RFI in an electronic device, the method comprising coupling a containment form to a printed circuit board, said containment form being a thermoform; grounding the containment form to a ground trace; and compressing the containment form against the ground trace by contacting a portion of a housing of the electronic device against the containment form. 14. The method of claim 13 wherein the containment form is a metallized thermoform. claim 13 15. The method of claim 14 further comprising vacuum metallizing the thermoform. claim 14 16. The method of claim 13 wherein grounding comprises contacting a protruding lip of the containment form against the ground trace. claim 13 17. The method of claim 13 wherein grounding comprises creating a Faraday cage. claim 13 18. The method of claim 13 wherein compressing comprises forcing ribs of the housing against the containment form so as to urge the containment form against the ground trace. claim 13 19. The method of claim 18 wherein forcing comprises receiving the ribs in cavities in the containment form. claim 18 20. The method of claim 13 further comprising positioning a non-conductive gasket between the housing and the containment form. claim 13 21. The method of claim 20 further comprising urging a rib of the housing against the nonconductive gasket so as to urge the containment form against the ground trace. claim 20 22. A system for shielding EMI/RFI, the system comprising: a housing; a circuit board comprising a ground trace, the circuit board being positioned within the housing; a containment form comprising a lip which extends around a periphery of the containment form, said containment for being a thermoform; a vacuum metallized layer attached to the containment form, wherein the vacuum metallized layer is capable of shielding EMI/RFI radiation; wherein the containment form is positioned in the housing so that the housing urges the containment form into contact with the ground trace so as to shield the circuit board from the EMI/RFI radiation. 23. The system of claim 22 wherein claim 22 the housing comprises four side walls and ribs, the containment form is received within the housing between the four side walls and the ribs contact the containment form to urge the containment form against the ground trace. 24. The system of claim 23 wherein the containment form comprises at least one hollow wall to receive the ribs. claim 23 25. The system of claim 22 wherein the containment form comprises dimples disposed on the lip. claim 22 26. The system of claim 22 further comprises compressing a compressible gasket positioned between the housing and the containment form, wherein the housing contacts the gasket to resiliently urge the containment form against the ground trace. claim 22 27. The system of claim 23 wherein the containment form comprises a plurality of compartments. claim 23
	summary
050645757	description	FIG. 1a depicts schematically a first method for loading and sealing a double container system 2, consisting of a removable inner container 4 of steel and an outer shielding container 6 in six steps, designated A, B, C, D, E, and F. The inner container 4 has a screw-in inner cover 8 and a weld-on outer cover 10 and the shielding container 6 has a screw-on shielding cover 12. For loading and sealing, in a first step A, the empty double container system 2 is injected into a shielded chamber 14, for example, a so-called hot cell. In the second step B, in the shielded chamber, the open inner container 4 is loaded through the top opening of the shielded container 6 with radioactive material 16 which is enclosed in and is to be stored in a sheath (box, metal mould) 16'. In the third step C, the inner container while still in the hot cell is sealed with the screw-in inner cover 8, and the seal of the screw-in cover is tested. In the fourth step D, the ejection from chamber 14 of the double container system which is loaded and sealed with the inner cover 8 takes place. In step E, outside of the shielded chamber, the outer cover 10 is welded to the inner container 4, and after the welding is complete, the weld is tested. Finally, in the last step F, the shielding cover 12 is screwed onto the shielding container 6. FIG. 1b depicts schematically a second method for loading and sealing a double container system 2, consisting of a removable inner container 4 of steel and an outer shielding container 6 in six steps, A, B, C, D, E, and F. The inner container 4 has a screw-in inner cover 8 and a weld-on outer cover 10 and the shielding container 6 has a screw-on shielding cover 12. For loading and sealing in a first step A, the empty and open double container system 2 is locked or gripped from below the hot cell (shielded chamber) 14', specifically within an injection aperture 17 located in the floor of the cell. The seal of the injection aperture is not depicted, but it is understood that suitable, known transport and lifting devices are used and that the docked double container system 2 is arranged absolutely sealed and shielded in the injection aperture 17, as is indicated by means of the seal/shield 19. In the second step B, the inner container 4 is loaded from the hot cell 14' with the radioactive material 16 which is to be stored and which is enclosed in a sheath 16'. In the third step C, the inner container 4 is sealed with the screw in cover 8 while the double container system 2 is still locked in the injection aperture 17, and the screw-in cover seal is tested. In the fourth step D, the sealing of the injection aperture 17 and the loosening and removal from the hot cell 14 of the loaded double container system 2, now sealed with the inner cover 8, takes place. After this, in step E, outside of the shielded region, the outer cover 10 is welded to the inner container 4, and after the welding is complete, the weld is tested. Finally, in the last step F, the shielding cover 12 is screwed to the shielding container 6. The inner container 4 which is depicted in greater detail in FIG. 2 consists of a cylindrical jacket 18, a floor 20, and a seal 22. The seal 22 consists of the inner cover 8 which is designed as a sealing plug, and can be screwed into the jacket 18 against bottom and side seals 24, and the outer cover 10, which is designed as a sealing plug with a handle 26, which outer cover is welded to the jacket 18 of the inner container 4. A welding gap 28 is left between the outer cover 10 and the jacket 18 for the application of a weld 29 between the cover and the jacket by means of narrow-gap welding. The sealed container is further provided with a welded-on plasma hot wire cladding layer 30 for corrosion protection. FIG. 3 shows the weld in more detail. The welding gap 28 between the outer cover 10 and the container jacket 18 widens slightly toward the top and is limited on the bottom by means of two surrounding welding flanges 32 and 34 which lie opposite one another, of which one is located on the jacket and the other on the outer cover 10. The welding flanges 32 and 34 are canted to attain a clean weld root 36. The cant 38 amounts to about 45.degree. and is preferably provided both on the top and on the bottom sides of the welding flanges. The canting has the advantage that heat dissipation from the weld point is improved. Also, the canting facilitates the introduction and positioning of the outer cover 10. The root welding is performed preferably by means of an inert gas-shielded arc welding device with tungsten electrodes ("WIG" welding device), with which a very precise weld can be performed. Onto the weld root 36, further weld layers 40 are welded with the inert gas-shielded arc welding device with tungsten electrodes, the purpose being to securely prevent burn-through during the following welding of the remaining weld layers 42 by means of a submerged arc welding system ("UP" welding system), with which large quantities of welding metal can be applied. The welding of the root thereby proceeds preferably with the help of at least two "WIG" welding heads, which lie opposite one another and operate simultaneously. This prevents distortion of the cover and thus a disruption of the uniformity of the welding gap during the welding process, and finally the danger of forming an uneven welded root and possible fissures. The final welding performed as narrow-gap submerged arc welding has the advantage, since it welds thicker cross-sections than has presently been the case, that it leads to lower production of heat and leads to a more uniform build-up of the weld layers and thus of the weld itself. The weld material of the edge layers molds to the edges of the welding gap between the cover and the container jacket, whereby the coarse grain is almost completely converted to fine grain by the following layer. Thus the necessary condition is created for eliminating resulting voltage warming and cooling, and it is assured that the material values of the weld lie within the framework of the material values of the base material. The production of the seal takes place with the help of a device depicted in FIGS. 4 through 6. The shielding container 6, into which the inner container 4 is inserted, is set on a horizontal rotary table 44, which is anchored to the floor 46 of a foundation pit 48. Where applicable, the inner container 4 can be set on the rotary table alone. The rotary table is equipped with a spherical turning connection to absorb horizontal and axial forces. The rotary table is driven by means of a motor within a low rotational speed range. The top of the table has a mechanical stage which is adjustable with a motor for the precision positioning of the container under the welding and testing movable bridge 50. Since the foundation pit 48 should be able to be used for containers of various sizes, spacing pieces 52 (drawn in dot-and-dash lines) are placed on the rotary table for equalization of length so that the sealing weld will always take place at the same height above the floor. A pit cover 54 (FIG. 5) is provided which is put on during loading of the rotary table with the help of a crane. It is moved over the concrete pit and can be walked on during the loading process. All welding, testing and other devices are mounted on the bridge 50. The movable bridge 50 has several interior tracks (not depicted) which serve for guiding transport carriages 56 and the welding and testing devices 58, 60, 62, 64 which are fastened on them. The bridge 50 is equipped with a traveling gear 66 (FIG. 4) with flanged wheels and a direct current drive by which the bridge can be driven at both positioning speed and rapid traverse speed along raised tracks 68 which are arranged on both sides of the foundation pit. The drives of the transport carriages are supplied with power separately, and each transport carriage has a height adjustment with which the complete welding or testing device can be moved into operating or waiting position. The welding and testing devices encompass an inert gas-shielded arc welding device with tungsten electrodes (WIG) 58, submerged arc welding device ("UP") 60, a plasma hot-wire welding device ("PH") 62 and a testing device 64. FIG. 6 shows a part of the welding and testing bridge 50 in greater detail with a testing device 64 in the operating position. The weld 29 and the application of the corrosion protection layer 30 have already taken place. The operating position of the welding devices 58 or 60 are quite analogous in appearance. It can be easily seen from FIG. 6 that the welding and testing of the outer cover 10 is to be performed with the inner container 2 inserted in the shielding container 6. Reference number 70 indicates a moderator. The complete welding and testing process preferably occurs automatically, for which purpose a control device 72 (FIG. 5) is provided. Otherwise, control panels are provided, from which the welding and testing devices can be controlled. If desired, the welding and testing devices can be mounted on bracket arms (not depicted), rather than on bridge 50. To guarantee optimum testing of the weld 29, the shielding container has a step-shaped, annular mouth 74 extending above the weld so that a sufficient free space arises between the shielding container and the inner container to insert and move the testing device 64 (FIG. 6). Furthermore, the outer cover 10 of the inner container 4 has provided on its bottom side below the welding flange 34 an annular, step-shaped recess 78, by means of which an annular chamber 80 is formed between the outer cover 10 and the inner cover 8, which continues in an annular groove 82 formed below the welding flange 32 of the container jacket in the inner wall of this jacket.
052987590	abstract	Ultraviolet (UV) light from a lamp (12) or laser (38) is provided for cracking Group V and Group VI species comprising clusters (dimers and tetramers) or metal-organic molecules to form monomers (atoms). The UV radiation interacts with a molecular beam (14) of Group V and Group VI species following their generation in a source cell (16), which may be an effusion source in molecular beam epitaxy (MBE) apparatus, a thermal cracker cell in gas-source MBE apparatus, or a gas injector cell in metal-organic MBE apparatus (MOMBE). As configured, the UV light source and associated elements comprise a unit, termed herein a "photo-cracker cell" (10). The photo-cracker cell includes an elliptical reflective cavity (18), which defines two foci. The source of UV light is located along one focus and the path of the molecular beam is located along the other focus substantially parallel thereto. The photo-cracker cell may be provided on existing MBE or MOMBE apparatus, between the present source cell and the growth chamber (36) in which III-V, IV, and II-VI semiconductor layers on substrates are deposited.
044341334	abstract	Organic hydrocarbon materials are produced from plentiful inorganic limestone type materials by: (1) reacting the limestone type materials with molten lithium metal to produce Li.sub.2 C.sub.2 (2) hydrolyzing the Li.sub.2 C.sub.2 to produce C.sub.2 H.sub.2, (3) catalytically reacting the C.sub.2 H.sub.2 with steam to produce CH.sub.3 COCH.sub.3, (4) pyrolyzing the CH.sub.3 COCH.sub.3 to provide ketene and methane, and separating the ketene. The ketene may then be decomposed to provide methylene, which can be reacted with an alkane, such as methane in an insertion chain reaction, to provide organic hydrocarbon materials. An in-place nuclear reactor can provide energy for the endothermic reactions of the system.
046481060	claims	1. An X-ray lithographic system in which an X-ray mask including a patterned mask membrane is positioned in spaced alignment with a semiconductor wafer substrate having an X-ray sensitive resist material thereon, said system comprising an X-ray source, a chamber though which X-rays are passed to an exit therein to a patterned mask membrane sealing said exit, means for flowing a low X-ray-attenuation gas into said chamber at a pressure near the ambient surrounding the outside of the chamber and means for venting said gas from said chamber in a controlled manner such that the gas pressure at said mask prevents ingress of contaminating gas into said chamber and does not significantly deflect said mask membrane with respect to the semiconductor wafer substrate and the velocity of venting gas is sufficient to prevent back diffusion of contaminating gases into said chamber, said means for venting comprising an inverted U-shaped vent tube extending at a first end from a position in said chamber adjacent a bottom portion of said chamber upwardly to a position outwardly of said chamber, said tube then extending downwardly to a second end of said tube exterior of said chamber and terminating in an orifice at a level approximate the level of the mask membrane. 2. An X-ray lithographic system in which an X-ray mask including a patterned mask membrane is positioned in spaced alignment with a semiconductor wafer substrate having an X-ray sensitive resist material thereon, said system comprising an X-ray source, a chamber through which X-rays are passed to an exit therein to a patterned mask membrane sealing said exit, means for flowing a low X-ray-attenuation gas into said chamber at a pressure near the ambient surrounding the outside of the chamber and means for venting said gas from said chamber in a controlled manner such that the gas pressure at said mask prevents ingress of contaminating gas into said chamber and does not significantly deflect said mask membrane with respect to the semiconductor wafer substrate and the velocity of venting gas is sufficient to prevent back diffusion of contaminating gases into said chamber; and wherein said chamber is oriented vertically and in which said means for venting includes an inverted U-shaped vent tube extending first upwardly from said chamber to a position outwardly of said chamber, said tube then extending downwardly to an end orifice of said tube exterior of said chamber at a level approximate the level of the mask membrane. an X-ray source, a low attenuation chamber through which X-rays are passed to an exit therein to a patterned mask membrane sealing said exit; means for flowing a low X-ray-attenuation gas into said chamber at a pressure near the ambient surrounding the outside of the chamber and means for venting said gas from said chamber in a controlled manner such that the gas pressure at said mask prevents ingress of contaminating gas into said chamber and does not significantly deflect said mask membrane with respect to the semiconductor wafer substrate and the velocity of venting gas is sufficient to prevent back diffusion of contaminating gases into said chamber; means for maintaining a process gas atmosphere in a zone surrounding a side of said mask membrane, facing away from said X-ray source, and said substrate to prevent ingress of contaminants and ambient atmosphere to said zone, said process gas atmosphere being at a gas pressure that does not significantly deflect said mask membrane with respect to said substrate; and wherein said means for maintaining a process gas atmosphere includes a gas flange having a planar surface extending parallel to and spaced from said mask membrane and means for flowing a process gas into said zone and past said gas flange. flowing a low X-ray attenuating gas into an upper portion of said chamber; venting said gas from a lower portion of said chamber at a level adjacent to said mask, said gas having a slight positive pressure in said chamber adjacent said mask; flowing said vented gas from said chamber for discharge at an exterior position level approximate the level of said mask; and tuning the exterior position level of gas venting discharge to minimize pressure differential across a top surface of said mask facing said chamber and a zone encompassing said substrate and a bottom surface of said mask spaced from and facing said substrate. 3. The invention as set forth in claim 2 further including means for tuning the level of said end orifice to vary the pressure on a side of the patterned mask membrane facing said X-ray source and adjust pressure differentials across the patterned mask membrane. 4. The invention as set forth in claim 1 further comprising means for maintaining a process gas atmosphere in a zone surrounding a side of said mask membrane, facing away from said X-ray source, and said substrate to prevent ingress of contaminants and ambient atmosphere to said zone, said process gas atmosphere being at a gas pressure that does not significantly deflect said mask membrane with respect to said substrate. 5. An X-ray lithographic system in which an X-ray mask including a patterned mask membrane is positioned in spaced alignment with a semiconductor wafer substrate having an X-ray sensitive resist material thereon, said system comprising: 6. The invention as set forth in claim 5 in which said substrate is mounted on a chuck and wherein said gas flange is sealed by a flexible seal to said chuck and said mask includes a mask frame sealed by a flexible seal to a mask holder supporting said mask. 7. The invention as set forth in claim 6 in which said gas flange is spaced from said mask frame forming a gap extending peripherally around said gas flange, said gap having a height of from about 10 to about 1000 microns. 8. The invention as set forth in claim 6 in which said gas flange and said chuck includes interdigitated circular bosses to block transmission of stray X-rays. 9. The invention as set forth in claim 4 wherein said means for maintaining a process gas atmosphere includes means for flowing process gas into said zone and vent means extending from said zone to flow process gas exteriorly of said zone. 10. A method for fabricating a semiconductor substrate by irradiating the substrate by an X-ray source passing through an exposure chamber mounting a patterned mask at an exit of said chamber and wherein said substrate is mounted for fabrication in a fixedly spaced parallel plane exteriorly from said chamber and in alignment with said mask, said method comprising 11. The method as set forth in claim 10 further comprising flowing a substrate fabrication process gas into said zone and venting said gas from said zone at a velocity to prevent ingress of contaminating gas to said zone and at a pressure to avoid deflection of said mask.
06069937&	description	DETAILED DESCRIPTION OF THE INVENTION The present invention relates to illumination apparatus and exposure apparatus, and in particular to illumination apparatus for use with soft X-ray projection exposure apparatus. A soft X-ray projection exposure apparatus transfers the circuit pattern on a photomask (i.e., a mask or reticle) onto a substrate, such as a wafer, through a reflective-type image-forming apparatus by means of a mirror projection system, such as an X-ray optical system. The present invention has the objective to provide a high-performance illumination apparatus, wherein the illumination efficiency is markedly higher than is conventional, and the numerical aperture of the X-rays at the illumination area formed in a circular arc (i.e., an arcuate area) is nearly uniform, independent of the illumination position. With reference to FIG. 1, the illumination apparatus 5 of the present invention comprises an excitation energy light generation unit 10, a target member 12, and an illumination optical system 13. Excitation energy light rays 15 are emitted from unit 10 and irradiate a plurality of locations 16 on target member 12. Target number 12 is capable of generating X-rays upon being irradiated with light of the appropriate wavelength. X-rays 17 are generated from locations 16 on target member 12 irradiated by excitation energy light rays 15. Thus, locations 16 become microscopic X-ray light sources 16. X-rays 17 emitted from sources 16 pass through illumination optical system 13 and irradiate a mask 14 as X-rays 18. With reference now to FIG. 2, by arranging a plurality of sources 16 (i.e., individual sources 61-66) on target member 12 having a curved surface S, widths p1-p3 of X-ray beams 71-73, respectively, constituting X-rays 17 can be controlled. For example, beam 71 includes X-rays emitted from sources 16, and in particular, includes X-rays emitted from sources 61 and 64, as well as from sources 62 and 63 therebetween (sources outside of source 62 and light source 63 are not illustrated). Similarly, beam 72 also includes the X-rays emitted from sources 16, and in particular, includes X-rays emitted from sources 62 and 65, as well as from sources 63 and 64 therebetween (sources outside of sources 63 and 64 are not illustrated). Likewise, beam 73 also includes X-rays emitted from sources 16, and in particular, includes X-rays emitted from sources 63 and 66, as well as from sources 64 and 65 therebetween (sources outside of sources 64 and 65 are not illustrated). With continuing reference to FIG. 2, width p1 of beam 71 in the sagittal direction is determined by the space between sources 61 and 64. Likewise, width p2 of beam 72 in the sagittal direction is determined by the space between sources 62 and 65, and width p3 of beam 73 in the sagittal direction is determined by the space between sources 63 and 66. Accordingly, by forming sources 16 on the desired curved surface S of target member 12, widths p1-p3 of beams 71-73 emitted toward illumination optical system L13 can be made a desired width. In a preferred embodiment, curved surface S of target member 12 is shaped such that X-rays 18 illuminate mask 14 at a roughly uniform numerical aperture (see FIG. 1). In particular, if an optical system such as the one shown in FIGS. 6 and 7 is used as illumination optical system 13 in the present invention, it is preferable that curved surface S be made cylindrical. In so doing, widths p1-p3 of beams 71-73 emitted from light sources 16 are equal. Thus, arcuate illumination area 140 (see FIG. 7) can be illuminated at a uniform numerical aperture. With reference again to FIG. 2, target member 12 can be considered to have a cylindrically curved surface S. Accordingly, if light sources 61-66 are arranged on cylindrical surface S, then light beams 71-73 have the same widths p1-p3 (i.e., p1=p2=p3), since the spaces between light source 61 and 64, between light source 62 and 65, and between light source 63 and 66, respectively, are equal to the diameter of the circular cross section of target member 12. In other words, sources 16 form the shape of curved surface S such that widths p1-p3 of each beam 71-73 is equal, or nearly so. Thus, with reference now to FIGS. 7 and 11, beams 124 and 125 converging on arcuate illumination area (field) BF have a cone shape, while constantly extending an equal angle with respect to the convergent point at every part (i.e., have a uniform cross-section), and the numerical aperture at the illumination area is equal and is independent of the illumination position. With reference now to FIG. 3, target member 12 on which X-ray sources 16 are formed can comprise a solid member 81 having curved surface S with a given shape, as shown. The shape in which sources 16 is arranged can be determined by the shape of curved surface S. Since member 81 can be designed to have curved surface S be of any desired shape, it is relatively straightforward to control the shape (i.e., curvature) of curved surface S on which light sources 16 are arranged. Generally, target member 12 tends to wear down when X-ray sources 16 are used over an extended period of time. Thus, with reference now to FIG. 4, it is preferable that a new target member be supplied. By providing a tape-shaped target member 82 moveable along curved surface S of a substrate 83, the target member can be continuously supplied. Tape-shaped target member 82 can be easily created in a desired curved surface shape by pressing or winding it onto substrate 83 having a desired shape of curved surface S, for example. With reference now to FIG. 5, a particulate target member 84 may also be employed. Particulate target member 84 is supplied by a particulate target member supply unit 85 (e.g., a nozzle). Further, particulate target member 84 is preferably arranged to have a desired curved surface shape. Since the position of particulate target member 84, e.g., the nozzle position, can control the angle at which the particles are emitted from the nozzle, and can control the emission speed of the particles and the like, it is easy to arrange sources 16 to have a desired curved surface shape. This method also has advantage that particulate target member 84 can be supplied continuously. Moreover, the amount of dispersed matter from particulate target member 84 can be reduced. In addition, if a pulsed laser is included in excitation energy light generation unit 10 so that X-rays 17 are generated in pulses (see FIG. 1), the pulsed laser light can illuminate particulate target member 84, which is supplied continuously. X-rays can thus be generated by synchronizing the supply of particulate target member 84 from nozzle 85 with the timing of the generation of pulsed laser light. With reference again to FIG. 1, it is preferable that illumination optical system 13 illuminate mask 14 with X-rays 18 at a uniform intensity and at a uniform divergence angle. In particular, it is preferable to arrange target member 12 in the vicinity of the front focal position of illumination optical system 13. In this case, X-rays 17 emitted from each X-ray source 16 pass through illumination optical system 13, and then are transformed into parallel light and the like and irradiate mask 14. Further-more, X-rays 17 emitted from each light source 16 irradiate mask 14 at various angles. Accordingly, mask 14 is illuminated by Kohler illumination or in a manner similar to Kohler illumination. In addition, it is preferable that illumination optical system 13 comprise a reflector (not shown), which preferably includes a multilayer film to increase the reflectance of the reflector surface. As previously discussed, excitation energy light rays 15 irradiate a plurality of locations 16 on target member 12, and X-rays 17 are generated from each of these locations, which become light sources 16. In this case, excitation energy light rays 15 can irradiate all locations (sources) 16 simultaneously, or can irradiate them separately or sequentially. The locations (sources) 16 on target member 12 to be irradiated should take into consideration the specifications required by the illumination optical system, such as, the intensity and divergence angle of X-rays 18, which illuminate the mask, the uniformity thereof, and the like. With continued reference to FIG. 1, when performing static (i.e., non-scanning) illumination of the mask, as in conventional exposure apparatuses, light sources 16 should be arranged two-dimensionally. In addition, in the case of soft X-ray exposure apparatus M, a band-shaped or belt-shaped area (illumination field) may be scanned and exposed due to design limitations of the image-forming optical system. In this case, it is preferable to perform critical Kohler illumination, as described in Japanese Patent Application Kokai No. Hei 7-235470, and light sources 16 should be arranged one-dimensionally. In other words, target member 12 should be arranged one-dimensionally, or when target member 12 is arranged in a plane, it should be irradiated by a one-dimensional (line-shaped) excitation energy light rays 15. In another preferred embodiment of the present invention, it is preferable to use an excitation energy light generation unit 10 that can change the path of excitation energy light rays 15, that can split the excitation energy light rays into a plurality of paths or beams, or that a plurality of excitation energy light generation sources, and the like. Changing the ray path of the excitation energy light rays 15 may be achieved, for example, by moving (e.g., oscillating) an optical element. In another preferred embodiment of the present invention, the excitation energy light rays 15 may be split into a plurality of excitation energy light beams. This may be accomplished using, for example, an optical element, like a beam splitter or a micro-lens array. The intensity of X-rays emitted from sources 16 is determined principally by the intensity of excitation energy light rays 15. Accordingly, the intensity of each of sources 16 can be flexibly adjusted by controlling the intensity of each light beam. In an additional preferred embodiment of the present invention, excitation energy light generation unit 10 includes a laser as a light generation source. If a plurality of laser light generation sources is employed, the intensity of the laser light can be increased to each of sources 16. In this manner, the intensity of X-rays 17 generated from sources 16 can be increased. In other words, this method can be employed if it is desired to increase the throughput of the soft X-ray projection exposure apparatus. The present invention is not limited to light generation sources and optical arrangements as described above. Rather, these are just examples. Although there are cases wherein optical elements are needed when changing the path of laser light or splitting laser light, these do not reduce the intensity of the laser light. This is because the reflectance and transmittance of laser light through laser optical elements is generally nearly 100%. Accordingly, high-intensity X-rays are emitted from each of sources 16. Also, in the illumination apparatus of the present invention, the material comprising target member 12 varies according to the X-rays to be generated. Generally, it is preferable to use a material with a high X-ray generation efficiency. For example, to generate X-rays of a 13 nm wavelength, tin, antimony, lead, tungsten, tantalum, and gold and the like are preferred. In addition, materials with a low melting point and materials that are liquid or gas at room temperature may be used. Materials that are solidified or condensed by cooling and the like can also be used and, moreover, liquids or gases can be used as is. The latter case will often be preferable since the generation of dispersed matter (debris) can be reduced. With reference again to FIG. 1, if excitation energy light rays 15 irradiate target member 12 to generate soft X-rays, it is preferable to use a laser plasma X-ray source as a light generation source in excitation energy light generation unit 10. In this regard, high-intensity soft X-rays can be generated by using laser light as the excitation energy light. Furthermore, in a preferred embodiment, a high-intensity laser may be used to provide laser light. For example, a YAG laser, excimer laser, glass laser or titanium sapphire laser, may be used to obtain a separate high-intensity X-ray source. In addition, excitation energy light rays 15 that illuminate target member 12 are not limited to laser light rays. For instance, the excitation energy light rays 15 may be from a source that can generate soft X-rays, such as an electron beam from an electron beam unit, and the like. Soft X-ray optical systems are often placed in a vacuum, since the absorption of X-rays due to air and the like is large. Accordingly, such systems are well-suited to using an electron beam in illumination apparatus 5. The illumination apparatus according to the present invention can also form a light source equivalent to a secondary light source of a conventional illumination apparatus, without using an X-ray integrator. Moreover, such a light source is superior in that the widths of the light beams emitted from the light sources can be controlled. Since the loss in intensity of the laser light due to the optical system is extremely small, a high-intensity X-ray light source can be obtained. Furthermore, every illumination point on the mask can be illuminated at the desired numerical aperture. By using the exposure apparatus according to the present invention, it is possible to supply high-intensity uniform illumination light, and a soft X-ray exposure apparatus that can expose a large area at high throughput can be provided. WORKING EXAMPLES The present invention is now described in greater detail based on three Working Examples. With reference to FIGS. 1 and 2, in each of the Working Examples below, the relevant illumination apparatus 5 comprises three principal components: excitation energy light generation unit 10, target member 12 and illumination optical system 13. A plurality of excitation energy light rays 15 are emitted from excitation energy light generation unit 10, and irradiate a plurality of locations 16 on target member 12. A YAG laser light is included in light generation unit 10 as a light generation source and a beam splitter (not shown) is used to split the beam emanating therefrom, thereby generating excitation energy light rays 15 as light beams. With reference to FIG. 1, in each of the Working Examples, X-rays 17 emitted from light sources 16 pass through illumination optical system 13 and irradiate mask 14 as X-rays 18. Illumination optical system 13 comprises a reflector (not shown) provided with a molybdenum and silicon multilayer film on the surface, which reflects X-rays having a wavelength in the vicinity of 13 nm. Also, a portion of mask 14 spanning an area 120 mm long and 5 mm wide is irradiated by X-rays 18 over an arcuate illumination area. At this point, mask 14 is Kohler illuminated by arranging target member 12 in the vicinity of the front focal position of illumination optical system 13. As a result, the entire arcuate illumination area (field) on mask 14 can be illuminated at a uniform numerical aperture. In addition, upon exposing a photoresist-coated substrate with an exposure apparatus provided with the illumination apparatus according to the present invention, a photo resist pattern of the desired shape can be obtained across the entire exposure area (field). However, a resist pattern of the desired shape could not be obtained over the entire exposure area with an exposure apparatus provided with a conventional illumination apparatus. Working Example 1 In Working Example 1, target member 12 is cylindrical and is comprised of tin (see FIG. 3). X-rays with a wavelength of at least 13 nm are generated from the locations 16 irradiated by excitation energy light rays 15. X-rays 17 are generated from a plurality of locations (i.e., light sources) 16 on target member 12. Working Example 2 In Working Example 2, target member 82 is a tape-shaped thin plate, as shown in FIG. 4. Target member (Plate) 82 is comprised of tungsten. Target member 82 is wound on cylindrical substrate 83 and arranged so that the shape of surface S forms a part of the cylinder shape. Furthermore, target member 82 is made so that it can be supplied continuously by offsetting its position in the longitudinal direction thereof. X-rays with a wavelength of at lease 13 nm are generated from the part irradiated by excitation energy light rays 15, and X-rays 17 are generated from a plurality of locations (i.e., light sources) 16 on target member 82. Working Example 3 In Working Example 3, target member 84 is made of ice particles (particulates) 16 having diameter on the order of 100 .mu.m, as shown in FIG. 5. Particulate target member 84 is supplied from nozzle 85, and made so that a cylindrical surface is formed by plurality of ice particles 16. If excitation energy light rays 15 irradiate particles 16, X-rays with a wavelength of at least 13 nm are generated. X-rays 17 are generated from a plurality of "locations" (i.e., particulates) 16 on target member 84, thereby becoming sources 16, as discussed above. As explained above, illumination apparatus 5 according to the present invention includes an excitation energy light generation unit 10 that generates excitation energy light rays 15. Light rays 15 irradiate a plurality of locations on target member 12, thereby forming a plurality of X-ray sources 16 corresponding to the aforementioned plurality of locations. Illumination apparatus 5 further includes illumination optical system 13 that irradiates the object to be illuminated with X-rays from the plurality of X-ray sources. The object to be irradiated can be Kohler illuminated at a high illumination efficiency and uniform numerical aperture, since a plurality of sources 16 are arranged on a curved surface S. In other words, the object to be illuminated can be illuminated by X-rays having a uniform intensity and uniform divergence. In addition, since the illumination apparatus according to the present invention can form a light source equivalent to the two-dimensional light source of a conventional illumination apparatus without using an X-ray integrator, it has advantages in that the transmittance (utilization efficiency) of X-rays is high compared with conventional apparatus. Further, it is easy to manufacture. Consequently, the pattern of a mask can be transferred faithfully onto substrates at a high throughput. The above Working Examples are examples of the present invention, and do not limit the present invention. For instance, with regard to Working Example 2, which employs ice particles as the target member, the present invention is not so limited. For instance, gas can be discharged from a nozzle, and the discharged gas or clusters generated by adiabatic expansion may also be used as the target member. Indeed, it will be apparent to one skilled in the art that the number and arrangement of light sources 16 are also not limited to the ones shown in the Working Examples. Light sources 16 arranged one-dimensionally can also be easily created and used in the present invention. Accordingly, the present invention can also be applied to critical Kohler illumination. Thus, the present invention is intended to cover all alternatives, modifications and equivalents as may be included within the spirit and scope of the invention as defined in the appended claims.
	claims	1. A nuclear power generation system comprisinga nuclear reactor,a heat exchanging system for exchanging heat between a primary coolant for cooling the nuclear reactor and a secondary coolant comprised of carbon dioxide or light water, anda turbine power generation system for generating power using heat of the secondary coolant;the nuclear reactor comprising:a reactor core having fuel assemblies comprising a plurality of fuel rods, the fuel rods comprising cladding tubes containing metallic fuel including at least one selected from uranium-235, uranium-238 and plutonium-239;a reactor vessel containing the reactor core;the primary coolant being one of metallic sodium, lead, and lead-bismuth loaded into the reactor vessel and heated by the reactor core; andat least one neutron reflector disposed around the reactor core,wherein the at least one neutron reflector disposed around the reactor core has neutron reflection efficiency for achieving criticality in the reactor core with an effective multiplication factor of neutrons emitted from the reactor core being maintained at or above about 1, andwherein the at least one neutron reflector is coupled to metallic members having a coefficient of thermal expansion greater than a coefficient of thermal expansion of the at least one neutron reflector, and the neutron reflection efficiency of the at least one neutron reflector is changeable utilizing displacement thereof due to dimensional change of the metallic members coupled thereto in accordance with changes in the temperature in the reactor vessel, thereby achieving load following control. 2. The nuclear power generation system according to claim 1, wherein the at least one neutron reflector disposed around the reactor core has a height lower than a height of the reactor core, and is movable upwardly or downwardly along the reactor core with a movement mechanism. 3. The nuclear power generation system according to claim 1, wherein the at least one neutron reflector provided around the reactor core has substantially the same, or smaller, length compared with a full length of the fuel assemblies. 4. The nuclear power generation system according to claim 1, wherein the metallic members comprise at least one of elastic spring or spiral members extending around or above the fuel assemblies. 5. The nuclear power generation system according to claim 1,wherein the at least one neutron reflector comprises a first neutron reflector and a second neutron reflector each extending along a different concentric circle extending around a center of the reactor core, each of the first and second neutron reflectors divided into two or more sections extending along their respective concentric circles,wherein the sections of the first neutron reflector are coupled to a first spiral metallic member provided on a concentric circle about the reactor core, andwherein sections of the first neutron reflector and the sections of the second neutron reflector are moveable, with respect to one another, in a circumferential direction about the reactor core,the movement in the circumferential direction forming slits therebetween in a circumferential direction, wherein the widths of the slits in the circumferential direction are changeable as a result of dimensional change of the first spiral metallic member based on the temperature in the reactor vessel. 6. The nuclear power generation system according to claim 5, wherein each of the first and second neutron reflectors are further radially divided into two or more sections. 7. The nuclear power generation system according to claim 5, wherein the sections of the second neutron reflector are coupled to a second spiral metallic member disposed along a concentric circle of the reactor core, and the spiral directions of the first spiral metallic member and the second spiral metallic member are in opposite directions. 8. The nuclear power generation system according to claim 1, wherein a material of the at least one neutron reflector is selected from beryllium, beryllium oxide, graphite, carbon, and stainless steel. 9. The nuclear power generation system according to claim 5, wherein carbon is provided as a lubricant between the sections of the first neutron reflector and the sections of the second neutron reflectors of the second group. 10. The nuclear power generation system according to claim 5, wherein the sections of the first neutron reflector and the sections of the second neutron reflector overlap in the radial direction from the center of the reactor core, and the widths of the radial overlaps define the temperature at which the slits open and at which the criticality reaches 1. 11. The nuclear power generation system according to claim 1,wherein the metallic members compriseadjustment springs having opposed first and second ends, anda fixation cylinder, against which the first end of the adjustment springs contact, is provided circumferentially outside of the circumferential location of the at least one neutron reflector;the at least one neutron reflector is divided into two or more arcuate sections extending along a concentric circle extending around the reactor core; anda plurality of reflector moving jigs corresponding in number to the number of sections of the neutron reflector, each reflector jig comprising an adjustment spring support plate and a neutron reflector adjusting rod; andeach adjustment spring contacts, at its second end, the adjustment spring support plate provided outside the fixation cylinder,wherein each neutron reflector adjusting rod is coupled to a corresponding neutron reflector at one end thereof and fixed to the adjustment spring support plate at an opposite end thereof, and the reflector adjusting rod moves the neutron reflector with respect to the location of the fuel assemblies upon changes in the temperature of the adjustment springs whereby load following control for the energy output from the nuclear reactor is enabled. 12. The nuclear power generation system according to claim 1,wherein the at least one neutron reflector comprises a multi-layer plurality of rings, each ring comprising a plurality of ring segments extending circumferentially around the fuel rods; and,the metallic members are disposed radially outwardly of and around the multi-layer neutron reflectors rings, wherein different ring segments of the multi-layer plurality of rings are coupled to different portions of the metallic members,wherein, upon a change in temperature of the metallic members, slits are formed between the neutron reflector rings in the circumferential direction and the widths of the slits are dependent upon thermally induced dimensional change of the metallic members to adjust the neutron leakage, whereby load following control for output from the nuclear reactor is enabled. 13. The nuclear power generation system according to claim 1,wherein the at least one neutron reflector comprises a plurality of neutron reflector sections, each of the neutron reflector sections comprising a first end coupled to a supporting rod and a second end distal to the first end, the supporting rods lying on a circular path extending around the reactor core,wherein the first end of each neutron reflector section is rotatable about the supporting rod and thereby form a slit extending in the circumferential direction about the reactor core, and each of the metallic members comprise spiral metallic members connected at a first end thereof to one of the supporting rods and at a second end thereof to the neutron reflector section associate to the rod to which it is connected,wherein changes in the temperature of the spiral metallic members causes dimensional change in the spiral metallic members, thereby causing the first end of the neutron reflector section connected thereto to rotate about the supporting rod to cause the second end of that neutron reflector section to move away from first end of an adjacent neutron reflector section to form a slit therebetween, wherein the span of the slit between the second end of the neutron reflector section and the adjacent first end of the adjacent neutron reflector section varies based on the temperature of the spiral metallic member, whereby load following control for output from the nuclear reactor is enabled. 14. The nuclear power generation system according to claim 4,wherein the at least one of an elastic spring or a spiral metallic members comprise at least one of stainless steel, a nickel based superalloy, and a nickel-cobalt based superalloy, or a bimetal. 15. The nuclear power generation system according to claim 1, wherein a neutron absorber is provided outside the neutron reflector. 16. The nuclear power generation system according to claim 15, wherein the neutron absorber comprises actinoids. 17. The nuclear power generation system according to claim 1,wherein the reactor core has a plurality of fuel rods comprise cladding tubes comprising at least one of ferritic stainless steel or chromium-molybdenum steel containing metallic fuel selected from at least one of an alloy of zirconium, uranium-235, uranium-238, and plutonium-239 or an alloy of zirconium and one selected from uranium-235, uranium-238 and plutonium-239. 18. The nuclear power generation system according to claim 1, wherein the heat exchanging system comprises a main heat exchanger supplied with the primary coolant heated by the nuclear reactor through a conduit, the main heat exchanger including a circulating secondary coolant heated by heat exchange with the primary coolant. 19. The nuclear power generation system according claim 1, wherein the primary coolant comprises lead or lead-bismuth and the secondary coolant comprises light water heated by heat exchange with the primary coolant.
	abstract	A system for an extreme ultraviolet light source includes one or more optical elements positioned to receive a reflected amplified light beam and to direct the reflected amplified light beam into first, second, and third channels, the reflected amplified light beam including a reflection of at least a portion of an irradiating amplified light beam that interacts with a target material; a first sensor that senses light from the first channel; a second sensor that senses light from the second channel and the third channel, the second sensor having a lower acquisition rate than the first sensor; and an electronic processor coupled to a computer-readable storage medium, the medium storing instructions that, when executed, cause the processor to: receive data from the first sensor and the second sensor, and determine, based on the received data, a location of the irradiating amplified light beam relative to the target material in more than one dimension.
	description	The present invention pertains generally to the field of fluid recirculation systems incorporating suction strainers. More particularly, the present application pertains to strainers used to remove debris from water being sucked into a piping system, such as in nuclear power plants. A critical function of Emergency Core Cooling Systems (ECCS) and other recirculation systems of nuclear power plants is to move fluids quickly and in large volumes to critical areas of the nuclear power plant in the event of accidents and emergencies. Integral to this critical function is the ability of strainers, filters, screens and other such devices associated with the systems to remove solids from the moving fluids while at the same time maintaining a sufficiently large volume of fluid flow. Nuclear plants have various safety systems to ensure that the nuclear fuel in the reactor core remains cooled in all credible accident scenarios. One such scenario is a “loss of coolant accident,” (LOCA) in which an external pipe is postulated to break, allowing a large amount of water to escape from the reactor cooling system. This water may dislodge solid debris from neighbouring pipes or other reactor structures. The water, along with some of the dislodged debris, will flow to the lowest parts of the reactor building into a sump. Plants are equipped with safety systems that pump water from the sump back into various reactor cooling systems. Strainers on the pump intakes ensure that any debris large enough to clog equipment in these systems is prevented from entering. Depending on the type of debris, the first layer to deposit on the strainer may form a mat of fibers and collect finer particles, which would otherwise pass through the strainer, resulting in a thin layer of low porosity debris with high hydraulic resistance. This behaviour is referred to as the “thin-bed effect” where the head loss per unit thickness of debris is relatively high as compared to that of full (or thick-bed) debris formation where relatively high porosity debris allows the passage of flow with lower head losses. Thin-bed debris can cause head losses high enough to threaten the functionality of emergency core cooling system (ECCS) sump recirculation pumps. Thin-bed debris has occurred operationally at nuclear power plants and has been created during head-loss testing. One way of alleviating the thin-bend effect is to increase the surface area of ECCS strainers. Strainers must have enough screen area that the debris layer on the strainer is not too thick to cause unacceptably high restriction to flow. Strainers must also be as small as possible to fit into the available space. Therefore compactness, i.e., accommodating the most screen area in the smallest volume, is important. Conventional strainers in many nuclear plants are simple box-type devices that were mounted over the pump intakes. Newer more advanced strainers often have an irregular surface to increase the surface area. An example of an advanced strainer is Atomic Energy of Canada Limited's (AECL's) Finned Strainer®, which is described in International PCT publication number WO 06/50606. The Finned Strainer performs the filtering function through modular hollow fins attached to a header that directs the filtered water to pump intake. The Finned Strainer includes two different fin designs (1) Flat-Surface Fins and (2) Corrugated Surface Fins. These fins have porous filtering surfaces. There remains a need for an improved strainer or filtering element, for example one that minimizes the thin-bed effect described above, and which can be incorporated into existing systems. This background information is provided for the purpose of making known information believed by the applicant to be of possible relevance to the present invention. No admission is necessarily intended, nor should be construed, that any of the preceding information constitutes prior art against the present invention. An object of the present invention is to provide a vaned filtering element. The present invention is an improvement to the Flat-Surface Fin design with the addition of vanes that result in a more compact design. The large increase in filtration surface area over a Flat-Surface fin is a significant advantage to reduce the thin-bed effect. The increased area reduces the restriction to flow entering the strainer by decreasing the water velocity through the screen and reducing the thickness of debris (because it is spread over a larger area). The resistance of this thin layer to flow entering the strainer is reduced with the larger screen area achieved by the vanes. In accordance with an aspect of the present invention, there is provided a vaned filter element comprising one or more fluid permeable screens formed from at least one layer of porous material that is folded into a plurality of hollow vanes extending outwardly from the outer surface of the each of the one or more fluid permeable screens. In accordance with another aspect of the present invention, there is provided a filtering element comprising: a perimeter frame having one or more openings along one side edge of said frame, a pair of fluid permeable screens fixed to the perimeter frame in opposed spaced relation to one another, and at least one fluid flow channel is formed between the fluid permeable screens for fluid communication with a header or tube via the one or more openings in the side edge of said frame, wherein each of said fluid permeable screens is formed from one or more perforated metal sheet, metal mesh or a combination thereof, and includes folds to form a plurality of outwardly extending hollow vanes. In accordance with another aspect of the invention, there is provided a strainer for filtering debris from a fluid comprising: (a) a header defining an enclosed volume and having an outlet in fluid communication with a suction source, said header having a plurality of inlet aperture slots formed therein, and (b) fin-like filter elements projecting outwardly from each aperture slot for filtering debris from said fluid, each said filter element comprising a perimeter frame and a pair of fluid permeable screens fixed thereto in opposed spaced relation, and at least one fluid flow channel therebetween in fluid communication with said enclosed volume through a marginal side edge of said frame and said aperture slot, wherein each of said fluid permeable screens is formed from one or more perforated metal sheet, metal mesh or a combination thereof, and includes folds to form a plurality of outwardly extending hollow vanes. Unless defined otherwise, all technical and scientific terms used herein have the same meaning as commonly understood by one of ordinary skill in the art to which this invention belongs. As used in the specification and claims, the singular forms “a”, “an” and “the” include plural references unless the context clearly dictates otherwise. The term “comprising” as used herein will be understood to mean that the list following is non-exhaustive and may or may not include any other additional suitable items, for example one or more further feature(s), component(s) and/or ingredient(s) as appropriate. Briefly described, the present invention provides a vaned filter element comprising one or more fluid permeable screens formed from at least one layer of porous material that is folded into a plurality of hollow vanes extending outwardly from the outer surface of the each of the one or more fluid permeable screens. The vaned filtering element of the present invention is designed to reduce the space required for strainer installation by increasing strainer surface area per unit volume, while maximizing the quantity of debris that can be filtered from the water. In a specific embodiment, the vaned filter element comprises two layers of porous material that are in the form of two fluid permeable screens in opposed spaced relation to one another. This vaned filter element has at least one fluid flow channel formed between the two fluid permeable screens. In this embodiment, the vaned filter element is referred to as a “fin”. The incorporation of the outwardly extending vanes in the fluid permeable screens of fins increases the filtering surface area of the fins in comparison to a similarly dimensioned Flat-Surface or Corrugated Surface fin and permits the use of small filter holes while minimizing the thin bed effect. In an alternative embodiment, the vaned filter element is configured as a cylinder having outwardly projecting vanes formed (see, for example, FIG. 11). In another embodiment, the vaned filter element is configured such that the vanes are inwardly projecting; this configuration is referred to as a vaned internal-fin filter element (see, for example, FIG. 12). In another embodiment, the vaned filter element is planar, such that the vanes extend outwardly from one side of the filter element (see, for example, FIG. 13). The selection of the appropriate filter element will depend on the ultimate application. The present invention further provides a strainer system that includes one or more vaned filter elements. The following description is based on vaned filter elements that are configured as fins. However, it is understood that this description is not intended to limit the invention to vaned filter elements having a fin configuration. Referring now to the drawings, in which like reference characters indicate like parts throughout the several views, FIG. 1 depicts a vaned filter element as a component of a strainer system according to one embodiment of the present invention. The strainer system depicted in FIG. 1 includes four major components: one or more connection(s) 1 to the pump intake pipe 2; a duct, termed header 3, that collects incoming fluid and directs it to the pump intake, which may be in the floor or on a wall; and vaned filter element 4 with porous and hollow vanes 5. Although the system depicted in FIG. 1 includes only a single vaned filtering element 4, header 3 includes a plurality of slots 6 to accommodate multiple vaned filtering elements, or fins, and to allow the passage of fluid from each of the fins to the header. The vaned filter element, or fin, of the present invention can be designed as a modular attachment to larger structures that are connected to the pump intake, such as ducts and headers. As an alternate arrangement, the vaned filter element can be installed directly to an existing sump where the pump intake 2 is connected. Depending on the amount of debris, composition of debris, flow rate and pump suction head availability, additional fins are added in a modular manner to form a strainer system. The vaned filter element design incorporates spatial flexibility by customizing dimensions A, B and C of each fin as shown in FIG. 1. This flexibility ensures that the available space, for example, in existing nuclear reactors, is used optimally and the strainer system can be built around existing structures. An example of a modular use of the vaned filter element in a strainer system is shown in FIG. 2. FIG. 2 depicts only part of a strainer assembly that includes multiple headers 3 that are connected to corresponding connections 1 to corresponding pump intake pipes 2. Each header 3 collects incoming fluid and directs it to the corresponding pump intake pipe 2. Further, as depicted in FIG. 2, each header 3 includes a plurality of vaned filtering elements 4. Fins can be mounted on one side (as shown in FIG. 1), two sides (as shown in FIG. 2) top, bottom, or a combination of sides, top and/or bottom of the header. Fins can have differing dimensions, and uniform or variable spacing, depending on the particular spatial and filtering requirements of the application. When designing a strainer system that includes a number of vaned fins or filtering elements, it is necessary to optimize the design for the type and quantity of debris that the strainer needs to be able to handle. Two basic factors need to be considered: the filtration area required, and the potential volume of debris that must be accommodated within the strainer. Simplistically, the number of vaned fins is determined by the required filtration area, and then fin spacing can be varied to ensure that there is sufficient space between fins for the postulated debris volume. To prevent air ingestion, it is simplest to ensure that there is sufficient height of water above the fin; however, another option is to design the header such that the flow passage is always submerged. This prevents air ingestion provided (simplistically) that the submergence of the entrance to the header exceeds the head loss across the screen and debris. Another option is to add a horizontal cover over the fins. This cover allows the fins to be closer to the water surface without ingesting air or causing hollow-core vortices. The vaned filter element, or fin, of the present invention is a variation of the Flat Surface Fin of the Finned Strainer technology described in International PCT publication number WO 06/50606. Two configurations of the vaned filter element of the present invention are detailed in the next sections. FIGS. 3(a) and (b) depicts a vaned filtering element, or fin, 4 according to one embodiment of the present invention, in which the vaned fin 4 is easily installed or removed using one or more connection means, such as a pin and a bolt inserted in opening 10. Alternatively, the vaned filtering element of the present invention can be permanently attached to a header to so as to form a permanent component of a strainer system (not shown). As shown in FIG. 3(a), water enters the fin 4 (shown as inflow 19) through a porous filtration screen, leaving debris on the screen, and exits fin 4 through side opening 18 (shown as outflow 20). In FIG. 3(b) the structure of the porous filtration surface of fin 4 is shown to include a filtration surface formed in the shape of vanes 5, to increase the surface area. Two types of vanes can be used based on the design requirements: 1. Perforated Metal Vane: The vanes of the porous filtration screen are formed from a single perforated sheet metal (FIG. 5 and FIG. 6). 2. Layered Metal Mesh Vane: If the application requires fine-particle filtering capability, an alternate porous filtration screen, composed of two layers of woven metal mesh, is used (FIG. 9 and FIG. 10). These are described in more detail below. FIG. 3(b) depicts the interior of a vaned fin 4, without an outer frame. However, in use and as depicted in FIG. 3(a), vaned fin 4, will include outer frame 7 in order to provide mechanical stability and to define part of the at least one fluid flow channel within the vaned fin 4. In some embodiments, the frame itself can include perforations or be formed from a rigid porous material to provide additional filtering. Alternatively, the frame can be made from a rigid material that is impermeable to fluid. In each case, however, the frame includes an elongated opening 18, or a series of smaller openings (not shown) along one side, which corresponds with a slot or opening in the header to which it is, or is to be, attached. It is through this opening that fluid flows from the main flow channel 13 within vaned fin 4, into the header. FIG. 4 is a photograph of a vaned fin as depicted in FIG. 3(a), and which includes perforated metal vanes. As depicted in FIG. 5 and FIG. 9, outer frame 7 of vaned fin 4 can include an end cap 17 at the header end of vaned fin 4. End cap 17 is formed from a plate that is welded over one end of corrugated plate 15. End cap 17 includes a large opening 18 through which flow exits fin 4 and enters the header. End cap is also attached to C-shaped mounting frame 16, which can be unitary or made up of separate plates, to form the overall outer frame 7. The edges of the end cap are sealed into the collection header with flexible metal strips 21, which ensure a good fit of the fin into the header. Perforated Metal Vanes FIG. 5 shows an exploded view of vaned filter element 4 according to one embodiment of the present invention, which includes fluid permeable screens made from vaned perforated metal. In this configuration the porous vaned surfaces are formed from a single perforated metal plate. The perforated metal plate vane surface is simple to manufacture and adds significantly to the strength of the fin. The two vaned surfaces 14 on each side of the fin are separated by a corrugated plate 15 that provides stiffness and strength, and forms flow channels for fluid communication with the collection header. Edges of the vaned fin are covered by a mounting frame 16 around all sides except the edge that fits into the header. This edge includes one or more openings to facilitate fluid communication between the flow channels and the interior of the header. This frame also adds to the structural strength of the fin. The mounting frame 16 can be fully or partially perforated if extra screen area is needed. The details of the perforated metal vane are shown in FIG. 6. As illustrated in this figure, after entering the vanes, water flows through the hollow core 12 of the vanes towards the main flow channel 13 of the fin. With the optimal use of vane spacing, the thin-bed effect is minimized. Layered Metal Mesh Vanes If the application of the vaned filter element requires smaller filtration holes than are achievable using standard perforated metal plate, an alternative is to manufacture the vanes using woven metal mesh. As shown in FIG. 7(a), water enters the fin 4 (shown as inflow 19) through a porous filtration screen, leaving debris on the screen, and exits fin 4 through side opening 18 (shown as outflow 20). In FIG. 7(b) the structure of the porous filtration surface of fin 4 is shown to include a filtration surface formed in the shape of vanes 5, to increase the surface area. In the embodiment depicted in FIG. 7, the vaned surfaces are formed from a layered metal mesh. The two vaned surfaces are attached to the flat perforated plate 23. FIG. 7(b) depicts the interior of a vaned fin 4, without an outer frame. However, in use and as depicted in FIG. 7(a), vaned fin 4, will include outer frame 7 in order to provide mechanical stability and to define part of the at least one fluid flow channel within the vaned fin 4. In some embodiments, the frame itself can include perforations or be formed from a rigid porous material to provide additional filtering. Alternatively, the frame can be made from a rigid material that is impermeable to fluid. In each case, however, the frame includes an elongated opening 18, or a series of smaller openings (not shown) along one side, which corresponds with a slot or opening in the header to which it is, or is to be, attached. It is through this opening that fluid flows (identified as outflow 20) from the main flow channel 13 within vaned fin 4, into the header. FIG. 8 is a photograph of a vaned fin as depicted in FIG. 3(a), and which includes perforated metal vanes. FIG. 9 shows an exploded view of the vaned filter element according to one embodiment of the present invention, which incorporates layered metal mesh. Using this configuration, filtration of debris smaller than 200 μm is possible. After the initial accumulation of a thin fibre mat, much smaller size debris (nominal size of 10 μm) are filtered. The mesh size can be changed depending on the requirements of the specific application. The two vaned surfaces on each side of the fin are composed of two layers of woven metal mesh surfaces 24 and 25 as shown in FIG. 9. The outer mesh surface 25 provides the fine-particle filtering capability. This mesh is very flexible on its own and requires structural enhancement for consistent vane forming. The inner mesh surface 24 is selected to increase the stiffness and strength of the layered surface. The combined fine-mesh surface 25 and the course-mesh surface 24 provide fine filtering capability with sufficient structural strength. The two mesh layers are attached prior to forming the vanes at multiple locations such that they behave as one surface. This layered mesh surface is used to manufacture the two faces of the fin with formed hollow vanes. The details of the vanes are shown in FIG. 10. The two vaned surfaces are attached to the flat perforated plate 23. The combined vaned face is structurally strong. The corrugated plate 15 adds significant additional stiffness and strength, and forms flow channels for fluid communication with the collection header. Edges of the vaned fin are covered by a mounting frame 16 around all sides except the edge that fits into the header. This edge includes one or more openings to facilitate fluid communication between the flow channels and the interior of the header. This frame also adds to the structural strength of the fin. The mounting frame 16 can be fully or partially perforated if extra screen area is needed. The details of the layered metal mesh vane is shown in FIG. 10. As illustrated in this figure, after entering the vanes, water flows through the hollow core 12 of the vanes towards the main flow channel 13 of the fin. With the optimal use of vane spacing, the thin-bed effect is minimized. Vane Dimensions The standard vane dimensions are selected based on the typical debris composition expected in nuclear power plants following a LOCA. The spacing between vanes is determined with the consideration of maximizing the filtering surface area without any loss of filtering capability. If the vane spacing is too large, the surface area of the fin will be non-optimal, resulting in a non-compact strainer design. If the vane spacing if too small, deposited debris on the vane surfaces will bridge across the vanes resulting in a reduction in the effective filtering area and, hence, a reduction in filtering efficiency. For this reason, the proper selection of vane spacing is important for optimal design. It is commonly accepted that the typical thin bed debris thickness is about 3 mm. As an example, if the true thin bed for a particular application were 3 mm, a vane pitch of approximately 14 mm with 10 mm space between two adjacent vanes could be appropriate in order to allow the formation of the thin bed without bridging between vanes. These dimensions can be optimized for different applications by laboratory testing. The selection of vane height is also determined by consideration of maximizing filtering surface area while maintaining filtering efficiency. Although the fin surface area can be increased indefinitely by increasing filtration area, there is a limit beyond which additional vane height will not provide any benefit for debris filtration. When the debris head loss becomes dominated by the thick-bed head loss, the increased vane height provides no further improvement in filtering capability. This is because the thick-bed debris fills the space between vanes and the thick-bed head loss becomes a function of the projected area (e.g., fin surface area without the vanes) that is independent of the vane height. The standard vane height used for nuclear ECCS strainer applications is roughly 25 mm, but this dimension has to be optimized with lab testing to include the effects of debris composition and debris quantity per unit area. Increase in Surface Area Compared to the Flat-Surface Fin and Corrugated Surface Fin of the Finned Strainer technology, the present Vaned Fin invention provides about 4.5 and 2.5 times the surface area per unit volume, respectively. The significant increase in the surface area minimizes the “thin-bed effect” by distributing the thin bed debris deposition over a larger surface area. All publications, patents and patent applications mentioned in this Specification are indicative of the level of skill of those skilled in the art to which this invention pertains and are herein incorporated by reference to the same extent as if each individual publication, patent, or patent applications was specifically and individually indicated to be incorporated by reference. The invention being thus described, it will be obvious that the same may be varied in many ways. Such variations are not to be regarded as a departure from the spirit and scope of the invention, and all such modifications as would be obvious to one skilled in the art are intended to be included within the scope of the following claims.
	description	The present application is related to and claims the benefit of the earliest available effective filing date(s) from the following listed application(s) (the “Related Applications”) (e.g., claims earliest available priority dates for other than provisional patent applications or claims benefits under 35 USC §119(e) for provisional patent applications, for any and all parent, grandparent, great-grandparent, etc. applications of the Related Application(s)). All subject matter of the Related Applications and of any and all parent, grandparent, great-grandparent, etc. applications of the Related Applications is incorporated herein by reference to the extent such subject matter is not inconsistent herewith. The present application constitutes a continuation-in-part of U.S. patent application No. 12/386,524, entitled A NUCLEAR FISSION REACTOR FUEL ASSEMBLY AND SYSTEM CONFIGURED FOR CONTROLLED REMOVAL OF A VOLATILE FISSION PRODUCT AND HEAT RELEASED BY A BURN WAVE IN A TRAVELING WAVE NUCLEAR FISSION REACTOR AND METHOD FOR SAME, naming Charles E. Ahlfeld; John Rogers Gilleland; Roderick A. Hyde; Muriel Y. Ishikawa; David G. McAlees; Nathan P. Myhrvold; Clarence T. Tegreene; Thomas Allan Weaver; Charles Whitmer; Victoria Y. H. Wood; Lowell L. Wood, Jr.; and George B. Zimmerman as inventors, filed Apr. 16, 2009, which is currently co-pending, or is an application of which a currently co-pending application is entitled to the benefit of the filing date. This application generally relates to nuclear reactor fuel assemblies and more particularly relates to a nuclear fission reactor fuel assembly and system configured for controlled removal of a volatile fission product and heat released by a burn wave in a traveling wave nuclear fission reactor and method for same. It is known that, in an operating nuclear fission reactor, neutrons of a known energy are captured by nuclides having a high atomic mass. The resulting compound nucleus separates into fission products that include two lower atomic mass fission fragments and also decay products. Nuclides known to undergo such fission by neutrons of all energies include uranium-233, uranium-235 and plutonium-239, which are fissile nuclides. For example, thermal neutrons having a kinetic energy of 0.0253 eV (electron volts) can be used to fission U-235 nuclei. Fission of thorium-232 and uranium-238, which are fertile nuclides, will not undergo induced fission, except with fast neutrons that have a kinetic energy of at least 1 MeV (million electron volts). The total kinetic energy released from each fission event is about 200 MeV. This kinetic energy is eventually transformed into heat. Moreover, the fission process, which starts with an initial source of neutrons, liberates additional neutrons as well as transforms kinetic energy into heat. This results in a self-sustaining fission chain reaction that is accompanied by continued release of heat. For every neutron that is absorbed, more than one neutron is liberated until the fissile nuclei are depleted. This phenomenon is used in a commercial nuclear reactor to produce continuous heat that, in turn, is used to generate electricity. Attempts have been made to address fission product accumulation during reactor operation. U.S. Pat. No. 4,285,891, issued Aug. 25, 1981 in the names of Lane A. Bray et al. and titled “Method of Removing Fission Gases from Irradiated Fuel” discloses a method for removing volatile fission products from irradiated fuel by first passing a hydrogen-containing inert gas by the fuel which is heated to an elevated temperature of at least 1000° C. and then passing inert gas alone by the fuel which is at the elevated temperature. Another approach is disclosed in U.S. Pat. No. 5,268,947, issued Dec. 7, 1993 in the names of Bernard Bastide et al. and titled “Nuclear Fuel Elements Comprising a Trap for Fission Products Based on Oxide”. This patent discloses a nuclear fuel element comprising sintered pellets which are surrounded by a metallic sheath and permitting trapping of the fission products characterized in that the pellets contain or are coated with or that the sheath is internally coated with an agent for trapping the fission products. The fission products are trapped by forming with the trapping agent oxygenated compounds which are stable at high temperature. According to an aspect of this disclosure, there is provided a nuclear fission reactor fuel assembly configured for controlled removal of a volatile fission product released by a burn wave in a traveling wave nuclear fission reactor, comprising an enclosure adapted to enclose a porous nuclear fuel body and a fluid control subassembly coupled to the enclosure and adapted to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body. According to an aspect of this disclosure, there is provided a nuclear fission reactor fuel assembly configured for controlled removal of a volatile fission product released by a burn wave in the nuclear fission reactor fuel assembly, comprising an enclosure adapted to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of pores having the volatile fission product therein and a fluid control subassembly coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and for controllably removing at least a portion of the heat generated by the nuclear fuel body. According to an aspect of this disclosure, there is provided a system for controlled removal of a volatile fission product released by presence of a burn wave in a nuclear fission reactor fuel assembly, comprising an enclosure adapted to enclose a porous nuclear fuel body defining a plurality of pores having the volatile fission product therein and a fluid control subassembly coupled to the enclosure to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body. According to an aspect of this disclosure, there is provided a system for controlled removal of a volatile fission product released by presence of a burn wave in a nuclear fission reactor fuel assembly, comprising an enclosure adapted to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores having the volatile fission product therein and a fluid control subassembly coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and for controllably removing at least a portion of the heat generated by the nuclear fuel body. According to an aspect of this disclosure, there is provided a method of assembling a nuclear fission reactor fuel assembly configured for controlled removal of a volatile fission product released by a burn wave in a traveling wave nuclear fission reactor, comprising providing an enclosure to enclose a porous nuclear fuel body and coupling a fluid control subassembly to the enclosure to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality locations corresponding to the burn wave. According to an aspect of this disclosure, there is provided a method of assembling a nuclear fission reactor fuel assembly configured for controlled removal of a volatile fission product released by a burn wave in a traveling wave nuclear fission reactor, comprising providing an enclosure to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores and coupling a fluid control subassembly to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in regions of the traveling wave nuclear fission reactor proximate to locations corresponding to the burn wave. According to an aspect of this disclosure, there is provided a method comprising controlling removal of a volatile fission product at a plurality of locations corresponding to a burn wave of a traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. According to an aspect of this disclosure, there is provided a method of operating a nuclear fission reactor fuel assembly configured for controlled removal of a volatile fission product released by a burn wave in a traveling wave nuclear fission reactor, comprising using an enclosure enclosing a porous nuclear fuel body having the volatile fission product therein and using a fluid control subassembly coupled to the enclosure to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. According to an aspect of this disclosure, there is provided a method of operating a nuclear fission reactor fuel assembly configured for controlled removal of a volatile fission product released by a burn wave in a traveling wave nuclear fission reactor, comprising using an enclosure enclosing a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores and using a fluid control subassembly coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. A feature of the present disclosure is the provision, for use in a traveling wave nuclear fission reactor, of an enclosure adapted to enclose a porous nuclear fuel body having the volatile fission product therein. Another feature of the present disclosure is the provision, for use in a traveling wave nuclear fission reactor, of a fluid control subassembly coupled to the enclosure and adapted to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body. Yet another feature of the present disclosure is the provision, for use in a traveling wave nuclear fission reactor, of a fluid control subassembly coupled to the enclosure for controllably removing at least a portion of the heat generated by the nuclear fuel body. Still another feature of the present disclosure is the provision, for use in a traveling wave nuclear fission reactor, of a dual-purpose circuit coupled to the enclosure for selectively removing the volatile fission product and the heat from the nuclear fuel body. In addition to the foregoing, various other method and/or device aspects are set forth and described in the teachings such as text (e.g., claims and/or detailed description) and/or drawings of the present disclosure. The foregoing is a summary and thus may contain simplifications, generalizations, inclusions, and/or omissions of detail; consequently, those skilled in the art will appreciate that the summary is illustrative only and is not intended to be in any way limiting. In addition to the illustrative aspects, embodiments, and features described above, further aspects, embodiments, and features will become apparent by reference to the drawings and the following detailed description. In the following detailed description, reference is made to the accompanying drawings, which form a part hereof. In the drawings, similar symbols typically identify similar components, unless context dictates otherwise. The illustrative embodiments described in the detailed description, drawings, and claims are not meant to be limiting. Other embodiments may be utilized, and other changes may be made, without departing from the spirit or scope of the subject matter presented herein. In addition, the present application uses formal outline headings for clarity of presentation. However, it is to be understood that the outline headings are for presentation purposes, and that different types of subject matter may be discussed throughout the application (e.g., device(s)/structure(s) may be described under process(es)/operations heading(s) and/or process(es)/operations may be discussed under structure(s)/process(es) headings; and/or descriptions of single topics may span two or more topic headings). Hence, the use of the formal outline headings is not intended to be in any way limiting. Moreover, the herein described subject matter sometimes illustrates different components contained within, or connected with, different other components. It is to be understood that such depicted architectures are merely exemplary, and that in fact many other architectures may be implemented which achieve the same functionality. In a conceptual sense, any arrangement of components to achieve the same functionality is effectively “associated” such that the desired functionality is achieved. Hence, any two components herein combined to achieve a particular functionality can be seen as “associated with” each other such that the desired functionality is achieved, irrespective of architectures or intermedial components. Likewise, any two components so associated can also be viewed as being “operably connected”, or “operably coupled,” to each other to achieve the desired functionality, and any two components capable of being so associated can also be viewed as being “operably couplable,” to each other to achieve the desired functionality. Specific examples of operably couplable include but are not limited to physically mateable and/or physically interacting components, and/or wirelessly interactable, and/or wirelessly interacting components, and/or logically interacting, and/or logically interactable components. In some instances, one or more components may be referred to herein as “configured to,” “configurable to,” “operable/operative to,” “adapted/adaptable,” “able to,” “conformable/conformed to,” etc. Those skilled in the art will recognize that “configured to” can generally encompass active-state components and/or inactive-state components and/or standby-state components, unless context requires otherwise. Heat build-up during reactor operation may cause fuel assemblies to undergo expansion leading to misalignment of reactor core components, fuel cladding creep that can increase risk of fuel cladding rupture and fuel swelling during reactor operation. This may increase the risk that the fuel might crack or otherwise degrade. Fuel cracking may precede fuel-cladding failure mechanisms, such as fuel-clad mechanical interaction, and lead to fission gas release. The fission gas release results in higher than normal radiation levels. Fission products are generated during the fission process and may accumulate in the fuel. Accumulation of fission products, including fission gas, may lead to an undesirable amount of fuel assembly expansion. Such fuel assembly expansion may, in turn, increase the risk of fuel cracking and concomitant release of fission products into the surrounding environment. Although safety margins incorporated into the reactor design and precise quality control during manufacture reduce these risks to a minimal level, in some cases, it may still be appropriate to reduce these risks even further. Therefore, referring to FIG. 1, there is shown a first embodiment nuclear fission reactor fuel assembly and system, generally referred to as 10, for producing heat due to fission of a fissile nuclide, such as uranium-235, uranium-233 or plutonium-239, or due to fast-fission of a nuclide such as thorium-232 or uranium-238. It will be understood from the description hereinbelow that fuel assembly 10 is also capable of controlled removal of a volatile fission product 15 produced during the fission process. Volatile fission product 15 is produced by a traveling burn wave 16 that is initiated by a comparatively small and removable nuclear fission igniter 17. In this regard, nuclear fission igniter 17, that includes a moderate isotopic enrichment of nuclear fissionable material, such as, without limitation, U-233, U-235 or Pu-239, is suitably located at a predetermined location in fuel assembly 10. Neutrons are released by igniter 17. The neutrons that are released by igniter 17 are captured by fissile and/or fertile material within nuclear fission fuel assembly 10 to initiate a fission chain reaction. Igniter 17 may be removed once the chain reaction becomes self-sustaining, if desired. It may be appreciated that volatile fission product 15 can be controllably released in response to the controlled positioning of burn wave 16 in nuclear fission reactor fuel assembly 10. It should be understood that any of the embodiments of the fuel assembly described herein may be used as a component of a traveling wave nuclear fission reactor. Such a traveling wave nuclear fission reactor is disclosed in detail in co-pending U.S. patent application Ser. No. 11/605,943 filed Nov. 28, 2006 in the names of Roderick A. Hyde, et al. and titled “Automated Nuclear Power Reactor For Long-Term Operation”, which application is assigned to the assignee of the present application, the entire disclosure of which is hereby incorporated by reference, now abandoned. Still referring to FIG. 1, fuel assembly 10 comprises an enclosure 20 having enclosure walls 30 for sealingly enclosing a porous nuclear fuel body 40 therein. Fuel body 40 comprises the aforementioned fissile nuclide, such as uranium-235, uranium-233 or plutonium-239. Alternatively, fuel body 40 may comprise the aforementioned fertile nuclide, such as thorium-232 and/or uranium-238, which will be transmuted during the fission process into one or more of the fissile nuclides mentioned hereinabove. A further alternative is that fuel body 40 may comprise a predetermined mixture of fissile and fertile nuclides. As described in more detail hereinbelow, fuel body 40 is capable of producing volatile fission product 15, which may be isotopes of iodine, bromine, cesium, potassium, rubidium, strontium, xenon, krypton, barium and mixtures thereof or other gaseous or volatile materials. Referring again to FIG. 1, as previously mentioned, porous nuclear fuel body 40 may substantially comprise a metal, such as uranium, thorium, plutonium, or alloys thereof. More specifically, nuclear fuel body 40 may be a porous material made from an oxide selected from the group consisting essentially of uranium monoxide (UO), uranium dioxide (UO2), thorium dioxide (ThO2) (also referred to as thorium oxide), uranium trioxide (UO3), uranium oxide-plutonium oxide (UO—PuO), triuranium octoxide (U3O8) and mixtures thereof. Alternatively, fuel body 40 may substantially comprise a carbide of uranium (UCx) or a carbide of thorium (ThCx). For example, fuel body 40 may be a foam material made from a carbide selected from the group consisting essentially of uranium monocarbide (UC), uranium dicarbide (UC2), uranium sesquicarbide (U2C3), thorium dicarbide (ThC2), thorium carbide (ThC) and mixtures thereof. The uranium carbide or thorium carbide may be sputtered into a matrix of niobium carbide (NbC) and zirconium carbide (ZrC), so as to form fuel body 40. A potential benefit of using niobium carbide and zirconium carbide is that they form a refractory structural substrate for the uranium carbide or thorium carbide. As another example, fuel body 40 may be a porous material made from a nitride selected from the group consisting essentially of uranium nitride (U3N2), uranium nitride-zirconium nitride (U3N2—Zr3N4), uranium-plutonium nitride ((U—Pu)N), thorium nitride (ThN), uranium-zirconium alloy (UZr) and mixtures thereof. As best seen in FIGS. 2 and 2A, porous fuel body 40 may define a plurality of interconnected open-cell pores 50 spatially distributed within fuel body 40. As used herein, the terminology “open-cell pores” means that each pore 50 is interconnected with one or more neighboring pores 50, thereby permitting fluid, such as gas or liquid, to directly travel between pores 50. That is, open-cell pores 50 are disposed within fuel body 40 so as to form a fibrous, rod-like, web-like or honeycomb structure. Alternatively, fuel body 40 may comprise a porous fuel material formed by a collection of fuel particles 63 (such as sintered beads or packed spheres) that define a plurality of interstitial channels 65 therebetween. Also, open-cell pores 50 may be disposed within fuel material having a mixture of foam and porous characteristics. It should be understood that the description hereinbelow pertaining to pores 50 also applies to channels 65. Referring again to FIGS. 2 and 2A, it may be appreciated that volatile fission product 15 that is produced by burn wave 16 may initially reside in some or all of pores 50 and can naturally vaporize and diffuse through nuclear fuel body 40. It also may be appreciated that at least some of pores 50 are of a predetermined configuration for allowing at least a portion of volatile fission product 15 to escape pores 50 of porous nuclear fuel body 40 within a predetermined response time. The predetermined response time may be between approximately 10 seconds and approximately 1,000 seconds. Alternatively, the predetermined response time may be between approximately one second and approximately 10,000 seconds depending on the predetermined configuration of pores 50. Returning to FIG. 1, coupled to enclosure 20, such as by a first pipe segment 70, is a fluid control subassembly 80 that defines a first volume 90 containing a first fluid, such as pressurized helium gas. Alternatively, the first fluid may be any suitable pressurized inert gas, such as, without limitation, neon, argon, krypton, xenon, and mixtures thereof Another alternative is the first fluid may be a suitable liquid, such as liquid lead (Pb), sodium (Na), lithium (Li), mercury (Hg) or similar liquids or liquid mixtures. As described more fully hereinbelow, fluid control subassembly 80 assists in controllably removing volatile fission product 15 and heat from fuel body 40. In other words, fluid control subassembly 80 is capable of circulating the first fluid through porous nuclear fuel body 40. In this manner, heat and volatile fission product 15 are removed from fuel body 40 while the first fluid circulates through fuel body 40. Turning now to FIG. 3, a second embodiment nuclear fission reactor fuel assembly and system, generally referred to as 100, is there shown. This second embodiment fuel assembly 100 is substantially similar to first embodiment fuel assembly 10, except that a heat exchanger 110 is associated with enclosure 20. Heat exchanger 110 comprises a shell 120 defining an interior 130 capable of containing a second fluid for cooling the first fluid that is used to remove heat and volatile fission product 15 from fuel body 40. The second fluid has a temperature lower than the temperature of the first fluid. Disposed within interior 130 are a plurality of U-shaped tubes 132 (only one of which is shown) having two open ends. In this regard, one end of U-shaped tube 132 has an opening 134 and the other end of U-shaped tube 132 has another opening 136. Openings 134 and 136 are in fluid communication with the first fluid occupying first volume 90 of fluid control subassembly 80. It may be appreciated that there is a density difference between the cooled portion of first fluid residing within tubes 132 and the heated portion of the first fluid in porous nuclear fuel body 40. This temperature difference will give rise to a difference in density between the cooled portion of the first fluid residing within tubes 132 and the heated portion of the first fluid in porous nuclear fuel body 40. The difference in fluid densities will, in turn, cause the molecules of the cooler fluid portion to be exchanged with the molecules of the hotter fluid portion because the cooler fluid portion is located physically higher than or above the hotter fluid portion. Thus, an interchange of cooler and hotter fluid portions will occur and cause a natural convective current that will circulate the first fluid through fuel assembly 100 and nuclear fuel body 40. Moreover, tubes 132 are U-shaped to increase heat transfer surface area to enhance this natural convection. Thus, natural convection is relied upon to circulate the first fluid due to the substantial temperature difference between the cooler and hotter portions of the first fluid. As the first fluid circulates through tubes 132, the second fluid, which is at a substantially lower temperature than the first fluid, will be caused to enter interior 130 through an inlet nozzle 140, such as by means of a pump (not shown). The second fluid will then exit interior 130 through an outlet nozzle 150. As the second fluid enters and exits heat exchanger 110, the lower temperature second fluid will surround the plurality of U-shaped tubes 132. Conductive heat transfer, through the walls of tubes 132, will occur between the first fluid circulating in tubes 132 and the second fluid surrounding tubes 132. In this manner, the heated first fluid will give up its heat to the cooler second fluid. Referring again to FIG. 3, this second embodiment fuel assembly 100 may be operable with no pumps or valves to circulate the first fluid because the first fluid can be circulated by means of natural convection. Absence of pumps and valves may increase reliability of second embodiment fuel assembly 100 while reducing costs of manufacture and maintenance of second embodiment fuel assembly 100. Still referring to FIG. 3, heat exchanger 110 may serve as a steam generator, if desired. That is, depending on the temperature and pressure within heat exchanger 110, a portion of the second fluid can vaporize to steam (when the second fluid is water) which exits outlet nozzle 150. The steam exiting outlet nozzle 150 can be transported to a turbine-generator device (not shown) for producing electricity in a manner well known in the art of electricity generation from steam. Referring to FIG. 4, there is shown a third embodiment nuclear fission reactor fuel assembly and system, generally referred to as 190, intended primarily for removing heat and volatile fission products 15 from fuel body 40. Third embodiment nuclear fission reactor fuel assembly 190 comprises a second pipe segment 200 that is in communication with first volume 90 at one end of second pipe segment 200 and is integrally connected at the other end of second pipe segment 200 to an inlet of a first pump 210, which may be a centrifugal pump. Such a pump suitable for this purpose may be of a type that may be available, for example, from Sulzer Pumps, Ltd. located in Winterthur, Switzerland. An outlet of first pump 210 is connected to a third pipe segment 220, which in turn is in communication with fuel body 40. Moreover, heat exchanger 110 may be coupled to third pipe segment 220 for removing heat from the fluid flowing through third pipe segment 220. Still referring to FIG. 4, to remove heat from fuel body 40, first pump 210 is activated. First pump 210 will draw fluid, such as the previously mentioned helium gas, from second pipe segment 200 and thus from first volume 90, which is defined by fluid control subassembly 80. First pump 210 will pump the fluid through third pipe segment 220. The fluid flowing through third pipe segment 220 is received by the plurality (or multiplicity) of open-cell pores 50 that are defined by fuel body 40. The fluid flowing through open-cell pores 50 will acquire the heat produced by fuel body 40. The heat is acquired by means of forced convective heat transfer as the fluid is pumped through open-cell pores 50 by means of first pump 210. As first pump 210 is operated, the fluid flowing through fuel body 40 and that is experiencing the convective heat transfer, is drawn, due to the pumping action of pump 210, through first pipe segment 70, into first volume 90, through second pipe segment 200 and thence into third pipe segment 220 where the heat is removed by heat exchanger 110. Also, while fluid circulates between fuel body 40 and first volume 90, a portion of volatile fission products 15 originating in fuel body 40 can be scavenged and retained within first volume 90 thereby removing or at least lowering the amount of fission product 15 present in fuel body 40. In this regard, first volume 90 may be lined with a fission product scavenging material 225 which retains fission product 15 as the fission product removal fluid enters volume 90. The fission product scavenging material may be, with limitation, silver zeolite (AgZ) for removing Xenon (Xe) and Krypton (Kr) or the fission product scavenging material may be, without limitation, metallic oxides of silicon dioxide (SiO2) or titanium dioxide (TiO2) for removing radioisotopes of cesium (Cs), rubidium (Rb), iodine (IA tellurium (Te) and mixtures thereof. A benefit of using this third embodiment fuel assembly 190 is that only a pump 210 is required to circulate the first fluid. No valves are needed. Absence of valves may increase reliability of third embodiment fuel assembly 190 while reducing costs of manufacture and maintenance of third embodiment fuel assembly 190. Referring to FIG. 5, a fourth embodiment nuclear fission reactor fuel assembly and system, generally referred to as 230, is capable of further enhancing removal of the previously mentioned volatile fission product 15 as well as heat from fuel body 40. Fourth embodiment nuclear fission reactor fuel assembly 230 is substantially similar to third embodiment nuclear fission reactor fuel assembly 190, except that means is added for enhanced removal of heat and volatile fission product 15. In this regard, a fourth pipe segment 240 has an end thereof in communication with first volume 90 and another end thereof integrally coupled to an intake of a second pump 250. A discharge of second pump 250 is integrally coupled to a sixth pipe segment 260. The sixth pipe segment 260 in turn is in communication with a second volume 270 defined by a first fission product reservoir or holding tank 280. During operation of fourth embodiment fuel assembly 230, pump 210 will pump the first fluid from first volume 90, through second pipe segment 200, through third pipe segment 220, through fuel body 40, through first pipe segment 70 and back into first volume 90. As the first fluid flows through third pipe segment 220, the fluid will surrender its heat to the second fluid in heat exchanger 110. First pump 210 may then be caused to cease operation after a predetermined amount of time. Second pump 250 may then be operated to draw the fission product 15, including the first fluid intermingled therewith, through fourth pipe segment 240, through fifth pipe segment 260 and into second volume 270 that is defined by first fission product reservoir or holding tank 280. Thus, volatile fission product 15 will have been removed from fuel body 40 and then retained in first fission product reservoir or holding tank 280 for subsequent off-site disposal or the fission product 15 in reservoir or holding tank 280 may remain in situ, if desired. In this fourth embodiment fuel assembly 230 only pumps 210/250 are required. No valves are needed. Absence of valves may increase reliability of fourth embodiment fuel assembly 230 while reducing costs of manufacture and maintenance of fourth embodiment fuel assembly 230. Another benefit of fourth embodiment fuel assembly 230 is that volatile fission products 15 are isolated in second volume 270 and can be removed for subsequent off-site disposal or left in place. Referring to FIG. 6, there is shown a fifth embodiment nuclear fission reactor fuel assembly and system, generally referred to as 290. In this regard, there may be a plurality of fifth embodiment nuclear fission reactor fuel assemblies 290 (only three of which are shown). A sealable vessel 310, such as a pressure vessel or containment vessel, surrounds nuclear fission reactor fuel assemblies 290 for preventing leakage of radioactive particles, gasses or liquids from fuel assembly 290 to the surrounding environment. Vessel 310 may be steel, concrete or other material of suitable size and thickness to reduce risk of such radiation leakage and to support required pressure loads. Although only one vessel 310 is shown, there may be additional containment vessels surrounding vessel 310, one enveloping the other, for added assurance that leakage of radioactive particles, gasses or liquids from nuclear fission reactor fuel assembly 290 is prevented. Vessel 310 defines a well 320 therein in which is disposed fifth embodiment nuclear fission reactor fuel assemblies 290. Fifth embodiment nuclear fission reactor fuel assembly 290 is capable of controlled removal of heat build-up and also controlled removal of volatile fission product 15, as described more fully hereinbelow. Referring again to FIG. 6, fuel assembly 290 comprises a compact, combined, closed-loop, dual-purpose heat removal and volatile fission product removal circuit, generally referred to as 330. Dual-purpose circuit 330 is capable of selectively removing heat as well as volatile fission products 15 from fuel body 40. In this regard, circuit 330 may be operated to first remove volatile fission products 15 and then remove heat, or vice versa. Thus, circuit 330 is capable of consecutively removing heat and fission products 15. Referring yet again to FIG. 6, dual-purpose circuit 330 comprises the previously mentioned fluid control subassembly 80 that defines first volume 90 containing the fluid supply. First pipe segment 70 is in communication with fuel body 40 at one end of first pipe segment 70 and is integrally coupled at the other end of first pipe segment 70 to an inlet of a third pump 340, which may be a centrifugal pump. The outlet of third pump 340 is connected to a sixth pipe segment 350, which in turn is in communication with first volume 90. Second pipe segment 200 is in communication with first volume 90 at one end of second pipe segment 200 and is integrally connected to an inlet of first pump 210 at the other end of second pipe segment 200. It is appreciated that pumps 340 and 210 may be selected so that either pump 340 or pump 210 operating alone is capable of circulating a reduced but sufficient flow rate of the fluid within dual-purpose circuit 330. That is, even if either pump 340 or pump 210 is absent, turned off, or otherwise non-functioning, dual purpose circuit will still retain a capability of fluid circulation through dual-purpose circuit 330. A heat exchanger 355 is disposed in third pipe segment 220 between a seventh pipe segment 360 and enclosure 20 for removing heat from the fluid as the fluid circulates through dual-purpose circuit 330. Heat exchanger 355 may be substantially similar in configuration to heat exchanger 110. Connected to any one of the pipe segments 70/200/220/350, such as to seventh pipe segment 360, is a second volatile fission product reservoir or holding tank 370. Second reservoir or holding tank 370 defines a third volume 380 for holding and isolating volatile fission products 15 therein. Second reservoir or holding tank 370 is coupled to third pipe segment 220 by seventh pipe segment 360. Operatively connected to seventh pipe segment 360 is a motor-operated first back-flow prevention valve 390 for allowing flow of volatile fission products 15 into third volume 380; but, not for allowing reverse flow of volatile fission products 15 from third volume 380. Motor-operated first back-flow prevention valve 390 may be operable by action of a controller or control unit 400 electrically connected thereto. Alternatively, valve 390 need not be motor-operated, but may be operated by suitable other means. Such a back-flow prevention valve suitable for this purpose may be available from, for example, Emerson Process Manufacture, Ltd. located in Baar, Switzerland. As described in more detail hereinbelow, volatile fission products 15 produced by fuel body 40 will be captured and held within third volume 380 in order to isolate volatile fission products 15. Still referring to FIG. 6, operatively connected to third pipe segment 220 and interposed between first back-flow prevention valve 390 and enclosure 20 is a motor-operated second back-flow prevention valve 410. Second back-flow prevention valve 410 allows flow of fluid into enclosure 20; but, does not allow reverse flow of fluid from enclosure 20 back into third pipe segment 220. Motor-operated second back-flow prevention valve 410 may be operable by action of control unit 400 electrically connected thereto. Thus, first pipe segment 70, third pump 340, sixth pipe segment 350, heat exchanger 355, fluid control subassembly 80, second pipe segment 200, first pump 210, third pipe segment 220, seventh pipe segment 360, second fission product reservoir or holding tank 370, first back-flow prevention valve 390, second back-flow prevention valve 410, control unit 400 and fuel body 40 together define dual-purpose circuit 330. As described in more detail presently, dual-purpose circuit 330 is capable of circulating the fluid through open-cell pores 50 of fuel body 40, so that the heat and volatile fission products 15 are selectively removed from fuel body 40 either consecutively or simultaneously. It should be understood from the description herein that a benefit of this fifth embodiment nuclear fission reactor fuel assembly 290 is that dual-purpose circuit 330 can selectively consecutively remove volatile fission products 15 and heat by controlled operation of pumps 210/340, valves 390/410 and control unit 400. Referring again to FIG. 6, a plurality of sensors or neutron flux detectors 412 (only one of which is shown) may be disposed in fuel body 40 for detecting various operating characteristics of fuel body 40. By way of example only, and not by way of limitation, detector 412 may be adapted to detect the operating characteristics of neutron population level, power level and/or position of burn wave 16 in fuel body 40. Detector 412 is coupled to control unit 400, which control unit 400 controls operation of detector 412. In addition, a plurality of fission product pressure detectors 413 (only one of which is shown) may be disposed in fuel body 40 for detecting fission product pressure level in fuel body 40. Moreover, it should be appreciated that control unit 400 is capable of operating valves 390 and 410 to control release of volatile fission product 15 and heat according to the amount of time nuclear fission reactor fuel assembly 290 is continuously or periodically operated and/or according to any time schedule associated with nuclear fission reactor fuel assembly 290. A controller suitable for use as control unit 400 might be of a type that may be available from, for example, Stolley and Orlebeke, Incorporated located in Elmhurst, Ill., U.S.A. Moreover, neutron flux detectors suitable for this purpose may be available from Thermo Fisher Scientific, Incorporated located in Waltham, Mass. U.S.A. In addition, suitable pressure detectors may be available from Kaman Measuring Systems, Incorporated located in Colorado Springs, Colo. U.S.A. As shown in FIGS. 6A and 6B, a first embodiment diaphragm valve, generally referred to as 414a, having a hollow valve body 415 may be substituted for valves 390 and/or 410, if desired. Alternatively, the previously mentioned back-flow prevention valve 390 or 410 may be used in combination with first embodiment diaphragm valve 414a, as shown. Disposed within hollow valve body 415 is a plurality of breakable barriers or membranes 416, which may be made of a thin elastomer, or metal of thin cross-section. Membranes 416 break or rupture when subjected to a predetermined system pressure. Each membrane 416 is mounted on respective ones of a plurality of supports 417, such as by means of fasteners 418. Supports 417 are integrally connected to valve body 415. Alternatively, either of valves 390 or 410 may be a second embodiment diaphragm valve, generally referred to as 414b, having breakable barriers or membranes 416 that are breakable by means of a piston arrangement, generally referred to as 419. Second embodiment diaphragm valve 414b may be used in combination with back-flow prevention valve 390 or 410, as shown. Piston arrangement 419 has a piston 419a movable to break membrane 416. Each piston 419a is movable by means of a motor 419b. Motors 419b are connected to control unit 400, so that control unit 400 controls motors 419b. Thus, each piston 419a is capable of moving to break membrane 416 by means of operator action as an operator operates control unit 400. Valves 414b may be custom designed valves that may be available from Solenoid Solutions, Incorporated located in Erie, Pa., U.S.A. However, it may be appreciated that valves 414a and 414b may be check valves rather than diaphragm valves, if desired. Returning to FIG. 6, operation of dual-purpose circuit 330 for removal of volatile fission products 15 from fuel body 40 will now be described. As previously mentioned, circuit 330 can be operated to selectively consecutively remove volatile fission products 15 as well as heat from fuel body 40. To remove volatile fission products 15 from fuel body 40, first valve 390 is opened and second valve 410 is closed, such as by action of control unit 400 to which valves 390/410 are electrically connected. As previously mentioned, volatile fission products 15 are produced in fuel body 40 by burn wave 16 and reside in open-cell pores 50. Third pump 340 is selectively operable, such as by means of control unit 400, so that fission products 15 acquired by open-cell pores 50 are drawn through first pipe segment 70, into sixth pipe segment 350 and then into first volume 90. First pump 210 will then draw the fission products 15 from first volume 90 and then through second pipe segment 200. First pump 210 will pump the fission products 15 from second pipe segment 200 and through third pipe segment 220. The fission products 15 flowing along third pipe segment 220 will be diverted to second fission product reservoir or holding tank 370 because first valve 390 is open and second valve 410 is closed. After a predetermined amount of time, first valve 390 is closed and second valve 410 is opened to resume removal of fission products 15 from fuel body 40, if needed. Still referring to FIG. 6, operation of circuit 330 for removal of heat from fuel body 40 will now be described. To remove heat from fuel body 40, first valve 390 is closed and second valve 410 is opened, such as by action of control unit 400. First pump 210 and third pump 340 are activated, which also may be by action of control unit 400. First pump 210 will draw the fluid, such as the previously mentioned helium gas, through first pipe segment 200 and thus from first volume 90, which is defined by fluid control subassembly 80. First pump 210 will pump the fluid through third pipe segment 220. The previously mentioned heat exchanger 355 is in heat transfer communication with the fluid flowing through third pipe segment 220 for removing the heat carried by the fluid. The fluid flowing through third pipe segment 220 will not be diverted to reservoir or holding tank 370 because first valve 390 is closed. The fluid flowing through third pipe segment 220 is received by the plurality (or multiplicity) of open-cell pores 50 that are defined by porous fuel body 40. The fluid received by open-cell pores 50 will acquire the heat produced by fuel body 40. The heat is acquired by means of convective heat transfer as the fluid flows through open-cell pores 50. As convective heat transfer occurs within fuel body 40, third pump 340 is operated, such as by means of control unit 400. As third pump 340 is operated, the fluid residing in fuel body 40 and that is experiencing the convective heat transfer, is drawn through first pipe segment 70 and into first volume 90 A benefit of using fifth embodiment nuclear fission reactor fuel assembly 290 is that compact, dual-purpose circuit 330 can selectively consecutively remove volatile fission products 15 and then remove heat or vice versa. This result is accomplished by controlled operation of pumps 210/340 and valves 390/410 by means of control unit 400 and also by means of heat exchanger 355. Referring to FIG. 7, a sixth embodiment nuclear fission reactor fuel assembly and system are there shown, generally referred to as 420. Sixth embodiment fuel assembly 420 is substantially similar to fifth embodiment fuel assembly 290, except that the following components are disposed substantially externally to vessel 310: first pipe segment 70, third pump 340, sixth pipe segment 350, fluid control subassembly 80, second pipe segment 200, first pump 210, third pipe segment 220, first valve 390, heat exchanger 355, seventh pipe segment 360, second fission product reservoir or holding tank 370, second valve 410 and control unit 400. In some cases disposing these components externally to vessel 310 may make these components more readily accessible for easier maintenance without exposing maintenance equipment and reactor personnel to radiation levels within vessel 310 while performing such maintenance. As seen in FIG. 7A, a first fluid supply reservoir or first component 422, a second fluid supply reservoir or second component 423 and fluid control subassembly 80 are operatively coupled together by a Y-shaped pipe junction 424. First fluid supply component 422 is capable of supplying a fission product removal fluid to fluid control subassembly 80, so as to enable fluid control subassembly 80 to circulate the fission product removal fluid through the open-cell pores 50 of nuclear fuel body 40. In this manner, at least a portion of volatile fission product 15 acquired by pores 50 of nuclear fuel body 40 is removed from pores 50 while fluid control subassembly 80 circulates the fission product removal fluid through pores 50. In addition, second fluid supply component 423 is capable of supplying a heat removal fluid to fluid control subassembly 80, so as to enable fluid control subassembly 80 to circulate the heat removal fluid through the pores of nuclear fuel body 40. In this manner, at least a portion of the heat generated by nuclear fuel body 40 is removed from nuclear fuel body 40 while fluid control subassembly 80 circulates the heat removal fluid through nuclear fuel body 40. The fission product removal fluid may be, with limitation, hydrogen (H2), helium (He), carbon dioxide (CO2), and/or methane (CH4). The heat removal fluid may be, without limitation, hydrogen (H2), helium (He), carbon dioxide (CO2), sodium (Na), lead (Pb), sodium-potassium (NaK), lithium (Li), “light” water (H2O), lead-bismuth (Pb—Bi) alloys, and/or fluorine-lithium-beryllium (FLiBe). First component 422 and second component 423 may be substantially identical in configuration. A pair of back-flow prevention valves (not shown) may be integrally coupled to respective ones of components 422/423 for controlling flow of the fission product removal fluid and heat removal fluid into volume 90, but not reverse flow from volume 90 and back into either first component 422 or second component 423. In this manner, first component 422 and second component 423 are capable of supplying, respectively, the fission product removal fluid and the heat removal fluid to fluid control subassembly 80. In other words, first component 422 and second component 423 are capable of sequentially supplying, respectively, the fission product removal fluid and the heat removal fluid to fluid control subassembly 80. Moreover, a pair of pumps (not shown) is coupled to first component 422 and second component 423, respectively, for pumping the fission product removal fluid and the heat removal fluid to fluid control subassembly 80. Referring to FIG. 7B, a fluid control subassembly may alternatively comprise an inlet subassembly 426 for supplying the fission product removal fluid to fluid control subassembly 80. A valve 426′ may be interposed between inlet subassembly 426 and fluid control subassembly 80 for controlling flow of the fission product removal fluid from inlet subassembly 426 to volume 90. A fourth pump 340′, that is in communication with volume 90 and that is connected to fuel body 40 may thereafter pump the fission product removal fluid to porous nuclear fuel body 40. An outlet subassembly 427 is also provided for removing the fission product removal fluid from porous nuclear fuel body 40. In this regard, third pump 340 is operated to withdraw the fission product removal fluid from nuclear fuel body 40 and into fluid control subassembly 80. Thereafter, the fission product removal fluid flows into outlet subassembly 427. Another valve 427′ may be interposed between outlet subassembly 427 and fluid control subassembly 80 for controlling flow of the fission product removal fluid to outlet subassembly 427. During operation, when valve 427′ is closed and valve 426′ is opened, the fission product removal fluid in inlet subassembly 426 is drawn by pump 340′into volume 90 and then into fuel body 40. After the fission product removal fluid is substantially exhausted from inlet subassembly 426, pump 340′ is caused to cease operation. Valve 426′ is then closed and valve 427′ is opened. Pump 340 is then operated to draw the fission product removal fluid from fuel body 40 and into volume 90. The fission product removal fluid will thereafter travel to outlet subassembly 427. Heat exchanger 355 may be interposed between fluid control subassembly 80 and outlet subassembly 427 for removing heat from the fluid, if desired. Referring to FIG. 7C, a fluid control subassembly may alternatively comprise inlet subassembly 426 that is coupled to enclosure 20. Optional pump 340a pumps the fission product removal fluid from inlet subassembly 426 to fuel body 40 and through pipe 426′ and pipe 70a. The fission product removal fluid is drawn from fuel body 40 and through pipe 70b, such as by another optional pump 340b, and then flows to fluid control subassembly 80. From there, the fission product removal fluid is pumped by optional pump 340c so that the fission product removal fluid flows through pipe 427′ to outlet subassembly 427. If desired, some or all of the pumps 340a, 340b, and 340c may be omitted. If desired, heat exchanger 355 may be interposed between fluid control subassembly 80 and outlet subassembly 427 for removing heat from the fission product removal fluid. Referring to FIG. 7D, a fluid control subassembly may alternatively comprise a plurality of outlet subassemblies 428a/428b/428c for receiving the fission product removal fluid from porous nuclear fuel body 40 and may further comprise a plurality of pumps 429a/429b/429c coupled to respective ones of outlet subassemblies 428a/428b/428c. Pumps 429a/429b/429c are configured to pump the fission product removal fluid along pipes 70a/70b/70c to respective ones of the plurality of outlet subassemblies 428a/428b/428c. The fission product removal fluid flows to fluid control subassembly 80 through pipe 71 due to the pumping action of a pump 71′. From there, the fission product removal fluid flows through pipe 427′ to a reservoir 427 due to the pumping action of a pump 429d. If desired, either or all of the pumps 429a, 429b, 429c, 429d and 71′ may be omitted. If desired, heat exchanger 355 may be interposed between fluid control subassembly 80 and outlet subassembly 427 for removing heat from the fluid. Referring to FIG. 7E, there is shown a seventh embodiment nuclear fission reactor fuel assembly and system, generally referred to as 430, for producing heat due to fission of a fissile nuclide. This seventh embodiment nuclear fission reactor fuel assembly and system is similar to the first embodiment nuclear fission reactor fuel assembly and system 10, except that there are a plurality of enclosures 20a, 20b, and 20c. Each of the enclosures 20a, 20b and 20c is connected to fluid control subassembly 80 by means of respective ones of a plurality of pipe segments 72a, 72b and 72c. Seventh embodiment nuclear fission reactor fuel assembly and system 430 otherwise operates in the same manner as first embodiment nuclear fission reactor fuel assembly and system 10. Referring to FIG. 8, there is shown an eighth embodiment nuclear fission reactor fuel assembly and system, generally referred to as 438. This eighth embodiment nuclear fission reactor fuel assembly 438 differs from fifth embodiment nuclear fission reactor fuel assembly 290 and sixth embodiment nuclear fission reactor fuel assembly 420 in that dual purpose circuit 330 is replaced by a fission product flow path, generally referred to as 440 and by a separate heat removal flow path, generally referred to as 450. The purpose of heat removal flow path 450 is to remove heat from fuel body 40. The purpose of fission product flow path 440 is to remove and isolate volatile fission products 15 from fuel body 40. Heat removal flow path 450 comprises the previously mentioned fluid control subassembly 80 that defines first volume 90. The first volume 90 contains the fluid, such as helium gas, that is used to remove heat. First pipe segment 70 is in communication with fuel body 40 at one end of first pipe segment 70 and is integrally connected at the other end of first pipe segment 70 to the inlet of third pump 340. The outlet of third pump 340 is connected to sixth pipe segment 350, which in turn is in communication with first volume 90. Second pipe segment 200 is in communication with first volume 90 at one end of second pipe segment 200 and is integrally connected to the inlet of first pump 210 at the other end of second pipe segment 200. The outlet of first pump 210 is connected to third pipe segment 220, which in turn is in communication with fuel body 40. Heat exchanger 355 is coupled to third pipe segment 220 for removing heat from the fluid. Thus, first pipe segment 70, third pump 340, sixth pipe segment 350, fluid control subassembly 80, second pipe segment 200, first pump 210, third pipe segment 220, fuel body 40 itself and heat exchanger 355, together define heat removal flow path 450. As described in more detail hereinbelow, heat removal flow path 450 is capable of circulating the heat removal fluid through heat exchanger 355 and open-cell pores 50 of fuel body 40, so that heat is removed from fuel body 40. Still referring to FIG. 8, fission product flow path 440 comprises a first flow pipe 460 having one end thereof in communication with fuel body 40. The other end of first flow pipe 460 is connected to an inlet of a fifth pump 470, which may be a centrifugal pump. The outlet of fifth pump 470 is connected to a second flow pipe 480. Second flow pipe 480 is in communication with a fourth volume 490, which is defined by a third fission product reservoir or holding tank 500. As described in more detail hereinbelow, fission product flow path 440 is capable of removing and isolating fission products 15 from fuel body 40. Referring again to FIG. 8, operation of heat removal flow path 450 to remove heat from fuel body 40 will now be described. In this regard, to remove heat from fuel body 40, first pump 210 and third pump 340 are activated, which may be by means of control unit 400. First pump 210 will draw the heat removal fluid, such as the previously mentioned helium gas, through first pipe segment 200 and thus from first volume 90, which is defined by fluid control subassembly 80. First pump 210 will pump the fluid through third pipe segment 220. The fluid flowing through third pipe segment 220 is received by the plurality (or multiplicity) of open-cell pores 50 that are defined by fuel body 40. The fluid received by open-cell pores 50 will acquire the heat produced by fuel body 40. The heat is acquired by means of convective heat transfer as the fluid flows through open-cell pores 50. As convective heat transfer is occurring within fuel body 40, third pump 340 is operated, such as by means of control unit 400. As third pump 340 is operated, the fluid that is experiencing the convective heat transfer in fuel body 40 is drawn through first pipe segment 70 by third pump 340 and then pumped by third pump 340 into first volume 90. First pump 210, third pump 340 and fourth pump 470 may each be selectively operated by means of control unit 400. The previously mentioned heat exchanger 355 that is in heat transfer communication with the fluid flowing in third pipe segment 220 removes the heat from the fluid. Pumps 340 and 210 are selected such that heat removal flow path 450 may be implemented with pump 340 alone, with pump 210 alone, or with pumps 340 and 210 together. In other words, simultaneous operation of pumps 340 and 210 will remove heat at a maximum rate. On the other hand, operation of either pump 340 or 210 alone will pump the heat removal fluid at a reduced, but sufficient, rate if either of pumps 340 or 210 is non-functional or otherwise unavailable. Referring again to FIG. 8, operation of second flow path 440 for removal and isolation of volatile fission product 15 from fuel body 40 will now be described. In this regard, heat removal flow path 450 is caused to cease operation, such as by deactivating pumps 210 and 340. Then, as fifth pump 470 is operated, volatile fission product 15 will be drawn into first flow pipe 460 and then pumped into second flow pipe 480. As volatile fission product 15 is pumped through second flow pipe 480, the fluid will enter fourth volume 490 that is defined by third fission product reservoir or holding tank 500. Thus, volatile fission product 15 will have been removed from fuel body 40 and then retained in third fission product reservoir or holding tank 500 for subsequent off-site disposal or the fission products 15 in reservoir or holding tank 500 may remain in situ, if desired. Fission product flow path 440 and heat removal flow path 450 may be operated either simultaneously or consecutively, as desired. Moreover, it may be appreciated from the description hereinabove, that volatile fission product 15 may remove itself from open-cell pores 50 and travel to volume 90 without assistance of fifth pump 470 by vaporization due to the inherently volatile nature of volatile fission product 15. Accordingly, fission product flow path 440 may be implemented with or without pump 470. Fission product flow path 440 may utilize one or more controllable shut-off valves (not shown) or back-flow prevention valves (also not shown) disposed in flow path 440 and operatively connected to control unit 400 for further isolating fourth volume 490. Referring to FIGS. 9 and 10, a ninth embodiment nuclear fission reactor fuel assembly and system 510 are there shown. In this ninth embodiment, fuel assembly 510 comprises a generally cylindrical enclosure 515 having enclosure wall 516 for enclosing fuel body 40 therein. The fission product removal fluid, which has the volatile fission product 15 entrained therein, is drawn from fuel body 40 and into fluid control subassembly 80 by pump 340. Heat exchanger 355 may be provided in pipe 220 to remove heat from the fluid. A potential benefit to using the cylindrical enclosure 515 is its utility in shaping fuel profiles. The terminology “fuel profile” is defined herein to mean the geometrical configuration of fissile material, fertile material, and/or neutron moderating material. Turning now to FIG. 11, a tenth embodiment nuclear fission reactor fuel assembly and system are there shown, generally referred to as 520. In this tenth embodiment, fuel assembly 520 comprises a generally spherical enclosure 525 having an enclosure wall 526 for enclosing fuel body 40 therein. A potential benefit to using the spherical enclosure 525 is that its spherical shape reduces the amount of cladding or enclosure material 20 required. Another potential benefit to using the spherical enclosure 525 is its utility in shaping fuel profiles. Referring to FIG. 12, an eleventh embodiment nuclear fission reactor fuel assembly and system are there shown, generally referred to as 530. In this eleventh embodiment, fuel assembly 530 comprises a generally hemi-spherical enclosure 540 having an enclosure wall 545 for enclosing fuel body 40 therein. A potential benefit to using the hemi-spherical enclosure 540 is that it may increase fuel assembly packing densities in well 320 that is defined by vessel 310. Another potential benefit to using the hemi-spherical enclosure 540 is its utility in shaping fuel profiles. Referring to FIGS. 13 and 14, a twelfth embodiment fuel assembly and system are there shown, generally referred to as 550. In this twelfth embodiment, fuel assembly 550 comprises a generally disk-shaped enclosure 560 having an enclosure wall 565 for enclosing fuel body 40 therein. A potential benefit to using the disk-shaped enclosure 560 is its utility in shaping fuel profiles. Referring to FIGS. 15 and 16, a thirteenth embodiment fuel assembly and system are there shown, generally referred to as 570. In this thirteenth embodiment, fuel assembly 570 comprises a polygonal-shaped (in transverse cross-section) enclosure 580 having an enclosure wall 585 for enclosing fuel body 40 therein. In this regard, enclosure 580 may have a hexagon shape in transverse cross section. A potential benefit attendant to the hexagonally shaped cross section of enclosure 580 is that more fuel assemblies 570 can be packed into well 320 of vessel 310 than otherwise would be allowed by many other geometric shapes for the fuel assembly. Another potential benefit to using the hexagonally shaped enclosure 580 is its utility in shaping fuel profiles. Referring to FIGS. 17 and 18, a fourteenth embodiment fuel assembly and system are there shown, generally referred to as 590. In this fourteenth embodiment, fuel assembly 590 comprises a parallelepiped-shaped enclosure 600 having enclosure walls 605 for enclosing fuel body 40 therein. A potential benefit to using the parallelepiped-shaped enclosure 600 is that it may increase fuel assembly packing densities in well 320 of vessel 310. Another potential benefit to using the parallelepiped-shaped enclosure 600 is its utility in shaping fuel profiles. Referring to FIG. 19, a fifteenth embodiment nuclear fission reactor fuel assembly and system, generally referred to as 610, is there shown. In this regard, fuel body 40 may include one or more fuel pellets 620 embedded therein. Fuel pellet 620 may function as a higher density fuel component to increase the effective density of fuel body 40. Referring to FIG. 20, a sixteenth embodiment nuclear fission rector fuel assembly and system, generally referred to as 625, is there shown. In this regard, fluid control subassembly 80 is coupled to a plurality of enclosures 20. Illustrative Methods Illustrative methods associated with exemplary embodiments of nuclear fission reactor fuel assemblies and systems 10, 100, 190, 230, 290, 420, 430, 510, 520, 530, 550, 570, 590, 610, and 625 will now be described. Referring to FIGS. 21A-21CQ, illustrative methods are provided for assembling the nuclear fission reactor fuel assembly and system. Referring now to FIG. 21A, an illustrative method 630 for assembling the nuclear fission reactor fuel assembly starts at a block 640. At a block 650, an enclosure is provided that encloses a porous nuclear fuel body. At a block 660, a fluid control subassembly is coupled to the enclosure 20 for removal of at least a portion of a volatile fission product at locations corresponding to a burn wave. The fluid control subassembly controls fluid flow in regions of the reactor proximate to locations corresponding to the burn wave. The method 630 stops at a block 670. Referring to FIG. 21B, an illustrative method 671 for assembling the nuclear fission reactor fuel assembly starts at a block 672. At a block 673, an enclosure is provided that encloses a nuclear fuel body. At a block 674, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 675, a control unit is coupled to the fluid control subassembly to control operation of the fluid control subassembly. The method 671 stops at a block 676. Referring to FIG. 21C, an illustrative method 677 for assembling the nuclear fission reactor fuel assembly starts at a block 680. At a block 690, an enclosure is provided that encloses a nuclear fuel body in the manner previously mentioned. At a block 700, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 710, a control unit is coupled to the fluid control subassembly to control operation of the fluid control subassembly. At a block 715, the control unit is coupled to permit a controlled release of the volatile fission product in response to a power level in the traveling wave nuclear fission reactor. The method 677 stops at a block 720. Referring to FIG. 21D, an illustrative method 730 for assembling the nuclear fission reactor fuel assembly starts at a block 740. At a block 750, an enclosure is provided that encloses a nuclear fuel body in the manner previously mentioned. At a block 760, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 770, a control unit is coupled to the fluid control subassembly to control operation of the fluid control subassembly. At a block 780, the control unit is coupled to permit a controlled release of the volatile fission product in response to neutron population level in the traveling wave nuclear fission reactor. The method 730 stops at a block 790. Referring to FIG. 21E, an illustrative method 800 for assembling the nuclear fission reactor fuel assembly starts at a block 810. At a block 820, an enclosure is provided that encloses a nuclear fuel body in the manner previously mentioned. At a block 830, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 840, a control unit is coupled to the fluid control subassembly to control operation of the fluid control subassembly. At a block 850, the control unit is coupled to permit a controlled release of the volatile fission product in response to a volatile fission product pressure level in the traveling wave nuclear fission reactor. The method 800 stops at a block 860. Referring to FIG. 21F, an illustrative method 870 for assembling the nuclear fission reactor fuel assembly starts at a block 880. At a block 890, an enclosure is provided that encloses a nuclear fuel body in the manner previously mentioned. At a block 900, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 910, a control unit is coupled to the fluid control subassembly to control operation of the fluid control subassembly. At a block 920, the control unit is coupled to permit a controlled release of the volatile fission product in response to a time schedule associated with the traveling wave nuclear fission reactor. The method 870 stops at a block 930. Referring to FIG. 21G, an illustrative method 940 for assembling the nuclear fission reactor fuel assembly starts at a block 950. At a block 960, an enclosure is provided that encloses a nuclear fuel body in the manner previously mentioned. At a block 970, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 980, a control unit is coupled to the fluid control subassembly to control operation of the fluid control subassembly. At a block 990, the control unit is coupled to permit a controlled release of the volatile fission product in response to an amount of time the nuclear fission reactor is operated. The method 940 stops at a block 1000. Referring to FIG. 21H, an illustrative method 1010 for assembling the nuclear fission reactor fuel assembly starts at a block 1020. At a block 1030, an enclosure is provided that encloses a nuclear fuel body in the manner previously mentioned. At a block 1040, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 1050, the enclosure is provided so as to enclose the nuclear fuel body. The method 1010 stops at a block 1060. Referring to FIG. 21I, an illustrative method 1070 for assembling the nuclear fission reactor fuel assembly starts at a block 1080. At a block 1090, an enclosure is provided that encloses a nuclear fuel body in the manner previously mentioned. At a block 1100, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 1110, the enclosure is provided so as to enclose a fissile material forming the nuclear fuel body. The method 1070 stops at a block 1120. Referring to FIG. 21J, an illustrative method 1130 for assembling the nuclear fission reactor fuel assembly starts at a block 1140. At a block 1150, an enclosure is provided that encloses a nuclear fuel body in the manner previously mentioned. At a block 1160, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 1170, the enclosure is provided so as to enclose a fissile material forming the nuclear fuel body. The method 1130 stops at a block 1180. Referring to FIG. 21K, an illustrative method 1190 for assembling the nuclear fission reactor fuel assembly starts at a block 1200. At a block 1210, an enclosure is provided that encloses a nuclear fuel body in the manner previously mentioned. At a block 1220, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 1230, the enclosure is provided so as to enclose a fissile and fertile material forming the nuclear fuel body. The method 1190 stops at a block 1240. Referring to FIG. 21L, an illustrative method 1250 for assembling the nuclear fission reactor fuel assembly starts at a block 1260. At a block 1270, an enclosure is provided that encloses a nuclear fuel body in the manner previously mentioned. At a block 1280, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 1290, the enclosure is provided so as to permit a controlled release of the volatile fission product in response to a power level in the traveling wave nuclear fission reactor. The method 1250 stops at a block 1300. Referring to FIG. 21M, an illustrative method 1310 for assembling the nuclear fission reactor fuel assembly starts at a block 1320. At a block 1330, an enclosure is provided that encloses a nuclear fuel body in the manner previously mentioned. At a block 1340, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 1350, the enclosure is provided so as to permit a controlled release of the volatile fission product in response to a neutron population level in the traveling wave nuclear fission reactor. The method 1310 stops at a block 1360. Referring to FIG. 21N, an illustrative method 1370 for assembling the nuclear fission reactor fuel assembly starts at a block 1380. At a block 1390, an enclosure is provided that encloses a nuclear fuel body in the manner previously mentioned. At a block 1400, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 1410, the enclosure is provided so as to permit a controlled release of the volatile fission product in response to a volatile fission product pressure level in the traveling wave nuclear fission reactor. The method 1370 stops at a block 1420. Referring to FIG. 21O, an illustrative method 1430 for assembling the nuclear fission reactor fuel assembly starts at a block 1440. At a block 1450, an enclosure is provided that encloses a nuclear fuel body in the manner previously mentioned. At a block 1460, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 1470, the enclosure is provided so as to permit a controlled release of the volatile fission product in response to a time schedule associated with the traveling wave nuclear fission reactor. The method 1430 stops at a block 1480. Referring to FIG. 21P, an illustrative method 1490 for assembling the nuclear fission reactor fuel assembly starts at a block 1500. At a block 1510, an enclosure is provided that encloses a nuclear fuel body in the manner previously mentioned. At a block 1520, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 1530, the enclosure is provided so as to permit a controlled release of the volatile fission product in response to an amount of time the traveling wave nuclear fission reactor is continuously operated. The method 1490 stops at a block 1540. Referring to FIG. 21Q, an illustrative method 1550 for assembling the nuclear fission reactor fuel assembly starts at a block 1560. At a block 1570, an enclosure is provided that encloses a nuclear fuel body in the manner previously mentioned. At a block 1580, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 1590, the enclosure is provided so as to enclose a porous nuclear fuel body in the form of a foam defining a plurality of pores. The method 1550 stops at a block 1600. Referring to FIG. 21R, an illustrative method 1610 for assembling the nuclear fission reactor fuel assembly starts at a block 1620. At a block 1630, an enclosure is provided that encloses a nuclear fuel body in the manner previously mentioned. At a block 1640, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 1650, the enclosure is provided to enclose a nuclear fuel body defining a plurality of pores, the plurality of pores having a spatially non-uniform distribution. The method 1610 stops at a block 1660. Referring to FIG. 21S, an illustrative method 1670 for assembling the nuclear fission reactor fuel assembly starts at a block 1680. At a block 1690, an enclosure is provided that encloses a nuclear fuel body in the manner previously mentioned. At a block 1700, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 1710, the enclosure is provided to enclose a nuclear fuel body having a plurality of channels. The method 1670 stops at a block 1720. Referring to FIG. 21T, an illustrative method 1730 for assembling the nuclear fission reactor fuel assembly starts at a block 1740. At a block 1750, an enclosure is provided that encloses a nuclear fuel body in the manner previously mentioned. At a block 1760, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 1770, the enclosure is provided so as to enclose a porous nuclear fuel body having a plurality of particles defining the plurality of channels therebetween. The method 1730 stops at a block 1790. Referring to FIG. 21U, an illustrative method 1800 for assembling the nuclear fission reactor fuel assembly starts at a block 1810. At a block 1820, an enclosure is provided that encloses a nuclear fuel body in the manner previously mentioned. At a block 1830, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 1840, the enclosure is provided so as to enclose a porous nuclear fuel body having a plurality of pores, at least one of the pores being of a predetermined configuration for allowing at least a portion of the volatile fission product to escape the porous nuclear fuel body within a predetermined response time. The method 1800 stops at a block 1850. Referring to FIG. 21V, an illustrative method 1860 for assembling the nuclear fission reactor fuel assembly starts at a block 1870. At a block 1880, an enclosure is provided that encloses a nuclear fuel body in the manner previously mentioned. At a block 1890, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 1900, the enclosure is provided so as to enclose a porous nuclear fuel body having a plurality of pores for allowing at least a portion of the volatile fission product to escape within a predetermined response time of between approximately 10 seconds and approximately 1,000 seconds. The method 1860 stops at a block 1910. Referring to FIG. 21W, an illustrative method 1920 for assembling the nuclear fission reactor fuel assembly starts at a block 1930. At a block 1940, an enclosure is provided that encloses a nuclear fuel body in the manner previously mentioned. At a block 1950, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 1960, the enclosure is provided fuel body having a plurality of pores for allowing at least a portion of the volatile fission product to escape within a predetermined response time of between approximately 10 seconds and approximately 1,000 seconds. The method 1920 stops at a block 1970. Referring to FIG. 21X, an illustrative method 1971 for assembling the nuclear fission reactor fuel assembly starts at a block 1972. At a block 1973, an enclosure is provided that encloses a nuclear fuel body in the manner previously mentioned. At a block 1974, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 1975, the enclosure is provided so as to sealingly enclose a porous nuclear fuel body having a cylindrical-shaped geometry. The method 1971 stops at a block 1976. Referring to FIG. 21Y, an illustrative method 1980 for assembling the nuclear fission reactor fuel assembly starts at a block 1990. At a block 2000, an enclosure is provided that encloses a nuclear fuel body in the manner previously mentioned. At a block 2010, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 2020, the enclosure is provided so as to sealingly enclose a porous nuclear fuel body having a polygonal-shaped geometry. The method 1980 stops at a block 2030. Referring to FIG. 21Z, an illustrative method 2040 for assembling the nuclear fission reactor fuel assembly starts at a block 2050. At a block 2060, an enclosure is provided that encloses a nuclear fuel body in the manner previously mentioned. At a block 2070, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 2080, the enclosure is provided so as to enclose a porous nuclear fuel body having a plurality of pores for acquiring the volatile fission product released by the burn wave in the traveling wave nuclear fission reactor. The method 2040 stops at a block 2090. Referring to FIG. 21AA, an illustrative method 2100 for assembling the nuclear fission reactor fuel assembly starts at a block 2110. At a block 2120, an enclosure is provided that encloses a nuclear fuel body in the manner previously mentioned. At a block 2130, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 2140, the enclosure is provided so as to enclose a porous nuclear fuel body having a plurality of pores to transport the volatile fission product through the porous nuclear fuel body. The method 2100 stops at a block 2150. Referring to FIG. 21AB, an illustrative method 2160 for assembling the nuclear fission reactor fuel assembly starts at a block 2170. At a block 2180, an enclosure is provided that encloses a nuclear fuel body in the manner previously mentioned. At a block 2190, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 2200, a reservoir is coupled to the fluid control subassembly to receive the volatile fission product. The method 2160 stops at a block 2210. Referring to FIG. 21AC, an illustrative method 2220 for assembling the nuclear fission reactor fuel assembly starts at a block 2230. At a block 2240, an enclosure is provided that encloses a nuclear fuel body in the manner previously mentioned. At a block 2250, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 2260, the fluid control subassembly is coupled to permit a controlled release of the volatile fission product in response to a position of the burn wave in the traveling wave nuclear fission reactor. The method 2220 stops at a block 2270. Referring to FIG. 21AD, an illustrative method 2280 for assembling the nuclear fission reactor fuel assembly starts at a block 2290. At a block 2300, an enclosure is provided that encloses a nuclear fuel body in the manner previously mentioned. At a block 2310, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 2320, the fluid control subassembly is coupled so that the nuclear fission fuel assembly is configured to circulate a fission product removal fluid through the porous nuclear fuel body and so that at least a portion of the volatile fission product is removed from the porous nuclear fuel body while the fluid control subassembly circulates the fission product removal fluid through the porous nuclear fuel body. The method 2280 stops at a block 2330. Referring to FIG. 21AE, an illustrative method 2340 for assembling the nuclear fission reactor fuel assembly starts at a block 2350. At a block 2360, an enclosure is provided that encloses a nuclear fuel body in the manner previously mentioned. At a block 2370, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 2380, the fluid control subassembly is coupled so that the nuclear fission fuel assembly is configured to circulate a fission product removal fluid through the porous nuclear fuel body and so that at least a portion of the volatile fission product is removed from the porous nuclear fuel body while the fluid control subassembly circulates the fission product removal fluid through the porous nuclear fuel body. At a block 2390, an inlet subassembly is provided to supply the fission product removal fluid to the porous nuclear fuel body. The method 2340 stops at a block 2400. Referring to FIG. 21AF, an illustrative method 2410 for assembling the nuclear fission reactor fuel assembly starts at a block 2420. At a block 2430, an enclosure is provided that encloses a nuclear fuel body in the manner previously mentioned. At a block 2440, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 2450, the fluid control subassembly is coupled so that the nuclear fission fuel assembly is configured to circulate a fission product removal fluid through the porous nuclear fuel body and so that at least a portion of the volatile fission product is removed from the porous nuclear fuel body while the fluid control subassembly circulates the fission product removal fluid through the porous nuclear fuel body. At a block 2460, an inlet subassembly is provided to remove the fission product removal fluid from the porous nuclear fuel body. The method 2410 stops at a block 2470. Referring to FIG. 21AG, an illustrative method 2480 for assembling the nuclear fission reactor fuel assembly starts at a block 2490. At a block 2500, an enclosure is provided that encloses a porous nuclear fuel body in the manner previously mentioned. At a block 2510, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 2520, the fluid control subassembly is coupled so that the nuclear fission fuel assembly is configured to circulate a fission product removal fluid through the porous nuclear fuel body and so that at least a portion of the volatile fission product is removed from the porous nuclear fuel body while the fluid control subassembly circulates the fission product removal fluid through the porous nuclear fuel body. At a block 2530, a reservoir is provided to receive the fission product removal fluid from. The method 2480 stops at a block 2540. Referring to FIG. 21AH, an illustrative method 2550 for assembling the nuclear fission reactor fuel assembly starts at a block 2560. At a block 2570, an enclosure is provided that encloses a porous nuclear fuel body in the manner previously mentioned. At a block 2580, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 2590, the fluid control subassembly is coupled so that the nuclear fission fuel assembly is configured to circulate a fission product removal fluid through the porous nuclear fuel body and so that at least a portion of the volatile fission product is removed from the porous nuclear fuel body while the fluid control subassembly circulates the fission product removal fluid through the porous nuclear fuel body. At a block 2600, a reservoir is coupled to supply the fission product removal fluid. The method 2550 stops at a block 2610. Referring to FIG. 21AI, an illustrative method 2620 for assembling the nuclear fission reactor fuel assembly starts at a block 2630. At a block 2640, an enclosure is provided that encloses a porous nuclear fuel body in the manner previously mentioned. At a block 2650, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 2590, the fluid control subassembly is coupled so that the nuclear fission fuel assembly is configured to circulate a gas fluid through the porous nuclear fuel body and so that at least a portion of the volatile fission product is removed from the porous nuclear fuel. The method 2620 stops at a block 2670. Referring to FIG. 21AJ, an illustrative method 2680 for assembling the nuclear fission reactor fuel assembly starts at a block 2690. At a block 2700, an enclosure is provided that encloses a porous nuclear fuel body in the manner previously mentioned. At a block 2710, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 2720, the fluid control subassembly is coupled so that the fluid control subassembly is configured to circulate a liquid through the porous nuclear fuel body. The method 2680 stops at a block 2730. Referring to FIG. 21AK, an illustrative method 2740 for assembling the nuclear fission reactor fuel assembly starts at a block 2750. At a block 2760, an enclosure is provided that encloses a porous nuclear fuel body in the manner previously mentioned. At a block 2770, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 2780, the method comprises coupling a pump. The method 2740 stops at a block 2790. Referring to FIG. 21AL, an illustrative method 2800 for assembling the nuclear fission reactor fuel assembly starts at a block 2810. At a block 2820, an enclosure is provided that encloses a porous nuclear fuel body in the manner previously mentioned. At a block 2830, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 2840, a pump is integrally connected to the fluid control subassembly to circulate a fluid between the fluid control subassembly and the porous nuclear fuel body. The method 2800 stops at a block 2850. Referring to FIG. 21AM, an illustrative method 2860 for assembling the nuclear fission reactor fuel assembly starts at a block 2870. At a block 2880, an enclosure is provided that encloses a porous nuclear fuel body in the manner previously mentioned. At a block 2890, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 2900, the method comprises coupling a valve. The method 2860 stops at a block 2910. Referring to FIG. 21AN, an illustrative method 2920 for assembling the nuclear fission reactor fuel assembly starts at a block 2930. At a block 2940, an enclosure is provided that encloses a porous nuclear fuel body in the manner previously mentioned. At a block 2950, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 2960, a valve is interposed between the enclosure and the fluid control subassembly to control flow of a fluid between the enclosure and the fluid control subassembly. The method 2920 stops at a block 2970. Referring to FIG. 21AO, an illustrative method 2980 for assembling the nuclear fission reactor fuel assembly starts at a block 2990. At a block 3000, an enclosure is provided that encloses a porous nuclear fuel body in the manner previously mentioned. At a block 3010, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 3020, a valve is interposed between the enclosure and the fluid control subassembly to control flow of a fluid between the enclosure and the fluid control subassembly. At a block 3030, a back-flow prevention valve is interposed between the enclosure and the fluid control subassembly. The method 2980 stops at a block 3040. Referring to FIG. 21AP, an illustrative method 3050 for assembling the nuclear fission reactor fuel assembly starts at a block 3060. At a block 3070, an enclosure is provided that encloses a porous nuclear fuel body in the manner previously mentioned. At a block 3080, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 3090, the method comprises coupling a controllably breakable barrier. The method 3050 stops at a block 3100. Referring to FIG. 21AQ, an illustrative method 3110 for assembling the nuclear fission reactor fuel assembly starts at a block 3120. At a block 3130, an enclosure is provided that encloses a porous nuclear fuel body in the manner previously mentioned. At a block 3140, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 3150, a controllably breakable barrier is interposed between the enclosure and the fluid control subassembly. The method 3110 stops at block 3160. Referring to FIG. 21AR, an illustrative method 3170 for assembling the nuclear fission reactor fuel assembly starts at a block 3180. At a block 3190, an enclosure is provided that encloses a porous nuclear fuel body in the manner previously mentioned. At a block 3200, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 3210, a controllably breakable barrier is interposed between the enclosure and the fluid control subassembly. At a block 3220, a barrier breakable at a predetermined pressure is interposed between the enclosure and the fluid control subassembly. The method 3170 stops at a block 3230. Referring to FIG. 21AS, an illustrative method 3240 for assembling the nuclear fission reactor fuel assembly starts at a block 3250. At a block 3260, an enclosure is provided that encloses a porous nuclear fuel body in the manner previously mentioned. At a block 3270, a fluid control subassembly is coupled to the enclosure for removal of at least a portion of a volatile fission product as previously mentioned. The fluid control subassembly controls fluid flow in regions of the reactor proximate locations corresponding to a burn wave. At a block 3280, a controllably breakable barrier is interposed between the enclosure and the fluid control subassembly. At a block 3290, a barrier breakable by operator action is interposed between the enclosure and the fluid control subassembly. The method 3240 stops at a block 3300. Referring to FIG. 21AT, an illustrative method 3310 for assembling the nuclear fission reactor fuel assembly starts at a block 3320. At a block 3330, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 3340, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in regions of the traveling wave nuclear fission reactor proximate to locations corresponding to the burn wave. The method 3310 stops at a block 3350. Referring to FIG. 21AU, an illustrative method 3360 for assembling the nuclear fission reactor fuel assembly starts at a block 3370. At a block 3380, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defusing a plurality of interconnected open-cell pores. At a block 3390, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in regions of the traveling wave nuclear fission reactor proximate to locations corresponding to the burn wave. At a block 3400, a control unit is coupled to the fluid control subassembly to control operation of the fluid control subassembly. The method 3360 stops at a block 3410. Referring to FIG. 21AV, an illustrative method 3420 for assembling the nuclear fission reactor fuel assembly starts at a block 3430. At a block 3440, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 3450, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in regions of the traveling wave nuclear fission reactor proximate to locations corresponding to the burn wave. At a block 3460, the enclosure is provided so as to enclose the nuclear fuel body. The method 3420 stops at a block 3470. Referring to FIG. 21AW, an illustrative method 3480 for assembling the nuclear fission reactor fuel assembly starts at a block 3490. At a block 3500, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 3510, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in regions of the traveling wave nuclear fission reactor proximate to locations corresponding to the burn wave. At a block 3520, the enclosure is provided so as to enclose a fissile material forming the nuclear fuel body. The method 3480 stops at a block 3530. Referring to FIG. 21AX, an illustrative method 3540 for assembling the nuclear fission reactor fuel assembly starts at a block 3550. At a block 3560, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 3570, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in regions of the traveling wave nuclear fission reactor proximate to locations corresponding to the burn wave. At a block 3580, the enclosure is provided so as to enclose a fertile material forming the nuclear fuel body. The method 3540 stops at a block 3590. Referring to FIG. 21AY, an illustrative method 3600 for assembling the nuclear fission reactor fuel assembly starts at a block 3610. At a block 3620, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 3630, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in regions of the traveling wave nuclear fission reactor proximate to locations corresponding to the burn wave. At a block 3640, the enclosure is provided so as to enclose a mixture of fissile and fertile material forming the nuclear fuel body. The method 3600 stops at a block 3650. Referring to FIG. 21AZ, an illustrative method 3660 for assembling the nuclear fission reactor fuel assembly starts at a block 3670. At a block 3680, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 3690, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 3700, the fluid control subassembly is coupled so as to permit a controlled release of the volatile fission product in response to a position of the burn wave in the traveling wave nuclear fission reactor. The method 3660 stops at a block 3710. Referring to FIG. 21BA, an illustrative method 3720 for assembling the nuclear fission reactor fuel assembly starts at a block 3730. At a block 3740, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 3750, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 3760, the fluid control subassembly is coupled so as to permit a controlled release of the volatile fission product in response to a power level in the traveling wave nuclear fission reactor. The method 3720 stops at a block 3770. Referring to FIG. 21BB, an illustrative method 3780 for assembling the nuclear fission reactor fuel assembly starts at a block 3790. At a block 3800, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 3810, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 3820, the fluid control subassembly is coupled so as to permit a controlled release of the volatile fission product in response to a neutron population level in the traveling wave nuclear fission reactor. The method 3780 stops at a block 3830. Referring to FIG. 21BC, an illustrative method 3840 for assembling the nuclear fission reactor fuel assembly starts at a block 3850. At a block 3860, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 3870, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 3880, the fluid control subassembly is coupled so as to permit a controlled release of the volatile fission product in response to a volatile fission product pressure level in the traveling wave nuclear fission reactor. The method 3840 stops at a block 3890. Referring to FIG. 21BD, an illustrative method 3900 for assembling the nuclear fission reactor fuel assembly starts at a block 3910. At a block 3920, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 3930, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 3940, the fluid control subassembly is coupled so as to permit a controlled release of the volatile fission product in response to a time schedule associated with the traveling wave nuclear fission reactor. The method 3900 stops at a block 3950. Referring to FIG. 21BE, an illustrative method 3960 for assembling the nuclear fission reactor fuel assembly starts at a block 3970. At a block 3980, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 3990, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 4000, the fluid control subassembly is coupled so as to permit a controlled release of the volatile fission product in response to an amount of time the traveling wave nuclear fission reactor is operated. The method 3960 stops at a block 4010. Referring to FIG. 21BF, an illustrative method 4020 for assembling the nuclear fission reactor fuel assembly starts at a block 4030. At a block 4040, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 4050, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 4060, a reservoir is coupled to the fluid control subassembly to receive the volatile fission product. The method 4020 stops at a block 4070. Referring to FIG. 21BG, an illustrative method 4080 for assembling the nuclear fission reactor fuel assembly starts at a block 4090. At a block 4100, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 4110, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 4120, the fluid control subassembly that is configured to circulate a fission product removal fluid through the pores of the nuclear fuel body is coupled so that at least a portion of the volatile fission product is removed from the pores of the nuclear fuel body while the fluid control subassembly circulates the fission product removal fluid through the pores of the nuclear fuel body. The method 4080 stops at a block 4130. Referring to FIG. 21BH, an illustrative method 4140 for assembling the nuclear fission reactor fuel assembly starts at a block 4150. At a block 4160, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 4170, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 4175, the fluid control subassembly that is configured to circulate a fission product removal fluid through the pores of the nuclear fuel body is coupled so that at least a portion of the volatile fission product is removed from the pores of the nuclear fuel body while the fluid control subassembly circulates the fission product removal fluid through the pores of the nuclear fuel body. At a block 4180, an inlet subassembly is provided to supply the fission product removal fluid to the pores of the nuclear fuel body. The method 4140 stops at a block 4190. Referring to FIG. 21BI, an illustrative method 4200 for assembling the nuclear fission reactor fuel assembly starts at a block 4210. At a block 4220, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 4230, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 4240, the fluid control subassembly that is configured to circulate a fission product removal fluid through the pores of the nuclear fuel body is coupled so that at least a portion of the volatile fission product is removed from the pores of the nuclear fuel body while the fluid control subassembly circulates the fission product removal fluid through the pores of the nuclear fuel body. At a block 4250, an outlet subassembly is provided to remove the fission product removal fluid from the pores of the nuclear fuel body. The method 4200 stops at a block 4260. Referring to FIG. 21BJ, an illustrative method 4270 for assembling the nuclear fission reactor fuel assembly starts at a block 4280. At a block 4290, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 4300, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 4310, the fluid control subassembly that is configured to circulate a fission product removal fluid through the pores of the nuclear fuel body is coupled so that at least a portion of the heat generated by the nuclear fuel body is removed from the nuclear fuel body while the fluid control subassembly circulates the heat removal fluid through the pores of the nuclear fuel body. The method 4270 stops at a block 4320. Referring to FIG. 21BK, an illustrative method 4330 for assembling the nuclear fission reactor fuel assembly starts at a block 4340. At a block 4350, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 4360, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 4370, the fluid control subassembly that is configured to circulate a fission product removal fluid through the pores of the nuclear fuel body is coupled so that at least a portion of the heat generated by the nuclear fuel body is removed from the nuclear fuel body while the fluid control subassembly circulates the heat removal fluid through the pores of the nuclear fuel body. At a block 4380, a reservoir is coupled to the fluid control subassembly to receive the heat removal fluid. The method 4330 stops at a block 4390. Referring to FIG. 21BL, an illustrative method 4400 for assembling the nuclear fission reactor fuel assembly starts at a block 4410. At a block 4420, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 4430, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 4440, the fluid control subassembly that is configured to circulate a fission product removal fluid through the pores of the nuclear fuel body is coupled so that at least a portion of the heat generated by the nuclear fuel body is removed from the nuclear fuel body while the fluid control subassembly circulates the heat removal fluid through the pores of the nuclear fuel body. At a block 4450, a reservoir is coupled to the fluid control subassembly to supply the heat removal fluid. The method 4400 stops at a block 4460. Referring to FIG. 21BM, an illustrative method 4470 for assembling the nuclear fission reactor fuel assembly starts at a block 4480. At a block 4490, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 4500, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 4510, the fluid control subassembly that is configured to circulate a fission product removal fluid through the pores of the nuclear fuel body is coupled so that at least a portion of the heat generated by the nuclear fuel body is removed from the nuclear fuel body while the fluid control subassembly circulates the heat removal fluid through the pores of the nuclear fuel body. At a block 4520, a heat sink is coupled to the fluid control subassembly, so that the heat sink is in heat transfer communication with the heat removal fluid to remove heat from the heat removal fluid. The method 4470 stops at a block 4530. Referring to FIG. 21BN, an illustrative method 4540 for assembling the nuclear fission reactor fuel assembly starts at a block 4550. At a block 4560, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defusing a plurality of interconnected open-cell pores. At a block 4570, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 4580, the fluid control subassembly that is configured to circulate a fission product removal fluid through the pores of the nuclear fuel body is coupled so that at least a portion of the heat generated by the nuclear fuel body is removed from the nuclear fuel body while the fluid control subassembly circulates the heat removal fluid through the pores of the nuclear fuel body. At a block 4590, a heat exchanger is coupled to the fluid control subassembly, so that the heat exchanger is in heat transfer communication with the heat removal fluid to remove heat from the heat removal fluid. The method 4540 stops at a block 4600. Referring to FIG. 21BO, an illustrative method 4610 for assembling the nuclear fission reactor fuel assembly starts at a block 4620. At a block 4630, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 4640, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 4650, the fluid control subassembly is coupled so as to simultaneously circulate a fission product removal fluid and a heat removal fluid. The method 4610 stops at a block 4660. Referring to FIG. 21BP, an illustrative method 4670 for assembling the nuclear fission reactor fuel assembly starts at a block 4680. At a block 4690, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defusing a plurality of interconnected open-cell pores. At a block 4700, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 4710, the fluid control subassembly is coupled so as to sequentially circulate a fission product removal fluid and a heat removal fluid. The method 4670 stops at a block 4720. Referring to FIG. 21BQ, an illustrative method 4730 for assembling the nuclear fission reactor fuel assembly starts at a block 4740. At a block 4750, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 4760, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 4770, a pump is integrally connected to the fluid control subassembly to pump a fluid from the fluid control subassembly to the pores of the nuclear fuel body. The method 4730 stops at a block 4780. Referring to FIG. 21BR, an illustrative method 4790 for assembling the nuclear fission reactor fuel assembly starts at a block 4800. At a block 4810, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 4820, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 4830, the method comprises coupling a pump. The method 4790 stops at a block 4840. Referring to FIG. 21BS, an illustrative method 4850 for assembling the nuclear fission reactor fuel assembly starts at a block 4860. At a block 4870, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 4880, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 4890, a fission product reservoir is coupled to the fluid control subassembly to receive the volatile fission product. The method 4850 stops at a block 4900. Referring to FIG. 21BT, an illustrative method 4910 for assembling the nuclear fission reactor fuel assembly starts at a block 4920. At a block 4930, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 4940, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 4950, a plurality of first components are coupled so as to enable the fluid control subassembly to circulate a fission product removal fluid through the pores of the nuclear fuel body, whereby at least a portion of the volatile fission product is removed from the pores of the nuclear fuel body while the fluid control subassembly circulates the fission product removal fluid through the pores of the nuclear fuel body. The method 4910 stops at a block 4960. Referring to FIG. 21BU, an illustrative method 4970 for assembling the nuclear fission reactor fuel assembly starts at a block 4980. At a block 4990, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 5000, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 5010, a plurality of first components are coupled so as to enable the fluid control subassembly to circulate a fission product removal fluid through the pores of the nuclear fuel body, whereby at least a portion of the volatile fission product is removed from the pores of the nuclear fuel body while the fluid control subassembly circulates the fission product removal fluid through the pores of the nuclear fuel body. At a block 5020, a plurality of second components are coupled so as to enable the fluid control subassembly to circulate a heat removal fluid through the pores of the nuclear fuel body, whereby at least a portion of the heat generated by the nuclear fuel body is removed from the nuclear fuel body while the fluid control subassembly circulates the heat removal fluid through the pores of the nuclear fuel body. The method 4970 stops at a block 5030. Referring to FIG. 21BV, an illustrative method 5040 for assembling the nuclear fission reactor fuel assembly starts at a block 5050. At a block 5060, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 5070, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 5080, a plurality of first components are coupled so as to enable the fluid control subassembly to circulate a fission product removal fluid through the pores of the nuclear fuel body, whereby at least a portion of the volatile fission product is removed from the pores of the nuclear fuel body while the fluid control subassembly circulates the fission product removal fluid through the pores of the nuclear fuel body. At a block 5090, a plurality of second components are coupled so as to enable the fluid control subassembly to circulate a heat removal fluid through the pores of the nuclear fuel body, whereby at least a portion of the heat generated by the nuclear fuel body is removed from the nuclear fuel body while the fluid control subassembly circulates the heat removal fluid through the pores of the nuclear fuel body. At a block 5100, the method comprises operatively coupling the first components and the second components, so that at least one of the first components and at least one of the second components are identical. The method 5040 stops at a block 5110. Referring to FIG. 21BW, an illustrative method 5120 for assembling the nuclear fission reactor fuel assembly starts at a block 5130. At a block 5140, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 5150, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 5160, the method comprises coupling a dual-purpose circuit to selectively remove the volatile fission product and heat from the nuclear fuel. The method 5120 stops at a block 5170. Referring to FIG. 21BX, an illustrative method 5180 for assembling the nuclear fission reactor fuel assembly starts at a block 5190. At a block 5200, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 5210, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 5220, the fluid control subassembly is coupled so that the nuclear fission fuel assembly is configured to circulate a gas through the pores of the nuclear fuel body. The method 5180 stops at a block 5230. Referring to FIG. 21BY, an illustrative method 5240 for assembling the nuclear fission reactor fuel assembly starts at a block 5250. At a block 5260, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 5270, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 5280, the fluid control subassembly is coupled so that the nuclear fission fuel assembly is configured to circulate a liquid through the pores of the nuclear fuel body. The method 5240 stops at a block 5290. Referring to FIG. 21BZ, an illustrative method 5300 for assembling the nuclear fission reactor fuel assembly starts at a block 5310. At a block 5320, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 5330, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 5340, the enclosure is provided so as to enclose a nuclear fuel body in the form of a foam defining the plurality of pores. The method 5300 stops at a block 5350. Referring to FIG. 21CA, an illustrative method 5360 for assembling the nuclear fission reactor fuel assembly starts at a block 5370. At a block 5380, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 5390, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 5400, the enclosure is provided so as to enclose a nuclear fuel body having a plurality of channels. The method 5360 stops at a block 5410. Referring to FIG. 21CB, an illustrative method 5420 for assembling the nuclear fission reactor fuel assembly starts at a block 5430. At a block 5440, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 5450, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 5460, the enclosure is provided so as to enclose a nuclear fuel body having a plurality of channels. At a block 5470, the enclosure is provided so as to enclose a nuclear fuel body having a plurality of particles defining the plurality of channels therebetween. The method 5420 stops at a block 5480. Referring to FIG. 21CC, an illustrative method 5490 for assembling the nuclear fission reactor fuel assembly starts at a block 5500. At a block 5510, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 5520, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 5530, the enclosure is provided so as to enclose a nuclear fuel body defining the plurality of pores, the plurality of pores having a spatially non-uniform distribution. The method 5490 stops at a block 5540. Referring to FIG. 21CD, an illustrative method 5550 for assembling the nuclear fission reactor fuel assembly starts at a block 5560. At a block 5570, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 5580, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 5590, the enclosure is provided so as to enclose a nuclear fuel body having the plurality of pores for acquiring the volatile fission product released by the burn wave in the traveling wave nuclear fission reactor. The method 5550 stops at a block 5600. Referring to FIG. 21CE, an illustrative method 5610 for assembling the nuclear fission reactor fuel assembly starts at a block 5620. At a block 5630, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 5640, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 5650, the enclosure is provided so as to enclose a nuclear fuel body having the plurality of pores, one or more of the plurality of pores being of a predetermined configuration to allow at least a portion of the volatile fission product to escape the nuclear fuel body within a predetermined response time. The method 5610 stops at a block 5660. Referring to FIG. 21CF, an illustrative method 5670 for assembling the nuclear fission reactor fuel assembly starts at a block 5680. At a block 5690, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 5700, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At block a 5710, the enclosure is provided so as to enclose a nuclear fuel body having the plurality of pores to allow at least a portion of the volatile fission product to escape the nuclear fuel body within a predetermined response time of between approximately 10 seconds and approximately 1,000 seconds. The method 5670 stops at a block 5720. Referring to FIG. 21CG, an illustrative method 5730 for assembling the nuclear fission reactor fuel assembly starts at a block 5740. At a block 5750, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 5760, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 5770, the enclosure is provided so as to enclose a nuclear fuel body having the plurality of pores to allow at least a portion of the volatile fission product to escape the nuclear fuel body within a predetermined response time of between approximately one second and approximately 10,000 seconds. The method 5730 stops at a block 5780. Referring to FIG. 21CH, an illustrative method 5790 for assembling the nuclear fission reactor fuel assembly starts at a block 5800. At a block 5810, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 5820, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 5830, the enclosure is provided so as to enclose a nuclear fuel body having the plurality of pores to transport the volatile fission product through the nuclear fuel body. The method 5790 stops at a block 5840. Referring to FIG. 21CI, an illustrative method 5850 for assembling the nuclear fission reactor fuel assembly starts at a block 5860. At a block 5870, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 5880, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 5890, the enclosure is provided so as to sealingly enclose a nuclear fuel body having a cylindrical-shaped geometry. The method 5850 stops at a block 5900. Referring to FIG. 21CJ, an illustrative method 5910 for assembling the nuclear fission reactor fuel assembly starts at a block 5920. At a block 5930, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 5940, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 5950, the enclosure is provided so as to sealingly enclose a nuclear fuel body having a polygonal-shaped geometry. The method 5910 stops at a block 5960. Referring to FIG. 21CK, an illustrative method 5970 for assembling the nuclear fission reactor fuel assembly starts at a block 5980. At a block 5990, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 6000, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 6010, the method comprises coupling a valve. The method 5970 stops at a block 6020. Referring to FIG. 21CL, an illustrative method 6030 for assembling the nuclear fission reactor fuel assembly starts at a block 6040. At a block 6050, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 6060, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 6070, a valve is interposed between the enclosure and the fluid control subassembly to control flow of a fluid between the enclosure and the fluid control subassembly. The method 6030 stops at a block 6080. Referring to FIG. 21CM, an illustrative method 6090 for assembling the nuclear fission reactor fuel assembly starts at a block 6100. At a block 6110, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 6120, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 6130, a valve is interposed between the enclosure and the fluid control subassembly to control flow of a fluid between the enclosure and the fluid control subassembly. At a block 6140, the method comprises interposing a back-flow prevention valve. The method 6090 stops at a block 6150. Referring to FIG. 21CN, an illustrative method 6160 for assembling the nuclear fission reactor fuel assembly starts at a block 6170. At a block 6180, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 6190, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At block 6200, the method comprises coupling a controllably breakable barrier. The method 6160 stops at a block 6210. Referring to FIG. 21CO, an illustrative method 6220 for assembling the nuclear fission reactor fuel assembly starts at a block 6230. At a block 6240, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 6250, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 6260, a controllably breakable barrier is interposed between the enclosure and the fluid control subassembly. The method 6220 stops at a block 6270. Referring to FIG. 21CP, an illustrative method 6280 for assembling the nuclear fission reactor fuel assembly starts at a block 6290. At a block 6300, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 6310, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 6320, a controllably breakable barrier is interposed between the enclosure and the fluid control subassembly. At a block 6330, the method comprises interposing a controllably breakable barrier breakable at a predetermined pressure. The method 6280 stops at a block 6340. Referring to FIG. 21CQ, an illustrative method 6350 for assembling the nuclear fission reactor fuel assembly starts at a block 6360. At a block 6370, an enclosure is provided to enclose a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 6380, a fluid control subassembly is coupled to the enclosure to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body as previously mentioned. At a block 6390, a controllably breakable barrier is interposed between the enclosure and the fluid control subassembly. At a block 6400, the method comprises interposing a controllably breakable barrier breakable by operator action. The method 6350 stops at a block 6410. Referring to FIG. 22A, an illustrative method is provided for removal of a volatile fission product at a plurality of locations corresponding to a burn wave. In this regard, the illustrative method 6420 for removal of the volatile fission product starts at a block 6430. At a block 6440, removal of a volatile fission product is controlled at a plurality of locations corresponding to a burn wave of a traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. The method 6420 stops at a block 6450. Referring to FIGS. 23A-23CK, illustrative methods are provided for operating the nuclear fission reactor fuel assembly and system. Referring to FIG. 23A, an illustrative method 6460 for operating a nuclear fission reactor fuel assembly starts at a block 6470. At a block 6480, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 6490, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. The method 6460 stops at a block 6500. Referring to FIG. 23B, an illustrative method 6510 for operating a nuclear fission reactor fuel assembly starts at a block 6520. At a block 6530, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 6540, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 6550, operation of the fluid control subassembly is controlled by operating a control unit coupled to the fluid control subassembly. The method 6510 stops at a block 6560. Referring to FIG. 23C, an illustrative method 6570 for operating a nuclear fission reactor fuel assembly starts at a block 6580. At a block 6590, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 6600, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 6610, operation of the fluid control subassembly is controlled by operating a control unit coupled to the fluid control subassembly. At a block 6620, operation of the fluid control subassembly is controlled by operating the control unit to permit a controlled release of the volatile fission product in response to a power level in the traveling wave nuclear fission reactor. The method 6570 stops at a block 6630. Referring to FIG. 23D, an illustrative method 6640 for operating a nuclear fission reactor fuel assembly starts at a block 6650. At a block 6660, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 6670, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 6680, operation of the fluid control subassembly is controlled by operating a control unit coupled to the fluid control subassembly. At a block 6690, operation of the fluid control subassembly is controlled by operating the control unit to permit a controlled release of the volatile fission product in response to a neutron population level in the traveling wave nuclear fission reactor. The method 6640 stops at a block 6700. Referring to FIG. 23E, an illustrative method 6710 for operating a nuclear fission reactor fuel assembly starts at a block 6720. At a block 6730, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 6740, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 6750, operation of the fluid control subassembly is controlled by operating a control unit coupled to the fluid control subassembly. At a block 6760, operation of the fluid control subassembly is controlled by operating the control unit to permit a controlled release of the volatile fission product in response to a volatile fission product pressure level in the traveling wave nuclear fission reactor. The method 6710 stops at a block 6770. Referring to FIG. 23F, an illustrative method 6780 for operating a nuclear fission reactor fuel assembly starts at a block 6790. At a block 6800, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 6810, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 6820, operation of the fluid control subassembly is controlled by operating a control unit coupled to the fluid control subassembly. At a block 6830, operation of the fluid control subassembly is controlled by operating the control unit to permit a controlled release of the volatile fission product in response to a time schedule associated with the traveling wave nuclear fission reactor. The method 6780 stops at a block 6840. Referring to FIG. 23G, an illustrative method 6850 for operating a nuclear fission reactor fuel assembly starts at a block 6860. At a block 6870, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 6880, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 6890, operation of the fluid control subassembly is controlled by operating a control unit coupled to the fluid control subassembly. At a block 6900, operation of the fluid control subassembly is controlled by operating the control unit to permit a controlled release of the volatile fission product in response to an amount of time the traveling wave nuclear fission reactor is operated. The method 6850 stops at a block 6910. Referring to FIG. 23H, an illustrative method 6920 for operating a nuclear fission reactor fuel assembly starts at a block 6930. At a block 6940, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 6950, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 6960, the enclosure is used so as to enclose the porous nuclear fuel body. The method 6920 stops at a block 6970. Referring to FIG. 23I, an illustrative method 6980 for operating a nuclear fission reactor fuel assembly starts at a block 6990. At a block 7000, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 7010, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 7020, the enclosure is used so as to enclose a fissile material forming the porous nuclear fuel body. The method 6980 stops at a block 7030. Referring to FIG. 23J, an illustrative method 7040 for operating a nuclear fission reactor fuel assembly starts at a block 7050. At a block 7060, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 7070, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 7080, the enclosure is used so as to enclose a fertile material forming the porous nuclear fuel body. The method 7040 stops at a block 7090. Referring to FIG. 23K, an illustrative method 7100 for operating a nuclear fission reactor fuel assembly starts at a block 7110. At a block 7120, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 7130, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 7140, the enclosure is used so as to enclose a mixture of fissile and fertile material forming the porous nuclear fuel body. The method 7100 stops at a block 7150. Referring to FIG. 23L, an illustrative method 7160 for operating a nuclear fission reactor fuel assembly starts at a block 7170. At a block 7180, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 7190, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 7200, the fluid control subassembly is used to permit a controlled release of the volatile fission product in response to a position of the burn wave in the traveling wave nuclear fission reactor. The method 7160 stops at a block 7210. Referring to FIG. 23M, an illustrative method 7220 for operating a nuclear fission reactor fuel assembly starts at a block 7230. At a block 7240, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 7250, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 7260, the enclosure is used so as to enclose a porous nuclear fuel body in the form of a foam defining a plurality of pores. The method 7220 stops at a block 7270. Referring to FIG. 23N, an illustrative method 7280 for operating a nuclear fission reactor fuel assembly starts at a block 7290. At a block 7300, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 7310, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 7320, the enclosure is used to enclose a porous nuclear fuel body defining a plurality of pores, the plurality of pores having a spatially non-uniform distribution. The method 7280 stops at a block 7330. Referring to FIG. 23O, an illustrative method 7340 for operating a nuclear fission reactor fuel assembly starts at a block 7350. At a block 7360, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 7370, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 7380, the enclosure is used so as to enclose a porous nuclear fuel body having a plurality of channels. The method 7340 stops at a block 7390. Referring to FIG. 23P, an illustrative method 7400 for operating a nuclear fission reactor fuel assembly starts at a block 7410. At a block 7420, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 7430, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 7440, the enclosure is used so as to enclose a porous nuclear fuel body having a plurality of channels. At a block 7450, the enclosure is used so as to enclose a porous nuclear fuel body having a plurality of particles defining the plurality of channels therebetween. The method 7400 stops at a block 7460. Referring to FIG. 23Q, an illustrative method 7470 for operating a nuclear fission reactor fuel assembly starts at a block 7480. At a block 7490, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 7500, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 7510, the enclosure is used so as to enclose a porous nuclear fuel body having a plurality of pores, at least one of the pores being of a predetermined configuration for allowing at least a portion of the volatile fission product to escape the porous nuclear fuel body within a predetermined response time. The method 7470 stops at a block 7520. Referring to FIG. 23R, an illustrative method 7530 for operating a nuclear fission reactor fuel assembly starts at a block 7540. At a block 7550, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 7560, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 7570, the enclosure is used so as to enclose a porous nuclear fuel body having a plurality of pores for allowing at least a portion of the volatile fission product to escape within a predetermined response time of between approximately 10 seconds and approximately 1,000 seconds. The method 7530 stops at a block 7580. Referring to FIG. 23S, an illustrative method 7590 for operating a nuclear fission reactor fuel assembly starts at a block 7600. At a block 7610, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 7620, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 7630, the enclosure so as to enclose a porous nuclear fuel body having a plurality of pores for allowing at least a portion of the volatile fission product to escape within a predetermined response time of between approximately one second and approximately 10,000 seconds. The method 7590 stops at a block 7640. Referring to FIG. 23T, an illustrative method 7650 for operating a nuclear fission reactor fuel assembly starts at a block 7660. At a block 7670, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 7680, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 7690, the enclosure is used so as to sealingly enclose a porous nuclear fuel body having a cylindrical-shaped geometry. The method 7650 stops at a block 7700. Referring to FIG. 23U, an illustrative method 7710 for operating a nuclear fission reactor fuel assembly starts at a block 7720. At a block 7730, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 7740, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 7750, the enclosure is used so as to sealingly enclose a porous nuclear fuel body having a polygonal-shaped geometry. The method 7710 stops at a block 7760. Referring to FIG. 23V, an illustrative method 7770 for operating a nuclear fission reactor fuel assembly starts at a block 7780. At a block 7790, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 7800, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 7810, the enclosure is used so as to enclose a porous nuclear fuel body having a plurality of pores for acquiring the volatile fission product released by the burn wave in the traveling wave nuclear fission reactor. The method 7770 stops at a block 7820. Referring to FIG. 23W, an illustrative method 7830 for operating a nuclear fission reactor fuel assembly starts at a block 7840. At a block 7850, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 7860, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 7870, the enclosure is used so as to enclose a porous nuclear fuel body having a plurality of pores to transport the volatile fission product through the porous nuclear fuel body. The method 7830 stops at a block 7880. Referring to FIG. 23X, an illustrative method 7890 for operating a nuclear fission reactor fuel assembly starts at a block 7900. At a block 7910, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 7920, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 7930, the volatile fission product is received into a reservoir coupled to the fluid control subassembly. The method 7890 stops at a block 7940. Referring to FIG. 23Y, an illustrative method 7950 for operating a nuclear fission reactor fuel assembly starts at a block 7960. At a block 7970, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 7980, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 7990, the fluid control subassembly is used to circulate a fission product removal fluid through the porous nuclear fuel body, so that at least a portion of the volatile fission product is removed from the porous nuclear fuel body while the fluid control subassembly circulates the fission product removal fluid through the porous nuclear fuel body. The method 7950 stops at a block 8000. Referring to FIG. 23Z, an illustrative method 8010 for operating a nuclear fission reactor fuel assembly starts at a block 8020. At a block 8030, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 8040, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 8050, the fluid control subassembly is used to circulate a fission product removal fluid through the porous nuclear fuel body, so that at least a portion of the volatile fission product is removed from the porous nuclear fuel body while the fluid control subassembly circulates the fission product removal fluid through the porous nuclear fuel body. At a block 8060, the fission product removal fluid is supplied to the porous nuclear fuel body by using an inlet subassembly. The method 8010 stops at a block 8070. Referring to FIG. 23AA, an illustrative method 8080 for operating a nuclear fission reactor fuel assembly starts at a block 8090. At a block 8100, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 8110, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 8120, the fluid control subassembly is used to circulate a fission product removal fluid through the porous nuclear fuel body, so that at least a portion of the volatile fission product is removed from the porous nuclear fuel body while the fluid control subassembly circulates the fission product removal fluid through the porous nuclear fuel body. At a block 8130, the fission product removal fluid is removed from the porous nuclear fuel body by using an outlet subassembly. The method 8080 stops at a block 8140. Referring to FIG. 23AB, an illustrative method 8150 for operating a nuclear fission reactor fuel assembly starts at a block 8160. At a block 8170, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 8180, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 8190, the fluid control subassembly is used to circulate a fission product removal fluid through the porous nuclear fuel body, so that at least a portion of the volatile fission product is removed from the porous nuclear fuel body while the fluid control subassembly circulates the fission product removal fluid through the porous nuclear fuel body. At a block 8200, the fission product removal fluid is received into a reservoir coupled to the fluid control subassembly. The method 8150 stops at a block 8210. Referring to FIG. 23AC, an illustrative method 8220 for operating a nuclear fission reactor fuel assembly starts at a block 8230. At a block 8240, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 8250, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 8260, the fluid control subassembly is used to circulate a fission product removal fluid through the porous nuclear fuel body, so that at least a portion of the volatile fission product is removed from the porous nuclear fuel body while the fluid control subassembly circulates the fission product removal fluid through the porous nuclear fuel body. At a block 8270, the fission product removal fluid is supplied from a reservoir coupled to the fluid control subassembly. The method 8220 stops at a block 8280. Referring to FIG. 23AD, an illustrative method 8290 for operating a nuclear fission reactor fuel assembly starts at a block 8300. At a block 8310, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 8320, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 8330, the fluid control subassembly is used so that the nuclear fission fuel assembly is configured to circulate a gas through the pores of the porous nuclear fuel body. The method 8290 stops at a block 8340. Referring to FIG. 23AE, an illustrative method 8350 for operating a nuclear fission reactor fuel assembly starts at a block 8360. At a block 8370, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 8380, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 8390, the fluid control subassembly is used so that the nuclear fission fuel assembly is configured to circulate a liquid through the porous nuclear fuel body. The method 8350 stops at a block 8400. Referring to FIG. 23AF, an illustrative method 8410 for operating a nuclear fission reactor fuel assembly starts at a block 8420. At a block 8430, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 8440, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 8450, the method comprises operating a pump. The method 8410 stops at a block 8460. Referring to FIG. 23AG, an illustrative method 8470 for operating a nuclear fission reactor fuel assembly starts at a block 8480. At a block 8490, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 8500, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 8510, a fluid is circulated between the fluid control subassembly and the porous nuclear fuel body by operating a pump integrally connected to the fluid control subassembly. The method 8470 stops at a block 8520. Referring to FIG. 23AH, an illustrative method 8530 for operating a nuclear fission reactor fuel assembly starts at a block 8540. At a block 8550, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 8560, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 8570, the method comprises operating a valve. The method 8530 stops at a block 8580. Referring to FIG. 23AI, an illustrative method 8590 for operating a nuclear fission reactor fuel assembly starts at a block 8600. At a block 8610, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 8620, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 8630, flow of a fluid is controlled between the enclosure and the fluid control subassembly by operating a valve interposed between the enclosure and the fluid control subassembly. The method 8590 stops at a block 8640. Referring to FIG. 23AJ, an illustrative method 8650 for operating a nuclear fission reactor fuel assembly starts at a block 8660. At a block 8670, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 8680, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 8690, flow of a fluid is controlled between the enclosure and the fluid control subassembly by operating a valve interposed between the enclosure and the fluid control subassembly. At a block 8700, flow of a fluid is controlled between the enclosure and the fluid control subassembly by operating a back-flow prevention valve. The method 8650 stops at a block 8710. Referring to FIG. 23AK, an illustrative method 8720 for operating a nuclear fission reactor fuel assembly starts at a block 8730. At a block 8740, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 8750, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 8760, the method comprises operating a controllably breakable barrier. The method 8720 stops at a block 8770. Referring to FIG. 23AL, an illustrative method 8780 for operating a nuclear fission reactor fuel assembly starts at a block 8790. At a block 8800, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 8810, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 8820, a controllably breakable barrier interposed between the enclosure and the fluid control subassembly is used. The method 8780 stops at a block 8830. Referring to FIG. 23AM, an illustrative method 8840 for operating a nuclear fission reactor fuel assembly starts at a block 8850. At a block 8860, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 8870, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 8880, a controllably breakable barrier interposed between the enclosure and the fluid control subassembly is used. At a block 8890, a barrier breakable at a predetermined pressure is used. The method 8840 stops at a block 8900. Referring to FIG. 23AN, an illustrative method 8910 for operating a nuclear fission reactor fuel assembly starts at a block 8920. At a block 8930, an enclosure is used that encloses a porous nuclear fuel body having the volatile fission product therein. At a block 8940, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the porous nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 8950, a controllably breakable barrier interposed between the enclosure and the fluid control subassembly is used. At a block 8960, a barrier breakable by operator action is used. The method 8910 stops at a block 8970. Referring to FIG. 23AO, an illustrative method 8980 for operating a nuclear fission reactor fuel assembly starts at a block 8990. At a block 9000, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 9010, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. The method 8980 stops at a block 9020. Referring to FIG. 23AP, an illustrative method 9030 for operating a nuclear fission reactor fuel assembly starts at a block 9040. At a block 9050, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 9060, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 9070, operation of the fluid control subassembly is controlled by operating a control unit coupled to the fluid control subassembly. The method 9030 stops at a block 9080. Referring to FIG. 23AQ, an illustrative method 9090 for operating a nuclear fission reactor fuel assembly starts at a block 9100. At a block 9110, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 9120, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 9130, the enclosure is used so as to enclose the nuclear fuel body. The method 9090 stops at a block 9140. Referring to FIG. 23AR, an illustrative method 9150 for operating a nuclear fission reactor fuel assembly starts at a block 9160. At a block 9170, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 9180, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 9190, the enclosure is used so as to enclose a fissile material forming the nuclear fuel body. The method 9150 stops at a block 9200. Referring to FIG. 23AS, an illustrative method 9210 for operating a nuclear fission reactor fuel assembly starts at a block 9220. At a block 9230, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 9240, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 9250, the enclosure is used so as to enclose a fertile material forming the nuclear fuel body. The method 9210 stops at a block 9260. Referring to FIG. 23AT, an illustrative method 9270 for operating a nuclear fission reactor fuel assembly starts at a block 9280. At a block 9290, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 9300, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 9310, the enclosure is used so as to enclose a mixture of fissile and fertile material forming the nuclear fuel body. The method 9270 stops at a block 9320. Referring to FIG. 23AU, an illustrative method 9330 for operating a nuclear fission reactor fuel assembly starts at a block 9340. At a block 9350, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 9360, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 9370, the fluid control subassembly is used so as to permit a controlled release of the volatile fission product in response to a position of the burn wave in the traveling wave nuclear fission reactor. The method 9330 stops at a block 9380. Referring to FIG. 23AV, an illustrative method 9390 for operating a nuclear fission reactor fuel assembly starts at a block 9400. At a block 9410, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 9420, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 9430, the fluid control subassembly is used so as to permit a controlled release of the volatile fission product in response to a power level in the traveling wave nuclear fission reactor. The method 9390 stops at a block 9440. Referring to FIG. 23AW, an illustrative method 9450 for operating a nuclear fission reactor fuel assembly starts at a block 9460. At a block 9470, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 9480, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 9490, the fluid control subassembly is used so as to permit a controlled release of the volatile fission product in response to a neutron population level in the traveling wave nuclear fission reactor. The method 9450 stops at a block 9500. Referring to FIG. 23AX, an illustrative method 9510 for operating a nuclear fission reactor fuel assembly starts at a block 9520. At a block 9530, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defusing a plurality of interconnected open-cell pores. At a block 9540, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 9550, the fluid control subassembly is used so as to permit a controlled release of the volatile fission product in response to a volatile fission product pressure level in the traveling wave nuclear fission reactor. The method 9510 stops at a block 9560. Referring to FIG. 23AY, an illustrative method 9570 for operating a nuclear fission reactor fuel assembly starts at a block 9580. At a block 9590, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 9600, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 9610, the fluid control subassembly is used so as to permit a controlled release of the volatile fission product in response to a time schedule associated with the traveling wave nuclear fission reactor. The method 9570 stops at a block 9620. Referring to FIG. 23AZ, an illustrative method 9630 for operating a nuclear fission reactor fuel assembly starts at a block 9640. At a block 9650, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 9660, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 9670, the fluid control subassembly is used so as to permit a controlled release of the volatile fission product in response to an amount of time the traveling wave nuclear fission reactor is operated. The method 9630 stops at a block 9680. Referring to FIG. 23BA, an illustrative method 9690 for operating a nuclear fission reactor fuel assembly starts at a block 9700. At a block 9710, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 9720, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 9730, the volatile fission product is received into a reservoir coupled to the fluid control subassembly. The method 9690 stops at a block 9740. Referring to FIG. 23BB, an illustrative method 9750 for operating a nuclear fission reactor fuel assembly starts at a block 9760. At a block 9770, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 9780, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 9790, the fluid control subassembly is used to circulate a fission product removal fluid through the pores of the nuclear fuel body, so that at least a portion of the volatile fission product is removed from the pores of the nuclear fuel body while the fluid control subassembly circulates the fission product removal fluid through the pores of the nuclear fuel body. The method 9750 stops at a block 9800. Referring to FIG. 23BC, an illustrative method 9810 for operating a nuclear fission reactor fuel assembly starts at a block 9820. At a block 9830, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 9840, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 9850, the fluid control subassembly is used so that the nuclear fission fuel assembly is configured to circulate a fission product removal fluid comprises supplying the fission product removal fluid to the pores of the nuclear fuel body using an inlet subassembly. The method 9810 stops at a block 9860. Referring to FIG. 23BD, an illustrative method 9870 for operating a nuclear fission reactor fuel assembly starts at a block 9880. At a block 9890, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 9900, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 9910, the fluid control subassembly is used so that the nuclear fission fuel assembly is configured to circulate a fission product removal fluid comprises removing the fission product removal fluid from the pores of the nuclear fuel body using an outlet subassembly. The method 9870 stops at a block 9920. Referring to FIG. 23BE, an illustrative method 9930 for operating a nuclear fission reactor fuel assembly starts at a block 9940. At a block 9950, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 9960, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 9970, the fluid control subassembly is used so that the nuclear fission fuel assembly is configured to circulate a heat removal fluid through the pores of the nuclear fuel body, so that at least a portion of the heat generated by the nuclear fuel body is removed from the nuclear fuel body while the fluid control subassembly circulates the heat removal fluid through the pores of the nuclear fuel body. The method 9930 stops at a block 9980. Referring to FIG. 23BF, an illustrative method 9990 for operating a nuclear fission reactor fuel assembly starts at a block 10000. At a block 10010, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 10020, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 10030, the fluid control subassembly is used so that the nuclear fission fuel assembly is configured to circulate a heat removal fluid through the pores of the nuclear fuel body, so that at least a portion of the heat generated by the nuclear fuel body is removed from the nuclear fuel body while the fluid control subassembly circulates the heat removal fluid through the pores of the nuclear fuel body. At a block 10040, the heat removal fluid is received into a reservoir coupled to the fluid control subassembly. The method 9990 stops at a block 10050. Referring to FIG. 23BG, an illustrative method 10060 for operating a nuclear fission reactor fuel assembly starts at a block 10070. At a block 10080, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 10090, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 10100, the fluid control subassembly is used so that the nuclear fission fuel assembly is configured to circulate a heat removal fluid through the pores of the nuclear fuel body, so that at least a portion of the heat generated by the nuclear fuel body is removed from the nuclear fuel body while the fluid control subassembly circulates the heat removal fluid through the pores of the nuclear fuel body. At a block 10110, the heat removal fluid is supplied from a reservoir coupled to the fluid control subassembly. The method 10060 stops at a block 10120. Referring to FIG. 23BH, an illustrative method 10130 for operating a nuclear fission reactor fuel assembly starts at a block 10140. At a block 10150, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 10160, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 10170, the fluid control subassembly is used so that the nuclear fission fuel assembly is configured to circulate a heat removal fluid through the pores of the nuclear fuel body, so that at least a portion of the heat generated by the nuclear fuel body is removed from the nuclear fuel body while the fluid control subassembly circulates the heat removal fluid through the pores of the nuclear fuel body. At a block 10180, heat is removed from the heat removal fluid by using a heat sink coupled to the fluid control subassembly, so that the heat sink is in heat transfer communication with the heat removal fluid. The method 10130 stops at a block 10190. Referring to FIG. 23BI, an illustrative method 10200 for operating a nuclear fission reactor fuel assembly starts at a block 10210. At a block 10220, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 10230, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 10240, the fluid control subassembly is used so that the nuclear fission fuel assembly is configured to circulate a heat removal fluid through the pores of the nuclear fuel body, so that at least a portion of the heat generated by the nuclear fuel body is removed from the nuclear fuel body while the fluid control subassembly circulates the heat removal fluid through the pores of the nuclear fuel body. At a block 10250, heat is removed from the heat removal fluid by using a heat exchanger coupled to the fluid control subassembly, so that the heat exchanger is in heat transfer communication with the heat removal fluid. The method 10200 stops at a block 10260. Referring to FIG. 23BJ, an illustrative method 10270 for operating a nuclear fission reactor fuel assembly starts at a block 10280. At a block 10290, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 10300, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 10310, the fluid control subassembly is used to simultaneously circulate a fission product removal fluid and a heat removal fluid. The method 10270 stops at a block 10311. Referring to FIG. 23BK, an illustrative method 10312 for operating a nuclear fission reactor fuel assembly starts at a block 10313. At a block 10314, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 10315, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 10316, the fluid control subassembly is used to sequentially circulate a fission product removal fluid and a heat removal fluid. The method 10312 stops at a block 10317. Referring to FIG. 23BL, an illustrative method 10318 for operating a nuclear fission reactor fuel assembly starts at a block 10319. At a block 10320, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defusing a plurality of interconnected open-cell pores. At a block 10330, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 10340, the method comprises operating a pump. The method 10318 stops at a block 10350. Referring to FIG. 23BM, an illustrative method 10360 for operating a nuclear fission reactor fuel assembly starts at a block 10370. At a block 10380, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 10390, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 10400, a fluid is pumped between the fluid control subassembly and the pores of the nuclear fuel body by operating a pump integrally connected to the fluid control subassembly. The method 10360 stops at a block 10410. Referring to FIG. 23BN, an illustrative method 10420 for operating a nuclear fission reactor fuel assembly starts at a block 10430. At a block 10440, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 10450, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 10460, a plurality of first components coupled to the fluid control subassembly are used to supply a fission product removal fluid to the fluid control subassembly, so as to enable the fluid control subassembly to circulate the fission product removal fluid through the pores of the nuclear fuel body, whereby at least a portion of the volatile fission product is acquired by the pores of the nuclear fuel body and is removed from the pores of the nuclear fuel body while said fluid control subassembly circulates the fission product removal fluid through the pores of the nuclear fuel body. The method 10420 stops at a block 10470. Referring to FIG. 23BO, an illustrative method 10480 for operating a nuclear fission reactor fuel assembly starts at a block 10490. At a block 10500, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 10510, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 10520, a plurality of first components coupled to the fluid control subassembly are used to supply a fission product removal fluid to the fluid control subassembly, so as to enable the fluid control subassembly to circulate the fission product removal fluid through the pores of the nuclear fuel body, whereby at least a portion of the volatile fission product is acquired by the pores of the nuclear fuel body and is removed from the pores of the nuclear fuel body while said fluid control subassembly circulates the fission product removal fluid through the pores of the nuclear fuel body. At a block 10530, a plurality of second components coupled to the fluid control subassembly are used to supply a heat removal fluid to the fluid control subassembly, so as to enable the fluid control subassembly to circulate a heat removal fluid through the pores of the nuclear fuel body, whereby at least a portion of the heat generated by the nuclear fuel body is removed from the nuclear fuel body while the fluid control subassembly circulates the heat removal fluid through the pores of the nuclear fuel body. The method 10480 stops at a block 10540. Referring to FIG. 23BP, an illustrative method 10550 for operating a nuclear fission reactor fuel assembly starts at a block 10560. At a block 10570, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 10580, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 10590, a plurality of first components coupled to the fluid control subassembly are used to supply a fission product removal fluid to the fluid control subassembly, so as to enable the fluid control subassembly to circulate the fission product removal fluid through the pores of the nuclear fuel body, whereby at least a portion of the volatile fission product is acquired by the pores of the nuclear fuel body and is removed from the pores of the nuclear fuel body while said fluid control subassembly circulates the fission product removal fluid through the pores of the nuclear fuel body. At a block 10600, a plurality of second components coupled to the fluid control subassembly are used to supply a heat removal fluid to the fluid control subassembly, so as to enable the fluid control subassembly to circulate a heat removal fluid through the pores of the nuclear fuel body, whereby at least a portion of the heat generated by the nuclear fuel body is removed from the nuclear fuel body while the fluid control subassembly circulates the heat removal fluid through the pores of the nuclear fuel body. At a block 10610, the first components and the second components are used so that at least one of the first components and at least one of the second components are identical. The method 10550 stops at a block 10620. Referring to FIG. 23BQ, an illustrative method 10630 for operating a nuclear fission reactor fuel assembly starts at a block 10640. At a block 10650, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 10660, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 10670, a dual-purpose circuit coupled to the enclosure is used to selectively remove the volatile fission product and heat from the nuclear fuel body. The method 10630 stops at a block 10680. Referring to FIG. 23BR, an illustrative method 10690 for operating a nuclear fission reactor fuel assembly starts at a block 10700. At a block 10710, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 10720, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 10730, the fluid control subassembly is used to circulate a gas through the pores of the nuclear fuel body. The method 10690 stops at a block 10740. Referring to FIG. 23BS, an illustrative method 10750 for operating a nuclear fission reactor fuel assembly starts at a block 10760. At a block 10770, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 10780, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 10790, the fluid control subassembly is used to circulate a liquid through the pores of the nuclear fuel body. The method 10750 stops at a block 10800. Referring to FIG. 23BT, an illustrative method 10810 for operating a nuclear fission reactor fuel assembly starts at a block 10820. At a block 10830, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 10840, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 10850, the enclosure is used so as to enclose a nuclear fuel body in the form of a foam defining the plurality of pores. The method 10810 stops at a block 10860. Referring to FIG. 23BU, an illustrative method 10870 for operating a nuclear fission reactor fuel assembly starts at a block 10880. At a block 10890, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 10900, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 10910, the enclosure is used so as to enclose a nuclear fuel body having a plurality of channels. The method 10870 stops at a block 10920. Referring to FIG. 23BV, an illustrative method 10930 for operating a nuclear fission reactor fuel assembly starts at a block 10940. At a block 10950, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 10960, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 10970, the enclosure is used so as to enclose a nuclear fuel body having a plurality of channels. At a block 10980, the enclosure is used so as to enclose a nuclear fuel body having a plurality of particles defining the plurality of channels therebetween. The method 10930 stops at a block 10990. Referring to FIG. 23BW, an illustrative method 11000 for operating a nuclear fission reactor fuel assembly starts at a block 11010. At a block 11020, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 11030, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 11040, the enclosure is used so as to enclose a nuclear fuel body defining the plurality of pores, the plurality of pores having a spatially non-uniform distribution. The method 11000 stops at a block 11050. Referring to FIG. 23BX, an illustrative method 11060 for operating a nuclear fission reactor fuel assembly starts at a block 11070. At a block 11080, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 11090, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 11100, the enclosure is used so as to enclose a nuclear fuel body having the plurality of pores for acquiring the volatile fission product released by the burn wave in the traveling wave nuclear fission reactor. The method 11060 stops at a block 11110. Referring to FIG. 23BY, an illustrative method 11120 for operating a nuclear fission reactor fuel assembly starts at a block 11130. At a block 11140, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 11150, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 11160, the enclosure is used so as to enclose a nuclear fuel body having the plurality of pores, one or more of the plurality of pores being of a predetermined configuration to allow at least a portion of the volatile fission product to escape the nuclear fuel body within a predetermined response time. The method 11120 stops at a block 11170. Referring to FIG. 23BZ, an illustrative method 11180 for operating a nuclear fission reactor fuel assembly starts at a block 11190. At a block 11200, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 11210, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 11220, the enclosure is used so as to enclose a nuclear fuel body having the plurality of pores to allow at least a portion of the volatile fission product to escape the nuclear fuel body within a predetermined response time of between approximately 10 seconds and approximately 1,000 seconds. The method 11180 stops at a block 11230. Referring to FIG. 23CA, an illustrative method 11240 for operating a nuclear fission reactor fuel assembly starts at a block 11250. At a block 11260, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 11270, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 11280, the enclosure is used so as to enclose a nuclear fuel body having the plurality of pores to allow at least a portion of the volatile fission product to escape the nuclear fuel body within a predetermined response time of between approximately one second and approximately 10,000 seconds. The method 11240 stops at a block 11290. Referring to FIG. 23CB, an illustrative method 11300 for operating a nuclear fission reactor fuel assembly starts at a block 11310. At a block 11320, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 11330, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 11340, the enclosure is used so as to enclose a nuclear fuel body having the plurality of pores to transport the volatile fission product through the nuclear fuel body. The method 11300 stops at a block 11350. Referring to FIG. 23CC, an illustrative method 11360 for operating a nuclear fission reactor fuel assembly starts at a block 11370. At a block 11380, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 11390, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 11400 the enclosure is used so as to sealingly enclose a nuclear fuel body having a cylindrical-shaped geometry. The method 11360 stops at a block 11410. Referring to FIG. 23CD, an illustrative method 11420 for operating a nuclear fission reactor fuel assembly starts at a block 11430. At a block 11440, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 11450, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 11460, the enclosure is used so as to sealingly enclose a nuclear fuel body having a polygonal-shaped geometry. The method 11420 stops at a block 11470. Referring to FIG. 23CE, an illustrative method 11480 for operating a nuclear fission reactor fuel assembly starts at a block 11490. At a block 11500, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 11510, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 11520, the method comprises operating a valve. The method 11480 stops at a block 11530. Referring to FIG. 23CF, an illustrative method 11540 for operating a nuclear fission reactor fuel assembly starts at a block 11550. At a block 11560, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 11570, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 11580, flow of a fluid is controlled between the enclosure and the fluid control subassembly by operating a valve interposed between the enclosure and the fluid control subassembly. The method 11540 stops at a block 11590. Referring to FIG. 23CG, an illustrative method 11600 for operating a nuclear fission reactor fuel assembly starts at a block 11610. At a block 11620, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 11630, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 11640, flow of a fluid is controlled between the enclosure and the fluid control subassembly by operating a valve interposed between the enclosure and the fluid control subassembly. At a block 11650, flow of a fluid is controlled between the enclosure and the fluid control subassembly by operating a back-flow prevention valve. The method 11600 stops at a block 11660. Referring to FIG. 23CH, an illustrative method 11670 for operating a nuclear fission reactor fuel assembly starts at a block 11680. At a block 11690, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 11700, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 11710, a controllably breakable barrier is used. The method 11670 stops at a block 11720. Referring to FIG. 23CI, an illustrative method 11730 for operating a nuclear fission reactor fuel assembly starts at a block 11740. At a block 11750, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 11760, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 11770, a controllably breakable barrier is interposed between the enclosure and the fluid control subassembly. The method 11730 stops at a block 11780. Referring to FIG. 23CJ, an illustrative method 11790 for operating a nuclear fission reactor fuel assembly starts at a block 11800. At a block 11810, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 11820, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 11830, a controllably breakable barrier is interposed between the enclosure and the fluid control subassembly. The method 11790 stops at a block 11840. Referring to FIG. 23CK, an illustrative method 11850 for operating a nuclear fission reactor fuel assembly starts at a block 11860. At a block 11870, an enclosure is used that encloses a heat-generating nuclear fuel body therein, the nuclear fuel body defining a plurality of interconnected open-cell pores. At a block 11880, a fluid control subassembly coupled to the enclosure is used to control removal of at least a portion of the volatile fission product from the pores of the nuclear fuel body and to control removal of at least a portion of the heat generated by the nuclear fuel body at a plurality of locations corresponding to the burn wave of the traveling wave nuclear fission reactor by controlling fluid flow in a plurality of regions of the traveling wave nuclear fission reactor proximate to the plurality of locations corresponding to the burn wave. At a block 11890, the method comprises interposing a barrier breakable by operator action. The method 11850 stops at a block 11900. One skilled in the art will recognize that the herein described components (e.g., operations), devices, objects, and the discussion accompanying them are used as examples for the sake of conceptual clarity and that various configuration modifications are contemplated. Consequently, as used herein, the specific exemplars set forth and the accompanying discussion are intended to be representative of their more general classes. In general, use of any specific exemplar is intended to be representative of its class, and the non-inclusion of specific components (e.g., operations), devices, and objects should not be taken as limiting. Moreover, those skilled in the art will appreciate that the foregoing specific exemplary processes and/or devices and/or technologies are representative of more general processes and/or devices and/or technologies taught elsewhere herein, such as in the claims filed herewith and/or elsewhere in the present application. While particular aspects of the present subject matter described herein have been shown and described, it will be apparent to those skilled in the art that, based upon the teachings herein, changes and modifications may be made without departing from the subject matter described herein and its broader aspects and, therefore, the appended claims are to encompass within their scope all such changes and modifications as are within the true spirit and scope of the subject matter described herein. It will be understood by those within the art that, in general, terms used herein, and especially in the appended claims (e.g., bodies of the appended claims) are generally intended as “open” terms (e.g., the term “including” should be interpreted as “including but not limited to,” the term “having” should be interpreted as “having at least,” the term “includes” should be interpreted as “includes but is not limited to,” etc.). It will be further understood by those within the art that if a specific number of an introduced claim recitation is intended, such an intent will be explicitly recited in the claim, and in the absence of such recitation no such intent is present. For example, as an aid to understanding, the following appended claims may contain usage of the introductory phrases “at least one” and “one or more” to introduce claim recitations. However, the use of such phrases should not be construed to imply that the introduction of a claim recitation by the indefinite articles “a” or “an” limits any particular claim containing such introduced claim recitation to claims containing only one such recitation, even when the same claim includes the introductory phrases “one or more” or “at least one” and indefinite articles such as “a” or “an” (e.g., “a” and/or “an” should typically be interpreted to mean “at least one” or “one or more”); the same holds true for the use of definite articles used to introduce claim recitations. In addition, even if a specific number of an introduced claim recitation is explicitly recited, those skilled in the art will recognize that such recitation should typically be interpreted to mean at least the recited number (e.g., the bare recitation of “two recitations,” without other modifiers, typically means at least two recitations, or two or more recitations). Furthermore, in those instances where a convention analogous to “at least one of A, B, and C, etc.” is used, in general such a construction is intended in the sense one having skill in the art would understand the convention (e.g., “a system having at least one of A, B, and C” would include but not be limited to systems that have A alone, B alone, C alone, A and B together, A and C together, B and C together, and/or A, B, and C together, etc.). In those instances where a convention analogous to “at least one of A, B, or C, etc.” is used, in general such a construction is intended in the sense one having skill in the art would understand the convention (e.g., “a system having at least one of A, B, or C” would include but not be limited to systems that have A alone, B alone, C alone, A and B together, A and C together, B and C together, and/or A, B, and C together, etc.). It will be further understood by those within the art that typically a disjunctive word and/or phrase presenting two or more alternative terms, whether in the description, claims, or drawings, should be understood to contemplate the possibilities of including one of the terms, either of the terms, or both terms unless context dictates otherwise. For example, the phrase “A or B” will be typically understood to include the possibilities of “A” or “B” or “A and B.” With respect to the appended claims, those skilled in the art will appreciate that recited operations therein may generally be performed in any order. Also, although various operational flows are presented in a sequence(s), it should be understood that the various operations may be performed in other orders than those which are illustrated, or may be performed concurrently. Examples of such alternate orderings may include overlapping, interleaved, interrupted, reordered, incremental, preparatory, supplemental, simultaneous, reverse, or other variant orderings, unless context dictates otherwise. Furthermore, terms like “responsive to,” “related to,” or other past-tense adjectives are generally not intended to exclude such variants, unless context dictates otherwise. While various aspects and embodiments have been disclosed herein, other aspects and embodiments will be apparent to those skilled in the art. For example, each of the embodiments of the nuclear fission reactor fuel assembly may be disposed in a thermal neutron reactor, a fast neutron reactor, a neutron breeder reactor or a fast neutron breeder reactor. Thus, each of the embodiments of the fuel assembly is versatile enough to be beneficially used in various nuclear reactor designs. Therefore, what are provided are a nuclear fission reactor fuel assembly and system configured for controlled removal of a volatile fission product and heat released by a burn wave in a traveling wave nuclear fission reactor and method for same. Moreover, the various aspects and embodiments disclosed herein are for purposes of illustration and are not intended to be limiting, with the true scope and spirit being indicated by the following claims.
	description	This application is a division of U.S. patent application Ser. No. 11/430,195 filed May 9, 2006, now U.S. Pat. No. 7,435,969 and claims priority of Japanese patent application No. 2005-286918 fled Sep. 30, 2005, each of which is incorporated herein by reference. The present invention relates generally to deflectors for deflecting electron beams, ion beams, and other electrically charged beams. More specifically, the invention relates to an electrostatic deflector for use in electron beam exposure apparatuses, ion implantation apparatuses, electron microscopes, and the like. Some of traditionally known deflectors are outlined below. Japanese Patent Laid-open No. 2-100250 describes an electrostatic deflector having four sector-form electrodes arranged, with rod-shaped structures as their supporting columns, inside a cylindrical insulator. Also, Japanese Patent Laid-open No. 8-171881 describes an electrostatic deflector constructed by machining eight split electrode pieces integrally into flange form and mounting these electrode pieces in or on an electrode supporter. In addition, Japanese Patent Laid-open No. 4-174510 describes a method of manufacturing an electrostatic deflector for an electron beam exposure apparatus. The method described in Japanese Patent Laid-open No. 4-174510 includes the steps of: bonding a cylinder formed of an electrode material, onto the inner surface of a cylinder formed of an insulator; cutting the cylinder formed of the electrode material, into a plurality of segments in the axial direction of the cylinder so that the cylinder formed of the insulator is invisible from the path of an electron beam; and using the remaining cut pieces as electrode pieces. Furthermore, Japanese Patent Laid-open No. 2-123651 describes a method of manufacturing an electrostatic deflection electrode having the required number of pole pieces. In the method described in Japanese Patent Laid-open No. 2-123651, after an integrated first component constructed of an electroconductive semiconductor or metal and having a hollow symmetrical shape has been readied for use, a second component constructed of an insulator is embedded in the outer surface or inner surface of the first component, then a plurality of slits each extending from one end of the first component to the other end thereof and terminating at the second component are formed to segment the first component at the slits. Moreover, Japanese Patent Laid-open No. 10-261376 describes a method of manufacturing an electrostatic deflection electrode for an electron beam lithography apparatus. The method described in Japanese Patent Laid-open No. 10-261376 includes: a first step of obtaining a cylindrical material formed of an electroconductive metallic; a second step of providing slits of a required width in the cylindrical material, each of the slits extending from the top of an independent line for sectioning the outer peripheral surface of the cylindrical material circumferentially into eight equal segments, to a radial halfway position on the cylindrical material in the direction of its axial center line; a third step of securing an independent, ring-shaped insulating jig internally to each of the regions provided with the slits at both edges of the cylindrical material in the direction of its axial center line; and a fourth step of extending the inner end side of each slit in the direction of the axial center line under the conditions where the ring-shaped jigs are mounted, and separating the cylindrical material circumferentially into eight electrode elements. Besides, Japanese Patent Laid-open No. 5-29201 describes a method of manufacturing an electrostatic deflection electrode in the manner below. A plurality of outer insulating grooves each extending from the side face of a block towards an electron beam passage region are formed, then an independent insulator is fittingly inserted into each outer insulating groove and bonded onto the inner wall thereof, and a plurality of intermediate insulating grooves are formed. This causes the outer insulating grooves to communicate with associated inner insulating grooves and thus forms a plurality of electrodes each surrounding the electron beam passage region. Such an electrostatic deflector as described in Japanese Patent Laid-open No. 2-100250, however, has a problem in that since four sector-form electrodes must be arranged with rod-shaped structures as their supporting columns inside a cylindrical insulator, too great a deal of working labor is required for efficient manufacture of the electrostatic deflector. Also, such electrostatic deflectors as described in Japanese Patent Laid-open Nos. 8-171881, 4-174510, 2-123651, and 10-261376 have a problem in that since electrodes must be mounted in or on an electrode supporter by means of bonding or the like, a great deal of working labor is required and the electrodes are extremely difficult to arrange in equally spaced form with respect to an electron beam so as not to cause a disturbance of a magnetic field and so as not to bring the electrodes into contact with one another. In addition, such an electrostatic deflector as described in Japanese Patent No. 5-29201, however, has a problem in that since independent insulators are fittingly inserted into outer insulating grooves and then bonded onto the inner walls thereof, too great a deal of working labor is required for efficient manufacture of the electrostatic deflector. An object of the present invention is therefore to provide an electrostatic deflector that can be manufactured easily, efficiently, and very accurately, without using a member for positioning. In an electrostatic deflector manufacturing method and electrostatic deflector according to the present invention, an electrode material formed with slits is connected to an insulator and then the electrode material is cut along the slits, whereby a plurality of electrode members are constructed. More specifically, the method of manufacturing an electrostatic deflector according to the present invention includes: forming a plurality of slits in an essentially conical electrode material to extend in the same direction as that of a bus bar of the electrode material, which has a large-diameter section formed with a flange section for installation on an insulator; and coupling the flange section with the insulator, and then cutting the electrode material to communicate with the slits for manufacture of an integrated electrode formed up of a plurality of electrode members electrically isolated from one another. Also, the electrostatic deflector according to the present invention is outlined below. The electrostatic deflector includes a plurality of electrode members arranged to put slits, extending along a bus bar, therebetween and to be formed into an essentially conical shape. The electrode members are installed on an insulator through a flange portion formed on a large-diameter side of the electrode members. Inn addition, the electrode members are manufactured by forming a plurality of slits in an essentially conical electrode material to extend in the same direction as that of a bus bar of the electrode material, coupling the flange section with the insulator, and then cutting the electrode material along extension lines of the slits for electrical isolation. The slits in the electrode material may be continuously formed spanning from the flange section to the conical section. The electrode material of the approximately conical shape can also be cut from a small-diameter section thereof. Additionally, electrical discharge machining can be employed to perform the above cutting operations. Furthermore, each of the electrode members can be an approximately conical member with required thickness. In the present invention, therefore, first assembling the electrode material into an insulating member integrally without separating the electrode material into each electrode member and by forming slits therein, and then splitting the electrode material allows an electrostatic deflector to be manufactured easily and very accurately without using a member for positioning the electrostatic deflector. A method of manufacturing an electrostatic deflector according to an embodiment of the present invention will be described below. The electrostatic deflector according to the invention is used in, for example, the scanning electron microscope shown in FIG. 1. This scanning electron microscope, after generating an electron beam 41 from an electron beam generator 11 provided in an upper section of a lens barrel 10, first deflects the electron beam via alignment coils 12 (a first deflector) and stigmatic coils 13 (a second deflector). Next, the scanning electron microscope adjusts a magnification using objective lens coils 14 (a magnification controller), and scans a sample 21. After this, the scanning electron microscope activates a detector 30 to detect an electrically charged particle 42 generated from the sample 21, such as a secondary electron or backscattered electron, and displays an image of the sample at an image display device not shown, such as a monitor. The image of the sample can thus be viewed. A detailed structure of this electrostatic deflector is shown in FIG. 2. In the present embodiment, inside the lens barrel 10, the electrostatic deflector 40 according to the embodiment is disposed spanning from a position internal to objective lens coils 50 equivalent to the objective lens coils 40 shown in enlarged view, to a position above the objective lens coils 50. In the present embodiment, electrodes 80 of the electrostatic deflector 40 are each attached to an annular installation member 70 with a screw 71, with an annular insulator 90 sandwiched between the electrode 80 and the annular installation member 70. As shown in FIGS. 2 to 4, each electrode 80 in the present embodiment is constructed of eight electrode members 81 to 88, and the electrode 80 has its entirety tapered as it goes downward, and is installed so as to form a conical shape having an electron beam penetration hole 89 at a front end. FIG. 4 is a view looking from installation flange portions 81a to 88a of FIG. 3. In the present embodiment, the eight electrode members 81 to 88 are of the same shape and as shown in FIG. 4, each of the members is formed symmetrical to an optical axis O and has a clearance 81b to 88b. Also, the flange portion 81a to 88a for installation through the insulator 90 is formed on a large-diameter side of each electrode member 81 to 88 of the conical shape, and the electrode member 81 to 88 provided extending downward from the flange portion 81a to 88a, along the conical shape. In addition, a slit 81d to 88d contiguous to the clearance 81b to 88b is formed spanning from the installation flange portion 81a to 88a to an electrode portion 81c to 88c. In the present embodiment, the installation flange portion 81a to 88a is installed on the insulator 90 by metallization. Next, a method of manufacturing the electrostatic deflector according to the present embodiment is described below. The electrostatic deflector 40 according to the present embodiment is manufactured by assembling into the insulator 90 an electrode material 100 which is an integrated body of the electrode members 81 to 88 and the installation flange portions 81a to 88a, and then cutting the electrode material 100 by electrical discharge machining. In the present embodiment, the electrode material 100 includes, as shown in FIGS. 5 and 6, eight installation flange portions 81a to 88a, a conical section 110 suspended in downward tapered form from a lower position of the installation flange portions 81a to 88a, and a cylindrical section 120 provided at a front end of the conical section 110. The electrode material 100 in the present embodiment is a metallic member and forms a spatial portion 111 inside the conical section 110, and the spatial portion 111 communicates with the electron beam penetration hole 89. Between the installation flange portions 81a to 88a in the present embodiment are also formed the slits 81d to 88d, each of which extends to a required section below, along a bus bar of the conical section 110. The insulator 90 is an annular member as shown in FIGS. 7A and 7B, and is formed with grooves 91 on its connection surfaces with respect to the installation flange portions 81a to 88a. Respective installation positions are set to achieve engagement with the above-mentioned slits 81d to 88d. In addition, through-holes 92 adapted for bolt insertion into the installation member 70 are provided in required portions of the insulator 90. Next, a description is given of manufacturing steps for the electrostatic deflector according to the present embodiment. First, the installation flange portions 81a to 88a of the electrode material 100 are installed on the insulator 90 by metallization. The insulator 90 can be of a material such as ceramics or resin. Also, an adhesive can be used to couple the insulator 90 with the installation flange portions 81a to 88a. After coupling between the electrode material 100 and the insulator 90, the electrode material 100 is divided into eight equal segments. This is accomplished by cutting the electrode material 100 from the cylindrical section 120 thereof, along the cylindrical section 120 and the bus bar of the conical section 110, by use of electrical discharge machining. The clearances 81b to 88b are formed as a result of the cutting operations. The clearances 81b to 88b are thus formed so that they lead to the slits 81d to 88d. In the method of manufacturing the electrostatic deflector according to the present embodiment, since each electrode member 81 is formed by cutting the electrode material 100 with each installation flange portion 81a to 88a and the insulator 90 remaining coupled with one another, a member for positioning is unnecessary and none of the electrode members requires assembly labor, either. The electrostatic deflector can therefore be manufactured easily and accurately. The electrostatic deflector manufactured is mounted in a required disposition section of an electron beam apparatus such as an electron beam exposure apparatus, ion implantation apparatus, or electron microscope. An insulator charge-up preventing component can also be installed internally to the disposition section in which the electrostatic deflector manufactured is mounted. In that case, since charge-up of the insulator can be prevented, this electrostatic deflector, unlike conventional types, makes it possible to avoid increasing the number of components required and complicating the shape of the electrodes. While an electrostatic deflector divided into eight equal segments has been described in the above embodiment, the deflector is not limited to such a structure and may be equally divided into a plurality of segments, such as two, three, or four segments. Also, the shape of the flange of the electrodes and the shape of the slits and clearances provided in the conical structure are not limited to a linear form and can be, for example, a zigzag form. Forming these sections into a zigzag shape makes it possible to prevent charge-up of the insulator, since the electron beam emitted is directly invisible from the insulator.
	abstract	In an X-ray image detector including an X-ray grid unit for transmitting primary X-rays and removing scattered X-rays, a fluorescent substance for emitting fluorescence through excitation by X-rays, and photoelectric conversion elements for photoelectrically converting the fluorescence, these X-ray grid unit, fluorescent substance and photoelectric conversion elements are constituted together as a single unit. The plurality of photoelectric conversion elements are arranged two-dimensionally between each adjacent two of which there is a predetermined insensitive region. The X-ray grid is composed of a plurality of X-ray absorption members for removing the scattered X-rays, and the X-ray absorption members are disposed substantially only on the predetermined insensitive regions when viewed from a direction from which X-rays are incident. Further, the fluorescent substance is disposed substantially only in the regions between the X-ray absorption members that are adjacent to each other when viewed from a direction from which X-rays are incident.
	summary
050680838	abstract	A dashpot in a control rod guide thimble for a nuclear fuel assembly includes a lower tubular portion of an elongated main tube of the guide thimble, an auxiliary hollow tube inserted in the lower portion of the main tube, and an end plug attached to a lower end portion of the auxiliary tube. The auxiliary tube has an outside diameter slightly less than an inside diameter of the main tube to permit a close fitting relationship between an exterior surface of the auxiliary tube and an interior surface of the main tube lower portion. The auxiliary tube also has an upper end portion with an inside surface portion in axial cross-section flaring upwardly and outwardly to provide a tapered transition extending between and connecting an interior surface of the auxiliary tube with the exterior surface thereof.
	claims	1. A fiber reinforced concrete cask formed by injecting into and solidifying concrete within a support frame, wherein reinforcement fiber sheets are disposed at least on an outside circumference surface of said cask, said reinforcement fiber sheets have a coefficient of thermal expansion less than a coefficient of thermal expansion of the concrete, and said support frame is sewn together into a cylindrical bag shape and made from reinforcement fiber sheets. 2. The fiber reinforced concrete cask according to claim 1, wherein said reinforcement fiber sheets are disposed on both the outside circumference surface and the inside circumference surface of said concrete cask, and said reinforcement fiber sheets on said outside and inside circumference surfaces are connected with strings. 3. The fiber reinforced concrete cask according to claim 1,whereinsaid reinforcement fiber sheets are carbon fibers and sewn together into the cylindrical bag shape to form bag-shaped cylindrical structures, andsaid bag-shaped cylindrical structures include hollow cylindrical shapes, hollow cylindrical shapes with a bottom, and structures where a base plate includes cylindrical forms. 4. The fiber reinforced concrete cask according to claim 1, wherein said support frame has an injection port in a lower part of said support frame. 5. A support frame for forming a concrete cask, wherein said support frame is made from reinforcement fiber sheets having a coefficient of thermal expansion that is less than a coefficient of thermal expansion of concrete used to form the concrete cask, and said support frame is sewn together into a cylindrical bag shape and made from the reinforcement fiber sheets. 6. The support frame for forming the concrete cask according to claim 5, wherein said support frame has a double walled structure made from said reinforcement fiber sheets comprising an outside sheet and an inside sheet joined together, and said outside sheet and inside sheet are joined by strings. 7. The support frame for forming the concrete cask according to claim 5, wherein said support frame has an injection port in a lower part of said support frame. 8. A method of fabricating a concrete cask, comprising:forming a support frame for injection of concrete, using reinforcement fiber sheets having a coefficient of thermal expansion less than a coefficient of thermal expansion of the concrete, andinjecting the concrete into said support frame. 9. The method of fabricating the concrete cask according to claim 8, wherein the forming the support frame includes using reinforcement fiber sheets that include an outside sheet and an inside sheet joined together by reinforcement fiber strings. 10. The method for the fabrication of fabricating the concrete cask according to claim 8, further comprisingfilling said formed support frame with a fluid that will maintain a shape of said support frame, andwherein the injecting the concrete is performed after the filling the formed support frame with the fluid and includes injecting the concrete from a bottom of said support frame to replace said fluid, which is pre-filled into said support frame to hold said shape, with the concrete. 11. The method of fabricating the concrete cask according to claim 8, wherein said injecting the concrete is performed so that tensile forces remain in said reinforcement fiber sheets of said support frame from pressure exerted upon said sheets during said injecting the concrete.
054886430	claims	1. A method of supporting a shroud in a pressure vessel of a nuclear reactor comprising the steps of: suspending a plurality of upper hanger rods on a structure disposed in said pressure vessel above the level at which said shroud is suspended; connecting the plurality of upper hanger rods to a multi-segment ring member which is disposed about said shroud by arranging hooks which are provided on the lower end of said upper hanger rods to engage with the ring member so that when said ring member is tightened portions of said hooks are pressed into engagement with said shroud; and tightening said ring member to clamp said portions of said hooks against said shroud. connecting lower hanger rods with said ring member; engaging lower ends of said lower hanger rods with a predetermined portion of said shroud; and retaining the lower ends of said lower hanger rods against the outer surface of said shroud in a manner wherein engagement between the lower ends of said hanger rods and said predetermined portion of said shroud is maintained, using a second ring member. 2. A method as set forth in claim 1, further comprising the steps of: 3. A method as set forth in claim 2, further comprising the step of adjusting the length of said upper hanger rods, before said steps of tightening and after said steps of hooking so as to bring the lower ends of said lower hanger rods into engagement with said predetermined portion of said shroud.
	claims	1. A method of adhering a layer to an imprint lithography substrate, said method comprising:(i) depositing a composition comprising a cross-linker and a multi-functional reactive compound upon said imprint lithography substrate, wherein the multi-functional reactive compound comprises:(a) an organic backbone group; and(b) a first functional group and a second functional group coupled to opposing ends of the organic backbone group, wherein the functional groups are different, and wherein the first functional group is acrylate and the second functional group is selected from the group consisting of a carboxylic group and an epoxy group;(ii) thermally curing said composition, wherein thermally curing comprises:(a) forming interactions between the imprint lithography substrate and the multi-functional reactive compound through the second functional group;(b) forming interactions between the substrate and the second functional group of the multi-functional reactive compound through the cross-linker; and(c) forming covalent or ionic bonds between the cross-linker and the multi-functional reactive compound through the second functional group;(iii) depositing a polymerizable material on said thermally cured composition to form said layer; and(iv) exposing the polymerizable material to actinic energy to polymerize the polymerizable material, wherein polymerizing the polymerizable material comprises forming covalent bonds between the polymerizable material and the multi-functional reactive compound in the thermally cured composition through the first functional group. 2. The method as recited in claim 1, wherein said interactions are selected from a set of mechanisms consisting of covalent bonds, ionic bonds and van der Waals forces. 3. The method as recited in claim 1, wherein depositing the polymerizable material comprises depositing discrete drops of the polymerizable material on the thermally cured composition. 4. The method of claim 1, wherein the cross-linker and the multi-functional reactive compound are different. 5. The method of claim 1, further comprising forming interactions between the cross-linker and the multi-functional reactive compound through the second functional group, wherein forming interactions between the cross-linker and the multi-functional reactive compound through the second functional group comprises cross-linking the composition. 6. A method of adhering a layer to an imprint lithography substrate, said method comprising:(i) depositing a composition comprising a cross-linker and a multi-functional reactive compound upon said imprint lithography substrate, wherein the multi-functional reactive compound comprises:(a) an organic backbone group; and(b) a first functional group and a second functional group coupled to opposing ends of the organic backbone group, wherein the first functional group is an acrylate group and the second functional group is selected from the group consisting of a carboxylic group and an epoxy group;(ii) solidifying said composition to form a solidified composition adhered to the imprint lithography substrate, wherein solidifying comprises:(a) forming an interaction between the second functional group and the imprint lithography substrate, and wherein the interaction is selected from the group consisting of covalent bonds, ionic bonds, and van der Waals forces; and(b) forming interactions between the substrate and the second functional group of the multi-functional reactive compound through the cross-linker; and(c) forming covalent or ionic bonds between the cross-linker and the multi-functional reactive compound through the second functional group;(iii) depositing discrete drops of a polymerizable material upon said solidified composition;(iv) contacting the polymerizable material with an imprint lithography mold;(v) polymerizing the polymerizable material, wherein polymerizing comprises forming covalent bonds between the polymerizable material and the first functional groups in the solidified composition during polymerization of the polymerizable material; and(vi) separating the imprint lithography mold from the polymerized material. 7. The method as recited in claim 6, wherein solidifying said composition comprises thermally curing said composition and wherein polymerizing the polymerizable material comprises exposing the polymerizable material to actinic energy. 8. The method of claim 6, wherein the cross-linker and the multi-functional reactive compound are different. 9. The method of claim 6, further comprising forming interactions between the cross-linker and the multi-functional reactive compound through the second functional group, wherein forming interactions between the cross-linker and the multi-functional reactive compound through the second functional group comprises cross-linking the composition. 10. A method of adhering a layer to an imprint lithography substrate, said method comprising:(i) depositing a composition comprising an organic multi-functional reactive compound upon said imprint lithography substrate, wherein the multi-functional reactive compound comprises:(a) an organic backbone group; and(b) a first functional group and a second functional group coupled to opposing ends of the organic backbone group, wherein the functional groups are different, and wherein the first functional group is an acrylate group and the second functional group is selected from the group consisting of a carboxylic group and an epoxy group;(ii) thermally curing said composition, wherein thermally curing comprises forming interactions between the imprint lithography substrate and the multi-functional reactive compound through the second functional group;(iii) depositing a polymerizable material on said thermally cured composition to form said layer; and(iv) exposing the polymerizable material to actinic energy to polymerize the polymerizable material, wherein polymerizing the polymerizable material comprises forming covalent bonds between the polymerizable material and the multi-functional reactive compound in the thermally cured composition through the first functional group. 11. The method of claim 10, wherein the second functional group is a carboxylic group. 12. The method of claim 10, wherein the second functional group is an epoxy group. 13. The method of claim 10, wherein thermally curing further comprises forming interactions between the cross-linker and the imprint lithography substrate. 14. A method of adhering a layer to an imprint lithography substrate, said method comprising:(i) depositing a composition comprising an organic multi-functional reactive compound upon said imprint lithography substrate, wherein the multi-functional reactive compound comprises:(a) an organic backbone group; and(b) a first functional group and a second functional group coupled to opposing ends of the organic backbone group, and wherein the first functional group is an acrylate group and the second functional group is selected from the group consisting of a carboxylic group and an epoxy group;(ii) solidifying said composition to form a solidified composition adhered to the imprint lithography substrate, wherein solidifying comprises forming an interaction between the second functional group and the imprint lithography substrate, and wherein the interaction is selected from the group consisting of covalent bonds, ionic bonds, and van der Waals forces;(iii) depositing discrete drops of a polymerizable material upon said solidified composition;(iv) contacting the polymerizable material with an imprint lithography mold;(v) polymerizing the polymerizable material, wherein polymerizing comprises forming covalent bonds between the polymerizable material and the first functional groups in the solidified composition during polymerization of the polymerizable material; and(vi) separating the imprint lithography mold from the polymerized material. 15. The method of claim 14, wherein the solidifying further comprises forming interactions between the cross-linker and imprint lithography substrate.
	claims	1. An X-ray fluorescence, XRF spectrometer, for measuring X-ray fluorescence emitted by a target, the XRF spectrometer comprising:an X-ray tube with an anode to emit a divergent X-ray beam;a capillary lens configured to focus the divergent X-ray beam on the target;an aperture system positioned between the anode of the X-ray tube and the capillary lens and comprising at least one pinhole; anda detector configured for detecting X-ray fluorescence radiation emitted by the target,wherein the at least one pinhole is configured for being inserted into the divergent X-ray beam and for reducing a beam cross section of the divergent X-ray beam between the anode and the capillary lens; andthe XRF spectrometer further comprising a control unit configured for adapting the focal depth dF of the XRF spectrometer by controlling the beam cross section of the divergent X-ray beam via the aperture system and based on a topography of the target and/or while scanning the target in at least one selected from among X- and Y-directions. 2. The XRF spectrometer of claim 1, wherein a front focal point of the capillary lens is placed at the anode. 3. The XRF spectrometer of claim 1, wherein the X-ray tube is a microfocus tube and/or wherein the XRF spectrometer is a micro XRF spectrometer. 4. The XRF spectrometer of claim 1,wherein the capillary lens has an entrance aperture, a front focal length, and a front aperture angle α that obey the equation: tan ⁢ ⁢ α = entrance ⁢ ⁢ aperture ⁢ ⁢ ( 11 ) front ⁢ ⁢ focal ⁢ ⁢ length ,and wherein the front focal length corresponds to a distance between the entrance aperture and the anode, and/orwherein the capillary lens has an exit aperture, a rear focal length, and a rear aperture angle β that obey the equation: tan ⁢ β = exit ⁢ ⁢ aperture ⁢ ⁢ ( 12 ) rear ⁢ ⁢ focal ⁢ ⁢ length ,and wherein the rear focal length corresponds to the distance between the exit aperture and the target. 5. The XRF spectrometer of claim 1, wherein the aperture system comprises at least one pinhole of adjustable size. 6. The XRF spectrometer of claim 1, wherein the aperture system comprises a revolver or a slider, each with a plurality of pinholes of different sizes that are each configured for being individually inserted into the divergent X-ray beam. 7. The XRF spectrometer of claim 6, wherein the revolver or the slider further comprises at least one filter for spectrally modifying the divergent X-ray beam. 8. The XRF spectrometer of claim 6, further comprising an additional revolver or slider with at least one filter for spectrally modifying the divergent X-ray beam. 9. A method for adjusting the focal depth dF of an X-ray fluorescence, XRF, spectrometer comprising an X-ray tube with an anode to emit a divergent X-ray beam, a capillary lens configured to focus the divergent X-ray beam on a target, an aperture system positioned between the anode of the X-ray tube and the capillary lens and comprising at least one pinhole, and a control unit configured for controlling the aperture system and for performing the steps of:inserting one of the at least one pinhole in the divergent X-ray beam between the anode and the capillary lens;reducing a cross section of the divergent X-ray beam and a front aperture angle α of the capillary lens with one of the at least one pinhole;increasing the focal depth dF of the XRF spectrometer;estimating a required target focal depth based on a topography of the target; andsetting the focal depth dF of the XRF spectrometer based on the estimated target focal depth. 10. A method for adjusting the focal depth dF of an X-ray fluorescence, XRF, spectrometer comprising an X-ray tube with an anode to emit a divergent X-ray beam, a capillary lens configured to focus the divergent X-ray beam on a target, an aperture system positioned between the anode of the X-ray tube and the capillary lens and comprising at least one pinhole, and a control unit configured for controlling the aperture system and for performing the steps of:inserting one of the at least one pinhole in the divergent X-ray beam between the anode and the capillary lens;reducing a cross section of the divergent X-ray beam and a front aperture angle α of the capillary lens with one of the at least one pinhole;increasing the focal depth dF of the XRF spectrometer;scanning the target in at least one selected from among X- and Y-directions; andadapting the focal depth dF of the XRF spectrometer while scanning the target.
	abstract	An object of the present invention is to provide a sample repairing apparatus, a sample repairing method and a device manufacturing method using the same method, which can reduce an edge roughness in a repaired pattern and also can provide the repairing of a sample by applying an electron beam-assisted etching or an electron beam-assisted deposition. There is provided a sample repairing method comprising: (a) a step of focusing an electron beam by an objective lens to irradiate a sample: (b) a step of supplying a reactive gas onto an electron beam irradiated surface of said sample: (c) a step of selectively scanning a pattern to be repaired on said sample with the electron beam so as to repair said pattern by applying an etching or a deposition; and (d) a step of providing a continuous exhausting operation by means of a differential exhaust system arranged in said objective lens so as to prevent the reactive gas supplied onto said electron beam irradiated surface from flowing toward an electron gun side.
055725597	summary	FIELD OF THE INVENTION This invention relates to a method and apparatus for performing radiography with high energy photons generated by activating water with 14-MeV deuterium-tritium (D-T) fusion neutrons via the .sup.16 O(n,p).sup.16 N reaction followed by the decay of .sup.16 N. More specifically, this invention involves a method and apparatus for studying thick dense objects which are not easily studied with lower energy X-rays or neutrons and which is capable of providing detailed information regarding the structure and composition of the object including the identification of such features as hidden holes and discontinuities in atomic number. BACKGROUND OF THE INVENTION The concept of using penetrating photons to examine the interior regions of objects that cannot be observed directly is about 100 years old. The revolutionary discovery of X-rays by Roentgen in 1895 led promptly to the development of non-destructive, non-invasive interrogation techniques applicable to various objects including the human body. Since the time of Roentgen, this method has developed enormously and now finds routine application in practically every aspect of modern life, e.g., manufacturing, construction, quality control, medicine, defense, transportation, security and basic and applied research. The fundamental principles of photon radiography are well known and widely described in the literature. The most widely used approach involves X-rays in the range of a few keV to several hundred keV that are produced at relatively low cost by electron bombardment of medium to high atomic number metals in sealed, evacuated X-ray tubes. While this approach is extremely versatile, there are limits based on the penetrating capacity of these photons and on attainable source intensities. Photons with higher energies and source intensities can be obtained from radioactive gamma-ray sources, e.g., .sup.60 Co (or .sup.137 Cs) and from electron accelerators such as linacs and synchrotons. Radioactive sources are difficult to handle and store safely. Also, the range of geometric configurations that are possible with these materials is somewhat limited, mainly due to safety considerations. Accelerator sources are capable of producing very high radiation intensities and relatively high photon energies, but like X-ray tubes, they involve continuous energy photon spectra. These machines are also generally rather costly to build and operate. Because photon transmission through matter is highly energy dependent, radiography with continuous energy sources generally suffers from lack of adequate contrast and the inability to select proper exposure. The present invention addresses the aforementioned limitations of the prior art by providing a radiographic method and apparatus which provides essentially monoenergetic, variable intensity, highly penetrating photons in an arrangement which is relatively inexpensive, safe and flexible in configuration for various applications. OBJECTS AND SUMMARY OF THE INVENTION Accordingly, it is an object of the present invention to provide one or more monoenergetic photon beams for use in the non-destructive, non-invasive analysis and testing of thick dense materials and objects. It is another object of the present invention to provide a photon source which is monoenergetic, of variable intensity, highly penetrating and is relatively safe and inexpensive to operate. Yet another object of the present invention is to provide a high energy photon source which employs the deuterium-tritium fusion reactor cooling process and does not present either chemical or radioactivity hazards. A further object of the present invention is to provide apparatus and method for determining the composition and structure of a solid object requiring only modest resolution, but substantial photon penetrating power and has the capability to contrast varying thicknesses of materials and elemental compositions, particularly for metals and higher atomic number materials. The present invention contemplates a method and apparatus for performing radiography with the high energy photons generated by activating water with 14-MeV D-T fusion neutrons via the .sup.16 O(n,p).sup.16 N reaction followed by the decay of .sup.16 N. More specifically, this invention involves a method and apparatus for performing scans of thick dense objects using highly monoenergetic photons produced by activating water with energetic neutrons. The apparatus thus includes a neutron source (normally a 14-MeV neutron generator), a sealed tube of rubber or flexible material in the form of a continuous loop, pure water which is placed inside the sealed tube for receiving the neutron radiation; a water pump; a water flow rate meter; a shielding and collimator system for forming the photon beam and a sodium iodide photon detector and associated electronics for detecting photons transmitted through the material or object being investigated; and for subsequently recording the signals. The water is continuously circulated between the region where it is bombarded with neutrons and becomes radioactive and the radiography portion of the system. The specific activity of the water (Curies per milliliter) depends upon the strength of the neutron field, the time the water spends in this field, and the transport time between the field region and the radiography portion of the system. In general, the intensity of the photon emission at the position of the radiography portion of the system depends on the water flow rate, the volume of water, the intensity of the neutron field and various geometrical factors. A portion of the water line is heavily shield, except for a collimator arrangement for forming the photon beam. The sodium iodide detector is also shielded and views the photon source through a similar collimator arrangement. The object or material to be studied by radiography is transported step-by-step through the gap between the photon source and the detector. The data recorded are photon transmissions, i.e., the ratio of incident photons per unit time and transmitted photons per unit time.
	claims	1. A boiling water reactor, comprising: a coolant system; and a reactor core cooled by water circulated by said coolant system, wherein said coolant system and said reactor core are configured so as to cool said reactor core during rated power operation primarily by natural-circulation of said water and wherein said reactor core has a void reactivity coefficient more than xe2x88x920.079% xcex94k/k/% void fraction and less than xe2x88x920.03% xcex94k/k/% void fraction during rated power operation. 2. A boiling water reactor, comprising: a pressure vessel; a reactor core in said pressure vessel, and having a plurality of fuel assemblies; wherein a fuel assembly includes a plurality of fuel rods arranged with spaces between said fuel rods and a water-rod having a path for coolant, surrounded by a channel box; said reactor core having a by-pass portion for said coolant between said fuel assemblies; and wherein said reactor core is cooled by natural-circulation of water during rated power operation and has a void reactivity coefficient during rated power operation that is more than xe2x88x920.07% xcex94k/k/% void fraction and is less than xe2x88x920.03% xcex94k/k/% void fraction; a main steam pipe for carrying steam generated in said pressure vessel by said reactor core; and a turbine for performing expansion work with said steam. 3. A boiling water reactor according to claim 2 , wherein a fuel rod includes uranium 235 and uranium 238 , and is divided into a plurality of parts in which enrichment of uranium 235 differs in a vertical direction, and wherein an enrichment difference between said parts is more than 0.3 wt % of uranium 235 . claim 2 4. A boiling water reactor according to claim 2 , wherein a fuel rod includes uranium 235 and uranium 238 , and is divided into a plurality of parts in which concentration of uranium 235 differs in a vertical direction, said fuel rod having an upper blanket area and a lower blanket area including at least one of natural uranium and depleted uranium, and having enriched uranium in between said upper blanket area and said lower blanket area, and wherein a length of said lower blanket area is longer than a length of said upper blanket area. claim 2 5. A boiling water reactor according to claim 2 , wherein a fuel rod includes uranium 235 and uranium 238 , and is divided into a plurality of parts in which enrichment of uranium 235 differs in a vertical direction, said fuel rod having a top area and a bottom area comprising a blanket area which includes at least one of natural uranium and depleted uranium. claim 2 6. A boiling water reactor according to claim 2 , wherein a ratio of a width of said by-pass portion to a width of said channel box is greater than or equal to 12%. claim 2 7. A boiling water reactor according to claim 2 , further comprising four water rods arranged at four corners in a fuel assembly. claim 2 8. A boiling water reactor according to claim 2 , further comprising four boiling water stream spaces, the same size as a fuel rod, at four corners of a fuel assembly. claim 2 9. A boiling water reactor according to claim 2 , further comprising: claim 2 a turbine governor valve controlling an amount of steam sent from said pressure vessel to said turbine; and a plurality of relief safety valves to reduce steam pressure of said steam, located upstream of said turbine governor valve, wherein at least one of said plurality of relief safety valves is opened upon receipt of a signal to close said turbine governor valve. 10. A boiling water reactor according to claim 2 , further comprising: claim 2 a circulating pump for making coolant circulate in said pressure vessel. 11. A boiling water reactor according to claim 10 , wherein a flow rate due to natural circulation when said circulating pump stops is at least 70% of total flow rate. claim 10 12. A boiling water reactor according to claim 10 , further comprising a circuit to stop said circulating pump when a turbine governor valve is closed. claim 10 13. A boiling water reactor according to claim 2 , further comprising: claim 2 a control rod to be inserted into said by-pass portion; a control rod guide to guide said control rod when said control rod is inserted into said by-pass portion; and a control rod drive to insert said control rod into a lower part of said pressure vessel from an upper part of said pressure vessel. 14. A method for operating a boiling water reactor, comprising: a pressure vessel; a reactor core in said pressure vessel, and having a plurality of fuel assemblies, at least one fuel assembly including a plurality of fuel rods arranged with spaces between said fuel rods, and a water-rod having a path for coolant, surrounded by a channel box; at least one fuel rod having a blanket area having at least one of natural uranium and depleted uranium in at least one of an upper part and a lower part of said fuel rod; said reactor core having a by-pass portion for said coolant between said fuel assemblies, and said reactor core being cooled by natural-circulation during rated power operations; a control rod to be inserted into said by-pass portion; a control rod guide to guide said control rod when said control rod is inserted into said by-pass portion; a control rod drive to insert said control rod into a lower part of said pressure vessel from an upper part of said pressure vessel; a main steam pipe for carrying steam generated in said pressure vessel; and a turbine for performing expansion work with said steam, comprising the step of: operating said boiling water reactor at or near rated power so as to maintain a void reactivity coefficient of xe2x88x920.07% xcex94k/k/% void fractionxe2x89xa6void reactivity coefficient xe2x89xa6xe2x88x920.03% xcex94k/k/% void fraction. 15. A method according to claim 14 , further comprising operating said boiling water reactor such that at least one control rod is inserted into said reactor core to a border position between a blanket area and an enriched area during a last stage of an operational cycle. claim 14 16. A boiling water reactor according to claim 13 , wherein said boiling water reactor is operated such that at least one control rod is inserted into said reactor core into a border position between a blanket area and an enriched area during a last stage of an operational cycle. claim 13
039705170	summary	BACKGROUND OF THE INVENTION The invention relates to a process of safely compacting a radio-active material into a solid body, said material being presented in a container which is enclosed in a vacuum chamber beforehand, as disclosed in applicant's co-pending U.S. application, having Ser. No. 370,513, filed June 15, 1973. It is known that the temperature of radio-active materials, emitting radiation energy, increases and that consequently they can be used as isotopic heat sources. Such an isotope is separated, as a rule, from fission products by means of chemical separation processes. The isotope to be separated is then bonded in a specific chemical compound, permitting the practical application of the radio-active isotope. As a rule the final product of the chemical separation processes applied, will be a product of average density, which however, is lower than the theoretical density of the chemical compound. Beginning with a radio-active material in the form of a granulated or pulverous chemical mass, such mass will be densified for practical purposes to the highest possible degree, by cold and/or hot compacting, and as the case may be, by a subsequent sintering process into bodies which can be easily handled and displaced. It is a well-known fact that the energy-output per unit of volume is in proportion to the number of radio-active atoms per unit of volume. Consequently, in order to realize optimum energy-output, the density of the solid body should approximate as closely as possible, the theoretical density of the selected chemical compound. SUMMARY OF THE INVENTION The present improvement provides a process, by which the aforementioned aim can be realized in a safe and economically advantageous manner. For this purpose and as described in the above-mentioned co-pending application within "hot cell", the evacuated compressible container is placed within a second safety-container, whereafter the entire arrangement of containers is compacted. Due to such process the density of the radio-active granulated or pulverous mass will be increased to a density, amounting to more than 95% of the chemical compound's theoretical density. According to the improvement, the second container constitutes a safety-buffer-element, preventing the spreading of the radio-active material during compacting. It is advantageous to fill up the second container with a high-pressure transmitting medium, preferably composed of a liquid metal; satisfactory results were obtained with liquid lead. It is also preferably to carry out the compacting of the container-arrangement under a pressure of at least 1000 bar and at a temperature above 1200.degree.C for certain period of time.
	summary
	abstract	A passive auxiliary condensing apparatus of a nuclear power plant includes a steam generation unit configured to heat a water supplied thereto into a steam by a heat produced when operating a nuclear reactor, a water cooled heat exchange unit connected to the steam generation unit and configured to store a cooling water therein to condense the steam provided from the steam generation unit, and a steam-water separation tank including a first side connected to the water cooled heat exchange unit and a second side connected to the steam generation unit, wherein a mixture of a water and a steam provided from the water cooled heat exchange unit is separated into the water and the steam to provide only the water to the steam generation unit.
	summary
054066115	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention is directed to a radiation gating device, of the type suitable for use in an x-ray diagnostics installation. 2. Description of the Prior Art German OS 1 441 312 discloses a gating device for an X-ray diagnostics installation which has two diaphragm plate pairs that are arranged offset by 90.degree. relative to one another in parallel planes. The diaphragm plate pairs can be adjusted in opposite directions via their mounts, which are attached at the end-faces of the diaphragm plates. The Siemens book "Bildgebende Systeme fuer die medizinische Diagnostik", edited by Erich Krestel, published by Siemens A. G., Berlin and Munich, 1980, pages 334 and 335, FIGS. 8.49 and 8.50 shows the beam path of a radiation apparatus in the form of a computer tomograph. The beam of the X-ray radiator is thereby gated to form a fan beam by a focus-proximate gating device and a gating device arranged in front of the arcuate radiation receiver composed of individual detectors. To that end, diaphragm plates arranged lying opposite one another form a slot-shaped opening that is variable via adjustment means. The adjustment of the diaphragm plate ensues in a direction perpendicular to its longitudinal axis, for which appropriate mounts and guides are provided. These mounts and guides are arranged at the end faces of the diaphragm plate aligned perpendicularly relative to the longitudinal axis of the diaphragm plate, so that a large structural width of the gating device is achieved. In order to accomplish the adjustment of the diaphragm plates, the adjustment means must be attached in the adjustment direction of the diaphragm plate, which results in a large structural width of the gating device and a mechanically complicated structure. SUMMARY OF THE INVENTION It is an object of the present invention to provide a radiation gating device having a compact structure and simple design, and which enables a precise adjustment of the diaphragm plate in the beam path. This object is achieved in a radiation gating device constructed in accordance with the principles of the present invention having at least one diaphragm plate adjustable in the beam path of a radiation transmitter, with at least one guide aligned obliquely relative to the longitudinal axis of the diaphragm plate is provided at the diaphragm plate, and adjustment means attached to the diaphragm plate so that the diaphragm plate is adjustable in the beam path in a manner defined by the guide. An advantage of the invention is that the adjustment means can be attached either along the longitudinal axis of the diaphragm plate or perpendicularly relative thereto, as a consequence of the guide extending obliquely relative to the longitudinal axis of the diaphragm plate, as a result of which the adjustment in the beam path can be effected. Only a single guide has to be provided for guiding the diaphragm plate, so that the structural outlay is low. The compact structure of the gating device derives in that the guide is provided at the diaphragm plate and not, as in the prior art, outside the edge region of the diaphragm plate. A structure that is not complicated is achieved in an embodiment wherein the adjustment means if a motor-adjustable spindle that attached to the diaphragm plate via an articulation for the adjustment of the diaphragm plate. In an embodiment which is uncomplicated in terms of fabrication technology, the guide is implemented as a recess provided in the diaphragm lamella into which, a pin, provided at the gating device housing, extends. In an embodiment which is advantageous for obtaining a fan-shaped ray beam, two diaphragm plates are arranged opposite one another and form a slot-shaped opening for gating the beam of the radiation transmitter, and the diaphragm plates are adjustable at their guide arranged obliquely relative to the longitudinal axis of the diaphragm plates, so that the spacing of the diaphragm plates from one another can be increased or diminished. Such a gating device can be especially advantageously utilized in a computer tomography apparatus when the gating device is arranged preceding the radiation receiver in the radiation direction. An economical format of the gating device thereby results because only a single adjustment means need be provided even if a plurality of diaphragm plates are arcuately joined to one another, with the adjustment means attached along the longitudinal axis of each diaphragm plate.
	summary
	abstract	The present invention relates to a radiation measurement instrument calibration facility with the abilities of lowering scattered radiation and shielding background radiation and it is capable of providing a suitable environment for performing performance test, calibration and experiment upon a radiation measurement instrument. In an embodiment, the calibration facility comprises: a shielding device, a collimator, a multi-source irradiator, a radiation baffle, a carrier, an electric door unit and a control unit. With the design of the calibration facility of the present invention, the interference coming from the background radiation and scattered radiation in the laboratory during the radiation measurement instrument calibration can be effectively reduced to enhance the accuracy of measurement or calibration for the instrument, and also the instrument calibration and testing can be performed in radiation fields of low-, medium- and high-dose rate levels to meet the requirements of ISO 4037-1 (1996) Standard.
042788927	description	SPECIFIC DESCRIPTION The drawing shows a radiation-shielding transport or storage receptacle 1 for radioactive wastes, especially for irradiated nuclear reactor fuel elements, which comprises a receptacle shell 2, a bottom 3 and a shielding cover 4. The receptacle shell 2 and the bottom 3 are formed unitarily from cast iron, especially spherolitic cast iron, of cast steel or the like. The shielding cover 4 is provided with a flange which is bolted to a shoulder inset in the mouth of the receptacle. The shell 2 comprises at least one cast-in-place conduit or passage 5 which communicates with the interior of the vessel close to the bottom thereof so that a fluid can be introduced or removed from a fitting 5' at the upper end of the vessel. In the embodiment illustrated and in the best-mode embodiment of the invention, a further passage or conduit 7 is cast in place in the thick wall 2 of the vessel. This conduit 7 opens into the upper end of the interior of the vessel and terminates in a chamber 8 in which a valve 9 is received or into which a valve 9 can be introduced. The conduits 7 and 5 and their valve or valves can be used for circulating a fluid through the interior of the vessel. The valve 9 can also be a pressure-relief valve to which a hose or length of tubing can be connected. As has especially been shown in FIG. 2, the passages 5 and 7 are located in the inner half of the thickness of the wall 2. This permits further passages 10 to be formed in the outer half of the thickness of the wall, the passages 10 extending the full length of the receptacle and along the bottom so that they can be filled with a material of higher radiation-adsorbing cross section, i.e. a so-called moderating material. This has been found to be especially advantageous when the container receives nuclear wastes having a high neutron activity. The passages 10, like the passages 5 and 7, can be closed at the top of the vessel by a safety cover 6 which overlies the shielding cover 4 and is applied after the shielding cover 4 has been bolted in place. The shielding cover 4 has the configuration of a plug to provide the necessary thickness for limiting the passage of radiation out of the interior of the vessel. As is also apparent from the drawing, the exterior of the shell 2 of the vessel is provided with cooling ribs 11 which can run parallel to the generatrix of the vessel wall. The individual cooling ribs 11 are cast unitarily with the wall and can be provided with gaps 12 along their lengths for expansion and contraction. The gaps 12, therefore, subdivide the cooling ribs 11 into elongated sections. According to the present invention, at the upper edge of the vessel wall 2, a continuous upstanding annular welding lip 13 is formed by an upwardly open groove 13' while the safety cover 6 is provided with a corresponding upstanding welding lip 15 along its outer periphery by an upwardly open groove 15'. The lips 13 and 15 are parallel to one another and terminate in a common plane P below the plane P' of the upper surface of the receptacle. The lips 13 and 14 define a welding crevice in which a bead of weldment 17 can be deposited to form the hermetic seal. The lip 13 with the shoulder 13" of the vessel wall 2 provides an annular space 14 in which the cover 6 is received. In the embodiment shown in the drawing, moreover, the weld seam 17 is located outwardly of the fitting 5' and the chamber 8 so that it hermetically seals the passages 5 and 7 as well as the passages 10 if the latter are similarly disposed within the perimeter of this weld seam. Prior to insertion of the cover 4 and the emplacement of the cover 6, water filling the interior of the vessel can be evacuated by the conduit 5. The conduit 5 can, however, be used for other purposes as well. For instance, it can be employed for introducing liquid radioactive wastes into the vessel or for supplying or circulating special coolants to the vessel or for passing a coolant through the vessel to abstract heat from the radioactive wastes contained therein. Any other passages or conduits required for this purpose can also be cast in place within the body of the vessel and closed similarly. As has been shown in FIGS. 1 and 4, the safety cover 6 can be provided with a bore 16 into which can be force-fitted a plug 16' or which can be welded shut. This bore can receive, once the plug 16' or the weldment is removed, a suction line to enable a gas detector to analyze withdrawn gases. When the interior of the vessel is pressurized with helium, the escape of helium into the space below the cover 6 and detected by withdrawal from the passage 16 indicates a failure of the seal between the shielding cover 4 and the remainder of the vessel. As the seal between the shielding cover 4 and the body of the vessel is monitored, any leakage can be detected so that replacement of the shielding cover 4 can be effected or repair of the seal ensured. To this end, the bead 17 of weldment can be simply burned off and the cover 6 removed to effect repair. With replacement of the cover 6, the hermetic seal by the formation of another deposit weld can be re-instituted.
050857096	description	DETAILED DESCRIPTION Mineral deposits or scale derived from subterranean waters form on natural gas handling equipment and media such as pipework, tubing, pumps, filters, screens, and sorption media such as charcoal, silica, alumina beds, as the gas passes through them and the water evaporates or is removed in the processing. The scale deposits frequently include radioactive components, especially the insoluble sulfate of radium, an alkaline earth metal, and of related metals, including thorium and thallium. The scale deposits usually include additional mineral components, for example, the sulfates of the other metals of the alkaline earth group, especially calcium, strontium and barium, which are of low solubility in conventional solvents, as described above. Once they are formed, these scale deposits cannot be readily removed by conventional means since they are both adherent and insoluble to the conventional solvents. Thus, they cannot be readily removed by washing or other simple remedies. The deposits therefore accumulate progressively on the equipment and because many of them are radioactive because of the presence of the radioactive species, increase the activity of the equipment over a period of time until it may no longer be acceptable according to the relevant regulatory standards. According to the present invention, deposits of scale on gas handling equipment and media which include water insoluble alkaline earth metal sulfates including radioactive contaminants such as radium sulfate, are removed by the use of a chemical composition which includes a chelant (chelating agent) in combination with a catalyst or synergist which increases the solubility of the alkaline earth metal sulfates in aqueous solution. The preferred catalyst or synergist is the oxalate anion as described in Ser. No. 07/369,897, but other synergists may also be used including the mono-carboxylate acid synergists as described in Ser. No. 07/431,114 and the thiosulfate or nitriloacetic acid synergists disclosed in Ser. No. 07/484,970 (Case 5710S). Reference is made to these applications for a description of suitable aqueous solvent compositions which may be used for the removal of these scale deposits according to the methods disclosed in this present application. Any of the scale removal compositions disclosed in the related applications identified above, together with other suitable compositions having the same or similar effect may be used in the present techniques and will be more or less preferred according to their effectiveness. The aqueous solvent composition which is used to remove the scale material from the earth comprises a polyaminopolycarboxylic acid such as ethylenediaminetetraacetic acid (EDTA) or diethylenetriaminepentaacetic acid (DTPA) as a chelant or chelating agent which is intended to form a stable complex with the cation of the alkaline earth scale-forming material. Of these chelants, DTPA is the preferred species since it forms the most soluble complexes at greatest reaction rate. EDTA may be used but is somewhat less favorable and, as noted below, may be less responsive to the addition of the catalyst or synergist. The chelant may be added to the solvent in the acid form or, alternatively, as a salt of the acid, preferably the potassium salt. In any event the alkaline conditions used in the scale removal process will convert the free acid to the salt. The concentration of the chelant in the solvent should normally be at least 0.1M in order to achieve acceptable degree of scale removal. Chelant concentrations in excess of 1.0M are usually not necessary and concentrations from about 0.3M up to about 0.6M will normally give good results; although higher concentrations of chelant may be used, there is generally no advantage to doing so because the efficiency of the chelant utilisation will be lower at excess chelant concentrations. In addition to the chelant, the scale removal compositions contain a catalyst or synergist for the dissolution of the scale. As described in the applications referred to above, the synergist is preferably the oxalate anion, a monocarboxylic anion such as mercaptoacetate, hydroxyacetate or aminoacetate or an aromatic acid, preferably salicylate, or thiosulfate or nitriloacetate. Generally these anions are added as salts or the free acid, depending on the stability and availability of the chosen synergist. In either case, however, the relatively alkaline conditions under which the process is operated, will result in the acid, if used, being converted to the salt form. The potassium salts are preferred in view of their greater solubility and for this reason, the solvent should preferably be brought to the desired pH value with a potassium base, preferably potassium hydroxide. The pH of the solvent is adjusted by the addition of a base, preferably potassium hydroxide, to the desired value, permitting scale removal to take place under alkaline conditions preferably at pH values of from about 8.0 to about 14.0, with optimum values being from about 11 to 13, preferably about 12. As noted above, the use of caustic potash is preferred to bring the composition to the desired pH since the potassium salts formed by its use are more soluble than the corresponding sodium salts: it is important to avoid the use of sodium cations when operating at high pH values, above pH 8, and instead, to use potassium or, alternatively, cesium as the cation of the scale-removing agent. Potassium is preferred for economy as well as availability. Thus, the normal course of making up the solvent will be to dissolve the chelant and the potassium salt of the selected synergist in the water to the desired concentration, after which a potassium base, usually potassium hydroxide is added to bring the pH to the desired value of about 12. The concentration of the catalyst or synergist in the aqueous solvent will be of a similar order to that of the chelant: thus, the amount of the synergist anion in the solvent should normally be at least 0.1M in order to achieve a perceptible increase in the efficiency of the scale removal, and concentrations from about 0.3M up to about 0.6M will give good results. Although higher concentrations of the synergist e.g. above 1.0M may be used, there is generally no advantage to doing so because the efficiency of the process will be lower at excess catalyst concentrations. Again, this economic penalty is particularly notable in oilfield operations. In the preferred scale removal compositions, a polyaminopolycarboxylic acid such as ethylenediaminetetraacetic acid (EDTA) or diethylenetriaminepentaacetic acid (DTPA) is used as the chelant, preferably in an amount of 0.1 to 1.0M as the chelant, typically about 0.5M giving good results. The preferred synergist or catalyst is the oxalate anion, as disclosed in Ser. No. 07/369,897. The oxalate is preferably used in an amount of about 0.1 to 1.0M, preferably about 0.5M, with a pH of 10 to 14, preferably 11 to 13, and usually about 12. The desired pH value is obtained by the addition of a base, preferably a potassium base such as caustic potash, potassium hydroxide. If the chelant is added in the form of a salt, the preferred cations for the salt will be potassium since these have been found to give better solubility. An alternative synergist or catalyst is a monocarboxylic acid anion, as described in Ser. No. 07/431,114, preferably salicylate. These anions have also been found to give fast rates of sulfate scale dissolution and are able to take up a high level of sulfate scale into solution so that they represent a particularly favored method of decontaminating has handling equipment and media. The thiosulfate or nitriloacetic acid synergists described in Ser. No. 07/484,970 Case 5107S) may also be used, as described in that application. The amounts of the chelant and synergist used with the moncarboxylic acid and other synergists are comparable to the amounts used with the oxalate synergists and comparable solution pH values are also used, i.e chelant and synergist concentrations from 0.1 to 10M, usually about 0.5M, solution pH from 10 to 14, usually 11 to 13 and for best results, about 12. The scale removal composition may be heated to a temperature between about 25.degree. C. to about 100.degree. C. (or higher if superatmospheric pressure can be employed), in order to improve the dissolution of the insoluble scale species in the composition. Contact time between the equipment and the scale-removing composition is typically from about ten minutes to about 7 hours, depending on the thickness of the scale deposits and the temperature, with faster dissolution of the scale being obtained at the higher temperatures. After remaining in contact with the equipment for the desired time, the composition containing the dissolved scale may be drained off and, if desired, recovered for removal of the dissolved scale species. In the treatment of the equipment and sorption media, the mineral deposits may be removed by washing with the selected solvent. The equipment may, if convenient, be washed with the solvent while still in place or, alternatively, removable items such as filters and minor pieces may be removed and washed with the solvent in a tank. Sorption media such as charcoal, alumina or silica, which are particulate in character, may be slurried with the solution after being unloaded from the sorption vessel or, alternatively, they may be treated in situ in the vessel if the loading of the medium and the mechanical features of the vessel permit this to be done. In either case, contact time will vary according to the thickness of the scale but at treatment temperatures of about 25.degree. to 100.degree. C., the duration of the treatment will normally be about 1 to 6 hours to reduce the radioactivity to acceptable levels. The solvent containing the dissolved scale components may then be treated to recover the dissolved radioactive materials for acceptable disposal methods, for example, by cation exchange onto a suitable cation-exchange resin to bring the radioactive components into solid form. EXAMPLE Samples were taken of a charcoal gas sorption medium, which had become contaminated with radium-226, thallium-208 and thorium 232. The samples contained these contaminants, accumulated over extended periods of time in gas processing, in amounts which precluded their disposal by normal methods. The activity was 24.9 pCi/g.(picocuries/gram.) for the radium component. The charcoal samples were slurried with an aqueous solution of 0.5M DTPA (diethylenetriamine pentaacetic acid) and 0.5M oxalic acid brought to pH=12 by the addition of caustic potash (potassium hydroxide). the slurry was held at a temperature of 90.degree.-100.degree. C. for approximately four hours, after which the charcoal was filtered off and dried. After drying, the activity of the samples was found to be 0.3 pCi/g.(radium-226), low enough to permit disposal of the charcoal by conventional methods.
	claims	1. An ion implanting apparatus which performs an ion implantation by irradiating an ion beam having passed through a separation slit to a substrate, the ion implanting apparatus comprising:(a) an ion source that generates plasma containing a desired type of ion to be implanted into the substrate;(b) an extraction electrode system that extracts an ion beam having a rectangular section and containing the desired type of ion from the plasma generated from the ion source;(c) a mass-separation electromagnet that performs a mass-separation by bending the extracted ion beam in a thickness direction so as to derive the ion beam containing the desired type of ion; and(d) the separation slit that receives the ion beam having passed through the mass-separation electromagnet and that allows the desired type of ion to selectively pass therethrough, wherein the separation slit is operable to vary a shape of a gap through which the ion beam passes, wherein the separation slit includes a first slit and a second slit that are disposed on both sides in a thickness direction of the ion beam so as to be opposed to each other with an interval therebetween, wherein the first slit and the second slit respectively include a plurality of small slits that are separated in a width direction of the ion beam, wherein the small slits are arranged so that a gap through which the ion beam passes is not formed between the small slits that are adjacent to each other in a width direction, and wherein the small slits are operable to move independently in a thickness direction. 2. An ion implanting apparatus which performs an ion implantation by irradiating an ion beam having passed through a separation slit to a substrate, the ion implanting apparatus comprising:(a) an ion source that generates plasma containing a desired type of ion to be implanted into the substrate;(b) an extraction electrode system that extracts an ion beam having a rectangular section and containing the desired type of ion from the plasma generated from the ion source;(c) a mass-separation electromagnet that performs a mass-separation by bending the extracted ion beam in a thickness direction so as to derive the ion beam containing the desired type of ion;(d) the separation slit that receives the ion beam having passed through the mass-separation electromagnet and that allows the desired type of ion to selectively pass therethrough; and(e) a variable slit that is disposed between the extraction electrode system and the mass-separation electromagnet so as to form a gap through which the ion beam passes and that is operable to vary a shape of the gap so as to shield a part of the ion beam extracted from the ion source, wherein the variable slit includes a first slit and a second slit that are disposed on both sides in a thickness direction of the ion beam so as to be opposed to each other with an interval therebetween, wherein the first slit and the second slit respectively include a plurality of small slits that are separated in a width direction of the ion beam, and wherein the small slits are operable to move independently in a thickness direction. 3. An ion implanting apparatus which performs an ion implantation by irradiating an ion beam having passed through a separation slit to a substrate, the ion implanting apparatus comprising:(a) an ion source that generates plasma containing a desired type of ion to be implanted into the substrate;(b) an extraction electrode system that extracts an ion beam having a rectangular section and containing the desired type of ion from the plasma generated from the ion source;(c) a mass-separation electromagnet that performs a mass-separation by bending the extracted ion beam in a thickness direction so as to derive the ion beam containing the desired type of ion;(d) the separation slit that receives the ion beam having passed through the mass-separation electromagnet and allows the desired type of ion to selectively pass therethrough, wherein the separation slit is operable to vary a shape of a gap through which the ion beam passes; and(e) a variable slit that is disposed between the extraction electrode system and the mass-separation electromagnet so as to form a gap through which the ion beam passes and that is operable to vary a shape of the gap so as to shield a part of the ion beam extracted from the ion source, wherein the variable slit includes a first slit and a second slit that are disposed on both sides in a thickness direction of the ion beam so as to be opposed to each other with an interval therebetween, wherein the first slit and the second slit respectively include a plurality of small slits that are separated in a width direction of the ion beam, and wherein the small slits are operable to move independently in a thickness direction. 4. The ion implanting apparatus according to claim 1, further comprising:(e) a beam profile monitor that is disposed on a downstream side of the mass-separation electromagnet in an ion beam travel direction so as to measure a shape of the section of the ion beam upon receiving the ion beam;(f) an ion monitor that is disposed on a downstream side of the separation slit in an ion beam travel direction so as to measure types and ratios of ions contained in the ion beam upon receiving the ion beam having passed through the separation slit; and(g) a control unit that is operable to independently control respective movements of the plurality of small slits and controls the respective small slits so as to obtain desired mass-separation resolution on the basis of measurement information obtained by the beam profile monitor and the ion monitor. 5. The ion implanting apparatus according to claim 2, further comprising:(f) a beam profile monitor that is disposed on the downstream side of the mass-separation electromagnet in an ion beam travel direction so as to measure a shape of the section of the ion beam upon receiving the ion beam; and(g) a control unit that is operable to independently control respective movements of the plurality of small slits, and that predicts a part of the ion beam of which current density becomes relatively high after passing the mass-separation electromagnet among the ion beam received by the variable slit on the basis of measurement information obtained by the beam profile monitor, and controls the respective small slits so as to shield the part of the ion beam by using the respective small slits disposed at a position corresponding to the predicted part. 6. An ion implanting apparatus which performs an ion implantation by irradiating an ion beam having passed through a separation slit to a substrate, the ion implanting apparatus comprising:(a) an ion source that generates plasma containing a desired type of ion to be implanted into the substrate;(b) an extraction electrode system that extracts an ion beam having a rectangular section and containing the desired type of ion from the plasma generated from the ion source;(c) a mass-separation electromagnet that performs a mass-separation by bending the extracted ion beam in a thickness direction so as to derive the ion beam containing the desired type of ion;(d) the separation slit that receives the ion beam having passed through the mass-separation electromagnet and allows the desired type of ion to selectively pass therethrough, wherein the separation slit is operable to vary a shape of a gap through which the ion beam passes, wherein the separation slit includes a first slit and a second slit that are disposed on both sides in a thickness direction of the ion beam so as to be opposed to each other with an interval therebetween, wherein the first slit and the second slit respectively include a plurality of small slits that are separated in a width direction of the ion beam, wherein the small slits are arranged so that a gap through which the ion beam passes is not formed between the small slits that are adjacent to each other in a width direction, and wherein the small slits are operable to move independently in a thickness direction; and(e) a variable slit that is disposed between the extraction electrode system and the mass-separation electromagnet so as to form a gap through which the ion beam passes and that is operable to vary a shape of the gap so as to shield a part of the ion beam extracted from the ion source. 7. The ion implanting apparatus according to claim 6, further comprising:(f) a beam profile monitor that is disposed on a downstream side of the mass-separation electromagnet in an ion beam travel direction so as to measure a shape of the section of the ion beam upon receiving the ion beam;(g) an ion monitor that is disposed on a downstream side of the separation slit in an ion beam travel direction so as to measure types and ratios of ions contained in the ion beam upon receiving the ion beam having passed through the separation slit; and(h) a control unit that is operable to independently control respective movements of the plurality of small slits and controls the respective small slits so as to obtain desired mass-separation resolution on the basis of measurement information obtained by the beam profile monitor and the ion monitor. 8. The ion implanting apparatus according to claim 3, further comprising:(f) a beam profile monitor that is disposed on the downstream side of the mass-separation electromagnet in an ion beam travel direction so as to measure a shape of the section of the ion beam upon receiving the ion beam; and(g) a control unit that is operable to independently control respective movements of the plurality of small slits, and that predicts a part of the ion beam of which current density becomes relatively high after passing the mass-separation electromagnet among the ion beam received by the variable slit on the basis of measurement information obtained by the beam profile monitor, and controls the respective small slits so as to shield the part of the ion beam by using the respective small slits disposed at a position corresponding to the predicted part.
	abstract	A core of a light water reactor having a plurality of fuel assemblies, which are loaded in said core, having nuclear fuel material containing a plurality of isotopes of transuranium nuclides, an upper blanket zone, a lower blanket zone, and a fissile zone, in which the transuranium nuclides are contained, disposed between the upper blanket zone and the lower blanket zone, wherein a ratio of Pu-239 in all the transuranium nuclides contained in the loaded fuel assembly is in a range of 40 to 60% when burnup of the fuel assembly is 0, sum of a height of the lower blanket zone and a height of the upper blanket zone is in a range of 250 to 600 mm, and the height of said lower blanket zone is in a range of 1.6 to 12 times the height of the upper blanket zone.
	abstract	The present invention relates generally to the field of compensation methods for nuclear reactors and, in particular to a method for fail-safe reactivity compensation in solution-type nuclear reactors. In one embodiment, the fail-safe reactivity compensation method of the present invention augments other control methods for a nuclear reactor. In still another embodiment, the fail-safe reactivity compensation method of the present invention permits one to control a nuclear reaction in a nuclear reactor through a method that does not rely on moving components into or out of a reactor core, nor does the method of the present invention rely on the constant repositioning of control rods within a nuclear reactor in order to maintain a critical state.
	summary

Subsets and Splits

No community queries yet

The top public SQL queries from the community will appear here once available.