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042882915	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention This invention relates to the construction of multisensor radiation detector systems designed for removable insertion into a nuclear flux region such as found in nuclear reactors. Such systems may be referred to as detector assemblies. 2. Description of the Prior Art Radiation detection devices of various configurations which are intended for insertion and use within the core of a nuclear reactor are well known in the prior art. The various designs may be typified by those shown in the U.S. Pat. No. 4,087,693, to Brown, et al., issued May 2, 1978; the U.S. Pat. No. 3,879,612, to Foster, et al., issued Apr. 22, 1975; and the U.S. Pat. No. 3,375,370, to Hilborn, issued Mar. 26, 1968. The radiation detection devices illustrated in these various patents include the type known as self-powered detectors wherein external power is not required. In such a detector an emitter, a collector, and an insulator material between the two are used to generate an electric current which is indicative of the intensity of the radiation present at the location of the emitter element as detailed in the Russian publication titled "Energy Transformation of Short-Life Radio Active Isotopes", by M. G. Mitelman, R. S. Erofeev and N. D. Rosenbloom, 1960. The U.S. patent to Hilborn is relevant for its disclosure of a self-powered neutron detector assembly. The U.S. Patent to Foster, et al, is relevant for its disclosure of a multi-sensor radiation detector system. That patent shows the use of a self-powered detector in combination with what is referred to in that patent as a fission chamber. The detector and the fission chamber are connected electrically in parallel requiring but two conductors extending out of the reactor to an external electrical circuit. Switching means are employed to switch from the detector to the fission chamber. The U.S. Pat. No. 4,087,693 to Brown, et al., discloses the use of silicon dioxide as a dielectric for insulation of the emitter element. While the various prior art radiation detection systems each has its particular advantages, typically all suffer from a common disadvantage resulting from its small diameter (typically 1/16 inch) and the long length of the detector. The typical radiation detector may range in length from 30 feet to over 100 feet. In order to insure accurate measurement of radiation levels, the manufacturing process must be carefully controlled. The detector element used in such detectors normally comprises a lead wire, which may be of a metal such as stainless steel, at the end of which is affixed an emitter element, typically of rhodium. The rhodium emitter and lead wire are insulated from an outer metallic conductor by a material which is typically aluminum oxide or magnesium oxide. The assembly comprising the central lead member, the emitter element, the insulating material and the outer metallic conductor is then swagged so as to compact the insulation material. The insulation is typically in the form of short cylinders of aluminum oxide or magnesium oxide slipped over the center wire and meant to be crushed around the wire during the swagging process. Since these cylinders are hard ceramics, the swagging can nick and break the lead member resulting in a high rejection ratio of sensors. The swagging and other steps in the production process may result in the central lead element and the rhodium emitter element deviating along their length from the center line of the outer conductor. This deviation from the center line may result in inaccuracies in measurements of radiation levels. Also such deviation from center would result in a different thickness of insulation material being present between the rhodium emitter element and the source of the radiation. As the thickness of the insulator material increases, the percentage of the charged particles emitted by the detector which is absorbed by the insulator material also increases. Charged particles so absorbed do not reach the outer conductor and thus are not measured. These inaccuracies in axial alignment of the emitter element can contribute to inaccuracy in measurement of radiation levels. One method of assuring the centering of the center conductor and rhodium emitter element along the axis of the outer conductor element is that disclosed in U.S. Pat. No. 4,087,693. That method comprises the use of a silicon dioxide insulation which is initially in the form of a cloth or woven material which has been formed into a loose sock and placed over the emitter element and the lead wire. This assembly is then inserted into a length of metal tubing forming a sheath which is subsequently drawn through a sizing die. It is stated in that patent that at high levels of compaction the silicon dioxide fibers are easily broken since they are relatively brittle and the broken pieces move or flow with some preference for axial alignment as the outer sheath is drawn through a die. The fragmented insulation then flows easily around the lead wire and the rhodium emitter element and centers it along the length of the detector. This method would appear to be suitable for centering a single lead wire and rhodium element within a single outer conductor. However, such a method would not insure proper centering and alignment of multiple rhodium emitter elements and lead wires as well as proper spacing of the multiple detectors throughout the length of a long outer sheath. In prior art detector systems the sensors are individually calibrated and then assembled into a multi-sensor detector; the individual sensors can not thereafter be calibrated. To insure accuracy of measurement, each sensor must be calibrated prior to assembly. It is thus an object of the invention to provide an easily installed, high yield, relatively flexible device for assuring accurate and consistent location of an emitter with respect to the center line of an outer sheath along the full length of the active radiation detecting zone. It is a further object of the invention to provide a multiple sensor detector system which accurately accounts for background radiation and does not require calibration of the sensors. Another object of the invention is to provide enhanced reliability through the absence of nicks and fractures of the emitters and lead wires. SUMMARY OF THE INVENTION The present invention comprises a multi-sensor radiation detector system wherein the alignment and spacing of the plurality of detectors is insured by the use of a common conductor for each of the sensors. The common conductor has a star-shaped cross-section configuration. This star-shaped common conductor element divides the entire active radiation detecting zone of the detector assembly, wherein the sensors are located, into five wedge-shaped sections. The multi-sensor detector is constructed by first providing an emitter element such as a length of rhodium wire. To each end of the rhodium emitter wire is suitably affixed a central lead wire which may be Inconel, nickel, stainless steel, or other suitable metal. Over this combination is extruded an envelope of silicon dioxide insulation. These assemblies are then fitted into the wedge-shaped compartments of the star-shaped common conductor element. Since the star-shaped common conductor element is preformed to divide the interior of the stainless steel outer sheath into equal portions, the axial location and centering of each of the sensors (lead wire plus emitter element) is accurately established and assured. In such an assembly having five wedge-shaped segments, four of the segments contain a lead wire plus emitter element and the fifth contains a lead wire of the same material as all other lead wires (no emitter element) to provide a measurement of background radiation. Each of the four sensors has its emitter element located at a different position along the length of the detector assembly system. This provides for simultaneously detecting and measuring the radiation at four different locations within the core of the reactor. This assembly is then fitted into the sheath of the final detector assembly and lightly swagged to form the sheath such that it is in contact with the star-shaped common conductor. The parts are sized such that the silicon dioxide insulation is only lightly compacted so as not to distort the emitters and lead wires. By use of a common conductor element, such as the star-shaped element shown herein, and the lack of any distortion of the emitter in combination with one of the five sensors actually being a background sensor, the four remaining sensor elements can provide accurate measurements even after assembly into the final configuration of the present invention without calibration. The magnitude of the background radiation is subtracted from the magnitude of the radiation indicated by each separate sensor. The common conductor also effectively shields each detector and its lead wires from the charged particles emitted from its neighboring detectors. (The detectors are emitters when subjected to nuclear radiation.) In this way cross-talk is minimized.
	abstract	A radiographic apparatus includes a radiation source for emitting radiation, a radiation detecting device with detecting elements arranged two-dimensionally, a radiation grid with absorbing foil strips for removing scattered radiation, a physical quantity acquiring device for calculating predetermined physical quantities based on outputs of the radiation detecting device, a physical quantity map generating device for generating a physical quantity map by mapping the predetermined physical quantities, and a physical quantity map smoothing device for smoothing the physical quantities arranged on the physical quantity map in a direction of extension of the absorbing foil strips, thereby to generate an average value map.
	claims	1. An atomic beam source comprising:a tubular cathode that includes an emission portion that includes an emission port through which an atomic beam can be emitted;a rod-shaped first anode disposed inside the cathode; anda rod-shaped second anode disposed inside the cathode and spaced from the first anode,wherein at least one selected from the group consisting of a shape of the cathode, a shape of the first anode, a shape of the second anode, and a positional relationship between the cathode, the first anode, and the second anode is predetermined so that emission of sputter particles resulting from collision of cations, which have been generated by plasma between the first anode and the second anode, with at least one selected from the cathode, the first anode, and the second anode is reduced. 2. The atomic beam source according to claim 1, wherein the first anode and the second anode are arranged parallel to each other so that center axes thereof are positioned on an installation plane parallel to the emission portion, and a value of (H+L)×H2/L is within a range of 750 or more and 1670 or less, where L (mm) represents a distance between the center axes of the first anode and the second anode, and H (mm) represents a distance between the installation plane and the emission portion. 3. The atomic beam source according to claim 1, wherein an inner side of the cathode has a rectangular shape with at least one corner having an edge-truncated shape in a cross section perpendicular to an axis direction of the cathode, or has a circular or elliptic shape in the cross section. 4. The atomic beam source according to claim 3, wherein the edge-truncated shape is either an R surface having a radius of 5 mm or more or a chamfer surface having a height and a width of 15 mm or more each. 5. The atomic beam source according to claim 3, wherein, in the cross section of the cathode, a minimum distance Xmin from a center to the inner side and a maximum distance Xmax from the center to the inner side satisfy 0.5≤Xmin/Xmax ≤1. 6. The atomic beam source according to claim 1, wherein the emission port is formed to have a tendency in which an opening area decreases from an outer surface of the cathode toward an inner surface of the cathode. 7. The atomic beam source according to claim 6, wherein the emission port includes a straight line connecting the outer surface to the inner surface and the straight line has a slope of 4° or more and 20° or less with respect to a direction perpendicular to the emission portion. 8. The atomic beam source according to claim 6, wherein the emission port includes a filter portion disposed at a side close to the inner surface of the cathode so as to have the tendency in which the opening area decreases from the outer surface of the cathode toward the inner surface of the cathode. 9. The atomic beam source according to claim 1, wherein the cathode includes a catching portion configured to catch a sputter component and a discharge portion connected to the catching portion and configured to discharge the sputter component to outside. 10. The atomic beam source according to claim 1, wherein each of the first anode and the second anode includes a projection disposed on a side opposite to a side on which the first anode and the second anode face each other.
	abstract	A device for handling a guide tube assembly. The assembly has an upper tube and a lower tube in each of which there are fixed horizontal guide plates arranged such that they are spaced apart in an axial direction. The handling device itself includes a gripper having a tubular body and two opposing arms located at one end of the tubular body and moveable between a retracted position and a deployed position, the arms bearing against a lower surface of a contacted guide plate when in the deployed position. A control member is located at an opposite end of the tubular body for controlling the arms. The length of the tubular body of the gripper selected to be: a) greater than the distance between an upper end of the upper tube and a first guide plate of the lower tube; and b) less than the distance between the upper end of the upper tube and a third guide plate of the second tube.
	abstract	A low-dose CT imaging system and method that operates according to a pulsed X-ray emission scheme according to a predefined sequence of rotation angles of the X-ray source, along with image reconstruction algorithms to achieve high spatial and temporal resolution for CT scans. The systems and methods involve high speed switching (on the order of milliseconds) to generate pulsed exposure of X-ray radiation to the patient, reducing radiation dose by 4-8 fold, or more.
047073268	abstract	An improved arrangement for attaching and reattaching a top nozzle of a reconstitutable fuel assembly includes a sleeve member associated with each guide thimble of the fuel assembly and complementary elements for attaching the sleeve member and the upper end portion of the guide thimble together. The sleeve member includes an inner tubular alignment sleeve portion which receives the guide thimble upper end portion and extends between the upper hold-down and lower adapter plates of the top nozzle. The sleeve member also includes an outer tubular shroud portion having a lower annular flange which underlies a coil spring surrounding the sleeve portion and interconnects the shroud portion and the sleeve portion. The outer shroud portion extends upwardly about a portion of the coil spring for protecting the spring from damage by coolant cross flow from adjacent fuel assemblies. The complementary elements include upper primary and lower secondary interior spaced annular grooves formed on the alignment sleeve portion and a primary exterior bulge formed on the guide thimble upper end portion. The primary bulge extends into the primary annular groove to connect the sleeve portion to the guide thimble. The secondary groove is adapted to receive a secondary exterior bulge, being formed on the guide thimble upper end portion after an initial severance of an upper segment of the upper end portion, for reconnection of the alignment sleeve portion and the severed guide thimble together.
	claims	1. A cask for loading at least one nuclear fuel assembly in a transport container, including a body with a longitudinal axis configured to cover sealably an upper end of a container, at least one aperture for letting through a fuel assembly, and at least one means configured to maintain a plug for sealing a chamber of the container inside the cask and at a distance from one entrance of said chamber during loading, wherein the maintaining means includes a first arm rotationally mobile around a first axis and a second arm attached to the first arm and rotationally mobile relatively to the first arm, said cask also including external control device for controlling the arms, and wherein the second arm includes a housing for receiving the sealing plug. 2. The cask according to claim 1, including an aperture for letting through a pneumatic tool configured to be connected to the container, and a housing for receiving a plug when the pneumatic tool is connected, said plug configured to seal the aperture for connecting the pneumatic tool to the container. 3. The cask according to claim 1, the control device includes means for controlling the first arm and means for controlling the second arm, the second arm being displaceable independently of the first arm. 4. The cask according to claim 3, wherein the control device is manually actuated and includes assistance means for displacing the arms along predetermined trajectories. 5. The cask according to claim 4, wherein the control device are formed by first and second handwheels firmly attached to first and second connecting shafts, respectively, said first and second connecting shafts being mechanically connected to the first and second arms, respectively. 6. The cask according to claim 5, wherein the assistance means include a fixed flange and a rotationally mobile flange with a connecting arm, the flanges being superimposed, a flange including at least one imprint delimiting two extreme positions of the associated arm and an abutment borne by the other flange, so as to limit the angle of rotation between both flanges. 7. The cask according to claim 6, wherein the mobile flange associated with the second arm is formed by the handwheel. 8. The cask according to claim 6, wherein the mobile flange is distinct from the handwheel for controlling the first arm and jointly rotatable with the latter. 9. The cask according to claim 8, wherein the fixed flange for the first arm is attached onto the upper surface of the cask. 10. The cask according to claim 6, wherein the assistance means include an imprint so as to define two extreme positions adapted to each chamber of the container. 11. The cask according to claim 1, wherein the control device includes indexation means for the position of each arm. 12. The cask according to claim 1, wherein the housing is placed at a free end of the second arm and in that it includes a sealed bottom so as to collect dusts and/or debris borne by the sealing plug. 13. The cask according to claim 1, wherein the arm also includes the housing for the plug of the orifice for pneumatically connecting the container. 14. The cask according to claim 1, wherein the cask includes several apertures each placed opposite to a respective chamber of the container, for loading and unloading nuclear fuel assemblies. 15. The cask according to claim 1, including at its lower end configured to come into contact with the upper end of the container, first and second annular bearing faces, configured to respectively bear against first and a second supporting faces of the container, the second bearing face being positioned radially towards the inside of the cask relatively to the first bearing face, both bearing faces being connected through a cylinder extending along the longitudinal axis. 16. The cask according to claim 1, including a plug for the aperture. 17. The cask according to claim 1, including an aperture sealed by a plug for letting through a pneumatic tool configured to be connected to the container, and a housing for receiving a plug when the pneumatic tool is connected, said plug configured to seal the orifice for connecting the pneumatic tool to the container. 18. A device for loading at least one nuclear fuel assembly in a transport container, including a cask for loading at least one nuclear fuel assembly in a transport container, including a body with a longitudinal axis configured to cover sealably an upper end of a container, at least one aperture for letting through a fuel assembly, and at least one means configured to maintain a plug for sealing a chamber of the container inside the cask and at a distance from one entrance of said chamber during loading, wherein the maintaining means includes a first arm rotationally mobile around a first axis and a second arm attached to the first arm and rotationally mobile relatively to the first arm, said cask also including an external control device for controlling the arms, and wherein the second arm includes a housing for receiving the sealing plug, and dynamic confinement means configured to be connected with a channel communicating with the inside of the chambers and configured to cause a flow of air from the outside to the inside of the cask, and then to the inside of the chambers for the whole opening period of a chamber. 19. The device according to claim 18, wherein the dynamic confinement means include a pneumatic suction exhauster configured to be connected to collector (206) for collecting particles sucked up into the container. 20. A method for loading a container with a nuclear fuel assembly, including the following steps:a) placing a cask for loading at least one nuclear fuel assembly in a transport container, including a body with a longitudinal axis configured to cover sealably an upper end of a container, at least one aperture for letting through a fuel assembly, and at least one means configured to maintain a plug for sealing a chamber of the container inside the cask and at a distance from one entrance of said chamber during loading, wherein the maintaining means includes a first arm rotationally mobile around a first axis and a second arm attached to the first arm and rotationally mobile relatively to the first arm, said cask also including an external control device for controlling the arms, and wherein the second arm includes a housing for receiving the sealing plug, on the upper end of a container,b) connecting dynamic confinement means in a lower portion of the container in order to create an air flow towards the inside of the container,c) removing the plug for the pneumatic tool and placing the pneumatic tool,d) removing the plug from the first cell,e) placing an assembly in the first cell,f) replacing the plug of the first cell,g) repeating steps d) to f) if necessary, for loading other assemblies in the other cells,h) removing the pneumatic tool and replacing the plug for the pneumatic tool,i) disconnecting the dynamic confinement means, andj) removing the cask. 21. The method according to claim 20, wherein the dynamic confinement means are actuated prior to the removal of the plug of the first cell. 22. The method according to claim 20, wherein a step for removing a plug from an associated passage of the cask is carried out prior to the removal of the pneumatic plug of the container or of the plugs of the cells.
058728244	description	DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT As noted above, a californium source produces heavy ions in a not-so-predictable manner. Thus, a feature of the invention is the generation of heavy molecular ions in a predictable or controllable way. According to the invention, a uranium 238 (.sup.238 U) foil is bombarded with neutrons from a pulsed neutron generator. The neutrons contact the foil and because of the high (n, f) reaction cross-section (neutron to fission fragment cross-section) of .sup.238 U, heavy ions come off. The pulsed neutron generator has an adjustable repetition rate, or pulse rate, such that timing information that correlates the masses of heavy molecular ions can be extracted. With this timing information, the heavy molecular ion generation can be characterized and used more effectively in studying a substance. The invention as described herein takes advantage of the n,f reaction in the 238 isotope of uranium. In doing so, the reliance on the random break-up of californium is removed. In the system according to the invention, fission fragment generation is directly related to the neutron burst. That is, the time mark for the secondary desorption event, which corresponds to the desorption of ions from the sample, can be related to the timing of the neutron pulse that caused the primary emission event. A burst of neutrons from a pulsed neutron generator is allowed to impinge on a .sup.238 U foil, causing multiple fission events. Of these, a considerable number of fission fragments called primary ions will propagate toward and fall incident on the thin-film sample target, thereby inducing desorption of the sample target. These multiple fission events occur at a rate of several orders of magnitude (i.e., hundreds or thousands), and cause desorption of the heavy ions from the sample target. This results in multiple, time-correlated events instead of the single time-correlated event of a spontaneous fission using the californium source. In some cases, it will be possible to reduce acquisition times from hours to minutes or even seconds by using the .sup.238 U source. Also, it is possible to optimize TOF windows, as well as the number of pulse cycles, by varying pulse parameters appropriately. According to one embodiment of the invention, a mass spectrometer source contains a pulsed neutron generator with associated control electronics to provided an appropriate neutron flux in the direction of a fissionable material with a high-enough n,f reaction cross-section. The preferred fissionable material is .sup.238 U. Resultant fission fragments, hereafter referred to as primary ions, are then used to desorb heavy molecular ions from a target. The desorbed ions are hereinafter referred to as secondary ions. These secondary ions are then analyzed by a TOF mass discriminator. The timing sequence allowing for TOF mass discrimination, and any decay particle suppression, are associated with the timing sequence of the neutron pulses, resulting in a significantly improved degree of correlation between primary ion generation, secondary ion generation, and detection. A block diagram of a preferred embodiment of the invention is given in FIG. 3. Neutron pulses are sent to a .sup.238 U source 110-3 by a neutron generator 10-3. The number and rate of the neutron pulses 120-3 emitted from the neutron source 10-3 are controlled by pulse out signals 180-3 sent by the control electronics 20-3. The control electronics 20-3 are controlled by control signals 130-3 sent by a computer/analyzer 40-3. By way of example but not by way of limitation, in the preferred embodiment, the computer/analyzer 40-3 can be an IBM-compatible personal computer. As a result of the bombardment of the .sup.238 U source 110-3 by the neutron pulses 120-3, heavy mass ions (also called fission fragments) 140-3 are emitted from the .sup.238 U source 110-3 at a controllable and predictable rate. There is no need to use a complement particle as a time mark (as is done with californium as the radioactive source), since a time zero mark signal 150-3 is sent from the control electronics 20-3 to a Time-to-Digital Converter (TDC) 100-3 in coincidence with each burst of neutron pulses 120-3 emitted from the neutron generator 10-3. The fission fragments 140-3 emitted from the .sup.238 U source 110-3 due to the neutron pulses 120-3 hit the sample stage 60-3. These fission fragments 140-3 then pass through the sample stage 60-3, and impinge on the thin film sample 50-3 attached on one side of the sample stage 60-3. The sample stage 60-3 is positioned close to the .sup.238 U source 110-3, such that a substantial majority of the fission fragments 140-3 emitted from the .sup.238 U source 110-3 will impinge on the sample 50-3. With the closeness in location between the .sup.238 U source and the sample stage 60-3, the solid angle of the dispersal of the fission fragments 140-3 being ionized from the .sup.238 U source will not present a problem with respect to a certain percentage of the fission fragments 140-3 missing the sample stage 60-3. As a result of their masses and high velocities, some of the fission fragments emitted from the my .sup.238 U source are able to deposit large amounts of energy as they impinge on the thin film sample 50-3, allowing for the desorption of high molecular weight species from the sample 50-3. Molecules 160-3 are desorbed off the thin film sample 50-3 and accelerate toward a grid 90-3 held at a fixed potential, typically a ground potential. The grid 90-3 may be made of any of several standard types of grid material, such as a screen material. The desorbed molecules 160-3 then enter a drift region 70-3 with velocities inversely proportional to the square root of their masses. The amount of time needed to travel through the drift region 70-3 determines the mass of each particle desorbed from the thin film sample 50-3. When the desorbed molecules 160-3 exit from the sample 50-3, each of the desorbed molecules 160-3 have a relatively small amount of energy associated with them, typically around 100 electron volts (eV) of kinetic energy. These low energy particles will typically be desorbed off the sample 50-3 in various directions. However, as these desorbed molecules 160-3 get pulled into the drift region 70-3 by the attraction to the high potential at the grid 90-3, the desorbed molecules 160-3 will be pulled in line, so that the desorbed molecules 160-3 will travel through the drift region 70-3 in essentially parallel paths with respect to each other. The high potential at the grid 90-3 causes the desorbed molecules 160-3 to be accelerated towards the grid 90-3, where the desorbed molecules 160-3 then pass through the grid 90-3, and enter a field-free region, also known as the drift region 70-3. In the drift region 70-3, there are no forces acting upon the desorbed molecules 160-3, and so the heavier ones of the desorbed molecules 160-3 lag behind the lighter ones of the desorbed molecules 160-3, due to the fact that each of the desorbed molecules 160-3 enters the drift region 70-3 with the same kinetic energy, and since kinetic energy=1/2massvelocity.sup.2, the heavier mass ions will have slower velocities than the lighter mass ions as these ions pass through the drift region 70-3. These particles are then detected by a detector 80-3 at the end of the drift region 70-3, and the instant in time when each particle arrives at the detector 80-3 is recorded as a multi-stop signal 170-3. By way of example but not by way of limitation, in the preferred embodiment, the detector 80-3 can be a Dual Microchannel Plate, #C-701, manufactured by R. M. Jordan Company. Other types of detectors 80-3 may be used in the invention by one of ordinary skill in the art in keeping within the scope of the invention. The information concerning the particles passes through the Multi-stop Time-to-Digital Converter (TDC) 100-3, and arrives at the computer/analyzer 40-3. Also, by way of example but not by way of limitation, in the preferred embodiment, the TDC 100-3 can be a TOF2 manufactured by Schmidt Industries, a division of SI Diamond Technology. Note that other similar devices may be substituted for the TDC 100-3 as used in the preferred embodiment and still keep within the scope of the invention. The time it takes the particles to travel through the drift region 70-3 and impinge on detector 80-3 is compared against a time zero mark as determined by the time zero mark signal 150-3 received from the control electronics 20-3, and the time difference determines the mass of each of the desorbed particles. This time difference corresponds to the instant in time of a particular neutron pulse being emitted from the neutron generator 10-3 subtracted from the instant in time of an ion being detected at the detector 80-3, wherein the ion detected at the detector 80-3 was desorbed off the sample 50-3 due to the particular neutron pulse. The time determination and comparison can be performed in the computer/analyzer 40-3, or any type of processor as otherwise convenient. As mentioned above, each time zero mark as determined by the time zero mark signals 150-3 are correlated to a corresponding one of the bursts of neutron pulses 120-3. The above-mentioned structure performs the TOF mass discrimination. As described earlier, the californium source provides a significant primary ion yield, but is not actively controllable. The present invention uses a pulsed neutron generator to provide an adequate neutron flux to a suitable fissionable material. .sup.238 U is one such suitable fissionable material, since it has a relatively high n,f reaction cross section, which is approximately 1.2 barns for 14 MeV neutrons. Based on this, one can obtain reasonably high fission fragment yields. These fission fragment yields are strongly dependent on the neutron flux applied to the .sup.238 U by a neutron generator, which can be highly controlled by using current neutron tube technology. One such neutron tube that can be used for implementing the present invention is a pulsed neutron tube developed by Martin Marietta Specialty Components, Inc., which can deliver neutron fluxes in a pulsed mode. In one embodiment, the neutron pulses would each cause about 10000 fission fragments due to a burst of approximately 5-100 nanoseconds. The burst repetition rate would be on the order of 2000 bursts/second. The burst repetition rate can be controlled by appropriate control signals 130-3 sent by the computer/analyzer 40-3 to the control electronics 20-3. Based on the control signals 130-3 received, the control electronics 20-3 sends pulse out signals 160-3 to the neutron generator 10-3 at instants in time corresponding to the desired neutron pulse repetition rate. Using an approach according to the invention, what previously took hours to perform mass spectrometry could be done in a manner of minutes or even seconds. The ion source would then be coupled to a TOF mass discriminator, allowing mass analyzing capability up to 100,000 amu, which is well beyond the range of current conventional mass spectrometers. The time zero mark for the TOF analysis can be derived from the electronics used to drive the neutron generator. Another advantage of a system according to the invention is that all of the fission fragments generated would be synchronized. As a result, the background noise due to fission fragments that produce ions while the ions from the previous fission fragments are being analyzed can be eliminated, since these fission fragments associated with the background noise do not have a time mark associated with them. While preferred embodiments of the invention have been described, modifications of the described embodiments may become apparent to those of ordinary skill in the art, following the teachings of the invention, without departing from the scope of the invention as set forth in the appended claims.
	description	This application claims priority under 35 U.S.C. §119(e) from Provisional Application Ser. No. 61/674,878, entitled “Passive Power Production During SBO With Thermo-Electric Generation,” filed Jul. 24, 2012. 1. Field This invention relates in general to nuclear power plants, and more particularly, to passively activated apparatus for providing auxiliary power to safety equipment in a nuclear power plant under emergency shutdown conditions where there is a loss of conventional onsite and offsite electrical power. 2. Related Art A nuclear reactor, such as a pressurized water reactor, circulates coolant at high pressure through a cooling circuit traversing a reactor pressure vessel containing nuclear fuel for heating the coolant and a steam generator operable to extract energy from the coolant for useful work. A residual heat removal system is typically provided to remove decay heat from the pressure vessel during shutdown. In the event of a loss of coolant, means are provided for adding additional coolant to the system. A coolant loss may involve only a small quantity, whereby additional coolant can be injected from a relatively small high pressure makeup water supply, without depressurizing the reactor coolant circuit. If a major loss of coolant occurs, it is necessary to add coolant from a low pressure supply containing a large quantity of the coolant. Since it is difficult using pumps to overcome the substantial operating pressure of the reactor coolant circuit, e.g., 2250 psi or 150 bar, the reactor coolant circuit is depressurized in the event of a major loss of coolant so that coolant water can be added from an in-containment refueling water storage tank at the ambient pressure within the nuclear reactor system containment shell. The primary circuit of an AP1000® nuclear reactor system 22 (shown in FIG. 1), offered by the Westinghouse Electric Company LLC, Cranberry Township, Pa., of which the present invention is a part, uses a staged pressure reduction system for depressurizing the primary coolant circuit, which is illustrated in FIGS. 1 and 2. A series of valves 72 couple the reactor outlet 56 (also known as the “hot leg” of the primary coolant circuit) to the inside of the containment shell 54 through spargers 74 in the in-containment refueling water storage tank 50, which vent and dissipate the energy of the hot leg coolant into the refueling water in the tank. When the tank heats up and emits steam, the steam is condensed on the inside of the containment shell. When initially commencing the depressurization, the coolant circuit 46 and the in-containment refueling water storage tank are coupled by the depressurization valves 72 through one or more small conduits 76 along the flow path with not insubstantial back pressure. As the pressure in the cooling circuit drops, additional conduits are opened by further actuation of the depressurization valves 72 in stages, each stage opening a larger and/or more direct flow path between the coolant circuit 46 and the containment shell 54. The initial depressurization stages couple a pressurizer tank 80 which is connected by conduits to the coolant circuit hot leg 56 and to spargers 74 in an in-containment refueling water supply tank 50. The spargers 74 comprise conduits leading to small jet orifices submerged in the tank, thus providing back pressure and allowing water to condense from steam emitted by the spargers into the tank 50. The successive depressurization stages have progressively larger conduit inner diameters. A final stage has a large conduit 84 that couples the hot leg directly into the containment shell 54, for example, at a main coolant loop compartment 40 through which the hot leg 56 of the reactor circuit 46 passes. This arrangement reduces the pressure in the coolant circuit expeditiously, substantially to atmospheric pressure, without sudden hydraulic loading of the respective reactor conduits. When the pressure is sufficiently low, water is added to the coolant circuit by gravity flow from the in-containment refueling water storage tank 50. Automatic depressurization in the AP1000® reactor system is a passive safeguard which ensures that the reactor core is cooled even in the case of a major loss of coolant accident such as a large breach in the reactor coolant circuit. Inasmuch as the in-containment refueling water storage tank drains by gravity, no pumps are required. Draining the water into the bottom of the containment building where the reactor vessel is located, develops a fluid pressure head of water in the containment sufficient to force water into the depressurized coolant circuit without relying on active elements such as pumps. Once the coolant circuit is at atmospheric pressure and the containment is flooded, water continues to be forced into the reactor vessel, where the boiling of the water cools the nuclear fuel. Water in the form of steam escaping from the reactor coolant circuit is condensed on the inside walls of the containment shell, and drained back into the in-containment refueling water storage tank to be injected again into the reactor coolant circuit. The AP1000® nuclear power plant has been designed such that in the event of a station blackout, i.e., the total loss of traditional onsite and offsite power, the plant can safely shut itself down and achieve a safe shutdown condition using only passive systems. By traditional onsite and offsite power, we are referring to electric power conventionally generated from onsite and offsite sources. A few simple valves align the passive safety systems when they are automatically actuated. In most cases, these valves are “fail safe.” They require power to stay in their normal, closed position. Loss of power causes them to open into their safe alignment. In all cases, their movement is made using stored energy from springs, compressed gas or batteries. The plant is designed to maintain this condition with no intervention for at least 72 hours after which some operator action is needed to extend the coping period. During the initial 72-hour period, battery banks are used to power any needed equipment and plant monitoring instrumentation, etc. It is desirable to explore additional passive means for extending this coping time beyond 72 hours by utilizing energy that is available within the plant at the time of and subsequent to shutdown. Accordingly, it is an object of this invention to use the resources within the plant to safely maintain the plant beyond 72 hours without operator intervention or the assistance of offsite power. It is a further object of this invention to so extend the coping period without altering the operation of existing plant systems. To achieve the foregoing objectives, this invention provides a nuclear power plant having a reactor with coolant circulating within a fissile nuclear core to carry heat generated within the core to a utilization circuit for creating useful work. The nuclear power plant includes a coolant residual heat removal circuit for dissipating residual heat generated in the core after the reactor has been shut down, especially in the unlikely event an abnormal operating condition is encountered. The residual heat removal circuit includes a residual heat removal conduit for conveying a volume of coolant from the reactor core through the residual heat removal circuit, wherein the residual heat removal conduit includes an uninsulated section. The residual heat removal circuit also includes a heat engine having a first component part in heat exchange relationship with the uninsulated section of the residual heat removal conduit and a second component part in heat exchange relationship with the environment surrounding the uninsulated section. The heat engine is responsive to a temperature difference between the residual heat removal conduit and the environment surrounding the uninsulated section to generate either electrical or mechanical power as an auxiliary power source to assist management of the abnormal operating condition. In one embodiment, the heat engine is a thermoelectric generator preferably fastened to an outer surface of the residual heat removal conduit with a heat conductive clamp. Desirably, the thermal electric generator is supported within a recess in the clamp; and preferably, the recess is in the outer surface of the clamp. In another embodiment, the heat engine is either a Rankine Cycle Engine or a Sterling Cycle Engine. Preferably, the residual heat removal circuit includes a passive residual heat removal heat exchanger having a channel head and the uninsulated section is on a piping section leading to the channel head and/or on the channel head. Typically, the nuclear power plant includes a heat removal and a monitoring system for managing a shutdown of the nuclear plant in the unlikely event of the abnormal operating condition. In accordance with another embodiment of this invention, one or more of the heat removal and the monitoring systems is at least in part operated by an onsite, independent, passively activated power source, wherein the auxiliary power source is connected to extend the operating life of the onsite independent, passively activated power source. Desirably, the auxiliary power source is only active when coolant flow has been initiated through the residual heat removal circuit. From FIG. 1, it can be appreciated that the residual heat removal system 98 includes a conduit 58 connected to the hot leg of the reactor coolant piping 56 that is connected to the passive residual heat removal heat exchanger (PRHR) 14 which is immersed in the in-containment refueling water storage tank (IRWST) 50. The outlet PRHR heat exchanger piping 12 is connected to the channel head of a steam generator 30 and the coolant flow is then directed into the cold leg piping 36 where it is returned to the reactor vessel 60. When this passive cooling system is activated flow will occur through the aforementioned circuit under the action of natural circulation only, carrying hot reactor coolant into the PRHR heat exchanger where it can dissipate a portion of that energy to the water in the IRWST. The reduced temperature coolant is then returned to the reactor vessel. The difference in temperature between the coolant in the hot leg and the coolant in the cold leg results in a corresponding density difference which causes the flow to occur passively. A portion 13 of the PRHR inlet piping 58 (shown in FIG. 2) is intentionally left un-insulated. As a result, when hot coolant flows through piping 58, the pipe surface achieves a temperature close to the hot coolant temperature. In the subject invention, a heat engine 148 including, one embodiment an array of thermo-electric generators 10 (shown in FIGS. 2, 5 and 8) is attached to the outside surface of piping 58 and produces power by operating between the piping temperature and the ambient containment 54 temperature, i.e., a first component part 152 of the heat engine 148 is in heat exchange relationship with the uninsulated pipe section 13 and a second component part 154 is in heat exchange relationship with the environment surrounding the uninsulated section. This invention turns the latent heat removed from the reactor as part of the passive residual heat removal system, into an auxiliary energy source that can be used to power many of the critical functions of the plant during a station blackout condition to maintain the plant in a safe state. In accordance with one embodiment of this invention, schematically illustrated in FIG. 2, thermoelectric generators are mounted on a portion of the piping 58 that is connected to the intake of the passive residual heat removal heat exchanger 14 which is located in the containment refueling water storage tank 50. The passive residual heat removal heat exchanger 14 is connected to the reactor coolant piping in such a way that core decayed heat removal can be accomplished by way of a natural circulation driven flow following a station blackout. A portion 13 of the piping 58 that is attached to the inlet channel head 16 of the passive residual heat removal heat exchanger 14 is intentionally left uninsulated (shown in FIG. 2 with the passive heat removal heat exchanger inlet nozzle, which is connected to the inlet pipe, shown as 58 in FIGS. 3 & 4. In addition, the surface of the channel head 16 itself is also uninsulated. Some of the available surface area of these components is shown in FIGS. 2, 3 and 4. When this heat exchanger 14 is placed into service during a station blackout, the uninsulated portion 13 of the inlet pipe 58 and the channel head 16 will be supplied with hot water from the reactor coolant system which will initially be at about 600° F. (316° C.), and remain at about 350° F. (177° C.) during the first 72-hour period. Thermoelectric generators are most efficient when they have a large temperature difference between the two sides of the thermoelectric generator (commonly referred to as the “hot side” and “cold side”). An exemplary thermoelectric device is illustrated in FIG. 5 and is generally designated by reference character 10. The thermoelectric device 10 generally consists of two or more elements of N and P-type doped semiconductor material 18 that are connected electrically in series and thermally in parallel. The N-type material is doped so that it will have an excess of electrons (more electrons than needed to complete a perfect molecular lattice structure) and the P-type material is doped so that it will have a deficiency of electrons (fewer electrons than are necessary to complete a perfect lattice structure). The extra electrons in the N material and the “holes” resulting from the deficiency of electrons in the P material are the carriers which move the heat energy from a heat source 20 through the thermoelectric material to a heat sink 24 which, in this case, is the environment surrounding the passive residual heat removal piping and/or channel head on which the thermoelectric generator is mounted. The electricity that is generated by a thermoelectric module, which may comprise one or more series or parallel connected thermoelectric elements, such as the element illustrated in FIG. 5, is proportional to the magnitude of the temperature difference between each side of the element 10. An array of thermoelectric generators are mounted on the passive residual heat removal system piping 58 or channel head 16 would experience a large delta temperature between the hot pipe surface and the ambient temperature within the containment and have the capability to generate significant power. FIG. 6 shows the typical power output of a single thermoelectric generator element as a function of time (since the hot temperature decays with time), and FIG. 7 shows the total power produced if all the available surface area of the pipe and the channel head is utilized. Full utilization of the available area in an AP1000® nuclear plant design will accommodate 2,130 thermoelectric generator panels. As can be seen, the total power produced is significant, starting out at over 25 kilowatts and remaining above 5 kilowatts for at least the first 24 hours. This power can be used to recharge batteries or directly power equipment as needed. This is of value to the utility operators of a nuclear plant, because this passively generated power can be used to extend the coping period for the plant by reducing the demand on the existing banks of batteries. Typically, the most efficient commercially available thermoelectric generators are manufactured as relatively small (2-3 inch (5.1-7-6 cm) square) panels. However, these square panels do not fit well around the circumference of a pipe. One way to solve this problem is to utilize a clamp that would fit around the pipe or extend around the circular section of the passive residual heat removal heat exchanger, i.e., the upper channel head in the area of the tube sheet. Such a clamp 138 is shown in FIGS. 2 and 8. The clamp comprises two semi-circular sections 140 which meet around the pipe at each end of the semi-circular sections and are coupled together by matching attachment flanges 142 at one end and at the other end by attachment flanges 144 that can be bolted together. Alternatively, the flanges 142 can be replaced by a hinge. The clamp 138 is constructed from a highly thermally conductive material such as an aluminum alloy, and has machined or cast recesses 146 around the circumference that the thermoelectric generator squares fit into and have good thermal contact with. These clamps would maximize the surface area of the hot side, allowing for the most power generation. Each of the thermoelectric generators may be connected to the others in an array, and the arrays can be connected in series, parallel, or a mix of both, to produce the voltage and current levels needed. The clamp 138, as an alternative to the mating bolted flanges shown in FIG. 8, can be made as a clam shell design to facilitate attachment to existing piping by incorporating a hinge joint in place of the bolted flange 142 at one end of the semi-circular sections and a bolted flange 144 opposite the hinge, or two bolted flanges respectively located on opposite sides of the clamp as shown in FIG. 8. This mounting concept can be extended to any curved surface including the passive residual heat removal heat exchanger channel head 16, with attachment points used as needed to secure the thermoelectric generator mounting hardware to the surface. Alternatively, a Stirling Cycle Engine or a Rankine Cycle Engine can be connected in heat exchange relationship with the piping 58 and/or channel head 16 to convert the delta temperature difference between those surfaces and the surrounding environment into mechanical power which can be used to directly drive pumps or drive a generator to create the auxiliary electric power that may be needed to operate valves and instrumentation. Such an alternate arrangement is figuratively illustrated in FIG. 2 with the block 148 representing either the Rankine Cycle or the Stirling Cycle Engine in heat communication with the piping 58 through a heat pipe 150. Accordingly, the embodiments described herein provide a true passive means for generating power for a nuclear plant from an independent source separate from the nuclear station's conventional source of power, following a station blackout. The heat engines, i.e., the thermoelectric generators or the Rankine Cycle or the Stirling Cycle Engines, are inactive under normal conditions since the piping and/or channel head that they are attached to are cold, but activate automatically when hot fluid passes through the pipe, as is the case when natural circulation driven flow starts. These devices can also be provided with protective coatings that enable them to operate in a steam environment which may be present inside the containment as the in-containment refueling water storage tank water boils off. While specific embodiments of the invention have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. Accordingly, the particular embodiments disclosed are meant to be illustrative only and not limiting as to the scope of the invention which is to be given the full breadth of the appended claims and any and all equivalents thereof.
063273209	summary	FIELD OF THE INVENTION This invention relates to methods and apparatus for loading objects into tubes and is especially, but not exclusively, concerned with the loading of nuclear fuel pellets into fuel tubes or pins. BACKGROUND TO THE INVENTION Cylindrical oxide fuel pellets are pressed form granulated uranium dioxide powder. The pellets are sintered in a hydrogen atmosphere furnace to obtain the required density and are finely ground to obtain the correct diameter. The grinding operation is required in order to carefully control the clearance between the pellet and the tubular cladding, since this clearance determines the heat transfer characteristics of the fuel pin. During the grinding operation, however, a greasy residue of the order of 2 .mu.m may become fixed to the surface of the fuel pellet. A critical operation in the assembly of an oxide fuel pin is the process of inserting the finished fuel pellets into the fuel tube. The insertion problem is concerned with pellet jamming as a result of the formation of chips, debris and the like. The total length of fuel pellets within a fuel pin may exceed 3 meters and, in some designs, 4 metres. It may be required to fill a fuel tube with the correct number of pellets within a short space of time, perhaps of the order of 30 seconds The fuel pellets may typically be manufactured in long and short lengths, nominally 13 mm and 10 mm. The use of both long and short pellets enables the correct pellet stack length to be made up for a particular fuel pin. Pellet jamming is likely to have a major impact upon the production of fuel pins since human access into the assembly areas is extremely difficult and the plant/equipment is normally operated remotely. Uranium dioxide pellets are susceptible to capping, a term used to describe damage to the integrity of the pressing, and chips are easily detached. As the pellets are loaded into the tube, the tube is vibrated simultaneously in both the horizontal and vertical planes, in order to complete the transfer of the pellets once they are inside the tube. It frequently happens that the griding residue which has adhered to the pellet becomes transferred to an orifice through which the pellets pass prior to entry into the fuel tube. Over time this residue builds up and reduces the diameter of the orifice, eventually causing an obstruction which requires manual intervention to clear. A further problem associated with physical contact of the pellet and the loading equipment is encountered when chip debris becomes detached from a pellet during its passage through the orifice. The chip creates a wedge preventing the remaining pellets from being loaded. In order to reduce or eliminate the above-mentioned problems, possible proposals which have been considered include the following: 1. Gravity loading the pellet into a vertically orientated fuel tube. Normally, height restrictions will preclude this possibility. 2. Positive air pressure to blow pellets into the tube. The use of compressed air will blow uranic contamination around the plant and, since the non-entry end of the tube is sealed, this approach is impractical. 3. Using two parallel wires to transport the pellets towards the tube. This method is in current use. The tension in the wire must be constantly checked to ensure the pellets are correctly supported. Otherwise they fall between the wires and must then be manually retrieved When the pellets arrive at the fuel tube they must be physically pushed into the open end. 4. Use of a robotic manipulator to "pick and place" the pellets. The large quantity of pellets which need to be transferred and the time taken for completion of such an operation make this approach impractical 5. Cushion Transfer. This method of transporting pellets using vibration and Floatex type carpet strips is in current use. However this process can release small numbers of fibres which must be prevented from becoming entrained within the fuel pellet stack. This method of trer is therefore not considered appropriate in the immediate vicinity of stack loading. 6. Removal of the end cap from the sealed end of the fuel tube and loading pellets into both ends. The resultant overall manufacturing changes would be relatively so significant hat this approach can not be contemplated In addition, it is necessary, at least for certain overall processes, to use fuel tubes sealed at one end. Removal of the end cap would involve unacceptable redesign. STATEMENTS OF INVENTION According to the present invention there is provided apparatus for successively loading solid objects into a tube via an open end thereof, the other end of said tube being closed, the apparatus comprising means for feeding said objects in a direction towards said open end of said tube, and means for producing a relatively reduced fluid pressure in a first region located between each said object and said open end, prior to the entry of an object into the tube, compared to that in a second region located on the side of said object emote from said open end of the said tube. It has been surprisingly found that it is possible to load objects such as fuel pellets into tubes closed at one end by the use of a reduced pressure or vacuum which draws the pellet into the fuel tube. The means for producing a relatively reduced pressure are, in use, operated continually while the solid objects are being loaded into the tube. Preferably, the tube is a nuclear fuel tube and the objects are substantially cylindrical fuel pellets. Preferably th e feed means includes a housing having a cylindrical bore extending at least partly therethrough, said bore being axially aligned with said tube. More preferably, one end of said bore, remote from said tube, includes a tapered entrance section. It is preferred that the other end of said bore terminates, in use, at a position spaced from the end of the tube. Preferably, the spacing between said other end of said bore and said open end of said tube is less than the length of said pellet i.e., between 1 and 5 mm. More preferably said spacing is about 3 mm. Preferably the means for producing a relatively reduced fluid pressure include means for lowering the pressure in said first region relative to ambient pressure. Preferably the above-mentioned bore extends from one end of said housing to a position within said housing where it opens into a chamber within which, in use, the open end of the tube is positioned. More preferably, means are provided for withdrawing fluid from the chamber. For instance, the chamber may be connected to an air line incorporating a venturi device which, in use, produces a reduced pressure within the chamber. The present invention also provides a method for succesively loading solid objects into a tub via an open end thereof, the other end of said tube being closed, the method comprising feeding the objects in a direction towards said open end, producing a relatively reduced fluid pressure in a first region located between each said object and said open end, prior to the entry of an object into the tube, compared to that in a second region located on that side of said object remote from said open end of said tube.
	description	The present application is related to U.S. patent application Ser. No. 10/851,040 (now U.S. Pat. No. 6,870,172), filed May 21, 2004, by inventors Marian Mankos et al., and entitled “Maskless Reflection Electron Beam Projection Lithography.” The present application is also related to U.S. patent application Ser. No. 10/851,041, filed May 21, 2004, by Harald F. Hess et al., and entitled “Reflective Electron Patterning Device and Method of Using Same.” The disclosures of the above-referenced patent applications are hereby incorporated by reference. 1. Field of the Invention The present invention relates generally to pattern generation for use in electron beam lithography and other applications. 2. Description of the Background Art Electron-Beam Direct Write Lithography As is understood in the art, a lithographic process includes the patterned exposure of a resist so that portions of the resist can be selectively removed to expose underlying areas for selective processing such as by etching, material deposition, implantation and the like. Traditional lithographic processes utilize electromagnetic energy in the form of ultraviolet light for selective exposure of the resist. As an alternative to electromagnetic energy (including x-rays), charged particle beams have been used for high resolution lithographic resist exposure. In particular, electron beams have been used since the low mass of electrons allows relatively accurate control of an electron beam at relatively low power and relatively high speed. Electron beam lithographic systems may be categorized as electron-beam direct write (EBDW) lithography systems and electron beam projection lithography systems. In EBDW lithography, the substrate is sequentially exposed by means of a focused electron beam, wherein the beam either scans in the form of lines over the whole specimen and the desired structure is written on the object by corresponding blanking of the beam, or, as in a vector scan method, the focused electron beam is guided over the regions to be exposed. The beam spot may be shaped by a diaphragm. EBDW is distinguished by high flexibility, since the circuit geometries are stored in the computer and can be optionally varied. Furthermore, very high resolutions can be attained by electron beam writing, since electron foci with small diameters may be attained with electron-optical imaging systems. However, it is disadvantageous that the process is very time-consuming, due to the sequential, point-wise writing. EBDW is therefore at present mainly used for the production of the masks required in projection lithography. In other words, EBDW lithography has the potential to achieve excellent resolution. However, EBDW has a traditional problem relating to its low throughput. For example, it may take ten to one hundred hours to inscribe an entire wafer using EBDW lithography. One previous approach to attempt to increase the throughput is by increasing the beam current. However, when the current density exceeds a certain threshold, electron-electron interactions cause the beam to blur. Conventional Electron-Beam Projection Lithography In electron-beam projection lithography, analogously to optical lithography, a larger portion of a mask is illuminated simultaneously and is imaged on a reduced scale on a wafer by means of projection optics. Since a whole field is imaged simultaneously in electron beam projection lithography, the attainable throughputs can be markedly higher in comparison with electron beam writers. Projecting the electron-beam over a relatively wide area enables use of a high beam current while keeping the beam current density at a level consistent with minimal electron-electron interactions. For example, an area roughly 0.1 millimeters (mm) wide may be illuminated. That area is several orders of magnitude larger than a traditional EBDW system that focuses the beam into a much smaller spot, for example, with a spot size on the order of tens of nanometers (nm) wide. A flood beam 0.1 mm wide would normally not provide a writing resolution sufficiently high for practical use in integrated circuit manufacturing. However, the system and method disclosed herein enables high-resolution writing by partitioning the flood beam into a multitude (for example, four million) of independently controllable beams. While others have tried building multiple columns with multiple sources to achieve multiple beams, there are various difficulties in that approach. For example, there is the difficulty of making the multiple columns behave uniformly. The system and method disclosed herein may be implemented using a single column and a single source. A conventional multi-beam system would require a large array of blankers to achieve a multitude of controllable beams from a single column, each blanker being a small and independently controllable element that can be switched on and off rapidly. However, it is quite problematic to build and control such a large array. For example, a blanker array for a conventional multi-beam system is not normally buildable using integrated circuits because such integrated circuits are opaque to electrons. Another disadvantage of conventional electron-beam projection lithography systems is that a corresponding mask is necessary for each structure to be exposed. The preparation of customer-specific circuits in small numbers is not economic, because of the high costs associated with mask production. One embodiment relates to a dynamic pattern generator for controllably reflecting charged particles. The generator includes at least a controllable light emitter array, an optical lens, and an array of light-sensitive devices. The controllable light emitter array is configured to emit a pattern of light. The optical lens is configured to demagnify the pattern of light. The array of light-sensitive devices is configured to receive the demagnified pattern of light and to produce a corresponding pattern of surface voltages. Another embodiment relates to a method of controllably reflecting charged particles. A pattern of light is controllably emitted from a light emitter array. The pattern of light is demagnified using a lens, and the demagnified pattern of light is received by an array of light-sensitive devices. The array of light-sensitive devices produce a corresponding pattern of surface voltages for reflecting the charged particles. Other embodiments and features are also disclosed. Reflective Electron-Beam Lithography One embodiment of the present invention relates to a technique for electron-beam lithography that overcomes the above-discussed disadvantages and problems with conventional electron beam lithography. This technique may be called reflective electron-beam lithography and is a form of electron-beam projection lithography. Reflective electron-beam lithography is described in U.S. patent application Ser. No. 10/851,040 (now U.S. Pat. No. 6,870,172), filed May 21, 2004, by inventors Marian Mankos et al., and entitled “Maskless Reflection Electron Beam Projection Lithography,” the disclosure of which is incorporated herein by reference. While conventional electron-beam projection lithography directly projects a partitioned flood beam onto a substrate, reflective electron-beam lithography re-directs the beam out of the direct line of sight between the electron source and the semiconductor wafer or other substrate. The beam is re-directed such that it impinges upon a reflective electron patterning device. The reflective electron patterning device may be a dynamic pattern generator for controllably reflecting electrons. Dynamic Pattern Generator for Controllably Reflecting Electrons Previous implementations of reflective electron patterning devices are described in U.S. patent application Ser. No. 10/851,041, filed May 21, 2004, by Harald F. Hess et al., and entitled “Reflective Electron Patterning Device and Method of Using Same,” the disclosure of which is incorporated herein by reference. These previous implementations include using CMOS-driven metallic pixels. The size of such pixels is largely determined as a multiple of the size of the transistors. As such, the size of the pixels depends upon the size of the transistors and in some cases may be too large for practical application in a lithography tool. Disadvantageously, these previous implementations rely on current (or even next generation) semiconductor fabrication processes and technologies to form the pixels of the dynamic pattern generator. The consequences of this include relatively low voltage differentials (typically 1.0 to 1.5 volts, or less) and relatively large pixels (2 microns×2 microns pixel size may be approximately the lower limit using current semiconductor fabrication processes and technologies). The relatively large pixel size for previous implementations requires a correspondingly high level of de-magnification to perform lithography. For example, a 2 micron size generator pixel would have to be de-magnified by at least 90 times to achieve an on-wafer pixels size of 22.5 nanometers (nm) for a target 45 nm resolution lithography. Such high de-magnifications may be limited due to the limitations of electromagnetic lens aberrations. In addition, for high throughput lithography, the relatively large pixel size requires a large illumination size, which may be disadvantageous due to the limitations of electromagnetic lens aberrations. Furthermore, the relatively small voltage differential of the previous implementations limits the level of contrast between on and off pixels. There is limited control available to tune (increase) the voltage differential because it is tied to the fabrication technology used. Means to expand the range of the voltage differential in previous implementations further increase the pixel size. In contrast to these previous implementations, the present application discloses a dynamic pattern generator which uses light to enable pattern transfer from a large integrated circuit structure to a smaller-area pattern generator. The smaller-area pattern generator may then be used to transform a single, spread electron beam into a patterned array, useful for lithography or other applications. The smaller pixel size of the smaller-area pattern generator reduces the de-magnification needed. In addition, the decoupling between the control circuitry and the physical pixels allows the pattern generator to produce higher voltage differentials. Larger-size transistors may be used to take advantage of less expensive fabrication technology. FIG. 1 is a perspective diagram depicting a dynamic pattern generator for controllably reflecting electrons in accordance with an embodiment of the invention. The dynamic pattern generator includes a light emitter array 106 and a photodiode array 110. The light emitter array 106 comprises a relatively dense array of semiconductor light emitters, and the photodiode array 110 comprises a separate, denser array. The light emitters on the light emitter array 106 may be driven by the directly-coupled driver array circuit (CMOS driver chip) 102. The driver array circuit 102 may be flip-chip bonded 104 to the light emitter array 106. An optical lens 108 focuses (de-magnifies) light from the emitter array 106 onto a photodiode array 110. In one embodiment, the light emitter devices of the light emitter array 106 may be vertical light-emitting diodes (LEDs). The use of vertical LEDs for the light emitters allows the formation of a dense array on a semiconductor wafer. In contrast to lasers, for example, the LEDs may be operated in a “sub-threshold” mode to generate a low light level required for this use. In addition, the LEDs are more readily compatible with a CMOS driver circuit 102. In an alternate embodiment, the emitter devices of the light emitter array 106 may be resonant-cavity light emitting diodes (RCLEDs). The use of RCLEDs may provide further advantages of more directional light output and lower driver current requirements for the same light power output. The structure and fabrication of an example RCLED array is described below in relation to FIGS. 3 and 4. The photodiode array 110 may be fabricated with the desired pixel size for the dynamic pattern generator. In accordance with an embodiment of the invention, the pixel size of the photodiode array 110 may be independent of the pitch of the light-emitter array 106. The optical lens 108 is configured to transfer a de-magnified image from the light-emitter array 106 to the photodiode array 110. In accordance with an embodiment of the invention, the photodiodes in the array 110 are configured to operate in an open circuit condition. When light is incident on the “back side” of the photodiode pixel in open circuit condition, excess electron-hole pair generation causes a potential (an open circuit voltage) to build up across the diode. In one implementation, with the back side of the photodiode array 110 held at the potential of the electron beam gun, a voltage differential is induced on the “top side” of those diodes which are illuminated on their bottom sides. The induced voltage on the top side of the back-side illuminated diodes may be used to cause reflection of incident electrons from the electron beam gun. Thus, a dynamic pattern in the reflected electron beam may be generated. Advantageously, the voltage differential between back-side illuminated and unilluminated photodiodes may be determined by the material used in fabricating the photodiode array 110, independent of the drive voltage used to operate the light-emitter array 106. In accordance with one embodiment of the invention, both the LEDs of the emitter array 106 and the photodiodes of the photodiode array 110 may be fabricated using an AlGaAs/GaAs materials system. The AlGaAs/GaAs materials system allows freedom in band gap engineering of the various layers in the device structures within the one to three volt range compatible with the potential differential required for electron reflection by the photodiode array 110 with an adequate contrast ratio. In addition, the AlGaAs/GaAs materials system is lattice matched over the entire composition range, allowing growth of structures with arbitrary thicknesses (particularly useful for distributed Bragg reflectors for RCLED devices) using well-controlled processes. AlGaAs with Al content below 40% is also a direct band gap semiconductor. This allows for the fabrication of efficient photodiodes for the photodiode array 110, as well as RCLEDs for the emitter array 106. Short carrier lifetimes and high carrier mobility advantageously allow the devices to operate at very high speeds. Advantageously, the photodiodes may be operable without amplifier circuits so that their bandwidth need not be limited by such amplifier circuits. The operating wavelength of light (and AlGaAs material compositions) depend on the desired voltage swing for the pattern generator (i.e. for the photodiode array 110. For illustrative purposes, consider an operating wavelength of 700 nanometers (1.77 electron-Volts). Such an operating wavelength provides a practical voltage swing of around 1.3 to 1.5 volts for the pattern generator and also allows for the use of high quality, stable materials for both active and transparent regions of the device structures. In other implementations, other wavelengths may be used. Driver Array Circuit FIG. 2 is a schematic perspective diagram of a driver array circuit (CMOS driver chip) 102 in accordance with an embodiment of the invention. The driver array circuit 102 may include an array of surface pads (digital hi/lo pads) 204 surrounded by a ground plane 202. The array of surface pads may be digitally driven between ground and a supply voltage (Vcc for CMOS). As shown schematically in FIG. 2, serial data may be input (multiplexed) 206 into the driver array circuit 102. The data may flow 208 across the array to the individual pads. The supply voltage (approximately 1 to 2 volts for CMOS) is sufficient to drive the LEDs, which are configured to operate “sub-threshold” with very low output power in the nanoWatt range. Here, “sub-threshold” refers to operation at very low current levels, near the elbow of the forward current-voltage characteristic. The switched voltage on the pads of the driver array circuit causes corresponding LEDs to light up, which in turn bias the photodiodes for pattern generation. This preferred embodiment utilizes digital switching with the photodiodes operating in saturation. Alternatively, analog control may be used. The pads 204 of the driver array circuit may be configured with a relatively small pitch between mesas, on the order of 5 to 20 microns, matching the pitch (and approximate size) of the LED mesas that are bonded to them. This pitch may be designed as small as practical to minimize the overall power consumption as well as minimize the optical demagnification onto the photodiode array, but the pitch need not approach fabrication limits. A target pitch of around 10 microns, for example, may be appropriate. The area surrounding the active pad array 204 may comprise a reasonably wide ground plane 202. Such a ground plane perimeter 202 may be used to facilitate flip-chip bonding 104 of the LED substrate 106 onto the driver array substrate 102 by using larger solder bumps for the taller space at the ground plane perimeter 202 (see FIG. 8). The fabrication of the pad array 204 may involve fabricating vias above each digital driver element (arrayed below the pads) to connect to the pads. In one implementation, a series of shift registers may be configured so as to be fed by a de-multiplexed serial data stream (see de-multiplexed data flow 208 in FIG. 2). The shift registers may comprise flip-flop circuits which may be readily fabricated on this size scale using presently available technology. The multiplexed serial data 206 may be fed into one end of the array structure, as illustrated in FIG. 2, and may be clocked down the array after de-multiplexing the data. Such a configuration may be advantageously compatible with certain electron beam exposure modes (for example, time delay integration, gray scaling, and so forth). This configuration also allows a large array of pixels to be fed by a single serial data stream that needs to be only as fast as the width of the array times the clock rate down the array. Such an approach is advantageously scalable in “width” by increasing the number of parallel rows in the array 204 and/or in “length” by incorporating additional input data streams 206. Resonant-Cavity Light Emitting Diode In one embodiment, the light emitter array 106 may be a resonant-cavity light emitting diode (RCLED) array. An RCLED array chip may be grown by molecular beam epitaxy (MBE) or by organometallic vapor phase epitaxy (OMVPE) or by another layered growth technique. FIG. 3 is a schematic cross-sectional diagram of an epitaxial layer structure 300 for a resonant-cavity light emitting diode (RCLED) in accordance with an embodiment of the invention. The epitaxial structure 300 may be fabricated on a GaAs substrate 302. For operation using light of around 700 nm wavelength, a GaAs substrate and the AlGaAs/GaAs material system may be utilized. For operation at other wavelengths, different materials may be used. The AlGaAs/GaAs material system allows distributed Bragg reflectors (DBRs) (306 and 316) to be integrated monolithically to create the resonant cavity. The DBRs may be configured, for example, with mirror reflectivities of about 90% for the higher reflector 316 and 50% to 85% for the out-coupler 306. AlGaAs quantum wells 314 may be utilized as the active light emitter. The AlGaAs quantum wells 314 may be sandwiched between an n-doped AlGaAs layer 308 and a p-doped AlGaAs layer 310. An intrinsic region may be defined 312 in the region in the vicinity of the junction between the n-doped and p-doped layers. FIG. 4 is a schematic perspective diagram of a fabricated RCLED structure 400 in accordance with an embodiment of the invention. The individual LEDs in the array are defined and isolated by etched mesas 406. To facilitate interfacing with the driver array circuitry, a bottom-emitting (through the substrate) configuration may be utilized. In other words, the LEDs emit light 410 through the substrate and out of its backside 404. Evaporated metal discs (patterned using lift-off, for example) may form the contact 408 on the mesa tops and may also be used as self-aligned etch masks to define the mesas 406. After etching to a lower contact layer, a bottom “ground plane” metal contact may deposited (again, using lift-off, for example). The chip may then be flip-chip bonded to the driver array chip, with the mesa tops 408 bonded to the individual pads on the driver array. Large indium solder bumps may be used in multiple locations around the periphery of the array to bond the ground planes of the two chips, so as to span the approximately 0.5 to 3.0 micron height differential between the top and bottom of the mesas. Mechanical stability and passivation of the flip-chip bonded structure may be achieved by flowing epoxy between the two chips. Since the GaAs substrate 302 absorbs 700 nm light, a substrate thinning step may be employed after flip-chip bonding. This thinning may be accomplished by growing a high Al content AlGaAs or AlAs etch stop layer 304 into the structure (see FIG. 3). A selective etch, such as citric acid:hydrogen peroxide, may then be used to remove the GaAs. The etchstop layer may also be removed with an additional selective etch, if desired, for example, by using a hydrochloric acid etch. A thin layer of GaAs (“Backside” Ground Plane) 402 is left on the chip after substrate etching to maintain the ground plane connection across the array (see FIG. 4). Photodiode Array The photodiode array 110 may include a back-illuminated structure to allow the top surface potential to be controlled by the light coming from the light emitters 106 below. Substrate removal may be performed if the substrate (for example, GaAs) is insufficiently transparent at the wavelengths appropriate to achieve sufficient voltage swings (for example, over one volt potential swing) at the top surface. However, if the substrate is removed, then a mechanical support may be required for the very thin membrane that will remain. A transparent structure such as quartz may be used, for example. FIG. 5 is a schematic cross-sectional diagram of an epitaxial layer (epilayer) structure 500 for a photodiode wafer in accordance with an embodiment of the invention. The epitaxial structure 500 may be fabricated on a GaAs substrate 502. The epitaxial device may be grown with MBE or OMVPE or other layered growth technique. The GaAs substrate 502 may be removed later and the structure flipped over so that the illumination will be incident upon the top layer 510 in FIG. 5. A p/n junction may be formed between two Al0.2Ga0.8As layers, one being a p+ doped layer 506 (for example, about 60 nm thick with doping concentration of about 2×1018 cm−3) and one being an n-doped layer 508 (for example, about 250 nm thick with doping concentration in a range of 1 to 2×1017 cm−3). The Al0.5Ga0.5As layers below 504 and above 510 the p/n junction serve as transparent layers. The heavily-doped p++ lower layer 504 (for example, in a range of 40 to 100 nm thick with doping concentration of roughly 5×1019 cm−3) serves to deplete carriers from the surface and minimize surface recombination. The n+ doped top layer 510 (for example, about 300 nm thick with doping concentration of roughly 2×1018 cm−3) serves as a transparent contact (to the quartz substrate). The overall thickness of the epilayer structure (not counting the GaAs substrate) 500 is preferably configured to allow for the practical etching of the isolation trenches between pixels, while providing sufficient light collection. Doping levels are preferably selected so that the diffusion lengths in each layer nearly match the layer thicknesses, allowing the full structure to efficiently contribute to the induced photovoltage. The specific design shown in FIG. 5 is bandgap engineered to operate with light of 700 nm wavelength. The absorbing layers (506 and 508) have a bandgap about 4 kT below the light energy, and the transparent layers (504 and 510) have a bandgap about 4 kT above the light energy. These bandgap estimates include bandgap renormalization effects from the high doping levels in some layers. The particular design shown in FIG. 5 has the p-type layer on top (exposed to the electron beam). This implies a positive bias with respect to the backside or ground plane when illuminated. In an alternate design, it may be desirable to design the inverse structure and impart a negative bias with illumination. Applicants believe that the particular design shown in FIG. 5 allows for higher efficiency operation and somewhat simpler fabrication. However, either polarity design may be used. In accordance with an embodiment of the invention, after the epitaxial device structure 500 is formed, the process to fabricate the photodiode array chip 110 may proceed as follows. A finished photodiode array chip 110 after the process is depicted in FIG. 6. (a) Deposit and anneal a standard n-type metal contact (for example, Au/Ge/Ni/Au) patterned with a window cover over the region that will become the active pixel area. This metal will become the backside contact to the photodiode array chip 110. (b) Deposit a thin oxide layer (for example, with plasma enhanced chemical vapor deposition or PECVD) over the metal contact and bond the wafer 500 face down (reverse of FIG. 5) onto a quartz wafer 600 using spin-on-glass or other suitable bonding agent. The oxide layer promotes adhesion to the quartz. The quartz 604 becomes transparent support for the epilayer structure 500 after substrate 502 removal. The epitaxial wafer 500 preferably overhangs 606 the quartz wafer 604 in order to allow contact to be made to the backside after fabrication is complete. (c) Perform substrate 502 removal. The substrate removal may be performed using, for example, citric acid:hydrogen peroxide selective GaAs etch that stops on the p++ Al0.5Ga0.5As layer 504. This etch stop and window layer is preferably selected so as to be sufficiently thick to accommodate any over-etching which may be required for complete substrate removal. (d) The isolation trench pattern to separate the individual pixels 602 may be defined on the new top surface using lithography. The lithography used may be, for example, conventional electron-beam lithography. This defines the array of pixels. (e) Vertical trenches are then etched through the lithographically-defined mask so as to form the floating, trench-isolated photodiode pixels 602. For example, the etching may be performed using inductively-coupled plasma etching to enable deep vertical trenches. (f) Top p-type metal (for example, Ti/Pt/Au) may then be evaporated onto the etched pixels at a steep angle. The steep-angled evaporation is used to prevent shorting between pixels. This top metal layer may not be needed if the bare semiconductor exhibits a homogeneous voltage potential across the surface. (g) A short oxide etch may be performed to remove the thin oxide layer from the backside metal overhanging the quartz so as to expose the back contact. This oxide etch may also remove a very slight thickness of the quartz wafer, possibly leaving a quartz surface that is not optical smooth. In such a case, polishing may be used to provide an optically-smooth surface. Alternatively, this etch step may be performed between steps (b) and (c). However, it is preferably performed after step (e) to keep the metal covered by the thin oxide layer during the etching in step (e) and avoid exposed metal in the semiconductor etch chamber. The finished photodiode array 110 may be configured with a high voltage (HV) 608 applied to both the overhanging back contact and the top metal. Alternatively, a higher voltage (HV+Voc+Δ) 609 may be applied to the top-side ground plane (while HV is applied to the back contact). This alternate applied voltage configuration provides a dark (electron absorbing) background for lithography in this positive-polarity design. Simulated pixel performance (voltage expression versus illumination intensity) of the epilayer structure 500 in FIG. 5 is shown in FIG. 7. In particular, FIG. 7 depicts graphs of open circuit voltage versus illumination intensity (plotted on (a) linear and (b) semi-log scales) in accordance with an embodiment of the invention. As a point of reference, 10 nanowatts (nW) of 700 nm wavelength light incident on a 0.25 micron×0.25 micron pixel is equivalent to 16 Watts/cm2 intensity, and at 10 MHz repetition rate represents 3,500 electrons per pulse. Hence, the data depicted in FIG. 7 shows that only nanowatts of light is needed per pixel to generate a sufficiently saturated voltage expression on the top surface of the photodiodes. Additionally, only a few to tens of electrons will be injected into each dark pixel from absorption of the electron beam during the refresh time. This small amount of injected electrons will have a negligible effect on the surface potential of the dark pixels. Assembly FIG. 8 is a schematic cross-sectional diagram showing an assembly 800 of the components for a dynamic pattern generator in accordance with an embodiment of the invention. The illustration is not to scale. As shown in FIG. 8, a lens 108 is positioned in between the RCLED chip 400 and the photodiode (PD) array 110. The lens 108 is configured to demagnify the image from the RCLED array 400. A mechanism is used to align the photodiode array to the RCLED array. For example, alignment marks in the wafer periphery may be made, and backside illumination may be employed for alignment under a microscope. The aligned components may then be fixed with epoxy, for example, or a mechanical mount that would allow easier replacement of individual components. The lens 108 in FIG. 8 effectively serves to optically scale down the pixels on the CMOS driver array chip 102 to a more practical size, such that less electron beam demagnification is needed from the PD chip 110 (the pattern generator) to the target wafer for a lithography application. Advantageously, the assembly 800 of FIG. 8 achieves the scale down (demagnification) with light optics which typically has less aberration than electron optics. In addition, this scale down (demagnification) is achieved while maintaining a compact assembly. For example, using 0.5 micron×0.5 micron pixels on the PD array and 10 micron pitch between LEDs requires a 20 times optical demagnification in the assembly, followed by a 22 times demagnification in the electron projection optics to achieve 22.5 nm size pixels on the wafer for 45 nm lithography. In contrast, an “all CMOS” implementation of the dynamic pattern generator using 5 micron×5 micron pixels would require 222 times demagnification in the electron projection optics to achieve the same 22.5 nm on-wafer pixel size. Also as shown in FIG. 8, solder bumps may be used to bond the emitter array chip 400 to the driver array chip 102 using flip-chip bonding techniques. The solder bumps may include flip-chip solder bumps 804 and a series of larger solder bumps 806 around the periphery to connect the ground planes of the two chips. In addition, the empty volume between the chips may be backfilled with epoxy after bonding to provide passivation and mechanical stability. The full metal coatings 408 over the mesa tops that form the array of contacts on the RCLED chip 400 may also prevent stray light from entering the driver array circuitry 102. Otherwise, such stray light may degrade performance. Furthermore, FIG. 8 illustrates how the high voltage supply needs to be connected only to the photodiode chip. This allows the floating pads of the photodiode array to form the programmable voltage pattern for the electron beam swath. Because the driver array chip 102 is electrically isolated from the photodiode array chip 110, no high voltage need be applied to the driver array chip 102. Hence, the difficulties and problems of high voltage supply design and shielding for the CMOS circuitry of the driver array chip 102 may be avoided. Furthermore, the driver array chip 102 and the RCLED chip 400 are protected from potentially damaging high voltage arcing events in the system. As an additional feature, an immersion high index fluid may be utilized between the lens 108 and the photodiode array 110. Such an immersion high index fluid may advantageously allow for high numerical aperture imaging for a large field of view and improved resolution of transfer of the pattern from the RCLED array to the photodiode array. Reflective Electron-Beam Lithography System FIG. 9 is a schematic diagram showing a reflective electron-beam lithography (REBL) system 900 including a dynamic pattern generator in accordance with an embodiment of the invention. As depicted, the system 900 includes illumination electron-optics 902 for providing the incident electron beam, an electron prism 904, an objective electron lens (magnifying and shaping) 906, a dynamic pattern generator 100, projection electron-optics 908 to project the patterned electron beam, and a stage 910 for holding a wafer or other target to be lithographically patterned. In accordance with an embodiment of the invention, the dynamic pattern generator 100 may be implemented as describe above, for example, using a photodiode chip 110, optical lens 108, RCLED chip 400, and CMOS driver 102. The other components of the system 100 are discussed in further detail in U.S. patent application Ser. No. 10/851,040 (now U.S. Pat. No. 6,870,172), filed May 21, 2004, by inventors Marian Mankos et al., and entitled “Maskless Reflection Electron Beam Projection Lithography,” the disclosure of which is hereby incorporated by reference. In the system 900, a relatively broad incident electron beam impinges on the pattern generator 100 with relatively low landing energy. The low landing energy may be accomplished by holding the photodiode chip 110 of the pattern generator 100 at nearly the same potential as the electron source. The low energy beam interacts with the dynamically controlled voltage pattern on the surface of the photodiode chip 110. An image of that pattern is reflected in the form of a patterned electron beam. The projection electron-optics 908 demagnifies the patterned electron beam to expose a patterned area on the wafer. As shown in FIG. 9, light optics is used to demagnify the pitch of the CMOS driver circuitry, and so lessens the demagnification needed by the projection electron-optics, as discussed above. The required size of the broad incident electron beam is also reduced, lessening the burden on the illumination and objective lens designs. The decoupling of the pattern generator pixel size from the pitch of the CMOS driver circuitry also readily allows for extension of this system to future lithography technology generations. The dynamic pattern generator disclosed herein provides various advantages. For example, it decouples the control electronics from the physical pixels that pattern the beam. This allows the pixel size to be independent of the fabrication and circuit technologies of the control electronics. Hence, sub-micron pixel sizes may be fabricated. This lessens the demagnification required by the projection electron-optics and the field size required of the illumination/objective optics. The decoupling between the control circuitry and the physical pixels also allows the pattern generator to produce voltage differentials that are independent of the supply voltage of the control circuitry. Furthermore, the decoupling isolates the control circuitry from the high voltage of the electron beam system, and so avoids the need to float the control circuitry at high voltage. The entire device may also be constructed so as to be compact and field-replaceable as a single unit. Finally, the dynamic pattern generator disclosed herein readily scales to future lithography technologies. Various fabrication and implementation alternatives are disclosed in the above description. For example, different material systems may be used for the LED and/or PD array chips so as to operate at different wavelengths and/or to optimize device lifetime, reliability and the like. The epitaxial structures may be grown by various methods, including MBE and OMVPE. A conventional LED structure may be used for the light emitters, instead of RCLEDs. It may also be possible to use vertical-cavity surface emitting lasers (VCSELs) or other laser structures for the light emitters. Multiple methods of attaching the LED chip, lens, and PD chip into an aligned unit may be used, including epoxy or mechanical mounts. As another alternative, the components may be kept separate in the system without creating a single replaceable unit. The attachment method, for example epoxy, may include means to control the index of refraction between the lens and the photodiode array and/or the lens and the emitter array. High index fluid may alternatively be used. The above-described diagrams are not necessarily to scale and are intended be illustrative and not limiting to a particular implementation. In the above description, numerous specific details are given to provide a thorough understanding of embodiments of the invention. However, the above description of illustrated embodiments of the invention is not intended to be exhaustive or to limit the invention to the precise forms disclosed. One skilled in the relevant art will recognize that the invention can be practiced without one or more of the specific details, or with other methods, components, etc. In other instances, well-known structures or operations are not shown or described in detail to avoid obscuring aspects of the invention. While specific embodiments of, and examples for, the invention are described herein for illustrative purposes, various equivalent modifications are possible within the scope of the invention, as those skilled in the relevant art will recognize. These modifications can be made to the invention in light of the above detailed description. The terms used in the following claims should not be construed to limit the invention to the specific embodiments disclosed in the specification and the claims. Rather, the scope of the invention is to be determined by the following claims, which are to be construed in accordance with established doctrines of claim interpretation.
055552803	claims	1. Process for producing a leakproof protective coating on a surface of a component of a nuclear reactor intended to bear cracking in the operating reactor comprising scanning the surface of the component with a jet of a semitransferred arc plasma torch, introducing a metal powder in said jet and controlling said plasma jet, conditions of introduction of metal powder and conditions of scanning, for ejecting and fixing the metal powder onto the surface of the component without melting the surface of the component, thus resulting in a dense, homogeneous and leakproof coating layer of a thickness of at most 1 mm. 2. Process according to claim 1 comprising maintening the distance between an outlet nozzle of the plasma torch and the surface of the component to be coated between 20 and 80 mm. 3. Process in accordance with claim 1, comprising fixing the metal powder on a surface of a zone of connection by welding of the component to a second component of the nuclear reactor. 4. Process in accordance with claim 1 wherein the nuclear reactor component is made of a material containing nickel and that the metal powder introduced into the plasma jet contains nickel and chromium. 5. Process in accordance with claim 1 wherein the coating layer deposited on the surface of the nuclear rector component has a thickness of between 0.5 and 1 mm. 6. Process in accordance with claim 1 comprising fixing the metal powder in the form of a coating layer on the outer surface of a region of connection between a primary pipe and a nozzle of a pressurized water nuclear reactor. 7. Process in accordance with claim 1 comprising fixing the metal powder in the form of a coating layer on the internal surface of an adapter passing through the reactor vessel head of a pressurized water nuclear reactor. 8. Process in accordance with claim 1 comprising fixing the metal powder in the form of a coating layer in a region of welding connection between an adapter passing through a closure head of a pressurized water nuclear reactor and the closure head. 9. Device for coating the outer surface of a region of connection between a nozzle of the vessel of a pressurized water nuclear reactor and a pipe of the primary system of the reactor, comprising an annular guiding rail fastened in a coaxial arrangement to the outer surface of the primary pipe in the vicinity of the region of connection, a trolley mounted so as to move on the annular rail, a column for guiding a platform in a radial direction of the primary pipe, a support mounted so as to move on the platform in an axial direction of the primary pipe and motor-driven means for moving the platform, the trolley and the support, so as to move the support in rotation about the axis of the primary pipe in a radial direction and in an axial direction in relation to the pipe, and means for fastening a plasma torch or an ultrasonic testing head to the support. 10. Device for producing a protective coating on the inner surface of a tubular adapter passing through the reactor vessel head of a pressurized water nuclear reactor, the reactor vessel head being deposited on a servicing stand and accessible from below through an opening in a protective wall, comprising a vertical column fastened to a support resting on a supporting surface of the stand below an adapter in vertical position, a platform mounted so as to move in the vertical direction on the column, a vertical support mounted so at to rotate about a vertical axis on the platform and a semitransferred arc plasma torch fastened to an upper end of the rotary support. 11. Device for producing a protective coating on a surface of a region of welding connection of an adapter passing through a closure head of a pressurized water nuclear reactor arranged on a servicing stand so that the connection region of the adapter is accessible from below the closure head through an opening in a biological isolation wall of the stand, comprising a column integrally attached to a support resting on a horizontal surface of the stand situated below the opening and the adapter, a platform movable in the vertical direction on the column, a vertical support movable in rotation about a vertical axis on the platform and a semitransferred arc plasma torch comprising a nozzle for ejecting a plasma jet and mounted so that it pivots at one end of the torch fastened in a vertical arrangement to the support.
	description	This application is a division of U.S. application Ser. No. 13/863,611 filed Apr. 16, 2013, now U.S. Pat. No. 9,911,512, which is a continuation-in-part of U.S. application Ser. No. 13/405,405 filed Feb. 27, 2012, now U.S. Pat. No. 9,805,832. U.S. application Ser. No. 13/863,611 also claims the benefit of U.S. Provisional Application No. 61/625,484 filed Apr. 17, 2012. The entire disclosure of the applications are hereby incorporated by reference herein. The following relates to the nuclear reactor arts, nuclear power generation arts, nuclear reactor control arts, nuclear reactor electrical power distribution arts, and related arts. In nuclear reactor designs of the integral pressurized water reactor (integral PWR) type, a nuclear reactor core is immersed in primary coolant water at or near the bottom of a pressure vessel. In a typical design, the primary coolant is maintained in a subcooled liquid phase in a cylindrical pressure vessel that is mounted generally upright (that is, with its cylinder axis oriented vertically). A hollow cylindrical central riser is disposed concentrically inside the pressure vessel. Primary coolant flows upward through the reactor core where it is heated, rises through the central riser, discharges from the top of the central riser, and reverses direction to flow downward back toward the reactor core through a downcomer annulus. The nuclear reactor core is built up from multiple fuel assemblies. Each fuel assembly includes a number of fuel rods. Control rods comprising neutron absorbing material are inserted into and lifted out of the reactor core to control core reactivity. The control rods are supported and guided through control rod guide tubes which are in turn supported by guide tube frames. In the integral PWR design, at least one steam generator is located inside the pressure vessel, typically in the downcomer annulus, and the pressurizer is located at the top of the pressure vessel, with a steam space at the top most point of the reactor. Alternatively an external pressurizer can be used to control reactor pressure. A set of control rods is arranged as a control rod assembly that includes the control rods connected at their upper ends with a spider, and a connecting rod extending upward from the spider. The control rod assembly is raised or lowered to move the control rods out of or into the reactor core using a control rod drive mechanism (CRDM). In a typical CRDM configuration, an electrically driven motor selectively rotates a roller nut assembly or other threaded element that engages a lead screw that in turn connects with the connecting rod of the control rod assembly. The control rods are typically also configured to “SCRAM”, by which it is meant that the control rods can be quickly released in an emergency so as to fall into the reactor core under force of gravity and quickly terminate the power-generating nuclear chain reaction. Toward this end, the roller nut assembly may be configured to be separable so as to release the control rod assembly and lead screw which then fall toward the core as a translating unit. In another configuration, the connection of the lead screw with the connecting rod is latched and SCRAM is performed by releasing the latch so that the control rod assembly falls toward the core while the lead screw remains engaged with the roller nut. See Stambaugh et al., “Control Rod Drive Mechanism for Nuclear Reactor”, U.S. Pub. No. 2010/0316177 A1 published Dec. 16, 2010 which is incorporated herein by reference in its entirety; and DeSantis, “Control Rod Drive Mechanism for Nuclear Reactor”, U.S. Pub. No. 2011/0222640 A1 published Sep. 15, 2011 which is incorporated herein by reference in its entirety. The CRDMs are complex precision devices which require electrical power to drive the motor, and may also require hydraulic, pneumatic, or another source of power to overcome the passive SCRAM release mechanism (e.g., to hold the separable roller nut in the engaged position, or to maintain latching of the connecting rod latch) unless this is also electrically driven. In existing commercial nuclear power reactors, the CRDMs are located externally, i.e. outside of the pressure vessel, typically above the vessel in PWR designs, or below the reactor in boiling water reactor (BWR) designs. An external CRDM has the advantage of accessibility for maintenance and can be powered through external electrical and hydraulic connectors. However, the requisite mechanical penetrations into the pressure vessel present safety concerns. Additionally, in compact integral PWR designs, especially those employing an internal pressurizer, it may be difficult to configure the reactor design to allow for overhead external placement of the CRDMs. Accordingly, internal CRDM designs have been developed. See U.S. Pub. No. 2010/0316177 A1 and DeSantis, “Control Rod Drive Mechanism for Nuclear Reactor”, U.S. Pub. No. 2011/0222640 A1 published Sep. 15, 2011 which is incorporated herein by reference in its entirety. However, placing the CRDMs internally to the reactor vessel requires structural support and complicates delivery of electrical and hydraulic power. Electrical conductors, which may be Mineral Insulated (MI) cable, that are usable inside the pressure vessel are generally not flexible and are not readily engaged or disengaged, or spliced, making installation and servicing of internal CRDM units time consuming and labor-intensive. Disclosed herein are improvements that provide various benefits that will become apparent to the skilled artisan upon reading the following. In some illustrative embodiments, an apparatus comprises: a nuclear reactor including a pressure vessel and a nuclear reactor core comprising fissile material disposed in the pressure vessel; an internal control rod drive mechanism (CRDM) including an electric motor disposed in the pressure vessel and a support surface including sealed electrical connectors electrically connected with the electric motor to deliver electrical power to the electrical motor; and a support element secured inside the pressure vessel on which the support surface of the internal CRDM is disposed to support the internal CRDM in the pressure vessel, the support element including sealed electrical connectors mating with the sealed electrical connectors on the support surface of the internal CRDM to deliver electrical power to the electric motor of the internal CRDM. In some embodiments the internal CRDM further comprises mineral insulated cables (MI cables) electrically connecting the electric motor to the sealed electrical connectors on the support surface, wherein each MI cable is connected to one of the sealed electrical connectors and the sealed electrical connectors are sealed glass connectors, sealed ceramic connectors, or sealed glass ceramic connectors. In some embodiments the sealed electrical connectors and the sealed electrical connectors are welded onto the ends of the MI cables. In some embodiments springs, e.g. wave springs, are disposed between the sealed electrical connectors of the support element and the mating sealed electrical connectors on the support surface of the internal CRDM. In some embodiments a purge line is integrated with each mated connection of a sealed electrical connector of the support element and the mated sealed electrical connector on the support surface of the internal CRDM. The internal CRDM may include a standoff mechanically secured with the internal CRDM, the support surface of the internal CRDM being a surface of the standoff. The support element may comprise a distribution plate including MI cables disposed on or in the distribution plate and terminating at the sealed electrical connectors of the distribution plate. In some illustrative embodiments, a method comprises providing an internal control rod drive mechanism (CRDM) including an electric motor and a support surface including sealed electrical connectors electrically connected with the electric motor to deliver electrical power to the electrical motor, and installing the internal CRDM inside a nuclear reactor, the installing including placing the support surface of the internal CRDM onto a support element inside the nuclear reactor, the placing causing sealed electrical connectors disposed on the support element to mate with the sealed electrical connectors on the support surface of the internal CRDM. In such a method, the nuclear reactor may contain coolant water and the installing may be performed with the internal CRDM submerged in the coolant water—the seals of the sealed electrical connectors of the internal CRDM and the support element are effective to prevent coolant water ingress into the sealed electrical connectors. The method may further comprise, after the placing is performed, purging space between the mated sealed electrical connectors of the internal CRDM and the support element through a purge line using an inert gas. Still further, the method may comprise sealing off the purge line after the purging in order to trap residual inert gas in the space between the mated sealed electrical connectors of the internal CRDM and the support element. In some illustrative embodiments, an internal control rod drive mechanism (CRDM) includes as a unitary assembly: an electric motor; a support surface; sealed glass, ceramic, or glass ceramic electrical connectors disposed on the support surface; and MI cables extending from the electric motor and having ends sealed inside the sealed glass, ceramic, or glass ceramic electrical connectors. The seals of the sealed glass, ceramic, or glass-ceramic electrical connectors are effective to allow the internal CRDM to be immersed in water without water ingress into the MI cables. Optionally, each sealed glass, ceramic, or glass-ceramic electrical connector further includes a purge line arranged to admit purge gas into space between the sealed glass, ceramic, or glass-ceramic electrical connector and an associated mating connector. FIG. 1 illustrates an integral Pressurized Water Reactor (integral PWR) generally designated by the numeral 10. A reactor vessel 11 is generally cylindrical and contains a nuclear reactor core 1 comprising fissile material (e.g. 235U), steam generators 2, and a pressurizer 3. Although an integral pressurized water reactor (PWR) is depicted, embodiments utilizing a boiling water reactor (BWR), PWR with external steam generators, or other type of nuclear reactor are also contemplated. Moreover, while the disclosed rapid installation and servicing techniques are described with reference to illustrative internal CRDM units, these techniques are readily adapted for use with other internal nuclear reactor components such as internal reactor coolant pumps. In the illustrative PWR, above the core 1 are reactor upper internals 12 of integral PWR 10, shown in inset. The upper internals 12 are supported laterally by a mid-flange 14, which in the illustrative embodiment also supports internal canned reactor coolant pumps (RCPs) 16. More generally, the RCPs may be external pumps or have other configurations, and the upper internals may be supported otherwise than by the illustrative mid flange 14. The upper internals include control rod guide frames 18 to house and guide the control rod assemblies for controlling the reactor. Control Rod Drive Mechanisms (CRDMs) 20 raise and lower the control rods to control the reactor. In accordance with one embodiment, a CRDM distribution plate 22 supports the CRDMs and provides power and hydraulics to the CRDMs. Control rods are withdrawn from the core by CRDMs to provide enough positive reactivity to achieve criticality. The control rod guide tubes provide space for the rods and interconnecting spider to be raised upward away from the reactor core. The CRDMs 20 require electric power for the motors which move the rods, as well as for auxiliary electrical components such as rod position indicators and rod bottom sensors. In some designs, the force to latch the connecting rod to the lead screw, or to maintain engagement of the separable roller nut, is hydraulic, necessitating a hydraulic connection to the CRDM. To ensure passive safety, a positive force is usually required to prevent SCRAM, such that removal of the positive force initiates a SCRAM. The illustrative CRDM 20 is an internal CRDM, that is, is located inside the reactor vessel, and so the connection between the CRDM 20 and the distribution plate 22 is difficult to access. Servicing of a CRDM during a plant shutdown should preferably be rapid in order to minimize the length of the shutdown. To facilitate replacing a CRDM in the field, a standoff assembly connected to the distribution plate 22 to provide precise vertical placement of the CRDM 20 is also configured to provide electrical power and hydraulics to the CRDM 20 via connectors that require no action to effectuate the connection other than placement of the standoff assembly onto the distribution plate 22. After placement, the standoff is secured to the distribution plate by bolts or other fasteners. Additionally or alternatively, it is contemplated to rely upon the weight of the standoff assembly and CRDM to hold the assembly in place, or to use welds to secure the assembly. The illustrative distribution plate 22 is a single plate that contains the electrical and hydraulic lines and also is strong enough to provide support to the CRDMs and upper internals without reinforcement. In another other embodiment, the distribution plate 22 may comprise a stack of two or more plates, for example a mid-hanger plate which provides structural strength and rigidity and an upper plate that contains electrical and/or hydraulic lines to the CRDMs via the standoff assembly. The motor/roller nut assembly of the CRDM is generally located in the middle of the lead screw's travel path. When the control rod is fully inserted into the core, the roller nut is holding near the top of the lead screw, and, when the rod is at the top of the core, the roller nut is holding near the bottom of the lead screw and most of the length of the lead screw extends upward above the motor/roller nut assembly. Hence the distribution plate 22 that supports the CRDM is positioned “below” the CRDM units and a relatively short distance above the reactor core. FIG. 2 illustrates the distribution plate 22 with one standoff assembly 24 mounted for illustration, though it should be understood that most or all openings 26 would have a standoff assembly (and accompanying CRDM) mounted in place during operation of the reactor. Each opening 26 allows a lead screw of a control rod to pass through and the periphery of the opening provides a connection site for a standoff assembly that supports the CRDM. The lead screw passes down through the CRDM, through the standoff assembly, and then through the opening 26. The distribution plate 22 has, either internally embedded within the plate or mounted to it, electrical power lines (e.g., electrical conductors) and hydraulic power lines to supply the CRDM with power and hydraulics. The illustrative openings 26 are asymmetric or keyed so that the CRDM can only be mounted in one orientation. As illustrated, there are 69 openings arranged in nine rows to form a grid, but more or fewer could be used depending on the number of control rods in the reactor. The distribution plate is circular to fit the interior of the reactor, with openings 28 to allow for flow through the plate. In some designs, not all openings may have CRDMs mounted to them or have associated fuel assemblies. The CRDMs are supported by the CRDM standoff assembly which is attached to a distribution plate that provides power to the CRDMs. The connectors for the CRDM's are integrated into the power distribution plate assembly and the CRDM standoff plate. They allow the disconnection of the power and signal leads when CRDM maintenance is required without splicing MI cable. FIG. 3 schematically illustrates a small cutaway view of one connection site of the distribution plate 22 for connecting a CRDM to the distribution plate. The connection site includes an opening 26 for passing the lead screw of a single CRDM. Located around the opening 26 are apertures 40 to accept bolts (more generally, other securing or fastening features may be used) and electrical connectors 42 for delivering electrical power to the CRDM. The illustrative CRDM employs hydraulic power to operate the SCRAM mechanism, and accordingly there is also a hydraulic connector 44 to accept a hydraulic line connection. The opening 26 and its associated features 40, 42, 44 create a connection site to accept the CRDM/standoff assembly. Internal to the distribution plate 22 may be junction boxes to electrically connect the connection sites to the electrical power lines running in between rows of connection sites. Similarly, the hydraulic connector 44 may connect to a common hydraulic line running through the distribution plate separated by depth. FIG. 4 illustrates a standoff 24 that suitably mates to opening 26 in the distribution plate 22. The standoff assembly has a cylindrical midsection with plates 45, 46 of larger cross-sectional area on either end of the midsection. The circular top plate 45 mates to and supports a CRDM 20. The square bottom plate 46 mates to the distribution plate 22. Although the illustrative bottom plate 46 is square, it may alternatively be round or have another shape. When the CRDM 20 and the top plate 45 of the standoff 24 are secured together they form a unitary CRDM/standoff assembly in which the bottom plate 46 is a flange for connecting the assembly to the distribution plate 22. Two bolt lead-ins 50 on diagonally opposite sides of the lower plate 46 mate to the apertures 40 of the distribution plate. The bolt lead-ins, being mainly for positioning the CRDM standoff, are the first component on the standoff to make contact with the distribution plate when the CRDM is being installed, ensuring proper alignment. Two electrical power connectors 52 on diagonally opposite corners of the bottom plate 46 mate to corresponding electrical power connectors 42 of the distribution plate 22. Each connector 52 is installed in a raised boss or collar 53 on the bottom plate 46 of the standoff 24 (e.g., see FIG. 5). A hydraulic line connector 54 on the bottom plate 46 mates to the corresponding hydraulic power connector 44 of the distribution plate 22. A central bore 56 of the standoff 24 aligns with the opening 26 of the distribution plate 22 and allows the lead screw to pass through. The connectors 42, 44 inside the distribution plate 22 (or connectors 52, 54 inside the standoff 24) optionally have internal springs to ensure positive contact, and the opposing bolts that attach at lead-ins 50 serve as tensioning devices to ensure proper seating of both the CRDM electrical connectors and hydraulic connectors. Illustrative flow slots 58 permit primary coolant to flow through the standoff. FIG. 5 illustrates a perspective view focusing on the top plate 45 of the standoff 24. The top plate 45 of the standoff mates to the CRDM and is attached via bolt holes 62. Bolt holes 62 may be either threaded or unthreaded. The CRDM and standoff can be attached to each other and electrical connections 52 and hydraulic connection 54 made before the CRDM is installed so as to form a CRDM/standoff assembly having flange 46 for connecting the assembly with the connection site of the distribution plate 22. The bottom plate 46 of the standoff 24 is secured to the connection site via bolts passing through holes 50 and secured by nuts, threads in the bolt holes 40, or the like. FIG. 6 shows standoff 24 connected to a CRDM 20 to form a CRDM/standoff assembly that can be mounted to the distribution plate. CRDM electrical cabling 80 extends upward to conduct electrical power received at the electrical connectors 52 to the motor or other electrical component(s) of the CRDM 20. In the embodiment of FIG. 5, each electrical connector 52 terminates two electrical cables 80. Similarly, a CRDM hydraulic line 82 extends upward to conduct hydraulic power received at hydraulic connector 54 to the hydraulic piston or other hydraulic component(s) of the CRDM 20 to maintain latching—removal of the hydraulic power instigates a SCRAM. The entire assembly including the CRDM and the standoff is then installed as a unit on a distribution plate, simplifying the installation process of a CRDM in the field. The interface points (i.e. CRDM electrical and hydraulic connectors) in the embodiment of FIG. 6 are at the standoff assembly but could be at any location along the length of the CRDM. For the illustrative examples, the interface point at which the CRDM is broken from the upper internals is at the bottom of the CRDM. The apparatus for which the interface points are located is the CRDM standoff and the CRDM power distribution plate. In one embodiment, the electrical cables 80 are mineral insulated cables (MI cables) which generally include one, two, three, or more copper conductors wrapped in a mineral insulation such as Magnesium Oxide which is in turn sheathed in a metal. The mineral insulation could also be aluminum oxide, ceramic, or another electrically insulating material that is robust in the nuclear reactor environment. MI cables are often sheathed in alloys containing copper, but copper would corrode and have a negative effect on reactor chemistry. Some contemplated sheathing metals include various steel alloys containing nickel and/or chromium, or a copper sheath with a protective nickel cladding. The electrical lines in the distribution plate 22 are also suitably MI cables, although other types of cabling can be used inside the distribution plate 22 if they are isolated by embedding in the plate. MI cables advantageously do not include plastic or other organic material and accordingly are well suited for use in the caustic high temperature environment inside the pressure vessel. The relatively rigid nature of the MI cables is also advantageous in that it helps ensure the integrity of the pre-assembled CRDM/standoff assembly during transport and installation. However, the rigidity of the MI cables limits their bending radius to relatively large radius turns, so that the MI cables inside the distribution plate 22 should be arranged as straight lines with only large-radius turns. The large area of the distribution plate 22, which spans a substantial portion of the inner diameter of the pressure vessel, facilitates a suitable arrangement of the MI cables inside the plate 22. Additionally, some types of MI cables are susceptible to degradation if the mineral insulation is exposed to water. Accordingly, the ends of the MI cables, e.g. at the coupling with the connector 52 in the standoff and the coupling of the power lines 30 with the electrical connectors 42 in the distribution plate 22, should be sealed against exposure to the primary coolant water. However, advantageously, the connectors 42, 52 themselves can be immersed in water. This makes installation, to be further described, readily performed even with the reactor core immersed in primary coolant. FIG. 7 shows the mating electrical connectors 42, 52 of the distribution plate 22 and CRDM/standoff assembly flange 46, respectively. The female electrical connector 52 (with sockets 48A) of the standoff assembly 24 lowers onto and covers the male electrical connector 42 of the distribution plate. The connectors 42, 52 preferably include glands or other features to prevent ingress of water to the mineral insulation of the MI cables 30, 80 at the junctions of these cables with the respective connectors 42, 52. For example, a glass seal or crushed metal seal may be employed. In this way, the connectors 42, 52 can be mated underwater without exposing the mineral insulation, so as to facilitate installing the CRDM/standoff assembly at the connection site of the distribution plate 22 while keeping the reactor core and the distribution plate 22 submerged in primary coolant. To ensure a good electrical connection, the connection between connectors 42, 52 can be purged to evacuate any trapped water. Alternatively, the electrical connectors could be mated and not purged, albeit typically with some increased resistance due to wet connectors. The connector body has integrated features in both the receptacle and socket for the brazing of the MI cable directly to them. The connector body also has fill holes to allow for insulation packing after the MI cable is spliced to it. The receptacle housings weld-on base is designed such that the entrance angle of the MI cable can be adjusted for. The socket housing also has integrated purge lines for the insertion of the inert gas. Alignment features are integrated into both the receptacle and socket that engage before the pin and sockets to ensure alignment and minimize stress. These alignment features optionally include a compliance feature such as a wave spring to help in allowing for multiple degrees of freedom with the sockets when mating. Alternatively, an elastomer component can be used to drive water out of any voids instead of purging with an inert gas. Multiple MI cables can be routed to a single connector instead of a single connector feeding a single MI cable. FIG. 8 is an enlarged isolation perspective view of the connector 42 that is mounted in the distribution plate 22. Visible are the five male prongs 48B, which in some embodiments are gold-plated pins to reduce electrical contact resistance, penetrating a glass seal plate 49. The hermetically sealed connector formed by the prongs 48B and seal plate 49 may in general be a sealed glass connector, sealed ceramic connector, a sealed glass ceramic connector, or so forth. The connector has a trunk 64 that houses the splice or brazing to the MI cable. FIG. 9 shows an alternative embodiment of the connector. The connector 66 corresponds to the connector 52 of the earlier embodiment, but terminates a single MI cable 80. Connector 66 has a purge line 70 to pump in an inert gas such as nitrogen or argon to evacuate any liquid when mating to connector 67. MI cable 80 is connected to connector 66 via a splicing sleeve 68. The purge line 70 may be alternatively located on the lower connector. FIG. 10 shows a connector 66 installed in the boss or collar 53 in cutaway. Purge line 70 runs into the area between the connectors, allowing the connection to be purged of fluid. The external connection to the purge line 70 optionally comprises a self-sealing connector or other mechanism enabling the purge line 70 to be sealed off after purging so as to trap residual purge gas in the space between the connectors and prevent water ingress after the purging. The female connector 76 mounted in the distribution plate 22 corresponds to the connector 42 of the previous embodiment. Connector 66 on the standoff has a male extension 74 extending from the connector 66 down into the female connector 76 to make the mating. FIGS. 11 and 12 show exploded and assembled perspective isolation views, respectively, of the connectors 66 and 76 of the standoff assembly and distribution plate, respectively. The extending portion 74 of the connector 66 extends down into connector 76. Connector 76 is formed of three parts: a top part 76A, a MI cable weld prep part 76B, and an adjustable weld-on base 76C. The top part 76A receives connector extension 74. The MI cable weld prep portion 76B receives the MI cable and allows the cable to be welded in place. The adjustable weld on base is welded into place, protecting the junction of the MI cable conductors to the top portion. Instead of or in addition to a welded metallic seal, other sealing configurations such as a glass-to-metal seal or a crushed metal seal may be employed. FIG. 12 also illustrates a wave spring 77 optionally used to provide compliance when connecting the connectors 66, 76. FIG. 13 diagrammatically illustrates a method of connecting a CRDM to a standoff to form a preassembled CRDM/standoff assembly and then connecting the CRDM/standoff assembly to the distribution plate. In step S1310, the method starts. In step S1320, the CRDM is bolted to the standoff assembly by a plurality of bolts. In step S1330, the electrical cable(s) are connected the electrical connection(s). In step S1340, the standoff plate, with CRDM bolted on top of it, is lowered onto the distribution plate, with the bolt holes 50 making contact first to ensure proper alignment of the standoff assembly and CRDM. In step S1350, the hold-down bolts are installed and torqued to attach the standoff assembly to the distribution plate and to ensure positive contact in the hydraulic and electrical connectors. At step S1360, the electrical connectors are optionally purged. At step S1370, the method ends. FIG. 14 illustrates a method of removing a CRDM from a distribution plate. In step S1410, the method starts. In step S1420, the hold-down bolts are removed. In step S1430, the CRDM and connected standoff assembly are lifted away from the distribution plate. In step S1440, the CRDM is optionally removed from the standoff assembly for repair or replacement. In step S1450, the method ends. The disclosed approaches advantageously improve the installation and servicing of powered internal mechanical reactor components (e.g., the illustrative CRDM/standoff assembly) by replacing conventional in-field installation procedures including on-site routing and installation of power lines (e.g. MI cables or hydraulic lines) and connection of each power line with the powered internal mechanical reactor component with a simple “plug-and-play” installation in which the power lines are integrated with the support plate and power connections are automatically made when the powered internal mechanical reactor component is mounted onto its support plate. Wet mating is enabled by the use of sealed male and female connectors and optional purging of space between the joined male and female connectors. The disclosed approaches leverage the fact that most powered internal mechanical reactor components are conventionally mounted on a support plate in order to provide sufficient structural support and to enable efficient removal for servicing (e.g., a welded mount complicates removal for servicing). By modifying the support plate to also serve as a power distribution plate with built-in connectors that mate with mating connectors of the powered internal mechanical reactor component during mounting of the latter, most of the installation complexity is shifted away from the power plant and to the reactor manufacturing site(s). The example of FIGS. 1-14 is merely illustrative, and numerous variations are contemplated. For example, the CRDM/standoff assembly can be replaced by a CRDM with an integral mounting flange, that is, the standoff can be integrally formed with the CRDM as a unitary element (variant not shown). With reference to FIGS. 15 and 16, as another illustrative example the disclosed approaches are applied to internal reactor coolant pumps (RCPs) 1400, such as are disclosed in Thome et al., U.S. Pub. No. 2010/0316181 A1 published Dec. 16, 2010 which is incorporated herein by reference in its entirety. For placement of the internal RCPs 1400 in the cold leg (i.e. the downcomer annulus), the RCPs 1400 are envisioned to be mounted on an annular pump plate 1402 disposed in the downcomer annulus. The pump plate 1402 serves as structural support for the RCPs 1400 and also as a pressure divider to separate the upper suction volume and the lower discharge volume. In the illustrative embodiment there are eight connection sites with six of these shown in FIG. 14 as containing RCPs 1400, and the remaining two being unused to illustrate the connection sites. The pump plate 1402 is modified to include MI cables 1404, 1405 disposed in or on the pump plate 1402. The annular shape of the pump plate 1402 precludes long straight runs of MI cable; however, the illustrative MI cables 1404, 1405 are oriented circumferentially with a large bend radius comparable with (half of) the inner diameter of the pressure vessel 11. Bolt apertures 1440 and electrical connectors 1442 are analogous to bolt apertures 40 and electrical connectors 42 of the illustrative CRDM embodiment, respectively. The opening 26 of the connection site of distribution plate 22 translates in the pump plate 1402 to be a generally circular opening 1426 (optionally keyed by a suitable keying feature, not shown) through which the RCPs 1400 pump primary coolant downward. As yet another contemplated modification, it will be appreciated that the female connector can be located in the supporting power distribution plate while the male connector can be located in the flange, standoff or other mounting feature of the internal mechanical reactor component. The preferred embodiments have been illustrated and described. Obviously, modifications and alterations will occur to others upon reading and understanding the preceding detailed description. It is intended that the invention be construed as including all such modifications and alterations insofar as they come within the scope of the appended claims or the equivalents thereof.
046817298	summary	BACKGROUND OF THE INVENTION This invention relates to the monitoring of temperature within a vessel, particularly vessels used for the long term storage of or transfer of nuclear fuel materials. In detecting outside a vessel a change in the temperature inside the vessel it is known to use a temperature probe to measure the temperature of the external surface of the vessel; to use an infra-red detector responsive to the surface temperature of the vessel; to receive outside the vessel radio signals from a temperature detection system in the vessel. The suitability of these known methods is in doubt especially in terms of long term storage of nuclear fuel materials. FEATURES AND ASPECTS OF THE INVENTION According to the invention there is provided a vessel provided internally with a device for enabling external detection of temperature changes within the vessel, said device comprising a permanent magnet, means mounting the magnet within the vessel for movement towards and away from a boundary wall of the vessel, means having sufficient magnetic coupling with the magnet to restrain the magnet against movement away from a predetermined position with respect to said boundary wall of the vessel, and temperature-responsive means for urging the magnet away from said predetermined position whereby, in the event of a substantial rise in temperature within the vessel, the latter means becomes effective to move the magnet relative to the boundary wall to vary the magnetic field strength detectable by detector means located on the external side of said boundary wall. The vessel may be open or closed. The vessel may for example be a nuclear fuel transport flask or a canister for nuclear fuel or a containment vessel for a reactor core. The detector means may comprise a Hall effect transducer responsive to the magnetic field produced by the permanent magnet outside the container. The detector means may alternatively comprise means for generating a time varying magnetic flux in said boundary wall of the vessel which interacts with the magnetic flux of the permanent magnet, and means for sensing magnetic flux external to said boundary wall. The temperature responsive means may include a thermal link so that on breaking of the link the permanent magnet is moved relative to said boundary wall by spring means. The link may be broken by an increase in temperature above a predetermined value or by corrosion.
	claims	1. An illumination system, comprising:a collector configured to concentrate the EUV radiation emitted by a source by reflection in the direction of an optical axis, along which the EUV radiation is guided by successive optical elements,a first optical element configured to generate secondary light sources in the illumination system,a second optical element at the location of the secondary light sources, the second optical element configured to guide the EUV radiation, reflected by the first optical element, along the optical axis, and the second optical element configured to image the first optical element into the illumination field arranged in an image plane which coincides with the object surface,wherein:a maximum of five reflecting elements are arranged between the collector and the illumination field;the optical axis meets the reflecting optical elements at an angle of incidence which is either greater than 60° or less than 30°,an axis angle of deflection between a source axis portion of the optical axis, which runs between the collector and the first optical element, or of the projection of this source axis portion of the optical axis onto an illumination main plane, which lies vertically on the image plane, and in which a field axis portion, which with the image plane encloses an angle less than 90°, of the optical axis runs between the last element of the optical device and the illumination field, and the field axis portion of the optical axis or a projection of the field axis portion of the optical axis onto the illumination main plane is greater than 30°; andat least an axis portion of the optical axis is inclined between at least two of the optical elements relative to the illumination main plane. 2. An illumination system according to claim 1, wherein a size of the illumination field of at least 100 mm2. 3. An illumination system according to claim 1, wherein the second optical element is part of an optical device which includes further optical elements, and which guides the EUV radiation reflected by the first optical element along the optical axis, and images the first optical element into the illumination field being arranged in the image plane, which coincides with the object surface. 4. An illumination system according to claim 3, wherein the optical device includes at least fourth and fifth optical elements after the second optical element, and an axis portion of the optical axis is inclined between the third and the fourth optical elements relative to the illumination main plane. 5. An illumination system according to claim 3, wherein the optical device includes at least fourth and fifth optical elements after the second optical elements and an axis portion of the optical axis is inclined between the first and second optical elements relative to the illumination main plane. 6. An illumination system according to claim 4, wherein an axis portion between the second and the third element of the optical device is inclined relative to the illumination main plane. 7. A illumination system according to claim 1, wherein:after the second optical element, a maximum of two further optical elements are provided;an axis portion of the optical axis between the collector and the first optical element is inclined relative to the illumination main plane; andthe source of the EUV radiation is a plasma source. 8. An illumination system according to claim 1, wherein:after the second optical element only one further optical element is provided;an axis portion of the optical axis between the collector and the first optical element is inclined relative to the illumination main plane; andthe source of the EUV radiation is a plasma source. 9. An illumination system according to claim 1, whereinafter the second optical element, a maximum of two further optical elements are provided;an axis portion of the optical axis between the collector and the first optical element is inclined relative to the illumination main plane;the source of the EUV radiation is a plasma source; andthe collector is in such a form that the EUV radiation is concentrated by at most two reflections on the collector. 10. An illumination system according to claim 1, whereinafter the second optical element only one further optical element is provided;an axis portion of the optical axis between the collector and the first optical element is inclined relative to the illumination main plane;the source of the EUV radiation is a plasma source; andthe collector is in such a form that the EUV radiation is concentrated by at most by two reflections on the collector. 11. An illumination system according to claim 7, wherein an axis portion of the optical axis between the first and second optical elements is inclined relative to the illumination main plane. 12. An illumination system according to claim 1, wherein:the optical device, in addition to the second optical element, includes only two fiber optical elements; andan axis portion of the optical axis between the collector and the first optical element and an axis portion of the optical axis between the second optical element and the third optical element is inclined relative to the illumination main plane. 13. An illumination system, comprising:a collector configured to concentrate EUV radiation emitted by a source by reflection in the direction of an optical axis, along which the EUV radiation is guided by successive optical elements;a first optical element to generate secondary light sources in the illumination system;a second optical element at the location of the secondary light sources, the second optical element configured to guide the EUV radiation, reflected by the first optical element, along the optical axis, and configured to image the first optical element into the illumination field being arranged in an image plane, which coincides with the object surface,wherein:between the collector and the illumination field, a maximum of five reflecting optical elements are arranged,the optical axis meeting the reflecting optical elements at an angle of incidence which is either greater than 60° or less than 30°;an axis angle of deflection between a source axis portion of the optical axis, which runs between the collector and the first optical element, or of the projection of this source axis portion of the optical axis onto an illumination main plane, which lies vertically on the image plane, and in which a field axis portion, which with the image plane encloses an angle less than 90°, of the optical axis runs between the last element of the optical device and the illumination field, and the field axis portion of the optical axis or a projection of the field axis portion of the optical axis onto the illumination main plane is greater than 30°;the optical device, in addition to the second optical element, includes only a third optical element, a fourth optical element and a fifth optical element; andthe optical axis meeting the third, fourth and fifth optical elements at an angle of incidence which is greater than 60°. 14. An illumination system according to claim 1, wherein a numerical aperture of the illumination is at least 0.02. 15. An illumination system according to claim 1, wherein a numerical aperture of the illumination is at least 0.03. 16. An illumination system according to claim 1, wherein a size of the illumination field is at least 500 mm2. 17. An illumination system according to claim 1, wherein a size of the illumination field is at least 800 mm2. 18. A projection exposure system with an illumination system according to claim 1. 19. A method, comprising:providing a system that includes a substrate, a reticle and the projection exposure system of claim 18; andusing the projection exposure system to project at least a portion of the reticle onto an area of a light-sensitive layer of the substrate. 20. The method of claim 19, comprising using the method to produce a structured component.
	claims	1. A device for inspecting a fuel assembly containing a plurality of fuel rods in a pool of a nuclear plant equipped with a grapnel for displacing and suspending the assembly in the pool, comprising:a moveable image sensor including a field of observation, anda boom including a reference graduation extending along an axis parallel to a longitudinal axis of a fuel assembly containing a plurality of fuel rods, so that upon suspending the fuel assembly in the pool by the grapnel, the image sensor may observe in its field both the boom and the assembly, and the boom hangs in the pool beside the fuel assembly so as not to occupy the bottom of the pool, removable fastening means for fastening the boom on the assembly so as to maintain the assembly at a stationary position, wherein the assembly is supported and suspended in the pool by the grapnel at a stationary position while the image sensor is moving. 2. A device according to claim 1, wherein the boom comprises two ends, and the removable fastening means comprises a receptacle of a base of the assembly, connected to at least one end of the boom. 3. A device according to claim 2, wherein another end of the boom is fastened to a float. 4. A device according to claim 2, wherein the fastening means comprises a fork capable of being removably fastened to a top of the assembly. 5. A device according to claim 1, wherein the graduation comprises a tape measure ballasted with a weight. 6. A device according to claim 1, wherein the image sensor is a still camera. 7. A device according to claim 6, wherein the still camera is a high definition digital still camera provided with a motorized telephoto lens. 8. A device according to claim 6, further comprisingmeans for processing and controlling the images acquired by the image sensor, anda digital transmission chain between the still camera and the processing and control means. 9. A device according to claim 1, wherein the image sensor is mounted on a table capable of displacing the sensor in a plane perpendicular to the axis parallel to the longitudinal axis of the assembly on the one hand, and around an axis perpendicular to the axis parallel to the longitudinal axis of the assembly, on the other hand, in order to be able to adjust horizontality of the sensor. 10. A device according to claim 1, wherein the image sensor is mounted on a basket mounted on a rail parallel to the longitudinal axis of the assembly, the basket being capable of displacing the sensor parallel to the longitudinal axis of the assembly. 11. A device according to claim 10, wherein the basket and the rail are part of a lowerator. 12. A method for inspecting at least one fuel assembly in a pool of a nuclear plant, equipped with a grapnel for displacing and suspending the assembly in the pool, comprising:mounting a moveable image sensor including a field of observation, in proximity to the assembly containing a plurality of fuel rods, said method being characterized in that it further includes steps according to which:removably fastening a boom including a reference graduation and extending along an axis parallel to a longitudinal axis of the assembly to the assembly, such that when the assembly is suspended in the pool by the grapnel, the boom hangs in the pool beside the fuel assembly so as not to occupy the bottom of the pool, andobserving via the image sensor both the boom and the assembly in its field while the assembly is suspended by the grapnel.
	abstract	In an illustrative embodiment, a pressurized water nuclear reactor (PWR) includes a pressure vessel (12, 14, 16), a nuclear reactor core (10) disposed in the pressure vessel, and a vertically oriented hollow central riser (36) disposed above the nuclear reactor core inside the pressure vessel. A once-through steam generator (OTSG) (30) disposed in the pressure vessel includes vertical tubes (32) arranged in an annular volume defined by the central riser and the pressure vessel. The OTSG further includes a fluid flow volume surrounding the vertical tubes and having a feedwater inlet (50) and a steam outlet (52). The PWR has an operating state in which feedwater injected into the fluid flow volume at the feedwater inlet is converted to steam by heat emanating from primary coolant flowing inside the tubes of the OTSG, and the steam is discharged from the fluid flow volume at the steam outlet.
051587394	description	DESCRIPTION OF THE PREFERRED EMBODIMENT The figure shows the upper part of the vessel 1 of a pressurized-water nuclear reactor, which consists of a casing of tubular shape having a large wall thickness and arranged with its axial vertical within a vessel well formed in a concrete structure (not shown). The upper part of the tubular vessel comprises a flange 2 having a thickness greater than that of the wall of the vessel in its running part. The flange 2 is intended for receiving the vessel cover ensuring a sealing closure of the inner volume of the vessel during the operation of the reactor. In the vicinity of its upper part, vessel 1 also comprises connection pieces 3, allowing the vessel to be connected to the pipelines of the primary circuit. The drawing illustrates a device making it possible to carryout the dismantling of the vessel by machining its wall, with chips being removed. After final shutdown of the nuclear reactor, the primary circuit and the vessel are cooled and the vessel cover is removed, the reactor pool located above the vessel being filled with water. Both the core assemblies and the internal equipment of the vessel are unloaded and taken away. The reactor pool is subsequently emptied, as is the vessel which can nevertheless be partially filled with water while the dismantling is being carried out. A dismantling device making it possible to carry out the process according to the invention is installed on the upper part of the vessel 1 by means of the polar bridge of the power station or by other suitable lifting and handling means. The drawing shows dismantling device 4 in operating position on the upper part of the vessel 1. The device 4 comprises a tubular support 5 which, in the operating position, is arranged with its axis coinciding with the axis 6 of the vessel 1. Four arms of large cross-section, such as the arms 7a and 7b, are fastened rigidly to the lower part of the support 5 by means of fastening brackets 8 in radial directions perpendicular to the axis 6 arranged at 90.degree. relative to one another about the axis 6 of the support 5. The arms, such as 7a and 7b, are machined internally in their axial direction, to form jack chambers, in which move rods 9 carrying, at their ends, blocks 10 bearing on the inner wall of the vessel 1. The arms, such as 7a and 7b, ensure fastening and centering of the device 4 by flanging within the vessel when the jack rods 9 are in their extended position. Fastened to the upper part of each of the arms, such as 7a and 7b, is a bearing device 12 making it possible to cause the device 4 as a whole to rest on the upper annular surface of the vessel 1, in order to ensure its retention independently of the flanging obtained by the set of radially directed rams, such as 7a and 7b. Each of the bearing devices 12 comprises an arm 13 mounted pivotably on the corresponding arm by means of a joint 14 of horizontal axis. The pivoting arm 13 comprises, at its end opposite the joint 14, a bearing piece 15 coming to rest on the upper annular surface of the vessel 1 when the arm 13 is in its low position, as shown in the left-hand part of the drawing, above the arm 7a. The position of the bearing surface 15 in the direction of an axis 17 perpendicular to this bearing surface can be adjusted by means of a compensating device 16, the functioning of which will be explained later in the test. A jack 18 for actuating the arm 13 is fastened in an articulated manner to the outer surface of the tubular support 5. The rod 19 of the jack 18 is connected to the arm 13, likewise in an articulated manner. The axes of articulation of the jack 18 and of the rod 19 extend in a horizontal direction. As a result of the actuation of the double-acting jack 18 in one direction or the other, the arm 13 can be put in a low bearing position, as shown on the left in the drawing, or in a raised position 13', as shown in the right-hand part of the drawing, the movement of the arm 13 by pivoting between these two positions being represented schematically by the arrow 21. The tubular support 5 carries, in the vicinity of its upper part, a rotary bearing 20 coaxial with support 5 and vessel 1. The bearing 20 is bearing comprising a stationary inner ring fixed to the support 5 and a rotationally movable outer ring to which a support 24 is fastened. A radially directed arm 25 is mounted within the support 24 for movement in a direction 26 corresponding to its longitudinal direction. A geared motor 27 ensuring the drive of a rack-and-pinion assembly mounted in the support 24 makes it possible to move the arm 25 to and fro in the direction 26, as represented schematically by the arrow 28. The arm 25, at its end opposite the support 24, carries a milling head 30 forming the tool for eliminating the irradiated material of the wall of the vessel 1 by machining. The milling head 30 has a milling cutter 31 mounted at the end of a vertically directed spindle driven in rotation by means of a motor 32. The support 5, at its upper end, carries a stationary gear ring 34 above the bearing 20. A geared motor 35 fastened to the movable outer ring of the bearing 20 carries, at the end of its output shaft, a driven pinion 36 meshing with the stationary toothed ring 34. Setting the geared motor 35 and the pinion 36 in rotation makes it possible to drive the outer ring of the bearing 20, the support 24 and the milling head 30 in rotation about the axis 6 of the vessel. The arms, such as 7a and 7b, carry, by means of flexible fastening devices 38, a collecting hopper 39 of frustoconical shape having along its upper edge, a peripheral gasket 40 the diameter of which is substantially equal to the inside diameter of the vessel 1. The gasket 40 makes it possible to obtain a sealing connection between the outer upper edge of the collecting hopper 39 and the inner surface of the vessel 1. A vertically directed conveyor 41, connected in its lower part to an extension 5a of the support 5, is mounted vertically within the support 5, so as to discharge, through its upper part, into an outwardly flared conduit 42. The machining of the upper surface of the vessel 1 by milling in accordance with the procedure described below gives rise to the formation of metal chips and particles of metal 44 which are steered by a deflector 45 in the direction of the collecting hopper 39. The chips and particles 44 coming in contact with the inner surface of the hopper 39 travel by gravity towards the bottom of the hopper, this movement of the chips and particles being facilitated by the presence of a vibrator 47 in contact with the outer surface of the hopper 39. The chips and particles gathering in the lower central part of the hopper 39 are picked up by the conveyor 41 and transported within the support 5 as far as its upper part, in order to be discharged onto a handling device or into a hopper making it possible to feed an induction furnace carrying out the remelting of the chips and particles of irradiated material of the vessel wall. The flared conduit 42 makes it possible to ensure complete recovery of the chips and particles at the upper part of the conveyor 41, since some of these chips or particles can be thrown outside their normal transport path, thereby bringing about contamination of the milling device by these particles of radioactive material. To carry out an operation of dismantling a vessel 1 of a pressurized-water nuclear reactor, the device 4 is put in its operating position on the upper part 2 of the vessel, consisting of the fastening flange for the cover. The bearing devices 12, the arms 13 of which are put in the low position, come to rest with their bearing pieces 15 on the upper surface of the flange 2. The device described, and can comprise more than two arms by feeding the jacks formed in the arms, such as 7a and 7b. The blocks 10 come in contact with the inner surface of the vessel in order to obtain the flanging. The milling cutter 31 which, at the start of the operation, is in a position set back towards the inside of the vessel, is set in rotation, and the arm 25 is moved outwards, so that the milling cutter, the vertical position of which is adjusted by means of the compensating devices 16 of the bearing pieces 15 of the arms 13, can engage into the metal of the vessel wall 1 over a thickness corresponding to the thickness of a machining pass. The movable outer ring of the bearing 20, the support 24, the arm 25 and the milling head 30 are set in rotation about the axis 6 of the vessel by feeding the geared motor 35. The milling head 30, rotating about the axis 6, executes a machining pass over the upper annular surface of the vessel. When the milling cutter comes into the vicinity of a bearing device 12, the arm 13 of which is in the low position, a detector makes it possible to control the corresponding jack 18 by means of a servo valve. The arm 13 is moved by pivoting so as to assume a raised position, such as the position 13'. The compensating device 16 ensures the outward movement of the bearing piece 15 over a distance corresponding to the thickness of the pass. When the milling cutter has carried out the machining on the part of the upper surface of the vessel on which the piece 15 of the arm 13 comes to bear, a detector controls the movement of the jack 18 in the direction causing the arm 13 to be turned downwards, the bearing piece 15 coming in contact with the freshly machined surface of the vessel wall. The adjustment of the position of the bearing piece 15 makes it possible to ensure that this bearing piece is put in perfect contact with the upper surface of the vessel when the pivoting arm 13 is turned down and kept in position by the rod 19 of the jack 18 in its extended position. The machining pass is executed during a complete revolution of the milling head 30 about the axis 6 of the vessel, the arms 13 of the bearing devices 12 being moved into their high position at the moment when the milling cutter passes level with them. In the thickest parts of the vessel, for example in the region of the vessel flange 2 and the connection pieces 3, the complete machining of the upper surface of the vessel over the thickness of one pass may require a radial movement of the milling cutter and the execution of a plurality of machining passes. When the upper part of the vessel 1 has been machined over a thickness corresponding to a machining pass, the milling head 30 is returned to its initial position and the rods 9 of the jacks associated with the arms, such as 7a and 7b, are put in their retracted position, so as to release the blocks 10 for flanging the device 4 within the vessel. The device 4 rests on the bearing devices 12 which are maintained in the low position by means of the jacks 18. The compensating devices 16 are then reinitiated, so that they can make the adjustments, at the moment when the milling cutter passes, during the following machining pass. The jacks associated with the arms, such as 7a and 7b, are actuated so as to ensure the flanging of the device 4 within the vessel. A new machining pass is then executed, as before. The chips and particles formed by the milling cutter 31 are recovered continuously by the hopper 39 and the vertical conveyor 41, so as to be introduced continuously into a remelting induction furnace. The vessel is dismantled by the elimination of the metal of its wall during successive milling passes. All the operations described above are controlled automatically, with the result that the dismantling of the vessel is carried out within the concrete structure in which the vessel well is formed, without any human involvement. This avoids exposing operators in a highly contaminated zone. Moreover, the cutting state of the milling cutter is checked and tracked automatically, so as to make an automatic change of this milling cutter when its state is considered to be defective. There can also be programmed sequences for changing the tool after the device for dismantling by machining has been operating from some time. The tool change is carried out in the conventional way by means of a robotized auxiliary arm which picks up the milling cutter from the milling head in order to introduce it into a magazine or rack and then to install a new milling cutter having a satisfactory cutting state. The dismantling process and device according to the invention therefore make it possible to carry out the dismantling of a nuclear reactor vessel completely automatically, with the result that the period of time necessary for carrying out the machining is of only secondary importance. It is possible to use the same device for dismantling by machining in order to carry out successively the dismantling of all the vessels of a group of nuclear reactors. Moreover, the recovered irradiated material can usually be conditioned by remelting and casting, in order to form blocks of irradiated material of a mass and shape facilitating long-term storage. The machining of the vessel wall can be carried out by using a device different from the one described, e.g., a metal working machine other than a milling head. The means for moving, holding and centering the metalworking machine can be different from those described, and can comprise more than two arms, to ensure the flanging of the machine on the tubular wall. All the handling means making it possible to recover the chips or particles and transport them towards a melting or recovery device can also be different from those described. Finally, the invention is used for the dismantling of any component of a nuclear reactor having at least one part of tubular shape arranged with its axis vertical.
	abstract	The method of protectively coating metallic uranium which comprises dipping the metallic uranium in a molten alloy comprising about 20-75% of copper and about 80-25% of tin, dipping the coated uranium promptly into molten tin, withdrawing it from the molten tin and removing excess molten metal, thereupon dipping it into a molten metal bath comprising aluminum until it is coated with this metal, then promptly withdrawing it from the bath.
058898310	summary	CROSS-REFERENCE TO RELATED APPLICATION This application is a continuation of International Application PCT/EP96/05230, filed on Nov. 26, 1996, which designated the United States. BACKGROUND OF THE INVENTION 1. Field of the Invention The invention relates to a containment of a nuclear power station having a device for igniting hydrogen contained in a hydrogen/air mixture. In the ever so unlikely event of accident situations in which an oxidation of zirconium can occur, for example due to a core heat-up, for precautionary reasons it is expected that hydrogen gas can be formed in a nuclear power station and released inside the containment surrounding the reactor core. An explosive gas mixture, in particular a hydrogen/air mixture, can thereby be produced inside the containment. Various devices or methods are discussed for the purpose of preventing a formation of such an explosive gas mixture in the containment of a nuclear power station. Those include, for example, devices such as catalytic recombiners or catalytically and/or electrically operated ignition devices for igniting hydrogen. When hydrogen ignites, it reacts with oxygen present in the air in the containment to form water, and is thereby recombined. Such an ignition device is disclosed, for example, in German Published, Non-Prosecuted Patent Application DE 41 25 085 A1, corresponding to U.S. Pat. Nos. 5,492,686 and 5,592,521. Alternatively, consideration is also given to methods for permanently or subsequently inerting the containment. In a conventionally used electrically driven ignition device such as is described, for example, in German Patent DE 30 04 677 C, a multiplicity of igniters or spark plugs is disposed inside the containment of the nuclear power station. Each of those igniters is suitable for igniting the hydrogen of the hydrogen/air mixture in its immediate environment, and thus for recombining it. The action of such an igniter is therefore narrowly limited locally. It is therefore impossible to keep the hydrogen concentration inside the entire containment at a uniformly low level below the prescribed limiting value, by using such igniters. SUMMARY OF THE INVENTION It is accordingly an object of the invention to provide a containment of a nuclear power station equipped for the ignition of hydrogen contained in a hydrogen/air mixture, which overcomes the hereinafore-mentioned disadvantages of the heretofore-known devices of this general type in such a way that in the event of an accident it is possible to ensure the maintenance of a particularly low hydrogen concentration in its entire interior. With the foregoing and other objects in view there is provided, in accordance with the invention, a containment of a nuclear power station, comprising a device for igniting hydrogen contained in a hydrogen/air mixture, the device including a central electrode for lightning flash generation and a high-voltage source connected to the central electrode for generating a high voltage greater than a disruptive discharge voltage of air. In a containment that is constructed in such a way, each grounded internal part can act as a counter-electrode for lightning flash generation by spark discharge, because of the electrical potential difference between the central electrode and grounded internal parts. The central electrode, which is placed under a high voltage in the event of an accident, is then discharged through a lightning flash discharge to the counter-electrode. In this case the lightning flashes are chaotic per se, with the result that every possible counter-electrode becomes the target of a lightning flash discharge in accordance with statistical criteria. The lightning flash discharges thus penetrate the entire interior of the containment, so that it is possible for subspaces inside the containment to be penetrated particularly frequently, depending on the configuration of the internal parts. It is thereby ensured that no subspaces are formed from which the chaotic lightning flash discharges are excluded and in which the hydrogen concentration could exceed a critical value. The high-voltage source can be constructed as a DC voltage source, for example in the manner of a Van-De-Graaff generator. In accordance with another feature of the invention, in order to preclude endangering the internal parts of the containment, as well as putting operational staff at risk from the lightning flash discharges, the high-voltage source is constructed to generate a high voltage with a frequency of more than 1 kH. On one hand, specifically, the frequency-dependent disruptive discharge voltage of air is lower for high frequencies than for low frequencies. A lightning flash discharge at a high frequency can then already be triggered at a peak voltage which is low in comparison with a low frequency. On the other hand, because of the skin effect, the lightning flash discharges emanating from the central electrode connected to such a high-frequency high-voltage source are already screened at a very shallow depth of penetration into the counter-electrode or an internal part, and are therefore incapable of penetrating into the interior of the respective internal part. In accordance with a concomitant feature of the invention, there is provided a number of counter-electrodes disposed in the interior of the containment for conducting lightning flashes. The counter-electrodes can define a preferred direction for a lightning flash discharge. As a result, the containment subspace surrounding the counter-electrodes is preferably penetrated by the lightning flash discharges. Consequently, hydrogen which is, for example, preferably produced in a subspace of the containment known as a hydrogen source, can be ignited in a particularly effective manner by a number of counter-electrodes in this subspace. However, even in the case of a containment having a number of such counter-electrodes, the penetration of the subspace of the containment which is not provided with counter-electrodes, by lightning flash discharges is ensured, although with reduced probability. The result is that even for such a containment the ignition of hydrogen is ensured in its entire interior or volume. The advantages achieved with the invention are, in particular, that due to the central electrode for lightning flash generation which is connected to the high-voltage source, chaotic lightning flash discharges can be generated between the central electrode and the internal parts of the containment which act as counter-electrode. Such chaotic lightning flash discharges penetrate the entire volume of the containment, including those subregions or subspaces of the containment which cannot be adequately reached by conventional ignition devices. The result is to ensure volumetric bonding of the hydrogen, that is to say bonding of the hydrogen in the entire volume of the containment. Other features which are considered as characteristic for the invention are set forth in the appended claims. Although the invention is illustrated and described herein as embodied in a containment of a nuclear power station, 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.
052251470	summary	This application is accompanied by a microfiche appendix having 2 microfiche films. A portion of the disclosure of this patent document contains material which is subject to copyright protection. The copyright owner has no objection to the facsimile reproduction by any one of the patent disclosure, as it appears in the Patent and Trademark Office files or records, but otherwise reserves all copyright rights whatsoever. FIELD OF THE INVENTION This invention relates to analyzing light water reactor core neutronics in real-time, more specifically to determining core neutronics for simulation training and engineering analyzers. BACKGROUND OF THE INVENTION In the field of nuclear power facilities, it is important to analyze the reactor core neutronic properties for maintaining the nuclear power facility and training reactor operators to perform routine and emergency monitoring procedures. Heretofore, core neutronics for a light water reactor have been analyzed using engineering codes, such as the coarse mesh method described in Borresen, "A simplified, Coarse-Mesh, Three-Dimensional Diffusion Scheme for Calculating the Gross Power Distribution in a Boiling Water Reactor," Nucl. Sci. Engr., 44, 37, 1971, and methods of the RAMONA-3B code described in Wulff et al., "A Description and Assessment of RAMONA-3B Mod. O Cycle 4: A Computer Code with Three-Dimensional Neutron Kinetics for BWR System Transients," NUREG/CR-3664, Brookhaven National Laboratory, January 1984. These codes provide methods for providing a set of core neutronics parameters in a defined circumstance and analyzing or determining the resultant reactor core neutronics parameters in response to the given conditions. The Borresen reference refers to obtaining core neutronics data and solving modified two-group neutron diffusion equations for the two types of neutrons inside the core, namely the fast neutrons and the thermal neutrons. A thermal neutron may be considered as a fast neutron that has slowed down. More specifically, the reactor core is represented as a number of nodes that are spaced apart such that the fast neutrons have a relatively large mean free path (i.e., diffusion length) and the thermal neutrons have a low leakage from node to node. This permits using an approximation for the thermal neutrons leakage and a modification of the two-group equations to simplify the number of steps required to determine the core neutronics for the given conditions. RAMONA-3B, developed by the Brookhaven National Laboratory, uses the Borresen coarse mesh method and also relies on the fast neutrons as the determining criteria. However, the RAMONA-3B method relies on solving the two-group neutron diffusion equations by a finite difference method to determine the core neutronics for the given conditions. One of the problems with these known techniques is that the model does not have the capability to run the code from power plant start up to shutdown continuously in real time. They do not have the ability to analyze dynamic or static conditions in real-time. Consequently, they are limited in their application to selected transient conditions. Further, those known techniques are not sufficiently flexible to train operators under a wide variety of conditions or in real-time environments. It is therefore, an object of the present invention to provide for determining core neutronics in a real-time environment. It is another object to provide for a real-time analysis of core neutronics that can be used for simulation training of facility operators and for engineering analysis of core neutronics, separately or simultaneously. It is another object of the invention to provide for determining core neutronics in response to rapid transient conditions in a real-time environment. It is another object of the invention to provide a real-time analysis of core neutronics under normal and emergency operating conditions. It is another object of the invention to simulate real-time core neutronics under normal, emergency, and beyond design conditions continuously. SUMMARY OF THE INVENTION The present invention provides for methods and apparatus for sensing the core neutronic parameters of a nuclear reactor core and analyzing and determining the core neutronics in a real-time environment. One aspect of the present invention concerns real-time analysis of light water reactor core neutronics in detailed three-dimensional geometry. More specifically, a method is provided including modeling the reactor as a plurality of nodes in a conventional manner, monitoring the pertinent input core neutronics parameters for the type of reactor core, providing time dependent two group neutron diffusion equations that have been subjected to a space-time factorization of the neutron flux and delayed neutron precursors by amplitude and shape, substituting a coarse mesh finite difference approximation for fast neutron shape functions and determining the resulting core neutronics by application of the modified time dependent, two-group neutron diffusion equations, using a constant time step in the calculations. Preferably, the time step is not less than one quarter second. Another aspect of the invention concerns a method for determining the neutronics parameters of a reactor core. One such method comprises the steps of: representing the reactor core as a plurality of nodes; monitoring selected neutronic parameters of the reactor core; providing time-dependent two group neutron diffusion equations coupled to delayed neutron precursor concentrations that have been subjected to space-time factorization by shape and amplitude functions in response to the plurality of nodes, sensing the monitored parameters; and determining the core neutronics parameters in response to the sensed parameters and the provided two group neutron diffusion equations in constant time steps, the time steps being less than one quarter second. Preferably, the method includes the step of selecting a coarse nodal representation of the reactor core and a time step for sensing the monitored parameters and determining the core neutronics parameters in a real-time environment. The solution methods in solving the shape functions and the amplitude functions are in real time, thereby providing the capability for simulating the full range of operation of a core continuously. The reference to the full range of operation of a core continuously should be understood to include, without limitation, transient, steady states, malfunction, and shutdown operations of the core. Preferably, for a pressurized water reactor, (PWR), each fuel assembly is represented as a radial node and for a boiling water reactor (BWR), each control cell having four fuel bundles surrounding a control blade position is represented as a radial node. Each radial node should have the same size. For BWRs, nodes on the core periphery will consist of fewer bundles. It should be understood that each radial node may have a plurality of radial nodes. Preferably the number of radial nodes is the same for each radial node, for example, from 8 to 24 axial nodes. This is known as a coarse node or coarse mesh model which permits use in a real-time environment. In a preferred embodiment, the invention also provides for monitoring certain thermohydraulic parameters associated with the reactor core, non-condensibles, and soluble boron quantities, and analyzing xenon and samarium concentrations and decay heat in the reactor core. The foregoing sensed thermohydraulic parameters are preferably provided by a real-time thermohydraulic analysis which is described in the copending and commonly assigned U.S. patent application Ser. No. 07/761,000, filed Sep. 17, 1991, entitled "REAL-TIME ANALYSIS OF POWER PLANT THERMOHYDRAULIC PHENOMENA", in the names of Guan-Hwa Wang and Zen-Yow Wang, which application is hereby incorporated by reference herein. This provides for simulator training and engineering analysis of a wide range of power plant scenarios, such as feedbacks between thermohydraulics and neutronics, operational and severe transients, human factor research, and design modification analysis. One advantage of the present invention is that it provides for analyzing a wide variety of fast transients, including, for example, thermohydraulic transients in the nuclear steam supply system, control rod movement, soluble boron changes, and xenon effects. Consequently, the invention can be used to simulate and to analyze core neutronics during startup and normal operation, anticipated operational occurrences, design-basis accidents, and many beyond design-basis accidents. Another advantage of the present invention is that it can be incorporated into a modern minicomputer or engineering workstation and utilized in a real-time environment. This provides for increased flexibility, particularly for simulation training of operators and real-time engineering analysis. Further, such a computer can be located in, near or remote from the control room of the reactor, thus providing for real-time simulation without interfering with the supervision or operation of the reactor. Another advantage of the invention is that it is compatible with many NRC-approved safety engineering analysis codes that are currently used for fuel management and reload safety analysis, thus providing for enhanced core neutronics analysis and simulation. The input data required by those engineering analysis codes can be easily adopted as the input data for the present invention.
	summary
	abstract	A system includes a beam filter positioning device including a plate configured to support one or more beam filters, and one or more axes operable to move the plate relative to a beam line. A control mechanism is coupled to the one or more axes for controlling the movement of the axes and configured to automatically adjust the position of at least one of the one or more beam filters relative to the beam line.
	description	Integrated chip (IC) designs are complicated schematic layouts that contain millions or billions of semiconductor devices (e.g., transistors, capacitors, etc.) interconnected together by conductive wires. Typically, an IC design represents IC components as a plurality of polygons. Integrated chips are generated by operating on a semiconductor substrate with a plurality of processing steps (e.g., lithography, implantations, etching, etc.) to form on-wafer shapes corresponding to the polygons within the substrate. As the size of integrated chip components has decreased, it has become increasingly difficult to form on-wafer shapes that accurately correspond to designed shapes due to image errors and/or processing effects. To make on-wafer shapes more closely resemble designed shapes a number of resolution enhancement techniques are used in modern day fabrication processes. One such resolution enhancement technique is optical proximity correction (OPC). OPC procedures reduce optical proximity effects on a designed shape by moving one or more edges of the design shape before the shape is written to a mask. The description herein is made with reference to the drawings, wherein like reference numerals are generally utilized to refer to like elements throughout, and wherein the various structures are not necessarily drawn to scale. In the following description, for purposes of explanation, numerous specific details are set forth in order to facilitate understanding. It may be evident, however, to one skilled in the art, that one or more aspects described herein may be practiced with a lesser degree of these specific details. In other instances, known structures and devices are shown in block diagram form to facilitate understanding. Optical proximity correction (OPC) procedures are performed on integrated chip (IC) designs to correct for optical proximity effects by moving one or more edges of an IC design. OPC procedures may dissect a polygon of an original IC design into a plurality of separate edges and then selectively move one or more of the separate edges in a direction perpendicular to corresponding edges of the original IC design. By varying the shape of the polygon of the original design, on-wafer errors caused by optical proximity effects can be reduced. However, it has been appreciated that moving the separate edges in a direction perpendicular to edges of the original IC design limits the degrees of freedom of an OPC procedure. The limitation to the degrees of freedom may lead to poor convergence of correction results and/or may cause difficulties in achieving an optimal solution for complicated IC designs. For example, perpendicular movement of an edge having a 45° angle may result in the formation of an unwanted negative polygon that can cause an OPC procedure to crash and/or may cause a simulated litho contour to be off-target. Accordingly, the present disclosure relates to a method of performing an optical proximity correction (OPC) procedure that provides for a high degree of freedom by using an approximation design layer. In some embodiments, the method is performed by forming an integrated chip (IC) design comprising a graphical representation of an integrated chip having an original design layer with one or more original design shapes. An approximation design layer, which is different from the original design layer, is generated from the original design layer. The approximation design layer is a design layer that has been adjusted to remove features that may cause optical proximity correction (OPC) problems. An optical proximity correction (OPC) procedure is then performed on the approximation design layer. By performing the OPC procedure on the approximation design layer rather than on the original design layer, characteristics of the OPC procedure can be improved. FIG. 1 illustrates a flow diagram of some embodiments of a method 100 of performing an optical proximity correction (OPC) procedure to correct an original design layer, by operating upon an approximation design layer that is different than the original design layer. While method 100 is illustrated and described below as a series of acts or events, it will be appreciated that the illustrated ordering of such acts or events are not to be interpreted in a limiting sense. For example, some acts may occur in different orders and/or concurrently with other acts or events apart from those illustrated and/or described herein. In addition, not all illustrated acts may be required to implement one or more aspects or embodiments of the description herein. Further, one or more of the acts depicted herein may be carried out in one or more separate acts and/or phases. At 102, an integrated chip (IC) design comprising a graphical representation of an integrated chip is formed. The IC design comprises an original design layer having one or more original design shapes corresponding to structures that are to be formed on an integrated chip (i.e., on-wafer). In some embodiments, the original design shapes may comprise a plurality of polygons, wherein one or more of the polygons are connected together. At 104, an approximation design layer, different than the original design layer, is generated from the original design layer. The approximation design layer is a design layer that is generated based upon the original design layer, but which eliminates undesirable design features of the original design layer that may lead to OPC problems. For example, the approximation layer can be generated to have one or more geometric features that enable better convergence of an OPC procedure and/or better shape fidelity for on-wafer shapes, with respect to the original design layer. In some embodiments, the approximation design layer may be generated by modifying an original design layer to remove unwanted features. In some embodiments, an approximation design shape may be generated by modifying an original design shape to replace an angled edge (e.g., having a slope of 30°, 45°, 60°, etc.) with a vertical edge and a horizontal edge. For example, in some embodiments an approximation design shape may be generated by modifying an original design shape to replace a 45° edge having a substantially 45° slope with a vertical edge (having a 90° slope) and a horizontal edge (having a 0° slope). In such embodiments, replacement of the 45° edge with the vertical/horizontal edges eliminates problems that the 45° edge may cause during operation of a subsequent OPC procedure. For example, removal of the 45° edge may improve OPC convergence, fidelity of on-wafer shapes, or features formed on mask. At 106, an optical proximity correction (OPC) procedure is performed on the approximation design layer to form an OPC layer having one or more OPC'd shapes that produce on-wafer shapes that closely resembles the original design layer. The OPC procedure modifies one or more approximate design shapes within an approximation design layer to improve the process window of resulting OPC'd shapes. For example, in some embodiments, the OPC procedure may move one or more edges of the approximation design layer to improve the process window of the approximation design layer in a manner that causes a resulting on-wafer shape to more closely resemble an original design shape (e.g., that mitigates corner rounding, line end shortening, etc.). The difference between the original design layer and the approximation design layer allows for a different OPC procedure to be used for correction of the approximation design layer than that which would have been used for correction of the original design layer. In some embodiments, by performing OPC on the approximation design layer rather than on the original design layer, an OPC procedure having a greater number of degrees of freedom may be used. The greater number of degrees of freedom provided by the approximation design layer can improve a process window and/or convergence of an OPC procedure, for example. In some embodiments, the OPC procedure of act 106 may be performed by separating respective edges of an approximation design shape into a plurality of distinct edges, at 108. For example, in some embodiments, an edge of an approximation design shape having a length L may be separated into n contiguous, distinct edges respectively having a length L/n. One or more of the distinct edges are then selectively moved to adjust the shape of the approximation design shape in a manner that mitigates optical proximity effects, at 110. For example, in some embodiments, one or more distinct edges of an approximation design shape may be moved to form resolution enhancement features (e.g., serifs or hammerheads) within a resulting OPC'd shape. At 112, the OPC'd shapes generated from the approximation design layer are written to a photomask set. In various embodiments, the photomask set may comprise a single mask set or a multi-mask set containing a binary mask, an alternating phase-shift mask, or an attenuated phase-shift mask. In some embodiments, writing the OPC'd shapes results in a photomask having horizontal and vertical edges at a position corresponding to an angled edge (e.g., a substantially 45° edge) within the original design layer. Therefore, method 100 performs an OPC procedure on an approximation design layer rather than on an original design layer to enable the OPC procedure to operate according to a greater number of degrees of freedom. By operating according to a greater number of degrees of freedom, the OPC procedure can improve convergence of the OPC model to improve a corresponding on-wafer shape. FIGS. 2-5 illustrate some embodiments of an integrated chip design upon which a method of optical proximity correction is performed. Although FIGS. 2-5 are described in relation to method 100, it will be appreciated that the structures disclosed in FIGS. 2-5 are not limited to such a method. FIG. 2 illustrates some embodiments of a top-view corresponding to act 102. The top-view shows an integrated chip (IC) design 200 comprising a graphical representation of an integrated chip. The IC design 200 comprises an original design shape 202 formed on an original design layer. The original design shape 202 comprises a polygon having a number of contiguous edges. One of the edges comprises a 45° edge 204 having a 45° slope. In some embodiments, the IC design 200 may comprise additional design shapes (not shown) that may be located in proximity to original design shape 202 or which may be coupled to original design shape 202. In some embodiments, the IC design 200 may comprise a Graphic Database System (GDS) file, such as a GDS or GDSII file. In other embodiments, the IC design 200 may comprise a CIF file, an OASIS file, or some other similar file format, for example. In some embodiments, the IC design 200 may be formed by a designer using a design software program running on an electronic design automation (EDA) tool. In other embodiments, the IC design 200 may be formed by an automatic place and route tool configured to automatically place the original design shape 202 within the IC design 200. It will be appreciated that the original design layer may comprise any design layer. In some embodiments, the original design layer may comprise an original design shape 202 having a metal interconnect shape. For example, the original design shape 202 may comprise a metal shape on a first back-end-of-the-line (BEOL) metal layer. In other embodiments, the original design layer may comprise a polysilicon layer, for example. FIG. 3 illustrates some embodiments of a top-view 300 corresponding to act 104. As shown in top-view 300, an approximation design shape 302 (on an approximation design layer) is generated from the original design shape 202. Since the 45° edge 204 may cause convergence problems during a subsequent OPC procedure, the 45° edge 204 is replaced with a horizontal edge 304 and a vertical edge 306 within the approximation design shape 302. The horizontal and vertical edges, 304 and 306, will not cause convergence problems during the subsequent OPC procedure, as explained below. In some embodiments, the original design shape 202 may comprise a subset of the approximation design shape 302. For example, as shown in top-view 300 the original design shape 202 is contained within the bounds of the approximation design shape 302. In other embodiments, the original design shape 202 may extend outside of the approximation design shape 302. FIG. 4 illustrates some embodiments of a top-view 400 corresponding to act 106. As shown in top-view 400, during an OPC procedure one or more edges of the approximation design shape 302 are separated into a plurality of distinct edges 402a-402n. For example, a top edge of the approximation design shape 302, which has a length L is separated during an OPC procedure into distinct edges 402a-402d, which respectively have a length L/4. The plurality of distinct edges 402a-402n are contiguous, so that the plurality of distinct edges 402a-402n collectively form the approximation design shape 302. FIG. 5 illustrates some embodiments of a top-view 500 corresponding to act 108. As shown in top view 500, one or more of the distinct edges 402a-402n of the approximation design shape 302 are selectively moved to form an OPC'd shape 502 (located within an OPC'd layer) that corrects proximity effects of the original design shape 202. For example, edge 402a is moved outward to form OPC'd edge 504a, while edge 402b is moved inward to form OPC'd edge 504b. One or more of the distinct edges 402a-402n of the approximation design shape 302 are moved in a non-perpendicular direction with respect to edges of the original design shape 202. For example, edges 402d and 402e are moved at 45° angles with respect to the 45° edge 204 of the original design shape 202 to form edges OPC'd 504d and 504e, respectively. The OPC'd shape 502 will change during processing to form an on-wafer shape that meets the target points 506a-506n. In some embodiments, the OPC procedure selectively adds assist features to the approximation design shape 302 that enable the OPC'd shape 502 to meet the target points 506a-506n. For example, by moving edges 402a and 402n outward to respectively form OPC'd edges 504a and 504n, a hammerhead is formed that reduces corner rounding. The assist features are configured to improve the process window of the OPC'd shape 502, thereby allowing for on-wafer shapes to more closely correspond to original design shape 202. FIGS. 6A-6B illustrate some embodiments of top views, 600 and 608, of an IC design that illustrate the difference between an OPC procedure performed on an original design shape 602 and an OPC procedure performed on an approximation design shape 610. As will be more fully appreciated below, by performing the OPC procedure on the approximation design shape 610 rather than on the original design shape 602, the OPC procedure may have a greater number of degrees of freedom. For example, FIG. 6A illustrates a top-view 600 that shows the movement of one or more of the distinct edges of the original design shape 602 to correct proximity effects. The original design shape 602 comprises mirror imaged 45° edges, 604a and 604c, separated by an interconnecting vertical edge 604b. During an OPC procedure, edges 604a-604c cannot be moved outward without generating the negative polygon 606. Therefore, since edges 604a-604c can move inward but not outward, the original design shape 602 provides an OPC procedure with a limited number of degrees of freedom that can be used for OPC correction. In contrast, FIG. 6B illustrates a top-view 600 that shows the movement of one or more of the distinct edges of the approximation design shape 610 to correct proximity effects. To generate the approximation design shape 610, the mirror imaged 45° edges, 604a and 604c, and the interconnecting vertical edge 604b are replaced by horizontal and vertical edges 612a-612f. During an OPC procedure on the approximation design shape 610, edges 612a-612f can be moved outward without generating a negative polygon. Therefore, since edges 612a-612f can be moved both inward and outward, the approximation design shape 610 provides an OPC procedure with a larger number of degrees of freedom than the OPC procedure used on the original design shape 602. FIG. 7 illustrates a block diagram of some embodiments of an EDA (Electronic design automation) tool 700 configured to execute the disclosed method of optical proximity correction. The EDA tool 700 comprises a computation element 702 and a memory element 704. The computation element 702 comprises an OPC element 706 and an approximation design level generation element 708. The memory element 704 is configured to store an original integrated chip (IC) design 710 (e.g., a GDS or GDSII file, a CIF file, or an OASIS file), an approximation design 712, an OPC (optical proximity correction) model 714, and computer readable instructions (CRI) 716 that provide for a method of operating one or more components of the EDA tool according to disclosed method 100. In various embodiments, the memory element 704 may comprise an internal memory or a computer readable medium. The approximation design level generation element 708 is configured to generate the approximation design 712 from the original IC design 710. The approximation design 712 is different from a layout of the original IC design 710. The OPC element 706 is configured to selectively access the OPC model 714, and based thereupon to perform an OPC procedure on the approximation design 712 to reduce proximity effects in the original IC design 710. For example, the OPC element 706 may add assist features to one or more shapes within the approximation design 712. The EDA tool 700 further comprises a design tool 718 configured to generate the original IC design 710. In some embodiments, the design tool 718 may comprise an automatic place and route tool configured to selectively route shapes on a plurality of design levels to generate the original IC design 710. In other embodiments, the design tool 718 may comprise a user interactive design environment that allows for designers to generate the original IC design 710. In such embodiments, the EDA tool 700 may comprise an input device 720 and/or an output device 722. The input device 720 is configured to allow a user to interact with the original IC design 710 and in various embodiments may comprise a keyboard, mouse, and/or any other input device. The output device 722 is configured to provide a graphical representation of the original IC design 710 that can be viewed by a user. In various embodiments, the output device 722 may comprise a monitor, for example. In some embodiments, the EDA tool 700 may further comprise a mask writing tool 724. The mask writing tool is configured to generate a mask set comprising shapes corresponding to the approximation design 712. In various embodiments, the mask set may comprise a binary mask, an alternating phase-shift mask, or an attenuated phase-shift mask. In some embodiments, the mask set may comprise a single mask set or a multi-mask set (e.g., a double mask set, a triple mask set, etc.). It will be appreciated that while reference is made throughout this document to exemplary structures in discussing aspects of methodologies described herein (e.g., the IC design presented in FIGS. 2-5, while discussing the methodology set forth in FIG. 1), that those methodologies are not to be limited by the corresponding structures presented. Rather, the methodologies (and structures) are to be considered independent of one another and able to stand alone and be practiced without regard to any of the particular aspects depicted in the Figs. Additionally, layers described herein, can be formed in any suitable manner, such as with spin on, sputtering, growth and/or deposition techniques, etc. Also, equivalent alterations and/or modifications may occur to those skilled in the art based upon a reading and/or understanding of the specification and annexed drawings. The disclosure herein includes all such modifications and alterations and is generally not intended to be limited thereby. For example, although the figures provided herein, are illustrated and described to have a particular doping type, it will be appreciated that alternative doping types may be utilized as will be appreciated by one of ordinary skill in the art. In addition, while a particular feature or aspect may have been disclosed with respect to only one of several implementations, such feature or aspect may be combined with one or more other features and/or aspects of other implementations as may be desired. Furthermore, to the extent that the terms “includes”, “having”, “has”, “with”, and/or variants thereof are used herein, such terms are intended to be inclusive in meaning—like “comprising.” Also, “exemplary” is merely meant to mean an example, rather than the best. It is also to be appreciated that features, layers and/or elements depicted herein are illustrated with particular dimensions and/or orientations relative to one another for purposes of simplicity and ease of understanding, and that the actual dimensions and/or orientations may differ substantially from that illustrated herein. The present disclosure relates to a method of performing an optical proximity correction (OPC) procedure to correct an original design layer, by operating upon an approximation design layer that is different than the original design layer. In some embodiments, the present disclosure relates to a method of performing an optical proximity correction (OPC) procedure. The method comprises forming an integrated chip (IC) design comprising a graphical representation of an integrated chip, wherein the IC design has an original design layer comprising one or more original design shapes. The method further comprises generating an approximation design layer from the original design layer, wherein the approximation design layer is different than the original design layer. The method further comprises performing an optical proximity correction (OPC) procedure on the approximation design layer to form an OPC'd layer that produces on-wafer shapes that correspond to the one or more original design shapes. In other embodiments, the present disclosure relates to a method of performing an optical proximity correction (OPC) procedure. The method comprises forming an integrated chip (IC) design comprising a graphical representation of an integrated chip having an original design layer comprising one or more original design shapes corresponding to structures that are to be formed on an integrated chip. The method further comprises generating an approximation design layer, from the original design layer, by replace a substantially 45° edge in the original design layer with a vertical edge having a 90° slope and a horizontal edge having a 0° slope in the approximation design layer. The method further comprises separating edges of the approximation design layer into a plurality of distinct edges that are contiguously connected. The method further comprises selectively moving the distinct edges to adjust a shape of the approximation design layer to form an OPC'd layer that produces on-wafer shapes that correspond to the one or more original design shapes. In yet other embodiments, the present disclosure relates to an EDA (Electronic design automation) tool. The EDA tool comprises a memory element configured to store an integrated chip (IC) design comprising a graphical representation of an integrated chip, wherein the IC design has an original design layer comprising one or more original design shapes. The EDA tool further comprises an approximation design generation element configured to generate an approximation design layer from the original design layer, wherein the approximation design layer is different than the original design layer. The EDA tool further comprises an OPC element configured to perform an optical proximity correction (OPC) procedure on the approximation design layer to form an OPC'd layer that produces on-wafer shapes that correspond to the one or more original design shapes.
040428281	abstract	A rack for storing spent nuclear fuel elements in which a plurality of aligned rows of upright enclosures of generally square cross-sectional areas contain vertically disposed spent fuel elements. Each fuel element is supported at the lower end thereof by a respective support that rests on the floor of the spent fuel pool for a nuclear power plant. An open rack frame is employed as an upright support for the enclosures containing the spent fuel elements. Legs at the lower corners of the frame rest on the floor of the pool to support the frame. In one exemplary embodiment, the support for the fuel element is in the form of a base on which a fuel element rests and the base is supported by legs. In another exemplary embodiment, each fuel element is supported on the pool floor by a self-adjusting support in the form of a base on which a fuel element rests and the base rests on a ball or swivel joint for self-alignment. The lower four corners of the frame are supported by legs adjustable in height for leveling the frame. Each adjustable frame leg is in the form of a base resting on the pool floor and the base supports a threaded post. The threaded post adjustably engages a threaded column on which rests the lower end of the frame.
044903286	claims	1. A gas cooled, high temperature nuclear reactor installation comprising: a prestressed concrete pressure vessel having a reactor cavity; a steel liner for said reactor cavity attached to said pressure vessel, said steel liner having a bottom member; and a bottom shield mounted on said bottom member of said liner wherein said bottom shield comprises: first means for spacing said first plate from said second plate coaxially thereto; and second means for spacing said second plate member above said bottom member of said liner. a supporting ring formed of a plurality of interconnected supporting ring members, said supporting ring circumferentially surrounding said lower circular plate. 2. A nuclear reactor installation as claimed in claim 1 in which said upper circular plate is thinner than said lower circular plate. 3. A nuclear reactor installation as claimed in claim 2 in which said lower circular plate is provided with passage orifices to permit cooling gas to flow into the space created by said first spacing means and onto said upper circular plate. 4. A nuclear reactor installation as claimed in claim 3 wherein said first and second spacing means comprise a plurality of bearing supports. 5. A nuclear reactor installation as claimed in claim 4 further comprising: 6. A nuclear reactor installation as claimed in claim 5 wherein said.interconnecting plate members are secured by means of pins in said overlapping radial portions. 7. A nuclear reactor installation as claimed in claim 6 further comprising a plurality of sheet metal plates secured in spaced relation to one another on the upper surface of said upper circular plate by means of pins, thereby forming a plurality of expansion gaps. 8. A nuclear reactor installation as claimed in claim 7 wherein said upper circular plate has a waffle-like pattern on its upper surface. 9. A nuclear reactor installation as claimed in claim 8 wherein said upper circular plate has a ribbed lower surface. 10. A nuclear reactor installation as claimed in claim 9 wherein said distances between said bearing supports for said upper circular plate are smaller than the distances between said bearing supports for said lower circular plate.
	description	This application is a national phase application of International Application No. PCT/EP2009/058458, filed Jul. 3, 2009, designating the United States and claiming priority to European Patent Application No. 08159577.9, filed Jul. 3, 2008, both of which are incorporated by reference herein in their entirety. The invention is related to the field of charged Hadron Therapy, i.e. radiation therapy using strongly interacting particles. More particularly, the invention is related to a detector and method for measuring the beam range of a charged hadron beam in a target object as well as the particle dose distribution in the target object. It is well established that charged hadrons (i.e., protons, pions, ions such as carbon ions) have physical advantages with respect to X-rays or gamma rays in the field of radiation therapy. For example, protons of a given energy (i.e. forming a mono-energetic proton beam), have a certain range or penetration depth in a target object and do not penetrate beyond that range, and furthermore, they deposit their maximum amount of energy or dose in the so-called Bragg Peak, which corresponds to the point of greatest penetration of the radiation in the target volume. Since the Bragg peak position depends on the energy of the hadron beam, it is evident that by precisely controlling and modifying the energy one can place the Bragg Peak at a suited depth of a tumour so as to administer the greatest radiation energy to selected points and spare, by contrast, healthy tissue surrounding said points. As a consequence, the location of the Bragg peak must be precisely known since critical tissue localized near the target tumor could receive overdoses, whereas conversely the target tumor could receive underdoses. Even if existing radiation treatment planning tools calculate a “theoretical” or “planned” beam range and dose distribution based on CT images, it is almost impossible to fully take into account the complex atomic composition of surrounding tissues. As a consequence, a degree of uncertainty is affecting the particle range. There is a need therefore to obtain a direct on-line, i.e. during beam delivery, measurement of the particle range. The direct detection of the end of the dose deposition of a hadron beam is impossible since the hadron beam stops inside the treatment volume, even during the irradiation. A solution could be the detection of secondary particles giving quantitative information on the dose deposition of the particle beam. Such a strategy was disclosed in document ‘Comparison of In-Beam and Off-beam PET Experiments at Hard Photons’, Möckel et al, 2007 IEEE Nuclear Science Symposium Conference Record, p. 4113-4116. This document describes the measurement of β+ activity distribution simultaneously—on-line—with the irradiation with a hard photon beam. In brief, positron emitters are generated in a phantom by hard photon irradiation (energy above ˜20 MeV) due to (γ,n) reaction, leading to pair annihilation and subsequent emission of two coincident gammas. Möckel et al used a double head PET camera that moved linearly in the beam direction to detect these gammas and deduced a two-dimensional activity distribution. For proton beams this principle can not be used due to differences in the mechanisms for producing PET isotopes with proton beams when compared with carbon beams. Indeed the distribution in the target object of positron emitters produced by proton-induced nuclear reactions is not directly correlated with the spatial proton dose distribution in the target object. Moreover the cross sections for producing positron emitter PET isotopes with proton beams are much smaller when compared with the cross sections for producing the PET isotopes with carbon beams. As positron emitters generated by the particle beam—mainly 11C and 15O—have different half-life the delay between irradiation and the off-beam PET scanning can affect the determination of the dose deposition. Moreover, the leakage of positron emitters to the blood flow will also affect the dose deposition measurement by off-beam PET scanning. Another solution is disclosed in the document ‘Prompt gamma measurements for locating the dose falloff region in the proton therapy’, Chul-Hee Min and Chan Hyeong Kim, 2006 Applied Physics Letters, article 183517. Chul-Hee and Chan Hyeong Kim used a gamma scintillation camera equipped with one multilayered collimator system to measure prompt gamma generated by irradiation. Nevertheless, this device is only able to detect prompt gamma emitted from 90° of the beam direction. To obtain the prompt gamma distribution along the beam direction, the detector needs to be moved step by step to different measurement positions which makes this device not useful for practical on-line measurements. Chul-Hee Min et al in Journal of the Korean Physical Society, Vol 52, N° 3, March 2008, pp 888-891 disclosed a linear array of scintillation detectors and photodiodes for the online measurement of the proton beam range. One of the disadvantages of this device is the increased level of background as result of reduced collimator shielding. It is an object of the present invention to provide a device and a method for charged hadron therapy verification which overcomes the drawbacks of prior art detectors and methods. More particularly, it is an object of the present invention to provide an on-line detector, i.e. a detector which is capable of providing real-time measurements of the penetration depth or range of the charged hadron beam in an object or in a body irradiated by the charged hadron beam. Moreover, the spatial dose distribution in an object or in a body irradiated by a charged hadron beam can be determined. The present invention relates to a device for charged hadron therapy verification by detecting prompt gammas produced when irradiating an object or a body with a charged hadron beam, said device comprising a gamma-ray pin-hole camera arranged to acquire the number of said prompt gammas emitted while the said charged hadron beam is penetrating the said object or body. Preferably, the gamma-ray pin-hole camera comprises shielding means to avoid detection of unwanted particles. Preferably, the gamma-ray pin-hole camera comprises electronic means for data acquisition. Preferably, the device according to the present invention further comprises computing means connected to said electronic means enable to determine from the counted said prompt gammas a measured penetration depth or range of said charged hadron beam in said object or body. Preferably, said computing means enable to compare the said measured penetration depth with the theoretical or planned penetration depth. Preferably, said computing means connected to said electronic means enable to build an image representing the relative dose deposition. Preferably, said image is a two-dimensional or three-dimensional representation of the relative dose deposition. Preferably, the optical axis of the camera is perpendicular to the direction of the beam. Preferably, the inner diameter d of the pinhole is strictly superior to the value of the wavelength of the most energetic emitted prompt gammas. Preferably, the device according to the present invention further comprises electronic means to acquire the said number of said prompt gammas in synchrony with the time structure of said charged hadron beam. Preferably, the device according to the present invention comprises at least two camera for detecting prompt gammas. Another aspect of the present invention relates to a method for charged hadron therapy verification by detecting prompt gammas obtained by irradiating a phantom with a particle beam comprising the steps of: irradiating the phantom with a charged hadron beam; detecting the emitted prompt gammas during irradiation; deducing from the detected prompt gammas the range or penetration depth of the said charged hadron beam; measuring the range of the charged hadron beam in the phantom with a dedicated range measuring device; comparing the deduced range based on the prompt gamma detection with the range measured with the said dedicated range measuring device. Preferably, the method according to the present invention further comprises the step of: calculating the radiation dose given to the phantom based on the said detected prompt gammas. Preferably, the method according to the present invention further comprises before the irradiation step the step of: positioning a prompt gamma detector in a fixed position relatively to the phantom. Preferably, the prompt gamma detector used in the method according to the present invention is a gamma-ray pinhole camera. The present invention also relates to a hadron therapy device for charged hadron range verification by detecting and/or quantifying prompt gammas produced when irradiating an object or a body with a charged hadron beam, said device comprising a gamma-ray pin-hole camera arranged to acquire the number of said prompt gammas emitted while the said charged hadron beam is penetrating the said object or body. Preferably, said gamma-ray pin-hole camera comprises shielding means to avoid detection of particles different from prompt gammas. Preferably said gamma-ray pin-hole camera comprises electronic means for data acquisition. Preferably the hadron therapy device further comprises computing unit, connected to said electronic means, for determining—from the counted said prompt gammas—a measured penetration depth or range of said charged hadron beam in said object or body. Preferably said computing unit compares the said measured penetration depth with the theoretical or planned penetration depth. Preferably said computing unit, connected to said electronic means, builds an image representing the relative dose deposition. Preferably said image is a two-dimensional or three-dimensional representation of the relative dose deposition. Preferably the optical axis of the camera is perpendicular to the direction of the beam. Preferably the inner diameter d of the pinhole is strictly superior to the value of the wavelength of the most energetic emitted prompt gammas. Preferably the hadron therapy device further comprises electronic means to acquire the said number of said prompt gammas in synchrony with the time structure of said charged hadron beam. Another aspect of the present invention relates to a method for charged hadron range verification by detecting prompt gammas comprising the steps of: irradiating with a charged hadron beam; detecting the emitted prompt gammas; deducing from the detected prompt gammas the range or penetration depth of the said charged hadron beam; measuring the range of the charged hadron beam with a dedicated range measuring device; comparing the deduced range based on the prompt gamma detection with the range measured with the said dedicated range measuring device. The method according to the present invention further comprises the step of: calculating the radiation dose based on the said detected prompt gammas. The method according to the present invention further comprises before the irradiation step the step of: positioning a prompt gamma detector in a fixed position relatively to the treatment room. Preferably the prompt gamma detector used in the method is a gamma-ray pin-hole camera. The present invention relates to a device for charged hadron therapy verification by detecting prompt gammas when irradiating an object or a body with a charged hadron beam (e.g. proton beam, carbon beam), said device comprising counting means arranged to acquire the number of said prompt gammas emitted while the said charged hadron beam is penetrating the said object or body Preferably, said counting means comprise a camera comprising a plurality of collimators coupled to at least one scintillation crystal, said crystal being coupled to photomultiplier tubes and electronic means for data acquisition. More preferably, said counting means comprise a gamma-ray pin-hole camera comprising electronic means for data acquisition. Preferably, said counting means comprise shielding means to avoid detection of unwanted particles. Preferably, computing means connected to said electronic means enable to determine from the counted said prompt gammas a measured penetration depth or range of said charged hadron beam in said object or body. Furthermore, computing means enable to compare the said measured penetration depth with the theoretical or planned penetration depth. Preferably computing means connected to electronic means enable to build an image representing the relative dose deposition. Preferably said image is a two-dimensional or three-dimensional representation of the relative dose deposition. Preferably the optical axis of the camera is perpendicular to the direction of the beam. Moreover, the present invention relates to a method for charged hadron therapy verification by detecting prompt gammas obtained by irradiating a phantom with a particle beam comprising the steps of irradiating of a phantom with a charged hadron beam, detecting the emitted prompt gamma during irradiation, deducing from the detected prompt gammas the range or penetration depth of the said charged hadron beam, measuring the range of the charged hadron beam in the phantom with a dedicated range measuring device and comparing the deduced range based on the prompt gamma detection with the range measured with the said dedicated range measuring device. In addition, the dose delivered to the object can be calculated based on the measured prompt gammas. The present invention is based on the detection of prompt gamma obtained by irradiation of human body, animal body or any object such as a water phantom. Following irradiation with a charged hadron beam, such as a proton beam, a nucleus in an excited state can return to his ground state thanks to several nuclear reactions, several of them being depicted in FIG. 1. Among these reactions the emission of a prompt gamma can be observed directly after interaction with a proton from the beam, as shown in FIG. 1a. The correlation between the energy deposition of charged hadron beams and the distribution of prompt gammas was theorically assessed by Monte Carlo simulation using both PHITS (Proton and Heavy Ion Transport System) and MCNPX 2.5.0 (Monte Carlo N-Particle eXtended version 2.5.0) codes. Monte Carlo simulation results for a 230 MeV mono-energetic proton beam—irradiating a water phantom—are displayed at FIG. 2. The depth-dose distribution (FIG. 2a) displays the relative energy deposition or dose deposition in the water phantom as function of the penetration depth of the hadron beam in the water phantom. A maximum in the dose distribution, the so-called Bragg peak, is observed at a depth of about 32 cm in the water phantom. The 90% value of the dose distribution at the right side of the Bragg peak is called the penetration depth or range of the hadron beam in the object (e.g. water phantom, body). Other definitions of the range of the hadron beam exist (eg 80% value, . . . ). The depth-dose distribution of the proton beam (FIG. 2a) is tightly correlated to the prompt gamma yield (FIG. 2b). The prompt gamma yield (FIG. 2b) displays the relative number of prompt gammas emitted as function of the charged hadron penetration depth in the water phantom. A peak in the prompt gamma spectrum (FIG. 2b) is observed at about the same depth position as the Bragg peak position in the depth-dose distribution spectrum (FIG. 2a). The water phantom considered in these simulations is a 40 cm long cylinder with 20 cm diameter. Beside the emission of prompt gamma, other particles are emitted while irradiating human body, animal body or any object. In particular, as shown in FIG. 1b, fast neutrons are produced and constitute the major source of background affecting the prompt gamma counting signal. Simulated spectra of fast neutrons—for different proton beam energies—are displayed at FIG. 3, PHITS code being used for Monte Carlo simulation. Although fast neutrons are mainly forward oriented, they largely contribute to the deterioration of the signal after being scattered on the wall of the treatment room. An appropriate shielding of the detector is therefore required against neutrons. Shielding against other particles such as X-ray generated by bremmstrahlung is also required. Preferably, the device according to the present invention comprises at least two camera for detecting prompt gammas. Said cameras can either by identical or different. Preferably the prompt gammas counting occurs in synchrony with the time structure of the particle beam provided by the accelerator. In the present invention, the term “time structure” refers to the variation of the beam intensity as function of time. In most cases, said beam intensity varies in time and depends on the characteristics of the particle accelerator. Two examples of time structures will be described hereafter for two types of particle accelerators suitable for particle therapy. The time structure of a particle accelerator (e.g. cyclotron, synchrocyclotron, linear accelerator, . . . ) is determined by the Radiofrequency (RF) accelerator system. For example a 230 MeV proton cyclotron using a constant RF frequency of 100 MHz will deliver a beam pulse every 10 nsec (=beam pulse repetition period) and have a pulse width of typically 1 nsec. Another example of a particle accelerator is a 250 MeV synchrocyclotron using a time varying RF frequency whereby the frequency changes from high to low in order to take into account the effect of the relativistic mass increase of the particle. During the acceleration cycle of a particle in the synchrocyclotron, the RF frequency can for example vary between 86 MHz and 66 MHz. This acceleration cycle is then repeated for each beam pulse to be produced. The typical repetition frequency of the acceleration cycle (corresponding to the RF modulation frequency) is between 200 and 1000 Hz. As a result, the time structure of a particle beam produced with a synchrocyclotron has a macro-level and a micro-level time structure. The macro-level time structure corresponds to a macro-level pulse produced every acceleration cycle e.g. every 2 ms (i.e. for a RF modulation frequency of 500 Hz) and has a typical width of a few hundred nanoseconds (e.g. 0.3 microseconds). This macro-level pulse comprises a series of micro-level pulses where the frequency of the micro-level pulses corresponds to the RF acceleration frequency. For an RF frequency of 66 MHz, this means that a micro-level pulse is produced every 15.2 nanoseconds with a pulse width of typically 1.5 nanoseconds. For example, a macro-level pulse can comprise a series of 20 micro-level pulses. To improve the ratio between the prompt gammas and the background radiation one can advantageously make use of the time structure of a particle accelerator as described above. After interaction of the particle beam with a target (human body, animal body or any object such as a water phantom), the time it takes before a radiation event is detected by a said gamma camera will depend on the type of radiation event. The prompt gammas, travelling at the speed of light, will promptly generate a radiation invent in the gamma camera while the slower neutrons will generate a radiation event in the gamma camera at a later time. After production of prompt gammas (following the interaction of the particle beam with the target), the time window during which prompt gammas can be detected in the gamma camera is small (for example of the order of a few nanoseconds or less). The neutrons not only interact at a later moment in time with the gamma camera when compared to the fast prompt gammas but the time window during which neutrons (having a broad energy spectrum) can be detected in the gamma camera, after the beam has hit the target, is much larger and can be several tens of nanoseconds and more (depending on the target to gamma camera geometry). It is this difference in time of flight (TOF) between the prompt gammas and neutrons before they generate a radiation event in the gamma camera that can be used to enhance the prompt gamma signal intensity with respect to the background radiation intensity (neutrons and neutron induced radiation). The time structure of the particle accelerators discussed above producing for example pulses every 10 ns and having a pulse width of 1 ns are well suited to optimize the signal to background ratio of the prompt gammas detected with a gamma camera. By measuring the prompt gammas in synchrony with the RF accelerator frequency of the particle accelerator, prompt gammas only need to be acquired during a small time window after the beam has hit the target and prompt gammas were produced. The amount of background reduction will depend on the time window selected for analysing the prompt gammas with respect to the beam pulse repetition period. In the case of a cyclotron as mentioned above where the beam pulse repetition period is 10 nanoseconds and the pulse width is 1 nanosecond, the background radiation can be reduced by a maximum factor of 10 when applying for example a prompt gamma time window of 1 ns. For the case of a synchrocyclotron where for example 20 beam pulses having a duration of 1.5 nanoseconds are delivered for each acceleration cycle period of 2 milliseconds, the background level can be reduced by a maximum factor of 66666 when setting for example the prompt gamma window to 1.5 ns. Preferably the device according to the present invention comprises electronic means to measure the intensity of the prompt gammas in synchrony with the RF frequency of the particle accelerator. The electronic means for measuring the prompt gammas in synchrony with the RF frequency of the particle accelerator can be, for example, conventional electronics used in nuclear physics. For example timing NIM modules (Nuclear Instrumentation Modules, NIM) can be used. The basic NIM module that can be used for timing analysis is the TAC module (time to analogical convertor) which produces an output pulse with amplitude directly proportional to the time between a start input signal and a stop input signal. This TAC can be used to measure the time between the start of an RF pulse signal (start input) and the moment a radiation event is detected in the gamma camera (stop input). Depending on the geometry of the facility where the prompt gammas are measured in a treatment room, the RF signal can be delayed to take into account the time needed for the particles of the particle beam to travel from the accelerator to the treatment room. This delay can also be performed with a standard NIM delay module. The NIM electronics are linked with a data acquisition system and user interface. The basic data that are acquired for each radiation event in the gamma camera are its energy deposited and the characteristic time measured with the TAC (corresponding to a TOF measurement). All acquired radiation events can then for example be plotted as function of the time determined with the TAC, this is the time of flight spectrum. As discussed above due to the difference in time of flight (TOF) between prompt gammas and neutrons a discrimination between prompt gammas and neutrons can be made. By setting an additional window on the measured energy of the radiation events, the discrimination between neutrons and prompt gammas can be further increased. Alternatively, a non-intercepting particle beam detector which allows to detect the presence of a particle beam in the beam line can be used as a start signal for performing a TOF measurement as discussed above. Embodiment 1 In one preferred embodiment an Anger camera (10) is used to detect the prompt gammas emitted in a direction essentially at 90° with respect to the beam direction. The camera (10) is installed such that its optical axis (9) is perpendicular to the beam direction, in order to detect prompt gammas emitted from the object at 90°. The same principle can be used by putting the camera in a different geometry with respect to the beam direction and measuring prompt gammas at different angles. The Anger camera (10) is equipped with appropriate shielding against neutrons, X-ray and unwanted gamma-rays. As depicted in FIG. 4 the camera (10) comprises a plurality of collimators (1) that are made of high atomic number material such as Pb or W, on top of which at least one scintillating crystal (2) is laid. The at least one scintillating crystal is optically coupled to at least one photomultiplier tube (3), each of them being connected to appropriate readout electronics (4). Therefore, the said camera comprises means that enable counting of prompt gammas emitted by the irradiated body or by the irradiated object. Preferably, the photomultiplier tubes (3) are forming a linear array. More preferably, the photomultiplier tubes (3) are forming a two dimensional array. The appropriate readout electronics (4) are connected to computing means such as a PC (5). The collimator (1) comprises a set of thick sheets of high atomic material—usually 2.5 cm to 8 cm thick—that are all parallel to each other and perpendicular to the plane defined by the scintillation crystal(s). In this way, the camera comprises a plurality of collimators and can thus detect prompt gammas emitted from a whole body or object without performing any movement relatively to this body or object. The at least one scintillating crystal (2) is made of scintillation material such as sodium iodine with thallium doping. Each scintillating crystal is light-sealed avoiding signal cross-contamination between adjacent crystals. The shielding of neutrons is composed of two layers (6,7). The outer layer (6) is made of material that reduces the energy of fast neutrons, such as paraffin or high density polyethylene, whereas the inner layer (7) is made of material capturing the low energy neutrons by the (n,γ) neutron capture reaction, such as B4C powder, Cd layer or a plastic doped with Li or B. A third layer (8), e.g. Pb or W, allows to shield against unwanted photons such as photons generated by inelastic scattering reactions and photons generated by the said (n,γ) neutron capture reaction. The appropriate readout electronics (4) connected to computing means, such as a PC (5), enables to build a distribution of prompt gammas. This distribution is representative of the relative dose distribution in the irradiated object or body. The distribution is either a one dimensional (1-D) or a two dimensional (2-D) representation of the dose distribution in the object. Preferably, a second camera identical to the first one is used to detect prompt gammas and to build a second distribution of prompt gammas. The combination of these distributions enables to build a three dimensional (3-D) distribution of prompt gammas thanks to dedicated software running on computing means (5). Preferably, the 2-D and 3_D distribution are represented using 2-D or 3-D maps generated by imaging software running on computing means (5). Embodiment 2 In another preferred embodiment a pinhole camera is used to detect the prompt gammas emitted from the irradiated body or object. The camera is installed such that is optical axis (30) is perpendicular to the beam direction 50 in order to detect prompt gammas emitted from the object. The pinhole gamma camera is equipped with shielding against neutrons, X-ray and unwanted gamma-rays. As depicted in FIG. 5 the pinhole camera (20) is equipped with a pinhole collimator (21) made in a high atomic number material such as Pb or W. An advantage of using a pin-hole camera when compared with an Anger camera is that a pin-hole camera can be better shielded for background irradiation. A problem with the Anger camera is the fact that to obtain a high resolution when measuring the distribution of prompt gammas, the distance between the collimators is small (few mm, e.g. 2 or 3 mm) and hence the effective shielding is reduced. Preferably, the section following any plane comprising the optical axis (30) of the pinhole has a conical shape (22). The conical shape enables to detect prompt gammas that are not emitted from 90° of the beam direction. Therefore, the camera can detect prompt gammas 55 emitted from a whole body or an object without performing any movement relatively to this body or object. Preferably, the camera is motionless during the irradiation of the object or the body. Eventually, the camera is able to move during the irradiation of the object or the body. Preferably, the inner diameter d of the pinhole is strictly superior to the value of the wavelength of the most energetic emitted prompt gammas. Below this value diffraction of photons occurs. The pinhole camera comprises at least one scintillating crystal (23) optically coupled to at least one photomultiplier tube (24), each of them being connected to a appropriate readout electronics (25). Therefore, the said camera comprises means that enable counting of prompt gamma emitted by the irradiated body or by the irradiated object. Preferably, the photomultiplier tubes (24) are forming a linear array. More preferably, the photomultiplier tubes (24) are forming a two dimensional array. The at least one scintillating crystal (23) is made of scintillation material such as sodium iodine with thalium doping. Preferably, each scintillating crystal is light-sealed avoiding signal cross-contamination between adjacent crystals. Preferably, the shielding of neutrons is composed of two layers (27,28). The outer layer (27) is made of material that reduces the energy of fast neutrons such as paraffin or high density polyethylene, whereas the inner layer (28) is made of material capturing the neutrons by the (n,γ) reaction, such as B4C powder, Cd layer or a plastic compound doped with Li or B. Preferably, a third layer (29), e.g. Pb or W, allows to shield against unwanted photons such as photons generated by inelastic scattering reactions, photon generated by the said (n,γ) neutron capture reaction (cf. infra) and prompt gamma produced at large angle. We mean by prompt gamma produced at large angle prompt gamma that go through the pinhole with an angle—relatively to the optical axis (30)—which exceed the value given by the formula:β=arctg(R/(2f))Where f is the focal length, i.e. the distance between the pinhole and the surface of the scintillation material, and where R is the longitudinal length of the scintillation material. The appropriate readout electronics (25) linked to computing means (26), such as PC, enables to build a two dimensional (2D) distribution of prompt gammas. The 2D distribution is representative of the relative dose distribution in the irradiated object or body. Preferably, the pinhole camera is composed of a set of pinhole cameras (minimum two) installed around the beam direction. The appropriate readout electronics (4) connected to computing means, such as a PC (5), enables to build a distribution of prompt gammas. This distribution is representative of the relative dose distribution in the irradiated object or body. The distribution is either a one dimensional (1-D) or a two dimensional (2-D) representation of the dose distribution in the object. Preferably, a second camera identical to the first one is used to detect prompt gammas and to build a second distribution of prompt gammas. Alternatively, a second camera different of the first one is used to detect prompt gammas and to build a second distribution of prompt gammas. Advantageously, the optical axis of the two cameras are orthogonal. Each pinhole camera detects prompt gammas and builds a 2D distribution of prompt gammas. The combination of these 2D distributions enables to build a three dimensional (3-D) distribution of prompt gamma thanks to dedicated software. The combination of these distributions enables to build a three dimensional (3-D) distribution of prompt gammas thanks to dedicated software running on computing means (5). Preferably, the 2-D and 3_D distribution are represented using 2-D or 3-D maps generated by imaging software running on computing means (5). The device according to the invention can be verified and/or calibrated by performing a comparison measurement in a phantom (e.g. water phantom). The deduced range based on the prompt gamma detection can be compared with the range obtained by using a dedicated range measuring device (e.g. ionization chamber, diode, . . . ). Another aspect of the present invention relates to a method for verifying and/or calibrating the device of the present invention comprises the steps of: irradiating a phantom with a charged hadron beam; detecting the emitted prompt gammas during irradiation; deducing from the detected prompt gammas the range or penetration depth of the said charged hadron beam; measuring the range of the charged hadron beam in the phantom with a dedicated range measuring device; comparing the deduced range based on the prompt gamma detection with the range measured with the said dedicated range measuring device. The device according to the present invention can further be used in a method comprising the steps of: irradiating a patient with a charged hadron beam; detecting the emitted prompt gammas during patient irradiation, i.e. while the said hadron beam is penetrating the patient's body; deducing from the detected prompt gammas the range or penetration depth of the said charged hadron beam. In addition, the device according to the invention can also be used in a method comprising the steps of: detecting prompt gammas emitted during patient irradiation with a hadron beam; deducing from the detected prompt gammas the dose distribution in the patient resulting from said irradiation. In addition, the device according to the invention can also be used in a method comprising the steps of: detecting prompt gammas emitted during patient irradiation with a hadron beam; calculating the radiation dose given to the patient based on the detected prompt gammas.
051529572	description	DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS The embodiments of the present invention will be explained below with reference to the drawings. In FIG. 1, reference numeral 1 shows a foreign matter recovering apparatus (hereinafter referred to as a recovering apparatus) which, in a nuclear power generation equipment or plant, recovers screws, metal pieces etc., trapped or caught in a fuel assembly, as well as foreign matter, such as fur and dust, deposited on the fuel elements in the fuel assembly. The recovering apparatus 1 comprises a working unit (body) 3 fixed to the lower end of a pipe-like support pole 2 of a circular cross-section extending in a straight way and sunken in the body of the water, and a remote control section 5 having a controller 4 arranged over the water at a predetermined location. The controller 4 is located outside, for example, a fuel storage pool and adapted to control the working unit 3. The working unit 3 is sunken in the body of the water through the operation of the support pole 2 and brought down to a water depth of, for example, about 10 m where it is moved closer to the fuel assembly (not shown). As shown in FIG. 2, the working unit 3 has a rectangular table 6 as indicated by a dash dot line and guide shafts 7 and 8, upper and lower, mounted below the fixed table 6 and extending in a parallel array. In the working unit 3, a plurality of guide fins 9, are provided, as a body fixing section, in front of the working unit 3 and arranged, for example, at a predetermined interval, in a parallel array in a direction of the width of the working unit 3, that is, in a direction of an arrow X in FIG. 2. The guide fins 9, . . . extend from the working unit 3 and are inserted in the parallel array into clearances created among respective parallel bar-like fuel elements in the fuel assembly. With the front edges of the fixed table placed in contact with the fuel assembly, the guide fins 9, . . . prevent any position error or displacement of the working unit 3 in the X-direction and fix the working unit 3 in place relative to the fuel assembly. Here it may be considered that those portions in contact with the fuel elements, such as front edges and guide fins 9, . . . , of the fixed table 6 are made of, for example, a radiation-resistant resin. A movable table 10 is mounted on the working unit 3 and serves as a movable mechanism section comprised of a rectangular plate. The movable table 10 depends from the lower side of the fixed table 6 with its longitudinal section oriented in a front/back direction. The movable table 10 is provided midway across the width of the fixed table 6. The movable table 10 is slidably movable in the front/back direction of the working unit 3, that is, in a Y-direction in FIG. 2. Stated another way, the movable table 10 has a rack 11 mounted integral with its one surface as shown in FIGS. 2 and 3 and extending along its longitudinal direction. A pinion 13 of a motor 12 which is fixed on the upper side of the fixed table 6 and has its rotation shaft extending down on the lower surface side of the fixed table 6 is engaged with the rack 11 on the movable table 10. A rotational force of the motor 12 is transmitted to the movable table 10 by the rack and pinion 13 to enable the movable table 10 to be moved back and forth in the Y direction while being guided by the guide shafts 7 and 8. The movable table 10 can be stopped in any desired position during its movement in the Y direction. The motor 12 is covered by a motor casing 12a in an airtight fashion to prevent the penetration of water As the motor 12, used is made of, for example, a servo motor whereby it is possible to properly control an amount of displacement. A swingable plate 1 is coupled to the movable table 10 in a manner to face the lower edge portion of the movable table 10, the intermediate portion of the swingable plate 14 being swingably supported on the table 10. The swingable plate 14 has its upper end face facing a submergible air cylinder 15 located on the rear portion of the movable table 10 and fixed there. A coil spring 16 is anchored at one end to the movable table 10 and at the other end to the rear end portion of the swingable plate 14. The swingable plate 14 has its rear end portion depressed down by a piston 17 extending from the cylinder 15 and is swung in a direction of an arrow .theta. in FIG. 2, while extending and deforming the coil spring 16, to tilt its forward end upward. With the piston 17 retracted back and swingable plate 14 released from the depression force of the submergible air cylinder 15, the swingable plate 14 is returned back, for example, to tilt its forward end downward under the urging face of the coil spring 16. A recovery probe (hereinafter referred to as a probe) 18 is located at the forward end portion of the swingable plate 14 and serves as a recovering working area. The probe 18 has, for example, a needle-like tip and is, for example, below 2 mm in external diameter. The probe 18 has its base portion fixed to the forward end portion of the swingable plate 14. The probe 18 extends along the longitudinal direction of the swingable plate 14 and extends from the forward end portion of the swingable plate 14 with its tip oriented in a predetermined state relative to the swingable plate 14. The probe 18, together with the swingable plate 14, is moved as one unit and has its tip moved up and down by a predetermined amount in accordance with the extending and retraction of the piston 17. The probe 18, together with the movable table 10, is moved, as one unit, through the swingable plate 14 to allow the probe tip to be penetrated into clearances among the fuel elements. The probe 18, together with the movable table 10, is moved as one unit to orient it in the Y direction. Further, the probe 18, together with the movable plate 14, is swung as one unit to allow its direction to be oriented in the .theta. direction. It may be considered that, in cooperation with the motor 12 and cylinder 15, the probe 18 is driven to scrape off fur, dust, etc., deposited on the fuel elements so that they are recovered from there. A probe swinging mechanism 14a comprises the swingable plate 14, air cylinder 15, coil spring 16, etc., and serves as a recovering working section swinging mechanism. It may be considered that, as the probe 18, use is made of, for example, a vacuum probe for vacuum-sucking a foreign matter and a magnetic probe for attracting a foreign matter. For the case of the vacuum probe, a suction nozzle 19 is employed as the probe as shown in FIG. 4. It may be considered that a submergible pump 20, which is placed in the water for a nuclear reactor, sucks foreign matter through the suction nozzle 19 and recovers it by a recovery case 22 provided between the suction nozzle 19 and the submergible pump 20 and equipped with a filter 21. In FIG. 2, reference numeral 23 shows a fiberscope. The fiberscope 23 has a distal end portion with, for example, an array of imaging and illumination tubes of, for example, below 2 mm in diameter. The distal end portion 24 is fixed to the movable table 10. The fiberscope 23 is moved into proximity to the tip of the probe 18 with the distal end 24a oriented toward the probe's tip so that the tip of the probe 18 is covered within a viewing field of the fiberscope 23. The fiberscope 23 is so tilted as to have an optical end face of the imaging and illumination tubes, etc., oriented toward the axis of the fiberscope. The viewing field of the imaging fiber is illuminated. An image is taken by an optical element, such as a prism, etc., and transmitted to a CCD (charge coupled device) provided on the side of, for example, a remote control section. The fiberscope 23 has its distal end portion 24 moved, together with the movable table 10, as one unit in a manner to follow the motion of the probe 18. In this way, it is possible to photograph the tip of the probe 18, while the distal end portion of the fiberscope is thus moved. The fiberscope 23 enables the manner of sucking or magnetically attracting, for example, a metal piece or scraping off fur, dust, etc., deposited on the fuel elements to be viewed through an eyepiece 26 connected to the distal end portion 24 via a flexible cable 25 extending between the body of the water and the surface of the water and through a monitor 26a connected to the eyepiece 26 as shown in FIG. 1. Here, the image of the fiberscope 23 may be displayed directly on the monitor 26a. The remote control section 5 comprises a controller 4, eyepiece 26 disposed near the controller 4, and monitor 26a. The remote control section 5 is of such a type that joysticks 27, 27, etc., are provided on an operation panel 4a of the controller 4 connected to the working unit 3 through a plurality of flexible cables 28 . . . . The remote control section 5, for example, enables the motor 12 and cylinder 15 to be driven independently by operation of the controller 4 while viewing an image displayed on the monitor 26a. The recovering apparatus 1 is operated as will be set forth below. First, the support pole 2 is operated to allow the working unit 3 to be submerged in the body of the water. The guide fins 9, . . . are penetrated into the clearances among the fuel elements at a water depth of, for example, about 10 m where the working unit 3 is located relatively to the fuel assembly and fixed there. The controller 4 is operated at a location over the water surface distant from the fuel assembly, causing the motor 12 to be driven and hence the probe 18 and fiberscope 23 to be penetrated into the clearances among the fuel elements. The foreign matter and probe's tip are covered by the fiberscope 23 as one image and the controller 4 is operated while viewing that image on the eyepiece 26. For example, the motor 12 and submergible air cylinder 15 are properly and simultaneously driven, enabling the probe 18 to be properly located for the foreign matter to be recovered. In the recovering apparatus 1, the working unit 3 with the probe 18 and fiberscope 23 attached thereto is brought into proximity to the fuel assembly. The working state of the working unit 3 is photographed by the fiberscope 23 and the probe's tip is penetrated into small clearances among the fuel elements and recovers foreign matter there. For this reason, any small foreign matter present in a specific environment inaccessible by a worker or by hand can be recovered at the distant location while being visually examined. Since the motor 12 positioning the probe 18 and fiberscope 23, as well as the submergible air cylinder 15 etc., is mounted on the working unit 3, it is possible to efficiently, positively and very accurately accomplish the positioning of the probe 18 and fiberscope 23. Since the positioning of the probe 18 and fiberscope 23 in the Y direction is effected by the motor 12 and the positioning of the probe 18 in the .theta. direction is effected by the air cylinder 15, the probe 18 is positioned only in one direction by the controlled operation of the motor and can readily be moved closer to foreign matter. The probe 18 and fiberscope 23 can be moved toward a spot in the fuel assembly. In the case where, as the probe 18, use is made of a vacuum probe or magnetic probe, since the probe needs not be placed in contact with the foreign matter, any foreign matter can be positively recovered even if positioning of the probe is setting only in the Y direction. It is possible to employ an ultrasonic probe as the probe 18. As the probe 18 is positioned only in one direction by the controlled operation of the motor, only one motor is required, making the working unit 3 compact and light in weight. It is possible to simplify the arrangement of the controller 4. After the guide fins 9, . . . are retracted back from the fuel assembly, the working unit 3 can be moved in the X direction. In order to identify the presence of fur, etc., it may be possible to utilize the difference between it and the rest of the fuel element through the utilization of the state in which light is reflected on the surface of the fuel element. The contrast in light between them cannot exactly identify any fur on the fuel element, but the presence and position of any fur can be exactly identified by bringing the probe into contact with the fur or scraping it off the surface of the fuel element. Further it is possible for diagnostically making contact the fuel element and to thus to exactly known the state of the fuel assembly. It is also possible to jet a stream of water by the probe, or direct air bubbles by the probe, at the surface of the fuel elements so that any foreign matter can be eliminated or recovered. The foreign matter, once being eliminated, may again be deposited on the surface of the fuel element, but, if any means is provided for eliminating it from the surface of the fuel element, it can be recovered by that means to prevent redeposition of the foreign matter on the fuel element. The fiberscope image, depending upon the image resolution of the CCD, etc., can be displayed as a high accurate image on the remote control section because the fiberscope 23 and probe 18 are brought into close proximity to the foreign matter on the fuel element. The motor 12 is of such a type as to allow any proper displacement control operation and it is possible to accurately drive, displace and position the probe 18 and fiberscope 23. The recovering apparatus 1 can check and examine the fuel assembly. The working unit 3 can be positioned in a gravity direction by suspending it by means of a winch, etc. Although, in the aforementioned embodiment, a combination of the rack 11 and pinion 13 is utilized in the transmission of power to the movable table 10, the present invention is not restricted to it. For example, a trapezoidal screw thread can be utilized instead. The swingable plate 14 can be driven by utilizing a motor, in which case it is possible to impart an enhanced operability to the swingable plate 14. Incorporating the working unit 3 into a remotely controlled submergible craft eliminates the need to provide the support pole 2, etc. FIGS. 5 to 7 show a major area of a recovering apparatus according to a second embodiment of the present invention. The same reference numerals are employed to designate parts or elements corresponding to those described in conjunction with the preceding embodiment. Further explanation of them is, therefore, omitted for the sake of brevity. In the arrangement shown in FIG. 5, reference numeral 2 shows a support pole whose upper end portion is exposed over the surface of the water and 3, a working unit submerged beneath the surface of the water and fixed to the lower end of the support pole 2. The working unit 3 is fixed in place relative to the fuel assembly by guide fins 9, . . . provided at the front of the working unit and located in a lateral direction, that is, in a direction of X in FIG. 5. The working unit 3 includes a probe as a recovering working section, not shown, and fiberscope 23 for normally covering the distal end of the probe within a visual field. A TV camera and monitor are connected to the proximal end, not shown, of the fiberscope 23. The working unit 3 slidably moves a movable table 10 back and forth by a motor held in a motor casing 1a to allow the probe and distal end 24 of the fiberscope 23 to be penetrated into the clearances among the fuel elements in a fuel assembly. The working unit 3 is of such a type that the probe and fiberscope's distal end are controlled by operating a controller while being viewed on the monitor, not shown, to allow the recovery of foreign matter present in the clearances among, or on the surfaces, of the fuel elements. An attitude control means 31 is provided on the working unit 3 and comprised of four contact sensors 32a to 32d provided in front of the working unit 3 and four water stream generation sections 33a to 33d provided in back of the working unit 3 to jet water streams backward. The attitude control means 31 is of such a type that the contact sensors 32a to 32d and water stream generation sections 33a to 33d are provided ahead and behind it, respectively, at four corners for example. Upon the contact of the working unit's front end with the fuel assembly during the fixing of the working unit 3 to the fuel assembly, the attitude control means 31 can detect the tilt of the working unit 3 relative to the fuel assembly on the basis of the ON and OFF states of the contact sensors, that is, the relative position of those contact sensors turned ON upon contact with the fuel assembly and those contact sensors turned OFF upon separation with the fuel assembly. In order to bring those contact sensors spaced from the fuel assembly, into contact with the fuel assembly the attitude control means 31 selects the water stream generation sections corresponding to the OFF contact sensors so that they are driven toward the fuel assembly. The attitude control means 31 jets a water stream backward, varying the tilt of the working unit through the utilization of a thrust force originating from some of the water stream generation sections 33a to 33d or a difference in propulsion force between the water stream generation sections 33a to 33d. It is thus possible to control the attitude of the working unit 3. Stated in another way, in the recovering apparatus 36 including the attitude control means 31 the tilt of the working unit 3 relative to the fuel assembly is corrected by the attitude control means, automatically controlling the attitude of the working unit 3. For this reason, the working unit 3 accommodates its gravity variation and hence can be made lighter in weight. It is also possible to prevent the fuel elements from being injured by the working unit 3. It may be considered that, upon the movement of the movable table, etc., the gravity of the working unit 3 is shifted, that is, the working unit 3 is displaced into a tilted position to make the positioning of the working unit 3 relative to the fuel assembly difficult. In order to position the working unit 3 relative to the fuel assembly through the controlling of the working unit's tilt, an extensive study needs to be made on the connection position between the support pole 2 and the working unit so as to meet a shifting gravity of the working unit 3. It is also necessary, during an initial phase of positioning, to add an additional weight to the working unit 3 so that a balance can be obtained. In the case where the working unit 3 is tilted, it is difficult to uniformly push the working unit 3 against the fuel assembly by drawing the support pole 2 toward the operator. By a strong push, the fuel assembly is liable to be injured. The working unit 3 is liable to be tilted by a buoyant force of the motor casing 12a. The attitude control means 31, provided on the working unit 3, accommodates a shift of gravity automatically and prevents a tilt, thus maintaining the attitude of the working unit 3 constant. It is not necessary to add a weight to the working unit 3 during an initial phase of positioning, so that the working unit 2 can be made lighter in weight. The working unit 3 can be uniformly pressed with a given pressure against the fuel assembly by turning on all the contact sensors 32a to 32d, for example, and making the thrust forces as set out above equal and constant. The working unit 3 can be prevented from being pressed with an excessive and nonuniform pressure and being injured. It may be possible to use touch sensors, pressure sensors, etc., as contact sensors for the attitude sensor. FIG. 8 shows a major area of a third embodiment of the present invention. The same reference numerals are employed to designate parts or elements corresponding to those set out in connection with the previous embodiment. Further explanation of them is omitted for the sake of brevity. Although, in the second embodiment, a plurality of water stream generation sections are provided as a water jet type, one screw propeller 34 is provided as a water stream generation section in the recovering apparatus 37 according to the third embodiment of the present invention. The attitude of the working unit 3 is controlled by the crew propeller 34 and rudder 35. In FIG. 8, the working unit 3 has the screw propeller 34 at a substantially middle of the rear side and the rudder 35 located behind the screw propeller 34 and swingable up and down. The working unit 3 has its front area pushed against the fuel assembly by the thrust force of the screw propeller 34 and has its up and down tilt controlled by the rudder 35 whose direction varies based on an output signal of the touch sensor. The working unit 3 has its X-direction attitude held substantially constant by the guide fins 9, . . . and has its attitude controlled by the rudder 35 in one direction, that is, in a direction of Z. FIGS. 9 and 10 show a fourth embodiment of the present invention. The same reference numerals are employed in this embodiment to designate parts or elements corresponding to those shown in the respective previous embodiments. Further explanation of them i omitted for simplicity. In FIG. 9, reference numeral 41 shows a foreign matter recovering apparatus (hereinafter referred to a recovering apparatus) for a nuclear power generation plant or equipment, which can recover, for example, screws or metal pieces trapped in the fuel assembly as well as any fur, dust, etc., deposited on the fuel elements in the fuel assembly. The recovering apparatus 41 comprises a working unit 43 submerged beneath the surface of the water and fixed to a support pole 42 extending in a straight fashion and a remote control section 44 located on the water at a given position and outside, for example, a nuclear reactor. The working unit 43 is sunken beneath the surface of the water through the operation of the support pole 42 and brought down to the water depth of, about 10 m where it is moved closer to a fuel assembly, not shown. As shown in FIG. 10, the working unit 43 has a rectangular casing 45 as indicated by a dash-dot line and equipped with a body fixing section 46 (hereinafter referred to simply as a fixing section) projecting from the casing 45. The fixing section 46 includes plate-like guides 47, 47 passing near a fuel assembly in a manner to be interposed therebetween and four clamps 48, . . . provided outside the guides 47, 47 and having parallel arrays of grooves inside the clamps to correspond to the location of the fuel elements. The fixing section 46, being so interposed, clamps the fuel assembly by the clamps 48, . . . with the fuel elements of the fuel assembly fitted in their grooves so that the casing 45 is held in place relative to the fuel assembly. The working unit 43 includes forceps 49 as a recovering operation section. The forceps 49 is mounted on the distal end of a flexible wire 50 comprised of inner and outer wires and extending out from the interior of the casing 45 and is placed, for example, near the base ends of the clamps 48, . . . . The forceps 49 is opened and closed by, for example, the inner wire of the flexible wire 50 connected at its base end to a link 53 engaged with a cam 52 which is rotated by a forceps drive motor 51. The forceps 49 has its flexible wire 50 connected partway to a forceps rotation mechanism section, as shown by 54 in FIG. 10, and to a positioning mechanism section 55 as will be described later. The forceps 49 is rotated by a rotational force transmitted by a forceps rotation motor 56 in the forceps rotation mechanism section 54 through a flexible shaft 57, helical-type gear portion 58, etc. In FIG. 10, reference numeral 59 shows a fiberscope whose distal end provides a linear array of imaging and illumination areas, etc. The fiberscope 59 has its distal end 60 moved nearer and faced to the forceps 49 so that it covers the forceps 49 within its viewing field. The distal end 60 of the fiberscope 59 is fixed to the positioning mechanism section 55. The fiberscope 59 has an optical end face of the imaging and illumination areas oriented obliquely relative to its axis and the field of vision is illuminated by the illumination area. An image is taken through an optical element means, such as a prism, and transmitted to a CCD (charge coupled device) provided, for example, on the remote control section side. The fiberscope 59 has its distal end coupled to a fiberscope rotation mechanism section as shown by 61. The fiberscope 59 has its distal end 60 rotated around a direction of Y in FIG. 10, that is, in the right/left direction with its base end as a center. This can be done by transmitting a rotational force of a forceps rotation motor 62 which is arranged in juxtaposition with the forceps rotation motor 51 to the distal end of the fiberscope through a flexible shaft 63, gear 64, etc. The positioning mechanism section 55 has first to third feed motors 65, 66 and 67 and first to third movable bodies 68, 69 and 70 linearly movable by the drive forces of the feed motors 65, 66 and 67 through their rack-and-pinion mechanisms. The positioning mechanism section 55 has the flexible wire 50 of the forceps, as well as the distal end of the fiberscope 59, coupled to the third movable body 70 formed of a platelike element. The positioning mechanism 55 is of such a type that the movable bodies 68 to 70 and the two feed motors 66 and 67 are moved by the first feed motor 65 in the right/left direction, that is, in the X direction in FIG. 10, that the second and third movable bodies 69 and 70 and third feed motor 67 are moved by the second motor 66 in a front/back direction of the casing 45, that is, in the Y direction in FIG. 10, and that the third movable body 70 is moved by the third feed motor 67 in an up/down direction of the casing 45, that is, in the Z direction in FIG. 10. By selectively driving the first to third feed motors 65 to 67, he positioning mechanism section 55 allows the distal end portions of the forceps and fiberscope 59 which are provided integral with the third movable body 70 to be moved to any proper position among the fuel elements, while moving the distal end 50 in a manner to follow the forceps 49. By so doing, the forceps 49 is positioned, while normally covering the forceps within a visual field of the fiberscope 59. In this embodiment, as the first to third feed motors 65 to 67 and forceps rotation motors 51 and 62, use is made of those motors which can properly control a displacement, such as a servo motor. The remote control section 44 comprises, as shown in FIG. 9, an operation panel 71, controller (control section) 72 monitor (display section) 73 and connects a controller 72 to an operation panel 71 and a monitor 73 to the fiberscope 59. Further, the remote control section 44 connects the controller 72 to a connector box 74 for gathering conductive wires of the respective motors of the working unit 43 beneath the surface of the water. The remote control section 44 enables a image which is picked up by the fiberscope 59 to be sent to the monitor 73 where it is possible to display the state in which the forceps 49 penetrates the clearances among the fuel elements and recovers foreign matter trapped there. In the remote control section 44, a plurality of joysticks 75, . . . are mounted upright on the operation panel 71 and some of them are operated by an operation while viewing an image on the monitor 73, during which time an instruction signal is sent by the remote control section from the operation panel 71 to the controller 72. The remote control section 44 controls the respective associated motors, individually, in the submerged working unit by means of the controller 72 to allow the forceps 49 and fiberscope 59 to be positioned relative to each other, driven, etc. Stated another way, the recovering apparatus 41 is so operated that the forceps 49 and fiberscope 59 in the working unit 43 are moved into proximity to the fuel assembly and then into clearances among the fuel elements, while viewing the state of working there, to achieve recovery of a foreign matter. By so doing, it is possible to recover foreign matter, while visually locating it at a remote site, in a specific environment not inaccessible by an operator or operator's hand. In the working unit 43, the fixing section 46 is clamped to the fuel assembly, coarsely locating the working unit 43 relative to the fuel assembly and then finely locating the forceps 49 and fiberscope 5 under the operation of the positioning mechanism section 55. It is thus possible to position the forceps 49 and fiberscope 59 with high accuracy. In the movement of the forceps 49 from one clearance to another, the first to third feed motors 65 to 67 are respectively driven with the casing 45 fixed, so that necessary associated parts, that is, the forceps 49 and fiberscope 59 alone can be moved into a site of interest in the fuel assembly. It is, therefore, possible to readily move the forceps 49 from one clearance to another in the fuel assembly. It is possible to very exactly locate the forceps 49 and fiberscope 59 in any proper site of interest and limited part in the fuel assembly. The foreign matter thus located ca be exactly eliminated using the forceps 49 provided in the working unit. The re-deposition of the once-eliminated foreign matter can be prevented by providing a means for sucking or magnetically attracting it for recovery. Further, it is also possible to diagnostically contact with the fuel assembly for the presence of any foreign matter and to more exactly know the state of the fuel assembly. The foreign matter can be caught directly by the forceps and recovered positively to a proper position, for example, and it is possible to remove the foreign matter trapped or caught, for example, at the spacer of the fuel assembly 201 or o scrape off a fur deposited on the fuel element by means of the forceps 49. Further, for the motors 51, 56 and 62 as well as the motors 65, 66 and 67 use is made of those motors for enabling any proper displacement control, so that forceps 49 and fiberscope 59 can exactly be driven, displaced and positioned In this embodiment, although the fixing section 46 is mounted integral with the working unit 43, the working unit 43 may be displaced relative to the fixing section 46 instead and self-propelled along the fuel assembly with the fuel assembly held in place by the fixing unit 46. The motors, etc., of the working unit 43 can be properly eliminated in a selective way in accordance with the freedom with which the forceps 49 needs to be moved. Further, the working unit 43, being built into a remotely controlled submergible craft, eliminates the need for providing, for example, a support pole 42. FIGS. 11 and 12 show a major section of a recovering apparatus according to a fifth embodiment of the present invention. The same reference numerals are employed in FIGS. 11 and 12 to designate parts or elements corresponding to those shown in the preceding embodiment of the present invention. Any further explanation of them is, therefore, omitted. In FIG. 11, reference numeral 81 shows a foreign matter recovering apparatus (hereinafter referred to as recovering apparatus) for eliminating a foreign matter deposited or trapped in a fuel assembly for nuclear power generation and for recovering it. This apparatus includes, as referring FIG. 9, a working unit 43 fixed to the lower end of an upwardly extending straight support rod 42 and submerged as a body underneath the surface of the water and a remote control section 44 placed on the surface of the water at a predetermined position such that it is located outside a nuclear reaction. The working unit 43 includes a positioning mechanism section 55 as shown in FIG. 11. The positioning mechanism section 55 includes firs to third feed motors 65, 66 and 67 and first to third movable bodies 68, 69 and 70 adapted to be linear moved through their rack-to-pinion mechanisms upon receipt of the rotational forces of the three feed motors 65, 66 and 67. The positioning mechanism section 55 is of such a type that a flexible wire 50 of a forceps 49, as well as the distal end 60 of a fiberscope 59, is coupled to the third movable body 70 made up of a plate-like member. The positioning mechanism section 55 is arranged such that the third movable body 70 is coupled to the second movable body 69 and that the second and third movable bodies 69 and 70 are coupled to the first movable body 68. In the positioning mechanism section 55, the movable bodies 68 to 70 and feed motors 66 and 67 are moved by the first feed motor 65, as one unit, in a right/left direction of the casing 45, that is, in the X-direction in FIG. 11; the second and third movable bodies 69 and 70 and third feed motor 67 are moved in a front/back direction of the casing 45, that is, in the Y direction in FIG. 11; and the third movable body 70 is moved by the third feed motor 67 in an up/down direction of the casing 45, that is, in the Z direction in FIG. 11. The positioning mechanism section 49 is so constructed that a forceps 49 and fiberscope 59 are moved by the first feed motor 65 in the X direction and by the second feed motor 66 in the Y direction and that the forceps 49 and fiberscope 59 are moved by the third feed motor 67 in the Z direction in FIG. 11 In the positioning mechanism section 55, the rotation of the feed motors in a positive direction results in the movement of the forceps 49 and fiberscope 59 in the X-Y-Z (positive) directions and the rotation of the feed motors 65 to 67 in a reverse direction results in the movement of the forceps 49 and fiberscope 59 in the X-Y-Z (negative) positions. The selective rotation of the first to third feed motors 65 to 67 in the positioning mechanism section 55 allows the forceps 49 and distal end 60 of the fiberscope 59 to be moved to any proper site of interest among fuel elements in accordance with the forceps 49. The forceps 49 is positioned, while normally covering it within a visual field of the fiberscope 59. An interlocking mechanism section 76 is provided as a movement restriction section at the working unit 43. The interlocking mechanism section 76 comprises a guide pin 77 projected as an engaging projection from the second movable body 69 and a guide plate 78 provided as a comb-like guide member on the second movable body 69. The guide pin 77 is cylindrical in configuration and provided integral with the upper surface of the second movable unit 69. The guide pin 77, together with the second movable body 69, is moved, as one unit, in the X and Y directions, following the movement of the forceps 49 and fiberscope 59. The guide plate 78 is made up of a plate-like member. The guide plate is opened at one side and has a plurality of slits 79, . . . as a multi-U shaped array with each U-shaped slit closed at the other side. The guide plate 78 is so constructed that the slits 79, . . . are arranged at substantially the same pitch as those of the fuel elements in the fuel assembly. The guide plate 78 is fixed, for example, at a predetermined position to the inside of the casing 45 such that it is located close to the second movable body 69. The guide plate 78 has its slit array extending in the X direction and the longitudinal slit portions extending in the Y direction. The guide plate has its closed side oriented toward a fixing section 46 side. The guide plate 78 is provided with the fixing section 46 as a reference. As shown schematically in FIG. 12, when clamps 48, . . . of the fixing section 46 hold a fuel assembly 201 therebetween to set the casing 45 fixed, the guide plate 78 is so set that its slits 79, . . . are arranged in the Y direction in a manner to correspond to small clearances (for example, 2 to 3 mm) 203, . . . among fuel elements 202, . . . arranged at substantially the same pitch. When the forceps 49 is displaced in the Y direction from a proper front-facing position relative to one (203a) of the clearances 203, the interlocking mechanism section 76 has its guide pin 77 guided into a slit 79a corresponding to the clearance 203a, that is, it moves the guide pin 77 along the slit 79a to allow the forceps 49 to be moved in a positive Y direction. The interlocking mechanism section 76 restricts a range of the movement of the forceps 49 in the clearance 203a to one corresponding to the size of the slit 79a. When the forceps 49 is moved in the positive Y direction from an improper position displaced relative to the clearance 203a, the interlocking mechanism section allows the guide pin 77 to contact with a slit-to-slit area and stops the second movable body 69 to prevent the forceps 49 from being moved in the positive Y direction. The guide plate has a rigidity great enough to latchingly engage the guide pin 77 there. The fuel elements 202, . . . are schematically shown in FIG. 12. The recovering apparatus 81, being equipped with the interlocking mechanism section 76, can readily position the forceps 49 relative to the clearances 203, . . . , among the fuel elements 202, . . . , preventing the forceps 49 from moving toward the fuel assembly 201, while being displaced relative to the clearances 203, . . . , among the fuel elements, and moving into contact with the fuel elements 202, . . . . It is, thus, possible to exactly insert the forceps 49 into the fuel assembly 201 at all times. Further, even after the forceps 49 has been inserted into the clearance as set out above, the range over which the forceps 49 can be moved can be restricted, preventing the forceps 49 from contacting with the fuel elements 202, . . . . It is also possible to seek for the clearances 203, among the fuel elements 202, . . . and to position the forceps 49 relative to the clearance without using the fiberscope 59. Further, the fiberscope 59 can be prevented from being bent or injured upon abutting against the fuel elements 202, . . . , thus keeping the fuels 202, . . . from damage. In the case where the fiberscope 59 is employed on the recovering apparatus, operators have to rely upon their senses in positioning it in a proper location, offering a risk of introducing an operation error. In order to prevent such an operation error, a greater field of view can be secured to readily seek for any proper position against the fuel assembly. In the present embodiment, the aforementioned interlocking mechanism section 76 prevents an operation error by the operation and can set the visual field of the fiberscope 59 to a minimum possible extent required. FIGS. 13 to 15 show a major area of a recovering apparatus according to a sixth embodiment of the present invention. In these Figures, the same reference numerals are employed to designate parts or elements corresponding to those shown in the preceding embodiment. Any further explanation of them is, therefore, omitted. In FIG. 13, reference numeral 91 shows an interlocking mechanism section serving as a movement restricting section. The interlocking mechanism 91 performs an interlocking operation of the forceps 49 in an X-Y direction and comprises first and second proximity sensors 92 and 93 projected from a second movable body 69 and a comb-like guide plate 94 located over the second movable body 69 and serving as a guide member. The guide plate 94 has a plurality of slits 95, . . . arranged at an equal pitch. The shape, mount position, direction, etc., of the guide plate 94 are set in substantially the same way as those of, for example, the fifth embodiment of the present invention. The proximity sensors 92 and 93 are integrally provided on the upper surface of the second movable body 69 and, together with the second movable body 69, are displaced as one unit in the X and Y directions so as to follow the movements of the forceps 49 and fiberscope 59. The proximity sensors 92 and 93 are turned ON when a distance from an object to be detected becomes below a predetermined value and turned OFF when above the predetermined value. The output signals of the proximity sensors 92 and 93 are sent to the controller (see FIG. 9) serving as a control section of the remote control section 44. The proximity sensors 92 and 93 are so dimensioned that their width is made somewhat smaller than that of respective slits 95 of the guide plate 94. The proximity sensors 92 and 93 are provided as an oblique array relative to the direction in which the slits 95 extend. The first sensor 92 is located in both the positive X and the negative Y directions, and the second sensor 93 in both the negative X and positive Y directions. The proximity sensors 92 and 93 are situated below the guide plate 94, that is, in the negative Z direction. When the forceps 49 is moved in the positive Y direction from a proper front-facing position into one of clearances among the fuel elements, the proximity sensors 92 and 93 follow the motion of the forceps 49 and are moved along the slit 95a in the positive Y direction. The proximity sensors 92 and 93 are rendered ON and OFF in accordance with a distance to the guide plate 94, that is, are rendered ON when they are moved nearer the nearest area of the guide plate 94 within a distance below a predetermined value. As a material for the guide plate 94, use may be made of one which provides a dock for the proximity sensors 92 and 93. In practice, for example, SUS 303 is employed as a material for the guide plate 94. In FIG. 13, reference numeral 96 shows a stroke end a limit sensor, serving as limit sensor, which is provided integral with the second movable body 69. The limit sensor 96 is provided on the rear end portion of the second movable body 69 under the assumption that the aforementioned rear end portion is determined with the Y direction as the forward/back direction. The stroke and limit sensor 96 is turned ON and OFF in accordance with, for example, a distance relative to the first movable body 68 situated on the rear side of the second movable body 69. The limit sensor 96 is turned ON when the second movable body 69 is retracted, in the Y direction, back toward the first movable body 68 to allow the limit sensor 96 to be moved to the first movable body 68 within a distance corresponding to a predetermined value. The output signal of the limit sensor 96 is delivered to the controller 72. In FIG. 13, reference numeral 97 shows a position detection section which comprises a proximity sensor 98, serving as a movable body, of a cylindrical configuration provided integral with the first movable body 68 and a plate-like fixing section 99 secured to a casing 45 and disposed near the first movable body 68. The position detection section 97 has a plurality of nearly circular detection holes 100, . . . at the fixing section 99 which are arranged at a substantially equal pitch in the X direction in a linear array. The position detection section 97 is so dimensioned that the diameter of the detection hole 100 is substantially equal, or corresponds to the clearance 203 of the fuel elements 202, . . . . The position detection section 97 is such that the proximity sensor 98 and detection holes 100, . . . are situated at a substantially equal height in the Z direction. Upon the movement of the proximity sensor 98 nearer to the fixing section 99, the position detection section 97 detects whether the proximity sensor 98 is correctly oriented relative to any one of the detection holes 100 or displaced therefrom. The result of detection is sent as an output to, for example, the controller 72. The operation of the foreign matter recovering apparatus 101 including the stroke end limit sensor 96, movement restriction section 91, position detection section, etc., will be explained below, as shown in FIG. 14. When the first proximity sensor 92 is moved toward a portion of the guide plate 94 to render it ON, the controller 72 sends to the first feed motor 65 an instruction not to rotate it in a positive direction even if the forceps 49 is operated by the remote control section 44 to move it in the positive X direction. When, with the first proximity sensor 92 ON, the forceps 49 is operated by the remote control section 44 so as to move it in the negative X direction, the first feed motor 65 is rotated in a reverse direction in accordance with that operation and the forceps 49 is moved in the negative X direction. With the second proximity sensor 93 ON, on the other hand, the first feed motor 65 is not rotated in the reverse direction and is rotated in a positive direction only. The forceps 49 is moved in the positive X direction only, not in the negative direction. With either one of the proximity sensors 92 and 93 ON, the controller 72 sends an instruction to the second feed motor 66 so as not to move the forceps 49 in the positive Y direction. When the second movable body 69 is retracted back in the negative Y direction to allow the forceps 49 to be withdrawn off the clearance 203 between the fuel elements 202, . . . and hence apart from the fuel assembly 201, then the proximity sensors 92 and 93 are also retracted away from the guide plate 94 in the Y direction and the stroke end limit sensor 96 is moved toward the first movable body 68 and, when the aforementioned predetermined distance between the limit sensor 96 and the first movable body 68 reaches a predetermined value, the stroke end limit sensor 96 is turned ON. In this case, the controller 72 imparts no restriction to the first feed motor 65 and the forceps 49, proximity sensors 92, 93, etc., are freely moved in the X direction in accordance with the operation of the remote control section 44. It may be possible to initially determine the stroke end of the forceps 49 in the positive Y direction so that it may not be penetrated too deeply. In practical use, for example, the closed end of the slits 95 are detected at their position through the use of the proximity sensors 92 and 93, and the forceps 49 is stopped from being further moved in accordance with the result of detection. It is thus possible to prevent any error of operation done while viewing the forceps 49 on the monitor 73, and to obtain high reliability. In the case where the forceps 49, etc., are placed in a state freely movable in the X direction, the position detection section 97 is employed to detect the forceps 49 for a proper position. That is, the second movable body 69 is retracted back to allow the proximity sensor 98 of the position detection section 97 to be moved nearer the fixing section 99. When the proximity sensor 98 faces the detection hole 100 on the front-facing side, the result of detection by the position detection section 97 is informed to the operator. As a way of informing the result of detection by the position detection section 97 to the operator, an LED may be provided on the operation panel 71 so that light is emitted upon the alignment of the proximity sensor 98 with the detection hole 100. After the forceps 49 is properly positioned with the proximity sensor 98 aligned with the detection hole 100, it is advanced toward the fuel assembly 201 and then into a predetermined clearance 203a of a fuel element array as shown in FIG. 12. Even after the advance of the forceps 49 into that clearance 203a, the movement of the forceps 49 is restricted by the interlocking mechanism 91 and the forceps 49 is moved through the clearance 203a without being brought into contact with the fuel elements 202, . . . . This embodiment can obtain the same advantage as that of the fifth embodiment and more exactly control the movement of the forceps 49. Since the interlocking mechanism section 91 is comprised of the proximity sensors 92, 93 and guide plate 94, the proximity sensors make no contact with the guide plate 94. Therefore, the guide plate 94 is set smaller in rigidity than that in the fifth embodiment, permitting the use of a thinner plate as the guide plate 94. FIGS. 16 and 17 schematically show a major area on a seventh embodiment of the present invention. In these Figures, the same reference numerals are employed to designate parts or elements corresponding to those shown in the preceding embodiment. Any further explanation of them is, therefore, omitted. In FIGS. 16 and 17, reference numeral 111 denotes a fiberscope oscillation mechanism hereinafter referred to simply as an oscillation mechanism. The oscillation mechanism 111 includes, as shown in FIG. 17, an oscillation mechanism body 112 and a rotation support 113 integrally projecting on the oscillation mechanism body 112. On the oscillation mechanism body 112 is mounted a drive shaft 115 having an externally threaded section 114 provided on its outer peripheral portion. A rotational force is transmitted to the drive shaft 115 from a drive source, not shown. A rotational force is transmitted to the drive shaft 115 through a flexible shaft. In FIG. 16, reference numeral 116 shows a swing plate, and 117, a fiberscope retaining section coupled integral with the swing plate 116. A ball 118 is mounted on one end of the swing plate 116 in a manner to be sandwiched between two flanges 119 and 119 mounted on the drive shaft 115 in a substantially parallel fashion. A spring 120 is anchored to the other end portion of the swing plate 116 which is coupled to the oscillation mechanism body 112 through the spring 120. The fiberscope retaining section 117 is coupled to a distal end 60 of a fiberscope 59 to retain the fiberscope's distal end. The retaining section 117 engages the rotation support 113 and is pivotally supported by the rotation support 113. The fiberscope retaining section 117, together with the ball 118 and spring 120, is arranged substantially along the axial direction of the swing plate 116 at a location between the ball 118 and the spring 120. The mounting relation among the oscillation mechanism body 112, swing plate 116 and fiberscope retaining section 117 is so adjusted that the swing plate 116 can be swung in a direction of an arrow B in FIG. 16 with the ball 118 inserted between the flanges 119 and 119 and that this is done without the fiberscope retaining section 11 being detached from the rotation support 113. With a rotational force applied to the drive shaft 115, the shaft 115 is rotated, moving back and forth in its axial direction in accordance with the direction in which it is rotated. The flanges 119 and 119, together with the drive shaft 115, are displaced as one unit and the ball 118 of the drive shaft 115 is pushed by the flanges 119 and 119. The ball 118 is rolled in accordance with an amount of displacement of the flanges 119, 119 in a manner to be sandwiched between the flanges 119 and 119. The swing plate 116 is swingable in a direction of an arrow B in FIG. 16 with the ball 118 as a center while compressing or stretching the spring 120. Upon the swinging of the swing plate 116, the fiberscope retaining section 117 is displaced as one unit with the swing plate 116 while retaining an engaging relation to the rotation support 113. The fiber retaining section 117 is inclined with the pivotal point of the rotation shaft 113 as a center to allow the distal end 60 of the fiberscope 59 to be axially rotated in accordance with the direction and displacement amount in and to which the swing plate 116 is swung. The spring 120 urges the swing plate 116 and distal end 60 of the fiberscope 59 so that their neutral attitudes may be taken. That is, the oscillation mechanism 111 imparts to the fiberscope 59 the degree of freedom of its rotation direction (.theta. direction). As shown in FIG. 17, the distal end 60 of the fiberscope 59 is rotated around its axis and the direction of the fiberscope 59 is changed in a direction intersecting with the axial direction (and entry direction) of the distal end 60. The oscillation mechanism 111 swings the fiberscope 59 with its axis as a center to oscillate the fiberscope 59. In this case, an imaging and an illuminating array, etc., constituting the distal end 60 of the fiberscope 59 are oscillated as one unit. The fiberscope 59 has its direction changed, as required, allowing an enlarged viewing field and observation area to be imparted thereto. In the case where the fiberscope 59 is so fixed that it cannot be rotated around its axis, a field of vision, C, of the fiberscope 59 is restricted by its predetermined range as well as by the range within which the distal end 60 can be moved in the clearances 203, among the fuel elements 202, . . . In the case where in the clearances of the fuel assembly 201 a site D is observed from a position spaced apart from, for example, a viewing field C in the X direction, (see FIG. 18), it is necessary to separate the aforementioned working unit recovering apparatus (3, 43) away from the fuel assembly 201, to move the distal end 60 of the fiberscope 59 in the Y direction and to once remove the distal end 60 from the fuel assembly 201. The working unit (3, 43) is moved with its direction changed so that it approaches the fuel assembly 201 from the X direction. The distal end 60 of the fiberscope 59 enters the fuel assembly 201 in the X direction and covers the site D within its visual field. Since the oscillation mechanism 111 as set out above is provided on the aforementioned embodiment, the field of view of the fiberscope 59 can be readily enlarged and an observation area covered by the fiberscope 59 without moving the working unit (3, 43, etc.,) is enlarged in the X direction to an area E as shown in FIG. 17. In the present embodiment, the way of obtaining a broader area of observation from a method of approach in two-way direction can be replaced by an approach in a single direction, shortening a working period of time required. The use of a plurality of fiber cables enables a plurality of clearances 203, . . . provided, for example, as a parallel array, in the same direction to be observed. Mounting the aforementioned oscillation mechanism 111 on the fourth to sixth embodiments (41, 81, 101) enables the fiberscope 59 to be moved in the X, Y, Z and .theta. directions. If a positional relation between the fiberscope 59 and the fiberscope retaining section 117 is adjusted, the fiberscope 59 can be rotated not only around its axis but also in an eccentric way. The present invention is not restricted to the aforementioned respective embodiments and may be changed or modified in various ways without departing from the spirit and scope of the present invention.
058833943	claims	1. A layered radiation shield, said layered radiation shield comprising: (a) an inner layer comprising at least one flexible sheet of solid radiation shielding material; and (b) an outer layer comprising a flexible, cohesive elastomeric coating, said elastomeric coating flexibly coating said inner layer, and resistant to discoloration and coating degradation when said layer is exposed to sunlight. (a) a sequence of layers of radiation shield portions, said sequence of layers of said radiation shield portions comprising a first shield layer S.sub.1 through an Nth shield layer S.sub.N, wherein N is a positive integral number corresponding to the number of radiation shield layers provided; and wherein in each of said layers S.sub.1, through S.sub.N, one or more shield portions is provided and wherein the number of shield portions in each of said sequence of layers may be described, in sequence, as shield portions S.sub.N (1) through S.sub.N (X), wherein X is a positive integer corresponding to the number of radiation shield portions in any layer N; and (b) an elastomeric outer coating layer, said elastomeric coating outer layer comprising a flexible, elastomeric coating, said elastomeric coating being resistant to discoloration and coating degradation when said layer is exposed to sunlight. (a) an inner layer comprising at least one sheet of solid radiation shielding material; (b) an outer stainless steel casing; (c) a sealant, said sealant sealing located between at least portions of said at least one sheet of solid radiation shielding material and said outer stainless steel casing; (d) said sealant and said stainless steel casing cooperating to effectively seal said solid radiation shielding material against leakage outward through said outer stainless steel casing. an obverse side comprising a reverse side comprising wherein said left flange of said obverse side, and said left flange of said reverse side each are sized to extend outward from said at least one sheet of solid radiation shielding material by a clearance distance sufficient to allow said left flange of said obverse side and said left flange of said reverse side to be joined by an internally protruding mechanical fastening device, without said internally protruding mechanical fastening device intruding into said at least one sheet of radiation shielding material. (a) a right flange on said obverse side, and (b) a right flange on said reverse side, (c) and wherein said right flange of said obverse side and said right flange of said reverse side each are sized to extend outward from said at least one sheet of solid radiation shielding material by a clearance distance sufficient to allow said right flange of said obverse side and said left flange of said reverse side to be joined by an internally protruding mechanical fastening device, without said internally protruding mechanical fastening device intruding into said at least one sheet of radiation shielding material. (a) a bottom flange on said obverse side, and (b) a bottom flange on said reverse side, (c) and wherein said bottom flange of said obverse side and said bottom flange of said reverse side each are sized to extend downward below the bottom of said at least one sheet of radiation shielding material by a clearance distance sufficient to allow said bottom flange of said obverse side and said bottom flange of said reverse side to be joined by an internally protruding mechanical fastening device, without said internally protruding mechanical fastening device intruding into said at least one sheet of radiation shielding material. (a) a top flange on said obverse side, and (b) a top flange on said reverse side, (c) and wherein said top flange of said obverse side and said top flange of said reverse side each are sized to extend upward above the top of said at least one sheet of radiation shielding material by a clearance distance sufficient to allow said top flange of said obverse side and said top flange of said reverse side to be joined by an internally protruding mechanical fastening device, without said internally protruding mechanical fastening device intruding into said at least one sheet of radiation shielding material. (a) a radiation shield, said shield comprising a body with opposing substantially planar surface portions, and, through said body, at least one through passageway having edge portions, said at least one through passageway in said radiation shield structurally adapted to allow said radiation shield to be held by a supporting structure protruding at least partially into said through passageway; and (b) a hanger, said hanger comprising (a) supporting radiation shields between an ionizing radiation source, and (b) an area in which exposure to said ionizing radiation is be attenuated, wherein said radiation shields comprise 2. The radiation shield as set forth in claim 1, wherein said elastomeric coating has a minimum percent elongation, as measured by ASTM Method D-838, of sixty percent. 3. The radiation shield as set forth in claim 1, wherein said elastomeric coating has a Shore D hardness, as measured by ASTM Method D-2240, of 37. 4. The radiation shield as set forth in claim 1, wherein said elastomeric coating has a tensile shear strength, as measured by ASTM Method D-1002, of approximately 347 pounds per square inch. 5. The radiation shield as set forth in claim 1, wherein said elastomeric coating comprises a Bisphenol A epoxy coating. 6. The radiation shield as set forth in claim 5, wherein said elastomeric coating is cross-linked with a modified cycloaliphatic amine curing agent. 7. The radiation shield as set forth in claim 1, wherein said radiation shield is of the type designed to be held up by a supporting structure, and wherein said radiation shield further comprises at least one grommet, said at least one grommet defining through passageways in said radiation shield, whereby said radiation shield may be upheld by a supporting structure protruding through said grommet. 8. The radiation shield as set forth in claim 1, wherein each of said at least one flexible solid sheets of radiation shielding material is provided in substantially planar form. 9. The radiation shield as set forth in claim 1, wherein each of said at least one solid sheets of radiation shielding material is provided in the shape of a segmented annulus. 10. A method for radiation shielding, said method comprising supporting the radiation shields as set forth in claim 1 between (a) a radiation source, and (b) an area in which a radiation exposure is to be attenuated. 11. The method as set forth in claim 10, wherein said radiation shields are supported by S-hook type fasteners. 12. The method as set forth in claim 10, wherein said radiation shields are supported by J-hook type fasteners. 13. A radiation shield, said radiation shield comprising: 14. The radiation shield as set forth in claim 13, wherein the number of radiation shield portions X in any one of said sequence of layers N equals the number of radiation shield portions X in an adjacent layer N-1 therebelow. 15. The radiation shield as set forth in claim 14, wherein the number of layers N is equal to two. 16. The radiation shield as set forth in claim 14, wherein said radiation shield further comprises a plurality of mechanical fasteners, and wherein said plurality of mechanical fasteners are adapted to fasten radiation shield portions in said layer S.sub.N to said radiation shield portions in layer S.sub.N-1. 17. The radiation shield as set forth in claim 16, wherein said mechanical fasteners comprise deck screws. 18. A lightweight, portable radiation shield, said lightweight, portable radiation shield comprising: 19. The radiation shield as set forth in claim 18, wherein said outer stainless steel casing comprises 20. The radiation shield as set forth in claim 19, wherein said stainless steel casing further comprises 21. The radiation shield as set forth in claim 19, wherein said stainless steel casing further comprises 22. The radiation shield as set forth in claim 19, wherein said stainless steel casing further comprises 23. The radiation shield as set forth in claim 22, further comprising a stainless steel cap, said stainless steel cap comprising an elongate, generally U-shaped channel having a reverse side leg and an obverse side leg, said cap fitted downward over said top flange of said reverse side and said top flange of said obverse side in a manner where said reverse side leg is placed in an abutting relationship with said reverse side, and wherein said obverse side leg is placed in an abutting relationship with said obverse side leg, and wherein a first mechanical fastening device is used to join said reverse side leg of said cap to said reverse side, and a second mechanical fastening device is used to join said obverse side leg of said cap to said obverse side, and wherein each of said first and said second mechanical fastening devices do not intrude into said at least one sheet of radiation shielding material. 24. The radiation shield as set forth in claim 23, wherein each of said first and said second mechanical fastening devices comprises pop-type rivets. 25. The radiation shield as set forth in claim 23, wherein said radiation shield is of the type designed to be held up by a supporting structure, and wherein said radiation shield further comprises at least one grommet, said at least one grommet defining through passageways in said radiation shield, whereby said radiation shield may be upheld by a supporting structure protruding through said grommet. 26. The radiation shield as set forth in claim 18, wherein said sealant comprises a silicon caulking compound. 27. The radiation shield as set forth in claim 18, wherein said sealant comprises polyvinyl chloride filler. 28. The combination of a radiation shield and a hanger, said combination comprising: 29. The combination as set forth in claim 28, further comprising a J-shaped hook, said J-shaped hook located at the lower reaches of said elongate, flat bar, said J-shaped hook adapted to cradle therein said edge portion of said through passageway of said radiation shield. 30. A method for radiation shielding, said method comprising:
	description	The present invention relates to an operation control system for monitoring the operational state of a system. More particularly, the present invention relates to a technique for obtaining operation performance data from a monitored object in order to monitor the operational state of the system. According to a prior art technique, an operation control system periodically obtains various type of operation performance data from monitored computers by use of a control computer to monitor the operational state of the network system. The obtained operation performance data is displayed on the display of the control computer and used by the manager to execute pattern analysis on the operational state of the network system and failure analysis. To reduce the network load occurring when operation performance data is collected from a monitored object, Japanese Laid-Open Patent Publication No. 11-234274 discloses a technique for performing failure analysis by use of the monitored server. However, the control system disclosed in the above Japanese Laid-Open Patent Publication does not change the number and the types of monitored items (e.g., CPU usage rate, memory usage rate, etc.) after it is determined that the operational state of the system has become risky based on the operation performance (metric) value of a specific monitored item. On the other hand, the manager determines the degree of risk involved with the operational state of the system and the risk factors by checking the operation performance value of a specific monitored item whose operation performance value is within a risk range set based on a certain threshold value and the operation performance values of its related monitored items. Thus, the monitored items used to actually monitor the operational state of the system are limited to those whose operation performance value is within the risk range and their related monitored items. The control system disclosed in the above Japanese Patent Laid-Open Publication obtains data of all predetermined, fixed monitored items, which increases both the capacity of the memory for storing the operation performance data and the use of the network (communication line) for transmitting/receiving the operation performance data and unduly reduces the processing performance of the CPU of the monitored computer, causing the problem of reduced processing performance for ordinary services. It is, therefore, an object of the present invention to provide an operation control system which imposes a monitoring load to the extent necessary to carry out pattern analysis on the system operation and failure analysis and does not apply any excessive monitoring load. To accomplish the above object, a control system according to an embodiment of the present invention includes a control computer, and a computer monitored by the control computer. The control computer includes an interface for receiving an operation performance metric value of each of a plurality of first monitored items from the monitored computer, and a control section for, based on the operation performance metric value of each first monitored item, determining a second monitored item whose data should be obtained and issuing an acquisition instruction instructing the monitored computer to obtain an operation performance metric value of the second monitored item which is associated with each first monitored item. The monitored computer includes an interface for receiving the acquisition instruction from the control computer, and a control section for, based on the acquisition instruction, obtaining the operation performance metric value of the second monitored item and transmitting it to the control computer. The present invention also provides a method and computer program each of which includes steps corresponding to the functions of the control system described above. FIG. 1 shows an overall configuration of a system according to the present invention. The system is made up of an operation control server 110 connected to a network 100, and a plurality of monitored servers 120. It should be noted that the network 100 is a communications line such as a local area network (LAN), WAN, or storage area network (SAN). Each monitored server 120 is a monitored computer which includes components such as a CPU 141, a main memory 142, an input device 143, a display device 144, an external storage device 145, and an interface 146 connected to one another by way of a system bus 147. The monitored servers 120 may be host computers, application servers, database servers, or storage devices, for example. The external storage device 145 stores a basic control program 122, a monitored program 123, and an operation performance obtaining agent 121 for obtaining operation performance information on the monitored server 120 from the basic control program 122 and operation performance information on the monitored program 123 from the monitored program 123 itself. They are read into the main memory 142 as necessary. The CPU 141 executes each program in the main memory 142. The input device 143 is a keyboard, a mouse, or the like, while the display device 144 is a bitmap display or the like. The interface 146 is used to connect with a network. The operation control server 110 is a control computer which includes components such as a CPU 131, a main memory 132, an input device 133, a display device 134, an external storage device 135, and an interface 136 connected to one another by way of a system bus 137. The external storage device 135 stores: an operation performance data collecting program 111 for collecting operation performance data from the operation performance obtaining agent 121 installed on each monitored server 120; a database 116 for storing the collected operation performance data; and a monitoring/analyzing program 115 for referring to and processing the stored operation performance data to indicate the operational state to the system manager. They are read into the main memory 132 as necessary. The CPU 131 executes each program in the main memory 132. The input device 133 is a keyboard, a mouse, or the like, while the display device 134 is a bitmap display or the like. The interface 136 is used to connect with a network. The database 116 stores: acquisition monitored item information 117 which defines attribute information on each operation performance monitoring item obtained from all monitored servers 120; and operation performance data 118 collected by the operation performance data collecting program 111 at regular time intervals. The operation performance data collecting program 111 implements the functions of such components as: a data collecting section 113 for collecting operation performance data from the operation performance obtaining agent 121 in each monitored server 120 based on the acquisition monitored items defined by the acquisition monitored item information 117 and storing it in the database 116; an acquisition monitored item setting section 114 for, when it is determined that an operation performance value collected by the data collecting section 113 is within a risk range set using a threshold value defined by the acquisition monitored item information 117 as a reference, setting the acquisition setting (that is, “To Be Acquired” or “Not To Be Acquired”) of each related monitored item indicated by the acquisition monitored item information 117; and an activation (start) timer section 112 for activating the data collecting section 113 at regular time intervals. FIG. 2 shows the data format of the acquisition monitored item information 117 according to the present embodiment. The acquisition monitored item attribute table 200 holds the attributes for all monitored operation performance data. Each entry in the acquisition monitored item attribute table 200 includes: a monitored item number 201 which is a unique number set for each monitored item by the system; a monitored item name 202; an acquisition setting 203 set for the operation performance data of each monitored item; a monitored object number 204 set for each monitored object from which the operation performance data of each monitored item is obtained (in FIG. 2, for example, 10 indicates a host computer, 20 a program, 30 an application server, and 40 a database server); and a threshold value 205 for indicating a reference value used to determine whether the operation performance value of each monitored item is within a risk range. It should be noted that according to the present embodiment, the monitored object number 204 is set on an apparatus basis for apparatuses such as host computers. However, the monitored object number 204 may be set on a hardware component basis or a software component basis (that is, for each program or each part of a program) within each apparatus. The threshold value 205 has attached thereto a sign indicating whether the range over or under the threshold value is the risk range. Specifically, if an acquired operation performance value is supposed to be in the risk range when it exceeds the threshold value, the plus sign “+” is assigned; otherwise the minus sign “−” is assigned instead. A plurality of threshold values 205 may be assigned to each item number 201. Further, the acquisition setting 203 may not indicate whether data of each item is “To Be Acquired” or “Not To Be Acquired”. Instead, the acquisition setting 203 may provide information for changing, in steps, the number of pieces of operation performance data to be obtained per unit time (acquisition interval). With this arrangement, more operation performance data may be obtained in a riskier case. The acquisition setting monitored item table 210 lists each specific monitored item and its related monitored items. The data of the related monitored items should be obtained when the operation performance value of the specific monitored item is in the risk range. It should be noted that the acquisition item setting section 114 may receive an input value from the input device 133, etc. and set or change the contents of the (acquisition setting) monitored item table 210 based on the input value. Each entry in the acquisition setting monitored item table 210 includes: a monitored item number 211; a monitored object number 212 (indicating a monitored object) set for the monitored item indicated by the monitored item number 211; a setting monitored item number 213 for indicating a monitored item whose data should be obtained when (the operation performance value of) the monitored item indicated by the monitored item number 211 is in the risk range; and a setting monitored object number 214 (indicating a monitored object) set for the monitored item indicated by the setting monitored item number 213. Record {circle around (3)} in FIG. 2, for example, indicates that if the threshold value of the monitored item “200” of the monitored object “20” changes, then the acquisition setting of the monitored item “201” of the monitored object “20” must be set again. It should be noted that a plurality of threshold values 205 may be employed each corresponding to a different risk range (degree of risk). With this arrangement, the setting monitored item number 213 and the setting monitored object number 214 for each monitored item number 211 may be changed for each risk range. Furthermore, priority may be given to each setting monitored item number 213 and each setting monitored object number 214. Then, as the degree of the risk increases, (the acquisition settings) for more setting monitored item numbers 213 and more setting monitored object numbers 214 may be set in the order of decreasing priority. FIG. 3 shows the data format of the operation performance data 118 according to the present embodiment. The operation performance data table 300 holds the operation performance values of monitored items collected from each monitored server 120. Each entry in the operation performance data table 300 includes a monitored object number 301, a monitored item number 302, an acquisition time 303, and an operation performance value 304, collectively constituting collected operation performance data. It should be noted that the example shown in FIG. 3 obtains data at one minute intervals. However, a different interval may be employed for each item number 302. FIG. 4 shows the flow of the processing carried out by the operation performance obtaining agent 121. It should be noted that the agent 121 is constantly activated as a demon program and returns the operation performance value of a desired monitored item requested by the data collecting section 113 of the operation control server 110 through the network 100. At step 401, the agent 121 receives an operation performance data acquisition request specifying a monitored item from the operation performance data collecting program 111. At step 402, the processing by the agent 121 proceeds to either step 403 or 404 depending on the contents of the received acquisition request. If it is determined at step 402 that the acquisition request is for the operation performance data of a hardware component within the monitored server, the agent 121 instructs the basic control program 122 to obtain the operation performance data of the specified monitored item at step 403. Upon receiving this instruction, the basic control program 122 obtains the operation performance data of the specified monitored item. If it is determined at step 402 that the acquisition request is for the operation performance data of a software program within the monitored server, on the other hand, the agent 121 instructs the monitored program 123 to obtain the operation performance data of the specified monitored item at step 404. Upon receiving this instruction, the monitored program 123 obtains the operation performance data of the specified monitored item. At step 405, the agent 121 sends the operation performance data (received from the basic control program 122 or the monitored program 123) to the operation performance data collecting program 111 of the operation control server 110. It should be noted that after sending the operation performance data, the agent 121 assumes a wait state waiting for the next request. FIG. 5 shows the flow of the processing carried out by the data collecting section 113 of the operation performance data collecting program 111. The collecting program 111 is activated by the activation (start) timer section 112 at regular time intervals (for example, one minute intervals), and collects the operation performance data of each monitored item whose acquisition setting 203 is set to “To Be Acquired” from each monitored server 120 based on the acquisition item attribute table 200 and stores the collected operation performance data in the database 116. At step 501, the collecting program 111 reads a record (made up of items 201 and 202) from the acquisition monitored item attribute table 117. At step 502, the processing by the collecting program 111 proceeds to either step 600 or 503 depending on whether or not all records have been already read. At step 503, the collecting program 111 checks the acquisition setting of the read record. If the acquisition setting of the record is “Not To Be Acquired” at step 503, the processing returns to step 501. If the acquisition setting of the record is “To Be Acquired” at step 503, on the other hand, the collecting program 111 sends a request for operation performance data to the operation performance obtaining agent 121 of the monitored server(s) 120 corresponding to the monitored object number of the record through the network at step 504. At step 505, the collecting program 111 obtains the requested operation performance data from the agent 121. Based on the obtained operation performance data, the collecting program 111 stores a new record in the operation performance data table of the operation performance data 118 at step 506, the new record including the monitored object number 301, the monitored item number 302, the acquisition time 303, and the operation performance value 304. After the new record is stored, the processing returns to step 501. If it is determined that all records have been already read at step 502, the acquisition monitored item setting process 600 shown in FIG. 6 is performed. FIG. 6 shows the flow of the setting processing carried out by the acquisition monitored item setting section 114 of the operation performance data collecting program 111 according to the present embodiment. The item setting section 114 is activated after the data collecting process shown in FIG. 5 is completed. The item setting section 114 sets the acquisition setting of each setting monitored item listed in the acquisition setting item table 210 by use of the acquisition item attribute table 200. (Specifically, if it is determined that the operation performance value of the monitored item indicated by a monitored item number 211 in the acquisition setting item table 210 is in the risk range, the item setting section 114 sets the acquisition setting of each setting monitored item for the monitored item to “To Be Acquired”. At step 601, the item setting section 114 reads one record made up of items 301 to 304, such as record {circle around (1)} in FIG. 3, whose acquisition time coincides with the current time from the operation performance data table of the operation performance data 118 stored in the database 116. At step 602, it is determined whether all records have been processed. If it is determined that not all records have been processed, the processing by the item setting section 114 proceeds to step 603. At step 603, the item setting section 114 retrieves from the acquisition monitored item attribute table a record (made up of items 201 and 202) whose item number 201 coincides with the item number 302 of the record read at step 601 (for example, record {circle around (1)} in FIG. 3→record {circle around (2)} in FIG. 2), and obtains the threshold value 205 of the retrieved record. At step 604, the item setting section 114 compares the obtained (read) operation performance value 304 and the obtained threshold value 205. If the comparison result indicates that the operation performance value is in the risk range, the item setting section 114 sets the variable SetStatus to “To Be Acquired” at step 605. If the operation performance value is in the normal range, on the other hand, the item setting section 114 sets the variable SetStatus to “Not To Be Acquired” at step 606. It should be noted that the variable SetStatus is used to establish each acquisition setting 203 at step 610 and is temporarily stored in the main memory 132. At step 607, the item setting section 114 retrieves from the acquisition setting monitored item table all records (each made up of items 211 to 214) whose item number 211 coincides with the monitored item number 302 of the record (for example, record {circle around (1)} in FIG. 3) retrieved at step 601. At step 608, the item setting section 114 retrieves one record from the records retrieved at step 607 (for example, record {circle around (1)} in FIG. 3→record {circle around (3)} in FIG. 2). At step 609, the item setting section 114 determines whether the record retrieved at step 608 is an unprocessed record. If it is an unprocessed record, the processing by the item setting section 114 proceeds to step 610. If all the records retrieved at step 608 have been processed, on the other hand, the processing returns to step 601. At step 610, the item setting section 114 finds from the acquisition monitored item attribute table a record whose monitored item number 201 coincides with the setting monitored item number 213 of the record (record {circle around (3)} in FIG. 2) retrieved at step 608 (record {circle around (3)} in FIG. 2→record {circle around (4)} in FIG. 2), and sets the acquisition setting 203 of the found record to the value of the variable SetStatus set at step 605 or 606. After that, the processing returns to step 608. It should be noted that depending on the contents of the acquisition setting item table, the acquisition setting 203 for the same item number 201 may need to be set a plurality of times at step 610 during the process of processing all the records retrieved at step 601. In such a case, “To Be Acquired” is given priority for the acquisition setting 203 for the item number 201 over “Not To Be Acquired”. If it is determined at step 602 that all the records in the operation performance data table whose acquisition time coincides with the current time have been processed, the processing ends (at END). FIG. 7 shows a variation of the flow of the setting processing carried out by the acquisition monitored item setting section 114 shown in FIG. 6, wherein expected operation performance data is introduced. It should be noted that since this example employs the same steps as those shown in FIG. 6 except for step 604, FIG. 7 shows only the different portion (steps). That is, after step 603, the processing proceeds through the steps shown in FIG. 7 before returning to step 605 or 606 in FIG. 6. At step 701, the item setting section 114 finds from the operation performance data table a record whose acquisition time coincides with the previous acquisition time and whose item number 302 coincides with that of the record retrieved at step 601 and obtains the (previous) operation performance value 304 of the found record. At step 702, the item setting section 114 calculates an expected operation performance value based on the current and previous operation performance values 304. It should be noted that according to the present embodiment, the current value minus the previous value is obtained and simply added to the current value to produce the expected value. However, any method for statistically calculating an expected value may be employed to produce the expected value. At step 703, the item setting section 114 compares the calculated expected value and the threshold value obtained at step 603 to determine whether the expected value is in the risk range. If the item setting section 114 determines that the expected value is in the risk range, the processing proceeds to step 605. If the expected value is in the normal range, on the other hand, the processing proceeds to step 606. According to the embodiment shown in FIG. 7 described above, the degree of risk of a system is determined based on an expected operation performance data value, making it possible to obtain, in advance, the operation performance data of monitored items necessary for analysis conducted when the value of the target monitored item is in the risk range. It should be noted that according to the embodiment shown in FIG. 7 described above, an expected value is calculated based on the current and previous operation performance values. However, N number of past operation performance values may be used in a similar manner to calculate a more accurate expected value. Further, when a Web system is set to be a monitored object, a periodic usage pattern is detected on a daily, weekly, or monthly basis or the like in most cases. In such a case, an expected value may be calculated based on values obtained at the same hour on the previous two days, for example. Thus, an expected value can be calculated based on the periodicity of the operation performance values. Still further according to the present embodiment, the control server 110 side determines a monitored item whose data should be obtained based on its association with a monitored item whose operation performance value is in the risk range and instructs the monitored server 120 to obtain the operation performance data of the determined monitored item. However, the monitored server 120 side may determine a monitored item whose data should be obtained based on its association with a monitored item whose operation performance value is in the risk range, instead, and obtains the operation performance data of the determined item. This arrangement reduces the burden on the resources of the control server 110, such as the CPU, and on the network 100. Thus, according to the embodiment of the present invention, when an obtained operation performance value is within a risk range, it is possible to increase the number of monitored items whose data is to be obtained for analysis, thereby intensively monitoring the closely related monitored items. Or alternatively, the number of monitored items whose data is to be obtained may be reduced to give priority to the primary service, resulting in less intensive monitoring operation. This arrangement makes it possible to collect operation performance data necessary and sufficient for monitoring analysis without imposing any unnecessary load on the monitored system. The present invention can provide an operation control system which imposes a monitoring load to the extent necessary to carry out pattern analysis on the system operation and failure analysis and does not apply any excessive monitoring load.
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	claims	1. A yellow room system, comprising:a plurality of portable unit process apparatuses that are sealed-type treatment apparatuses configured to perform a single treating process in a device manufacturing process and have a same standardized outer shape, the unit process apparatuses including a light-shielding area configured to shield an exposure light to a photosensitive material formed on a workpiece;a conveyance container that is a sealed-type conveyance container configured to house a wafer as a workpiece target and convey the workpiece between the unit process apparatuses, the conveyance container being formed to shield the exposure light; anda light-shielding coupling structure that couples the unit process apparatuses and the conveyance container together, whereinthe wafer housed in the conveyance container has a wafer size for manufacturing a device in a minimized unit. 2. The yellow room system according to claim 1,wherein the conveyance container includes: a conveyance-container main body that forms a housing space for the workpiece; a conveyance-container door configured to shield the housing space; and a first sealing structure configured to seal the housing space by tight coupling between the conveyance-container main body and the conveyance-container door, each of the conveyance-container main body, the conveyance-container door, and the first sealing structure being formed of a member configured to shield a exposure light to the photosensitive material formed on the workpiece,the unit process apparatuses each include: a front chamber to be coupled to the conveyance container; and a treatment chamber to be coupled to the front chamber,the front chamber includes: a front-chamber main body formed of a member configured to shield the exposure light; an opening portion disposed at the front-chamber main body, the opening portion being opened to the treatment chamber; a front-chamber door configured to shield the front-chamber main body from the exposure light; and a second sealing structure configured to seal the front chamber by tight coupling between the front-chamber door and the front-chamber main body and configured to shield the exposure light, andthe conveyance container and the front chamber have a third sealing structure configured to: ensure sealing by tight coupling between the conveyance container and the front chamber; and shield the exposure light, the conveyance container and the front chamber having a structure configured to form one indivisible coupling chamber sealed by the third sealing structure only while the conveyance container and the front chamber are tightly coupled together so as to separate the conveyance-container door from the conveyance container. 3. The yellow room system according to claim 2,wherein a work area and a conveyance area for the workpiece are configured to shield the exposure light, the work area including at least a treatment position of the workpiece within the treatment chamber, the conveyance area being disposed from the work area to a door opening position of the conveyance-container door. 4. The yellow room system according to claim 3, further comprisinga structure configured to open the conveyance-container door by attracting the conveyance-container door to the front-chamber door using a magnetic force of a magnet of the front-chamber door. 5. The yellow room system according to claim 3,wherein the unit process apparatuses are sealed-type treatment apparatuses configured to perform a singular treatment process in a device manufacturing process, the unit process apparatuses being portable,the conveyance container is a sealed-type conveyance container configured to house one wafer as a workpiece target, andthe wafer housed in the sealed conveyance container has a wafer size for manufacturing a device in a minimized unit. 6. The yellow room system according to claim 3,wherein the unit process apparatuses are constituted as application apparatuses for a photosensitive material on a workpiece, the application apparatuses including: a sealed-type container main body configured to house the photosensitive material and is configured to shield the exposure light; a supplying member configured to supply the photosensitive material onto a workpiece; and a plug-in connector removably coupling the container main body and the supplying member together, andthe plug-in connector has a structure configured to shield the exposure light, the plug-in connector including a valve configured to open during coupling. 7. The yellow room system according to claim 2, further comprisinga structure configured to open the conveyance-container door by attracting the conveyance-container door to the front-chamber door using a magnetic force of a magnet of the front-chamber door. 8. The yellow room system according to claim 7,wherein a clearance is disposed between an magnetized object of the conveyance-container door and a magnetic material of the conveyance-container main body and a clearance is disposed between the magnetic material of the conveyance-container door and the magnet of the front-chamber door, so as to open and close the conveyance-container door. 9. The yellow room system according to claim 8,wherein the unit process apparatuses are sealed-type treatment apparatuses configured to perform a singular treatment process in a device manufacturing process, the unit process apparatuses being portable,the conveyance container is a sealed-type conveyance container configured to house one wafer as a workpiece target, andthe wafer housed in the sealed conveyance container has a wafer size for manufacturing a device in a minimized unit. 10. The yellow room system according to claim 8,wherein the unit process apparatuses are constituted as application apparatuses for a photosensitive material on a workpiece, the application apparatuses including: a sealed-type container main body configured to house the photosensitive material and is configured to shield the exposure light; a supplying member configured to supply the photosensitive material onto a workpiece; and a plug-in connector removably coupling the container main body and the supplying member together, andthe plug-in connector has a structure configured to shield the exposure light, the plug-in connector including a valve configured to open during coupling. 11. The yellow room system according to claim 7,wherein the unit process apparatuses are sealed-type treatment apparatuses configured to perform a singular treatment process in a device manufacturing process, the unit process apparatuses being portable,the conveyance container is a sealed-type conveyance container configured to house one wafer as a workpiece target, andthe wafer housed in the sealed conveyance container has a wafer size for manufacturing a device in a minimized unit. 12. The yellow room system according to claim 7,wherein the unit process apparatuses are constituted as application apparatuses for a photosensitive material on a workpiece, the application apparatuses including: a sealed-type container main body configured to house the photosensitive material and is configured to shield the exposure light; a supplying member configured to supply the photosensitive material onto a workpiece; and a plug-in connector removably coupling the container main body and the supplying member together, andthe plug-in connector has a structure configured to shield the exposure light, the plug-in connector including a valve configured to open during coupling. 13. The yellow room system according to claim 2,wherein the unit process apparatuses are sealed-type treatment apparatuses configured to perform a singular treatment process in a device manufacturing process, the unit process apparatuses being portable,the conveyance container is a sealed-type conveyance container configured to house one wafer as a workpiece target, andthe wafer housed in the sealed conveyance container has a wafer size for manufacturing a device in a minimized unit. 14. The yellow room system according to claim 2,wherein the unit process apparatuses are constituted as application apparatuses for a photosensitive material on a workpiece, the application apparatuses including: a sealed-type container main body configured to house the photosensitive material and is configured to shield the exposure light; a supplying member configured to supply the photosensitive material onto a workpiece; and a plug-in connector removably coupling the container main body and the supplying member together, andthe plug-in connector has a structure configured to shield the exposure light, the plug-in connector including a valve configured to open during coupling. 15. The yellow room system according to claim 1,wherein the minimized unit is one, andthe wafer size is 0.5 inches in diameter. 16. The yellow room system according to claim 15,wherein the unit process apparatuses are constituted as application apparatuses for a photosensitive material on a workpiece, the application apparatuses including: a sealed-type container main body configured to house the photosensitive material and is configured to shield the exposure light; a supplying member configured to supply the photosensitive material onto a workpiece; and a plug-in connector removably coupling the container main body and the supplying member together, andthe plug-in connector has a structure configured to shield the exposure light, the plug-in connector including a valve configured to open during coupling. 17. The yellow room system according to claim 1,wherein the unit process apparatuses are constituted as application apparatuses for a photosensitive material on a workpiece, the application apparatuses including: a sealed-type container main body configured to house the photosensitive material and is configured to shield the exposure light; a supplying member configured to supply the photosensitive material onto a workpiece; and a plug-in connector removably coupling the container main body and the supplying member together, andthe plug-in connector has a structure configured to shield the exposure light, the plug-in connector including a valve configured to open during coupling. 18. The yellow room system according to claim 1,wherein the unit process apparatuses are constituted as application apparatuses for a photosensitive material on a workpiece, the application apparatuses including: a sealed-type container main body configured to house the photosensitive material and is configured to shield the exposure light; a supplying member configured to supply the photosensitive material onto a workpiece; and a plug-in connector removably coupling the container main body and the supplying member together, andthe plug-in connector has a structure configured to shield the exposure light, the plug-in connector including a valve configured to open during coupling.
	summary
	claims	1. A self-contained, multi-function test and measurement apparatus used to test and measure the performance of a laser gun, comprising:internal control means for selecting and directing at least one of a plurality of selected test and measurement functions;means, operationally coupled to said internal control means, for measuring or verifying the capability of said laser gun to accurately measure at least two of the following: a double pulse; pulse frequency within a prescribed range of a designed pulse frequency; pulse width within a prescribed range of a designed pulse width; optical power within a prescribed maximum limit; actual wavelength within a prescribed range of a designed wavelength; speed simulation; and internal clock frequency;a housing including a certification unit containing the internal control means, said certification unit comprising:an integrated programmable display that provides a visual indicia of instructions or data or both to a user;a programmable control circuit including a plurality of functional blocks that provide control instructions for the apparatus, operationally coupled to the programmable display; anda plurality of integrated connectors that each provide a plug-and-play-type interface for a modular test head; andat least one modular, plug-and-play-type test head including a programmable test control circuit including a plurality of functional blocks that provide control operations of the test head,wherein the apparatus has at least one of an optical and an electrical interface that enables a respective operational connection with the laser gun, and further wherein the apparatus is self-contained and has size and weight characteristics that enable hand-held transport by the user. 2. The apparatus of claim 1, wherein the at least one modular, plug-and-play-type test head is an optical pulse characterization test head that enables a measurement of an optical pulse characteristic of the laser gun output being tested. 3. The apparatus of claim 2, wherein the optical pulse characteristic is pulse width. 4. The apparatus of claim 2, wherein the optical pulse characteristic is pulse frequency. 5. The apparatus of claim 2, wherein the optical pulse characteristic is a double-pulse condition. 6. The apparatus of claim 1, wherein the at least one modular, plug-and-play-type test head is an optical wavelength measurement test head that enables a measurement of an optical characteristic of the laser gun output being tested. 7. The apparatus of claim 6, wherein the optical characteristic is pulse wavelength. 8. The apparatus of claim 1, further comprising a coil of optical fiber having a known length, interfaced to the certification unit. 9. The apparatus of claim 8, wherein the at least one modular, plug-and-play-type test head is a distance measurement test head that enables a measurement of a distance calibration characteristic of the laser gun being tested. 10. The apparatus of claim 9, wherein the distance measurement test head is a short-range distance test head comprising a length of optical fiber. 11. The apparatus of claim 9, wherein the distance measurement test head is a long-range distance test head comprising a length of optical fiber. 12. The apparatus of claim 1, wherein the at least one modular, plug-and-play-type test head is an optical power measurement test head that enables a measurement of an optical power characteristic of the laser gun output being tested. 13. The apparatus of claim 1, wherein the at least one modular, plug-and-play-type test head is a speed simulation test head that enables a measurement of an object speed calibration characteristic of the laser gun being tested. 14. The apparatus of claim 1, wherein the at least one modular, plug-and-play-type test head is an internal clock frequency measurement test head that enables a measurement of an internal clock frequency characteristic of the laser gun being tested. 15. The apparatus of claim 1, further comprising a personal computer coupled to the certification unit. 16. A method for certifying the operation of a laser-gun-type speed measurement device, comprising:performing a ‘pulse width’ test of a laser, including the steps of:a) applying a first operating voltage to a laser gun of the laser-gun-type speed measurement device;b) directing an output of the laser gun into an optical detector;c) obtaining an output of the optical detector in the form of low-level current pulses;d) converting the low-level current pulses into a corresponding voltage;e) comparing the corresponding voltage to a reference voltage and determining whether an ‘in-range’ condition is present and, if yes;f) converting the low-level current pulses into high-level voltage pulses, wherein the high-level voltage pulses directly represent a pulse train output of the laser gun;g) providing a sampling circuit including eight parallel sampling paths that operate at phase differences which are successively 45 degrees apart, wherein each cycle of eight samples contains a series of ‘1’ or ‘0’ bits in an 8-bit data block that represent whether a pulse from the laser gun was present (‘1’) or not present (‘0’) at the time the sample was taken;h) accumulating 512 8-bit data blocks in a memory;i) counting the ‘1’ bits in all 512 data blocks;j) using a known sample time to calculate an effective pulse width of the laser gun output;k) determining whether the calculated effective pulse width is within a specified range of a laser gun reference pulse width and, if no,l) concluding the test; otherwise,m) determining whether the ‘pulse width’ test is to be repeated at a different operating voltage and, if yes,n) applying at least a second operating voltage to the laser gun; ando) repeating steps (b-o) until the test has been performed at all selected operating voltages. 17. The method of claim 16, further comprising performing an ‘optical power’ test of a laser-gun-type speed measurement device. 18. The method of claim 16, further comprising performing a ‘wavelength’ test of a laser-gun-type speed measurement device. 19. The method of claim 16, further comprising performing a ‘speed simulation’ test of a laser-gun-type speed measurement device. 20. The method of claim 16, further comprising performing a ‘horizontal beam width’ test of a laser-gun-type speed measurement device. 21. The method of claim 16, further comprising performing a ‘vertical beam width’ test of a laser-gun-type speed measurement device. 22. The method of claim 16, further comprising performing a ‘sight alignment’ test of a laser-gun-type speed measurement device. 23. The method of claim 16, further comprising performing a ‘double pulse’ test, a ‘pulse frequency’ test, an ‘optical power’ test, a ‘wavelength’ test, a ‘speed simulation’ test, a ‘horizontal beam width’ test, a ‘vertical beam width’ test, and a ‘sight alignment’ test of a laser-gun-type speed measurement device. 24. A method for certifying the operation of a laser-gun-type speed measurement device, comprising:performing a ‘double pulse’ test of a laser, including the steps of:a) applying a first operating voltage to a laser gun of the laser-gun-type speed measurement device;b) directing an output of the laser gun into an optical detector;c) obtaining an output of the optical detector in the form of low-level current pulses;d) converting the low-level current pulses into a corresponding voltage;e) comparing the corresponding voltage to a reference voltage and determining whether an ‘in-range’ condition is present and, if yes;f) converting the low-level current pulses into high-level voltage pulses, wherein the high-level voltage pulses directly represent a pulse train output of the laser gun;g) providing a sampling circuit including eight parallel sampling paths that operate at phase differences which are successively 45 degrees apart, wherein each cycle of eight samples contains a series of ‘1’ or ‘0’ bits in an 8-bit data block that represent whether a pulse from the laser gun was present (‘1’) or not present (‘0’) at the time the sample was taken;h) accumulating 512 8-bit data blocks in a memory;i) determining whether one or more ‘0’ bits are found between two ‘1’ bits and, if yes,j) concluding the test; otherwise,k) determining whether the ‘double pulse’ test is to be repeated at a different operating voltage and, if yes,l) applying at least a second operating voltage to the laser gun; andm) repeating steps (b-m) until the test has been performed at all selected operating voltages. 25. A method for certifying the operation of a laser-gun-type speed measurement device, comprising:performing a ‘pulse frequency’ test of a laser, including the steps of:a) applying a first power supply voltage to a laser gun of the laser-gun-type speed measurement device;b) directing an output of the laser gun into an optical detector;c) obtaining an output of the optical detector in the form of low-level current pulses;d) converting the low-level current pulses into a corresponding voltage;e) comparing the corresponding voltage to a reference voltage and determining whether an ‘in-range’ condition is present and, if yes;f) converting the low-level current pulses into high-level voltage pulses, wherein the high-level voltage pulses directly represent a pulse train output of the laser gun;g) providing a sampling circuit including eight parallel sampling paths that operate at phase differences which are successively 45 degrees apart, wherein each cycle of eight samples contains a series of ‘1’ or ‘0’ bits in an 8-bit data block that represent whether a pulse from the laser gun was present (‘1’) or not present (‘0’) at the time the sample was taken;h) providing a sampling circuit that includes a flip-flop function that toggles from a “0” to a “1” state on every other pulse in the pulse train from the laser unit under test;i) providing a counter to accumulate the number of 400 MHz clock pulses that occur when the flip-flop is in the “1” state;j) transferring the value in the counter to a latch when the flip-flop returns to the “0” state, creating a locked value, and resetting the counter for a next cycle;k) using the locked value to determine the pulse period of the output of the laser unit under test by multiplying said latched value by the period of the 400 MHz clock and generating a measured pulse frequency value;l) determining whether the measured pulse frequency value is within a predetermined range of a specified pulse frequency value and, if no,m) concluding the test; otherwise,n) determining whether the ‘pulse frequency’ test is to be repeated at a different power supply voltage and, if yes,o) applying at least a second power supply voltage to the laser gun; andp) repeating steps (b-p) until the test has been performed at all selected power supply voltages.
051494914	claims	1. A fuel management method for a dual-phase nuclear reactor, said method comprising: installing a fuel bundle at a first core location accessed by coolant through a relatively small aperture, each of said bundles having a predetermined group of fuel elements; operating said reactor a first time; shutting down said reactor; reinstalling said fuel bundle at a second core location accessed by coolant through a relatively large aperture; and operating said reactor a second time. shutting down said reactor; reinstalling said fuel bundle at a peripheral core location; and operating said reactor a third time. shutting down said reactor; and disposing of said fuel bundle so that said fuel bundle is installed exactly three times in said reactor. installing a fuel bundle at a first core location accessed by coolant through a relatively small aperture so as to establish a relatively low coolant flow rate and thus a relatively high conversion ratio in said bundle when said reactor is next operated; operating said reactor a first time; shutting down said reactor; reinstalling said fuel bundle at a second core location accessed by coolant through a relatively large aperture so as to establish a relatively high coolant flow rate and thus a relatively low conversion ratio when said reactor is next operated, said second core location being radially inward of said first core location; and operating said reactor a second time. shutting down said reactor; reinstalling said fuel bundle at a peripheral core location characterized by a relatively low coolant flow rate; and operating said reactor a third time. a core plate defining an array of fuel bundle locations, including conversion locations and power locations, said power locations being characterized by larger flow orifices than said conversion locations, said conversion locations being located radially outside said power locations; fuel bundles with relatively large quantities of fertile fuel disposed at said conversion locations; and fuel bundles with relatively large quantities of fissile conversion products at said power locations; whereby, when an incorporating reactor is operating, said fuel bundles with relatively large quantities of fertile fuel are exposed to relatively large voids and thus to fast neutrons that promote conversion of fertile fuel to fissile fuel, said fuel bundles with relatively large quantities of fissile conversion products are exposed to relatively small voids so that moderation is greater and more thermal neutrons are present to promote fissioning. fuel bundles with relatively little fissile fuel and relatively large quantities of fission byproducts at said peripheral locations, said fuel bundles at said peripheral locations each having a face not opposing any other of said fuel bundles in said core. 2. A method as recited in claim 1 wherein said second core location is radially inward of said first core location. 3. A method as recited in claim 2 further comprising additional steps following said step of operating said reactor a second time, said additional steps comprising: 4. A method as recited in claim 3 further comprising additional steps following said step of operating said reactor a third time, said additional steps comprising: 5. A fuel management method for a dual-phase nuclear reactor, said method comprising: 6. A method as recited in claim 5 further comprising additional steps following said step of operating said reactor a second time, said additional steps comprising: 7. A core for a dual-phase nuclear reactor comprising: 8. A reactor core as recited in claim 7 wherein said core plate also includes peripheral locations and said core further comprises:
	description	This application is a division of commonly assigned U.S. application Ser. No. 12/649,758 for a UVLED Apparatus for Curing Glass-Fiber Coatings, (filed Dec. 30, 2009, and published Jul. 22, 2010, as Publication No. 2010/0183821 A1), now U.S. Pat. No. 8,314,408, which itself claims the benefit of commonly assigned U.S. Patent Application No. 61/141,698 for a UVLED Apparatus for Curing Glass-Fiber Coatings (filed Dec. 31, 2008). Each of the foregoing patent applications and patent application publication is hereby incorporated by reference in its entirety. The present invention embraces an apparatus and a method for curing coatings on drawn glass fibers. Glass fibers are typically protected from external forces with one or more coating layers. Typically, two or more layers of coatings are applied during the optical-fiber drawing process (i.e., whereby a glass fiber is drawn from an optical preform in a drawing tower). A softer inner coating layer typically helps to protect the glass fiber from microbending. A harder outer coating layer typically is used to provide additional protection and to facilitate handling of the glass fiber. The coating layers may be cured, for example, using heat or ultraviolet (UV) light. UV curing requires that the coated glass fiber be exposed to high intensity UV radiation. Curing time can be reduced by exposing the coating to higher intensity UV radiation. Reducing curing time is particularly desirable to permit an increase in fiber drawing line speeds and thus optical-fiber production rates. Mercury lamps (e.g., high pressure mercury lamps or mercury xenon lamps) are commonly used to generate the UV radiation needed for UV curing. One downside of using mercury lamps is that mercury lamps require a significant amount of power to generate sufficiently intense UV radiation. For example, UV lamps used to cure a single coated fiber (i.e., one polymeric coating) may require a collective power consumption of 50 kilowatts. Another shortcoming of mercury lamps is that much of the energy used for powering mercury lamps is emitted not as UV radiation, but rather as heat. Accordingly, mercury lamps must be cooled (e.g., using a heat exchanger) to prevent overheating. In addition, the undesirable heat generated by the mercury lamps may slow the rate at which the optical fiber coatings cure. Furthermore, mercury lamps generate a wide spectrum of electromagnetic radiation, such as having wavelengths of less than 200 nanometers and greater than 700 nanometers (i.e., infrared light). Typically, UV radiation having wavelengths of between about 300 nanometers and 400 nanometers is useful for curing UV coatings. Thus, much of the electromagnetic radiation generated by mercury bulbs is wasted (e.g., 90 percent or more). Additionally, glass fibers possess an exemplary diameter of about 125 microns, which, of course, is much smaller than the mercury bulbs. Consequently, most of the UV radiation emitted by the mercury lamps does not reach the glass fiber's uncured coating (i.e., the energy is wasted). It may thus be advantageous to employ UVLEDs, an alternative to conventional mercury lamps, to cure glass-fiber coatings. UVLEDs typically require significantly less energy and correspondingly generate much less heat energy than conventional UV lamps. For example, U.S. Pat. No. 7,022,382 (Khudyakov et al.), which is hereby incorporated by reference in its entirety, discloses the use of UV lasers (e.g., continuous or pulsed lasers) for curing optical fiber coatings. U.S. Patent Application Publication No. 2003/0026919 (Kojima et al.), which is hereby incorporated by reference in its entirety, discloses the use of ultraviolet light emitting diodes (UVLEDs) for curing optical fiber coatings. The disclosed optical fiber resin coating apparatus includes a mold assembly in which a UV curable resin is coated onto an optical fiber. Also at the mold assembly, the coated optical fiber is exposed to UV radiation from a number of UVLEDs to cure the UV coating. A control circuit may be used to control the UV radiation output from the UVLED array. For example, the control circuit may reduce the current to one or more UVLEDs to reduce the intensity of emitted UV radiation. The control circuit may also be used to vary the intensity of the UV radiation as the optical fiber progresses through the mold assembly. Even so, UVLEDs, though more efficient than mercury lamps, still waste a significant amount of energy in curing glass-fiber coatings. In particular, much of the emitted UV radiation is not used to cure the glass-fiber coatings. Therefore, a need exists for a UVLED apparatus that, as compared with a conventional mercury-lamp device, not only consumes less power and generates less unwanted heat, but also is capable of curing glass-fiber coatings with improved curing efficiency. Accordingly, the invention embraces a UVLED apparatus and associated method for curing in situ optical-fiber coating. The apparatus and method employ one or more UVLEDs that emit electromagnetic radiation into a curing space. An incompletely cured, coated glass fiber passes through the curing space, thereby absorbing electromagnetic radiation to effect curing of the optical-fiber coating. An exemplary UVLED apparatus includes one or more UVLED-mirror pairs. Each UVLED-mirror pair includes one or more UVLEDs capable of emitting electromagnetic radiation and one or more mirrors (or other reflective surfaces) that are capable of reflecting electromagnetic radiation. The UVLED(s) and corresponding mirror(s) are positioned apart from one another so as to define a curing space. As noted, this curing space permits the passage of a coated glass fiber between the UVLED(s) and the mirror(s). Moreover, the mirror(s) are typically positioned opposite corresponding UVLED(s) to efficiently reflect the electromagnetic radiation emitted from the UVLED(s) (and not already absorbed by the glass-fiber coating) into the curing space. In another aspect, the present invention embraces a method for curing a coating on a glass fiber. UV radiation is emitted from one or more sources of electromagnetic radiation toward a curing space. A portion of the emitted UV radiation is transmitted entirely through the curing space. At least some of the UV radiation transmitted entirely through the curing space is reflected toward the curing space (e.g., with a mirror). A glass fiber having an incompletely cured coating is passed through the curing space to effect the absorption of both emitted and reflected UV radiation, thereby curing the coated glass fiber. The foregoing illustrative summary, as well as other exemplary objectives and/or advantages of the invention, and the manner in which the same are accomplished, are further explained within the following detailed description and its accompanying drawings. In one aspect, the present invention embraces an apparatus for curing glass-fiber coatings. The apparatus employs one or more UVLEDs that are configured to emit electromagnetic radiation toward a drawn glass fiber to cure its coating(s), typically polymeric coatings. In this regard, a plurality of UV lamps may be positioned in various configurations, such as an apparatus 10 containing opposing UVLEDs 11 schematically depicted in FIG. 1 and an apparatus 20 containing staggered UVLEDs 11 depicted in FIG. 2. The UVLEDs 11 define a curing space 15 having a central axis 14 along which an optical fiber (i.e., a glass fiber having one or more coating layers) may pass during the curing process. A heat sink 12 may be positioned adjacent to each UVLED 11 to dissipate generated heat energy. A mounting plate 13 provides structural support for the UVLEDs 11. In one exemplary embodiment, the apparatus for curing glass-fiber coatings includes at least a UVLED-mirror pair, which includes (i) an ultraviolet light emitting diode (UVLED) and (ii) a mirror (e.g., a concave mirror) that is positioned to reflect and focus the UV-radiation emitted by the UVLED. FIG. 3 schematically depicts an apparatus 30 containing a plurality of UVLED-mirror pairs. As depicted in FIG. 3, the UVLEDs 11 and corresponding mirrors 16 define a space through which the coated glass fiber can pass (i.e., a curing space 15). This curing space 15 further defines a central axis 14, typically the axis along which a drawn glass fiber passes during the curing process (i.e., the glass fiber's curing axis). Although the central axis 14 is typically vertical, non-vertical (e.g., horizontal) arrangements of the central axis 14 may also be used. Moreover, although the central axis may be centrally positioned in the curing space, a central axis that is not centrally positioned within the curing space is within the scope of the present invention. Each UVLED 11 may be positioned such that it emits UV radiation toward the central axis 14. In this regard, those of ordinary skill will appreciate that the power per unit area (i.e., the radiant flux) emitted by a UVLED decreases exponentially as the distance of the UVLED from the optical fiber increases. Accordingly, each UVLED may be positioned at a distance of between about 1 millimeter and 100 millimeters (e.g., typically between about 5 millimeters and 20 millimeters) from the optical fiber to be cured (e.g., from the central axis). Typically, a UVLED and its corresponding mirror are positioned so that a substantial portion of the UV radiation incident to the optical fiber is substantially perpendicular to the optical fiber (i.e., incident at about a 90 degree angle). Alternatively, the UVLED and/or its corresponding mirror may be positioned at an angle so that most of the UV radiation incident to the optical fiber is incident at an angle other than 90 degrees. FIG. 5 schematically depicts an apparatus 50 containing UVLEDs 11 and mirrors 16 positioned at an angle (i.e., an angle other than 90 degrees relative to the optical fiber and the central axis 14). It may be desirable for the power of the UV radiation incident to the optical fiber to vary as the optical fiber progresses through the apparatus. Varying the power of the UV radiation may aid in the curing of the glass-fiber coating. Depending on the curing properties of a particular coating, it may be desirable to initially expose the optical fiber to high intensity UV radiation. Alternatively, it may be desirable to initially expose the optical fiber to lower intensity UV radiation (e.g., between about 10 percent and 50 percent of the maximum exposure intensity) before exposing the optical fiber to high intensity UV radiation (e.g., the maximum intensity to which the optical fiber is exposed). In this regard, initially exposing the optical fiber to lower intensity UV radiation may be useful in controlling the generation of free radicals in an uncured coating. Those of ordinary skill in the art will appreciate that if too many free radicals are present, many of the free radicals may recombine rather than encourage the polymerization of the glass-fiber coating—an undesirable effect. Varying the intensity of the UV radiation may, for example, be achieved by positioning the UVLED at an angle. As noted, the intensity of the UV radiation incident to a portion of the optical fiber may vary depending upon the distance from that portion to the UVLED. Alternatively, in an apparatus containing a plurality of UVLEDs the intensity of the UV radiation output from the UVLEDs may vary. As noted, each UVLED may be positioned such that it emits UV radiation toward the central axis. That said, it will be appreciated by those of ordinary skill in the art that a UVLED does not emit UV radiation only toward a point or line, but rather emits UV radiation in many directions. Thus, most of the UV radiation emitted by a UVLED will not strike the glass-fiber coating to effect curing. However, in curing an optical fiber coating, it is desirable that as much UV radiation as possible strikes the optical fiber (i.e., a coated glass fiber). In this regard, it will be further appreciated by those of ordinary skill in the art that curing occurs when UV radiation is absorbed by photoinitiators in the glass-fiber coating. Thus, a UVLED may have one or more associated reflectors that focus emitted UV radiation toward the central axis. For example, FIG. 6 depicts a cross-sectional view of a UVLED 11 having an attached reflector 17. The reflector 17 may, for example and as depicted in FIG. 6, have the shape of a rotated teardrop curve. By having a teardrop shape, the reflector focuses UV radiation having various angles of emittance toward the central axis. That said, the present invention embraces UVLEDs with associated reflectors of various shapes (e.g., a spherical, elliptical, cylindrical or parabolic mirror). A UVLED may have one or more lenses attached for focusing UV radiation emitted by the UVLED toward the central axis. Typically, a lens for focusing electromagnetic radiation is convex (e.g., biconvex or plano-convex). In an alternative embodiment, a Fresnel lens may be employed. Moreover, the lens may be selected such that the lens has a focal point at the glass fiber's curing axis (e.g., the central axis). One or more mirrors may be positioned opposite a UVLED (i.e., on the opposite side of the central axis) so as to reflect the UV radiation emitted by the UVLED in the general direction of the central axis. In other words, a UVLED emits UV radiation toward the curing space and the central axis, and its corresponding mirror(s) reflect emitted UV radiation not initially absorbed by optical fiber coatings back toward the central axis (e.g., the glass fiber's curing axis). In this respect, FIG. 3 depicts UVLEDs 11 having a mirror 16 positioned opposite the central axis 14. Typically a mirror will be larger than its corresponding UVLED. For example, a mirror may have a height of between about one inch and 1.5 inches; however, other mirror sizes are within the scope of the present invention. The mirror may be formed from a suitable reflective material. For example, the mirror may be formed from polished aluminum, polished stainless steel, or metalized glass (e.g., silvered quartz). The mirror may have a concave shape (i.e., the mirror is curved inwards toward the curing space) so that the mirror focuses UV radiation emitted by the UVLED toward the central axis. By way of example, a concave mirror may have, inter alia, a cylindrical, elliptical, spherical, or parabolic shape (e.g., a paraboloid or a parabolic cylinder). A concave mirror can focus otherwise lost UV radiation (e.g., UV radiation not initially incident to an optical fiber to be cured) onto an optical fiber for curing, thus limiting the amount of wasted energy. A UVLED-mirror pair as used herein is not limited to a single UVLED paired with a single mirror in a one-to-one relationship. A UVLED-mirror pair may include a plurality of UVLEDs. Alternatively, a UVLED-mirror pair may include a plurality of mirrors. UVLEDs are capable of emitting wavelengths within a much smaller spectrum than conventional UV lamps. This promotes the use of more of the emitted electromagnetic radiation for curing. In this regard, a UVLED for use in the present invention may be any suitable LED that emits electromagnetic radiation having wavelengths of between about 200 nanometers and 600 nanometers. By way of example, the UVLED may emit electromagnetic radiation having wavelengths of between about 200 nanometers and 450 nanometers (e.g., between about 250 nanometers and 400 nanometers). In a particular exemplary embodiment, the UVLED may emit electromagnetic radiation having wavelengths of between about 300 nanometers and 400 nanometers. In another particular exemplary embodiment, the UVLED may emit electromagnetic radiation having wavelengths of between about 350 nanometers and 425 nanometers. As noted, a UVLED typically emits a narrow band of electromagnetic radiation. For example, the UVLED may substantially emit electromagnetic radiation having wavelengths that vary by no more than about 30 nanometers, typically no more than about 20 nanometers (e.g., a UVLED emitting a narrow band of UV radiation mostly between about 375 nanometers and 395 nanometers). It has been observed that a UVLED emitting a narrow band of UV radiation mostly between about 395 nanometers and 415 nanometers is more efficient than other narrow bands of UV radiation. Moreover, it has been observed that UVLEDs emitting UV radiation slightly above the wavelength at which a glass-fiber coating has maximum absorption (e.g., about 360 nanometers) promote more efficient polymerization than do UVLEDs emitting UV radiation at the wavelength at which the glass-fiber coating has maximum absorption. In this regard, although an exemplary UVLED emits substantially all of its electromagnetic radiation within a defined range (e.g., between 350 nanometers and 450 nanometers, such as between 370 nanometers and 400 nanometers), the UVLED may emit small amounts of electromagnetic radiation outside the defined range. In this regard, 80 percent or more (e.g., at least about 90 percent) of the output (i.e., emitted electromagnetic radiation) of an exemplary UVLED is typically within a defined range (e.g., between about 375 nanometers and 425 nanometers). UVLEDs are typically much smaller than conventional UV lamps (e.g., mercury bulbs). By way of example, the UVLED may be a 0.25-inch square UVLED. The UVLED may be affixed to a platform (e.g., a 1-inch square or larger mounting plate). Of course, other UVLED shapes and sizes are within the scope of the present invention. By way of example, a 3-millimeter square UVLED may be employed. Each UVLED may have a power output of as much as 32 watts (e.g., a UVLED having a power input of about 160 watts and a power output of about 32 watts). That said, UVLEDs having outputs greater than 32 watts (e.g., 64 watts) may be employed as such technology becomes available. Using UVLEDs with higher power output may be useful for increasing the rate at which optical fiber coatings cure, thus promoting increased production line speeds. Relative to other UV radiation sources, UVLED devices typically generate a smaller amount of heat energy. That said, to dissipate the heat energy created by a UVLED, a heat sink may be located behind the UVLED (e.g., opposite the portion of the UVLED that emits UV radiation). The heat sink may be one-inch square, although other heat sink shapes and sizes are within the scope of the present invention. The heat sink may be formed of a material suitable for conducting heat (e.g., brass, aluminum, or copper). The heat sink may include a heat exchanger that employs a liquid coolant (e.g., chilled water), which circulates within the heat exchanger to draw heat from the UVLED. Removing heat generated by the UVLED is important for several reasons. First, excess heat may slow the rate at which optical-fiber coatings cure. Furthermore, excess heat can cause the temperature of the UVLED to rise, which can reduce UV-radiation output. Indeed, continuous high-temperature exposure can permanently reduce the UVLED's radiation output. With adequate heat removal, however, the UVLED may have a useful life of 50,000 hours or more. In another exemplary embodiment, the apparatus for curing glass-fiber coatings includes two or more UVLEDs. For instance, the UVLEDs may be arranged in an array (e.g., a planar or non-planar array, such as a three-dimensional array). With respect to a three-dimensional array, two or more UVLEDs may be configured in two or more distinct planes that are substantially perpendicular to the central axis. As depicted in FIGS. 1 and 2, the UVLEDs 11, which emit UV radiation toward the curing space 15, are typically arranged approximately equidistant from the central axis 14. The UVLEDs, for instance, may be arranged to define one or more helixes (i.e., a helical array of UVLEDs). In an array containing more than one helix, the helixes may have the same chirality (i.e., asymmetric handedness). Alternatively, at least one helix may have the opposite chirality (e.g., one helix may be right-handed and a second helix may be left-handed). The three-dimensional array of UVLEDs may define a curing space that is suitable for the passage of a coated glass fiber for curing. As before, the curing space defines a central axis (e.g., the axis along which a drawn glass fiber passes during the curing process). The apparatus employing a three-dimensional array of UVLEDs for curing glass-fiber coatings may also include one or more mirrors for reflecting UV radiation into the curing space. For example, the apparatus may include a plurality of the foregoing UVLED-mirror pairs. By way of further example and as depicted in FIG. 4, a plurality of UVLEDs 11 may be embedded in a mirror 46 (e.g., a mirror in the shape of a circular, elliptical, or parabolic cylinder), the interior of which defines the curing space 15. It will be appreciated by those of skill in the art that the position of the UVLEDs in a three-dimensional array may be defined by the cylindrical coordinate system (i.e., r, θ, z). Using the cylindrical coordinate system and as described herein, the central axis of the curing space defines a z-axis. Furthermore, as herein described and as will be understood by those of ordinary skill in the art, the variable r is the perpendicular distance of a point to the z-axis (i.e., the central axis of the curing space). The variable θ describes the angle in a plane that is perpendicular to the z-axis. In other words and by reference to a Cartesian coordinate system (i.e., defining an x-axis, a y-axis, and a z-axis), the variable θ describes the angle between a reference axis (e.g., the x-axis) and the orthogonal projection of a point onto the x-y plane. Finally, the z variable describes the height or position of a reference point along the z-axis. Thus, a point is defined by its cylindrical coordinates (r, θ, z). For UVLEDs employed in exemplary configurations, the variable r is usually constant. Stated otherwise, the UVLEDs may be positioned approximately equidistant from the central axis (i.e., the z axis). Accordingly, to the extent the variable r is largely fixed, the position of the UVLEDs can be described by the z and θ coordinates. By way of example, a helical UVLED array may have a first UVLED at the position (1, 0, 0), where r is fixed at a constant distance (i.e., represented here as a unitless 1). Additional UVLEDs may be positioned, for example, every 90° (i.e., π/2) with a Δz of 1 (i.e., a positional step change represented here as a unitless 1). Thus, a second UVLED would have the coordinates (1, π/2, 1), a third UVLED would have the coordinates (1, π, 2), and a fourth UVLED would have the coordinates (1, 3π/2, 3), thereby defining a helical configuration. Those having ordinary skill in the art will appreciate that, as used in the foregoing example, the respective distances r and z need not be equivalent. Moreover, those having ordinary skill in the art will further appreciate that several UVLEDs in an array as herein disclosed need not be offset by 90° (e.g., π/2, π, 3π/2, etc.). For example, the UVLEDs may be offset by 60° (e.g., π/3, 2π/3, π, etc.) or by 120° (e.g., 2π/3, 4π/3, 2π, etc.). Indeed, the UVLEDs in an array as discussed herein need not follow a regularized helical rotation. It will be further appreciated by those of ordinary skill in the art that UVLEDs may absorb incident electromagnetic radiation, which might diminish the quantity of reflected UV radiation available for absorption by the glass-fiber coating. See FIG. 1. Therefore, in an apparatus for curing glass-fiber coatings having a plurality of UVLEDs, it may be desirable to position the UVLEDs in a way that reduces UV radiation incident to the UVLEDs. See FIGS. 2, 3, and 4. In an exemplary embodiment described using the cylindrical coordinate system, UVLEDs with a Δθ of π (i.e., UVLEDs positioned on opposite sides of a UVLED array) may be positioned so that they have a Δz that is at least the height of the UVLED. Thus, if each UVLED has a height of 0.5 inch, a UVLED with a Δθ of π should have a Δz of at least 0.5 inch. It is thought that this would reduce the absorption by one UVLED of UV radiation emitted by another UVLED, thereby increasing the availability of UV radiation for reflection by one or more mirrors and absorption by the glass-fiber coating. Alternatively (or in accordance with the foregoing discussion), the UVLEDs may employ a reflective surface (e.g., a surface coating) that promotes reflection of incident electromagnetic radiation yet permits the transmission of emitted electromagnetic radiation. In view of the foregoing, yet another exemplary embodiment employs a plurality of UVLED-mirror pairs that are arranged in a three dimensional configuration. In particular, the plurality of UVLED-mirror pairs share a common curing space defining a common central axis. In an exemplary configuration, the UVLED-mirror pairs may be helically arranged (e.g., configured in a 60°, 90°, or 120° helical array). In yet another exemplary embodiment, the apparatus for curing glass-fiber coatings includes one or more UVLEDs positioned within a cylindrical cavity (or a substantially cylindrical cavity) having a reflective inner surface (e.g., made from stainless steel or silvered quartz, or otherwise including a reflective inner surface). The interior of the cylindrical cavity defines the curing space. As before, the curing space defines a curing axis (e.g., a central axis) along which a drawn glass fiber passes during the curing process. Moreover, one or more UVLEDs may be positioned within the cylindrical cavity such that they emit UV radiation in the direction of the curing axis. In a typical embodiment, the cylindrical cavity has a non-circular elliptical cross-section. In other words, the cylindrical cavity typically has the shape of an elliptic cylinder. For an elliptic cylinder the curing axis may correspond with one of the two line foci defined by the elliptic cylinder. In addition, each UVLED may be positioned along the other line focus such that they emit UV radiation in the general direction of the curing axis. This arrangement is useful for improving curing efficiency, because any electromagnetic radiation that is emitted from one line focus (regardless of direction) will be directed toward the other line focus after being reflected at the inner surface of the cylinder. This principle is illustrated in FIG. 7, which depicts a cross-section of a reflective elliptic cylinder 55 having a first line focus 51 and a second line focus 52. As depicted in FIG. 7, each UV ray 53 emitted from the first line focus 51 will intersect the second line focus 52. That said, the UV radiation emitted from a UVLED is not emitted from a single point. Therefore and because of the small size of a coated glass fiber, it is desirable to use small UVLEDs (e.g., a 3-millimeter square UVLED or a 1-millimeter square UVLED), because a greater percentage of emitted and reflected light from a small UVLED will be incident to the coated glass fiber. In accordance with the foregoing, FIGS. 8-9 depict an exemplary apparatus 60 for curing a coated glass fiber 66. The apparatus 60 includes a substantially cylindrical cavity 65 having an elliptical shape and having a reflective inner surface. The cavity 65 defines a first line focus 61 and a second line focus 62. A plurality of UVLEDs 64 are positioned along the first line focus 61. The second line focus 62 further defines a curing axis along which a coated glass fiber 66 passes so it can be cured. As depicted in FIG. 9, UV rays 63 emitted from the UVLEDs 64 may reflect off the inner surface of the cavity 65 such that the reflected UV rays 63 are incident to the coated glass fiber 66. To facilitate uniform curing of the coated glass fiber 66, some of the UVLEDs 64 may be differently oriented. For example, a second apparatus 70 for curing a glass fiber could have a different orientation than the apparatus 60 (e.g., the second apparatus 70 may have UVLEDs positioned along a line focus 71 that differs from the first line focus 61). In an alternative exemplary embodiment, rather than placing the UVLEDs along one of the line foci, each UVLED may include a lens for focusing emitted UV radiation. In particular, each lens may have a focus at one of the two line foci (e.g., the line focus not defining a curing axis). By including a lens with each UVLED, the efficiency of the apparatus can be further improved. An apparatus as described herein may include a dark space between one or more UVLEDs. In other words, the apparatus may include a space in which substantially no UV radiation is incident to the optical fiber being cured. A pause in the curing process provided by a dark space can help to ensure even and efficient curing of the optical fiber coatings. In particular, a dark space may be useful in preventing too many free radicals from being present in a glass-fiber coating before it is cured (i.e., dark space helps to control free-radical generation). For example, it may be desirable to initially expose an optical fiber to low power UV radiation and then pass the optical fiber through a dark space. After the optical fiber passes through a dark space, it is exposed to higher power UV radiation to complete the curing process. A curing apparatus employing dark space is disclosed in commonly assigned U.S. Pat. No. 7,322,122 for a Method and Apparatus for Curing a Fiber Having at Least Two Fiber Coating Curing Stages, which is hereby incorporated by reference in its entirety. An apparatus for curing glass-fiber coatings may include a control circuit for controlling the UV radiation output from the UVLED. The control circuit may be used to vary the intensity of the UV radiation as an optical fiber progresses through the apparatus. For example, to ensure that the optical fiber receives a consistent dose of ultraviolet radiation, the UV radiation output of the UVLEDs may vary with the speed at which the optical fiber passes through the apparatus. That is to say, at higher speeds (i.e., the speed the optical fiber passes through the apparatus) the output intensity of the UVLEDs may be greater than the output intensity at lower speeds. The output intensity of the UVLEDs may be controlled by reducing (or increasing) the current flowing to the UVLEDs. In another aspect, the present invention embraces a method of employing the foregoing apparatus to cure a coating on a glass fiber (i.e., in situ curing). In an exemplary method, a glass fiber is drawn from an optical preform and coated with a UV curable material. UV radiation is emitted from one or more sources of electromagnetic radiation (e.g., one or more UVLEDs) toward a curing space (e.g., in the general direction of the coated glass fiber). A portion, if not most, of the emitted UV radiation is transmitted entirely through the curing space. Typically, at least some, if not most, of the UV radiation transmitted entirely through the curing space (i.e., at least some UV radiation that has not been absorbed) is reflected (e.g., with a mirror) toward the curing space. A glass fiber having an incompletely cured coating is continuously passed through the curing space to effect the absorption of emitted and reflected UV radiation. The absorption of the UV radiation cures the glass-fiber coating. Moreover, and in accordance with the foregoing, to improve the curing rate, at least a portion of the reflected UV radiation may be focused on the glass fiber (e.g., by using a concave mirror to reflect UV radiation toward the curing space's curing axis). In accordance with the foregoing, the resulting optical fiber includes one or more coating layers (e.g., a primary coating and a secondary coating). At least one of the coating layers—typically the secondary coating—may be colored and/or possess other markings to help identify individual fibers. Alternatively, a tertiary ink layer may surround the primary and secondary coatings. For example, the resulting optical fiber may have one or more coatings (e.g., the primary coating) that comprise a UV-curable, urethane acrylate composition. In this regard, the primary coating may include between about 40 and 80 weight percent of polyether-urethane acrylate oligomer as well as photoinitiator, such as LUCIRIN® TPO, which is commercially available from BASF. In addition, the primary coating typically includes one or more oligomers and one or more monomer diluents (e.g., isobornyl acrylate), which may be included, for instance, to reduce viscosity and thereby promote processing. Exemplary compositions for the primary coating include UV-curable urethane acrylate products provided by DSM Desotech (Elgin, Ill.) under various trade names, such as DeSolite® DP 1011, DeSolite® DP 1014, DeSolite® DP 1014XS, and DeSolite® DP 1016. Those having ordinary skill in the art will recognize that an optical fiber with a primary coating (and an optional secondary coating and/or ink layer) typically has an outer diameter of between about 235 microns and about 265 microns (μm). The component glass fiber itself (i.e., the glass core and surrounding cladding layers) typically has a diameter of about 125 microns, such that the total coating thickness is typically between about 55 microns and 70 microns. With respect to an exemplary optical fiber achieved according to the present curing method, the component glass fiber may have an outer diameter of about 125 microns. With respect to the optical fiber's surrounding coating layers, the primary coating may have an outer diameter of between about 175 microns and about 195 microns (i.e., a primary coating thickness of between about 25 microns and 35 microns) and the secondary coating may have an outer diameter of between about 235 microns and about 265 microns (i.e., a secondary coating thickness of between about 20 microns and 45 microns). Optionally, the optical fiber may include an outermost ink layer, which is typically between two and ten microns thick. In an alternative embodiment, the resulting optical fiber may possess a reduced diameter (e.g., an outermost diameter between about 150 microns and 230 microns). In this alternative optical fiber configuration, the thickness of the primary coating and/or secondary coating is reduced, while the diameter of the component glass fiber is maintained at about 125 microns. By way of example, in such embodiments the primary coating layer may have an outer diameter of between about 135 microns and about 175 microns (e.g., about 160 microns), and the secondary coating layer may have an outer diameter of between about 150 microns and about 230 microns (e.g., more than about 165 microns, such as 190-210 microns or so). In other words, the total diameter of the optical fiber is reduced to less than about 230 microns (e.g., about 200 microns). Exemplary coating formulations for use with the apparatus and method described herein are disclosed in the following commonly assigned applications, each of which is incorporated by reference in its entirety: U.S. Patent Application No. 61/112,595 for a Microbend-Resistant Optical Fiber, filed Nov. 7, 2008, (Overton); International Patent Application Publication No. WO 2009/062131 A1 for a Microbend-Resistant Optical Fiber, (Overton); U.S. Patent Application Publication No. US2009/0175583 A1 for a Microbend-Resistant Optical Fiber, (Overton); and U.S. patent application Ser. No. 12/614,011 for a Reduced-Diameter Optical Fiber, filed Nov. 6, 2009, (Overton). To supplement the present disclosure, this application incorporates entirely by reference the following commonly assigned patents, patent application publications, and patent applications: U.S. Pat. No. 4,838,643 for a Single Mode Bend Insensitive Fiber for Use in Fiber Optic Guidance Applications (Hodges et al.); U.S. Pat. No. 7,623,747 for a Single Mode Optical Fiber (de Montmorillon et al.); U.S. Pat. No. 7,587,111 for a Single-Mode Optical Fiber (de Montmorillon et al.); U.S. Pat. No. 7,356,234 for a Chromatic Dispersion Compensating Fiber (de Montmorillon et al.); U.S. Pat. No. 7,483,613 for a Chromatic Dispersion Compensating Fiber (de Montmorillon et al.); U.S. Pat. No. 7,555,186 for an Optical Fiber (Flammer et al.); U.S. Patent Application Publication No. US2009/0252469 A1 for a Dispersion-Shifted Optical Fiber (Sillard et al.); U.S. patent application Ser. No. 12/098,804 for a Transmission Optical Fiber Having Large Effective Area (Sillard et al.), filed Apr. 7, 2008; U.S. Patent Application Publication No. US2009/0279835 A1 for a Single-Mode Optical Fiber Having Reduced Bending Losses, filed May 6, 2009, (de Montmorillon et al.); U.S. Patent Application Publication No. US2009/0279836 A1 for a Bend-Insensitive Single-Mode Optical Fiber, filed May 6, 2009, (de Montmorillon et al.); U.S. patent application Ser. No. 12/489,995 for a Wavelength Multiplexed Optical System with Multimode Optical Fibers, filed Jun. 23, 2009, (Lumineau et al.); U.S. patent application Ser. No. 12/498,439 for a Multimode Optical Fibers, filed Jul. 7, 2009, (Gholami et al.); U.S. patent application Ser. No. 12/614,172 for a Multimode Optical System, filed Nov. 6, 2009, (Gholami et al.); U.S. patent application Ser. No. 12/617,316 for an Amplifying Optical Fiber and Method of Manufacturing, filed Nov. 12, 2009, (Pastouret et al.) U.S. patent application Ser. No. 12/629,495 for an Amplifying Optical Fiber and Production Method, filed Dec. 2, 2009, (Pastouret et al.); U.S. patent application Ser. No. 12/633,229 for an Ionizing Radiation-Resistant Optical Fiber Amplifier, filed Dec. 8, 2009, (Regnier et al.); and U.S. patent application Ser. No. 12/636,277 for a Buffered Optical Fiber, filed Dec. 11, 2009, (Testu et al.). To supplement the present disclosure, this application further incorporates entirely by reference the following commonly assigned patents, patent application publications, and patent applications: U.S. Pat. No. 5,574,816 for Polypropylene-Polyethylene Copolymer Buffer Tubes for Optical Fiber Cables and Method for Making the Same; U.S. Pat. No. 5,717,805 for Stress Concentrations in an Optical Fiber Ribbon to Facilitate Separation of Ribbon Matrix Material; U.S. Pat. No. 5,761,362 for Polypropylene-Polyethylene Copolymer Buffer Tubes for Optical Fiber Cables and Method for Making the Same; U.S. Pat. No. 5,911,023 for Polyolefin Materials Suitable for Optical Fiber Cable Components; U.S. Pat. No. 5,982,968 for Stress Concentrations in an Optical Fiber Ribbon to Facilitate Separation of Ribbon Matrix Material; U.S. Pat. No. 6,035,087 for an Optical Unit for Fiber Optic Cables; U.S. Pat. No. 6,066,397 for Polypropylene Filler Rods for Optical Fiber Communications Cables; U.S. Pat. No. 6,175,677 for an Optical Fiber Multi-Ribbon and Method for Making the Same; U.S. Pat. No. 6,085,009 for Water Blocking Gels Compatible with Polyolefin Optical Fiber Cable Buffer Tubes and Cables Made Therewith; U.S. Pat. No. 6,215,931 for Flexible Thermoplastic Polyolefin Elastomers for Buffering Transmission Elements in a Telecommunications Cable; U.S. Pat. No. 6,134,363 for a Method for Accessing Optical Fibers in the Midspan Region of an Optical Fiber Cable; U.S. Pat. No. 6,381,390 for a Color-Coded Optical Fiber Ribbon and Die for Making the Same; U.S. Pat. No. 6,181,857 for a Method for Accessing Optical Fibers Contained in a Sheath; U.S. Pat. No. 6,314,224 for a Thick-Walled Cable Jacket with Non-Circular Cavity Cross Section; U.S. Pat. No. 6,334,016 for an Optical Fiber Ribbon Matrix Material Having Optimal Handling Characteristics; U.S. Pat. No. 6,321,012 for an Optical Fiber Having Water Swellable Material for Identifying Grouping of Fiber Groups; U.S. Pat. No. 6,321,014 for a Method for Manufacturing Optical Fiber Ribbon; U.S. Pat. No. 6,210,802 for Polypropylene Filler Rods for Optical Fiber Communications Cables; U.S. Pat. No. 6,493,491 for an Optical Drop Cable for Aerial Installation; U.S. Pat. No. 7,346,244 for a Coated Central Strength Member for Fiber Optic Cables with Reduced Shrinkage; U.S. Pat. No. 6,658,184 for a Protective Skin for Optical Fibers; U.S. Pat. No. 6,603,908 for a Buffer Tube that Results in Easy Access to and Low Attenuation of Fibers Disposed Within Buffer Tube; U.S. Pat. No. 7,045,010 for an Applicator for High-Speed Gel Buffering of Flextube Optical Fiber Bundles; U.S. Pat. No. 6,749,446 for an Optical Fiber Cable with Cushion Members Protecting Optical Fiber Ribbon Stack; U.S. Pat. No. 6,922,515 for a Method and Apparatus to Reduce Variation of Excess Fiber Length in Buffer Tubes of Fiber Optic Cables; U.S. Pat. No. 6,618,538 for a Method and Apparatus to Reduce Variation of Excess Fiber Length in Buffer Tubes of Fiber Optic Cables; U.S. Pat. No. 7,322,122 for a Method and Apparatus for Curing a Fiber Having at Least Two Fiber Coating Curing Stages; U.S. Pat. No. 6,912,347 for an Optimized Fiber Optic Cable Suitable for Microduct Blown Installation; U.S. Pat. No. 6,941,049 for a Fiber Optic Cable Having No Rigid Strength Members and a Reduced Coefficient of Thermal Expansion; U.S. Pat. No. 7,162,128 for Use of Buffer Tube Coupling Coil to Prevent Fiber Retraction; U.S. Pat. No. 7,515,795 for a Water-Swellable Tape, Adhesive-Backed for Coupling When Used Inside a Buffer Tube (Overton et al.); U.S. Patent Application Publication No. 2008/0292262 for a Grease-Free Buffer Optical Fiber Buffer Tube Construction Utilizing a Water-Swellable, Texturized Yarn (Overton et al.); European Patent Application Publication No. 1,921,478 A1, for a Telecommunication Optical Fiber Cable (Tatat et al.); U.S. Pat. No. 7,570,852 for an Optical Fiber Cable Suited for Blown Installation or Pushing Installation in Microducts of Small Diameter (Nothofer et al.); U.S. Patent Application Publication No. US 2008/0037942 A1 for an Optical Fiber Telecommunications Cable (Tatat); U.S. Pat. No. 7,599,589 for a Gel-Free Buffer Tube with Adhesively Coupled Optical Element (Overton et al.); U.S. Pat. No. 7,567,739 for a Fiber Optic Cable Having a Water-Swellable Element (Overton); U.S. Patent Application Publication No. US2009/0041414 A1 for a Method for Accessing Optical Fibers within a Telecommunication Cable (Lavenne et al.); U.S. Patent Application Publication No. US2009/0003781 A1 for an Optical Fiber Cable Having a Deformable Coupling Element (Parris et al.); U.S. Patent Application Publication No. US2009/0003779 A1 for an Optical Fiber Cable Having Raised Coupling Supports (Parris); U.S. Patent Application Publication No. US2009/0003785 A1 for a Coupling Composition for Optical Fiber Cables (Parris et al.); U.S. Patent Application Publication No. US2009/0214167 A1 for a Buffer Tube with Hollow Channels, (Lookadoo et al.); U.S. patent application Ser. No. 12/466,965 for an Optical Fiber Telecommunication Cable, filed May 15, 2009, (Tatat); U.S. patent application Ser. No. 12/506,533 for a Buffer Tube with Adhesively Coupled Optical Fibers and/or Water-Swellable Element, filed Jul. 21, 2009, (Overton et al.); U.S. patent application Ser. No. 12/557,055 for an Optical Fiber Cable Assembly, filed Sep. 10, 2009, (Barker et al.); U.S. patent application Ser. No. 12/557,086 for a High-Fiber-Density Optical Fiber Cable, filed Sep. 10, 2009, (Louie et al.); U.S. patent application Ser. No. 12/558,390 for a Buffer Tubes for Mid-Span Storage, filed Sep. 11, 2009, (Barker); U.S. patent application Ser. No. 12/614,692 for Single-Fiber Drop Cables for MDU Deployments, filed Nov. 9, 2009, (Overton); U.S. patent application Ser. No. 12/614,754 for Optical-Fiber Loose Tube Cables, filed Nov. 9, 2009, (Overton); U.S. patent application Ser. No. 12/615,003 for a Reduced-Size Flat Drop Cable, filed Nov. 9, 2009, (Overton et al.); U.S. patent application Ser. No. 12/615,106 for ADSS Cables with High-Performance Optical Fiber, filed Nov. 9, 2009, (Overton); U.S. patent application Ser. No. 12/615,698 for Reduced-Diameter Ribbon Cables with High-Performance Optical Fiber, filed Nov. 10, 2009, (Overton); U.S. patent application Ser. No. 12/615,737 for a Reduced-Diameter, Easy-Access Loose Tube Cable, filed Nov. 10, 2009, (Overton); U.S. patent application Ser. No. 12/642,784 for a Method and Device for Manufacturing an Optical Preform, filed Dec. 19, 2009, (Milicevic et al.); and U.S. patent application Ser. No. 12/648,794 for a Perforated Water-Blocking Element, filed Dec. 29, 2009, (Parris). In the specification and/or figures, typical embodiments of the invention have been disclosed. The present invention is not limited to such exemplary embodiments. The figures are schematic representations and so are not necessarily drawn to scale. Unless otherwise noted, specific terms have been used in a generic and descriptive sense and not for purposes of limitation.
	abstract	Disclosed herein are zirconium-base alloys excellent in both corrosion resistance and hydrogen absorption property, useful as materials for nuclear reactors. Such a zirconium-base alloy for nuclear reactors comprises 0.5-2 wt. % Sn, 0.07-0.6 wt. % Fe, 0.03-0.2 wt. % Ni, 0.05-0.2 wt. % Cr, and the balance being zirconium and unavoidable impurities, wherein the Fe content (X wt. %) of the zirconium-base alloy and the mean size (Y nm) of precipitates in the zirconium-base alloy are present in a region on the x (Fe content X) and y (mean precipitate size) rectangular coordinates, surrounded by the following five lines: i) Y=xe2x88x92444xc3x97X+154, ii) Y=910xc3x97Xxe2x88x9246, iii) Y=0, iv) Y=300, and v) X=0.6.
	claims	1. An irradiator apparatus for radiation delivery, comprising:a single X-ray source for generating radiation in an X-ray beam to irradiate a product sample;an X-ray filter to filter the radiation generated by the single X-ray source prior to delivery of the generated radiation in the X-ray beam to the product sample;a reflector assembly to reflect radiation delivered by the single X-ray source to the product sample back to the product sample; anda rotation device associated with a sample holder configured to support the product sample and configured to rotate, flip or orient the product sample to a plurality of positions or orientations for delivery of radiation from the single X-ray source to the product sample at each of the plurality of positions or orientations to facilitate a substantially uniform irradiation of the product sample and a substantially uniform radiation exposure delivered to the product sample providing a substantial dose profile uniformity in the irradiated product sample. 2. The irradiator apparatus according to claim 1, wherein the reflector assembly comprises:a first part or portion communicatively associated with the sample holder and the rotation device to position and support the product sample at a corresponding one of the plurality of positions or orientations for delivery of radiation to the product sample; anda second part or portion configured to selectively move out of and into communication with the first part or portion of the reflector assembly to allow a space for rotation or orientation of the product sample to the corresponding one of the plurality of positions or orientations for delivery of radiation to the product sample. 3. The irradiator apparatus according to claim 2, further comprising:a first shaft assembly communicatively associated with the rotation device to selectively rotate the sample holder and the first part or portion of the reflector assembly to drive the rotation device to move the rotation device to position the product sample at the corresponding one of the plurality of positions or orientations for delivery of radiation to the product sample; anda second shaft assembly communicatively associated with the second part or portion of the reflector assembly to drive the selective movement of the second part or portion of the reflector assembly out of and into communication with the first part or portion of the reflector assembly. 4. The irradiator apparatus according to claim 1, wherein:the reflector assembly comprises a material of a low-Z atomic number and a high density material composition to facilitate X-ray radiation reflection and sample irradiation. 5. The irradiator apparatus according to claim 4, wherein:the material of a low-Z atomic number comprises a material selected from the group of consisting of beryllium, boron, carbon, and combination thereof. 6. The irradiator apparatus according to claim 1, wherein:the X-ray filter comprises a metal of a thickness to reduce low-energy X-rays and facilitate less attenuation of the generated X-ray beam to provide a substantially homogeneous dose distribution laterally and at depth throughout the irradiated product sample. 7. The irradiator apparatus according to claim 1, wherein:the X-ray filter comprises copper having a generally flat sheet configuration having a thickness in a range of about 75 microns to 130 microns to facilitate low-energy X-ray beam filtration and beam depth penetration of the irradiated product sample. 8. The irradiator apparatus according to claim 1, wherein:the X-ray filter comprises one or more metallic sheets formed in a step configuration to facilitate a selective spatial filtration of the generated X-ray beam and to attenuate the generated X-ray beam at a beam center to provide a substantially homogeneous and uniform dose distribution in the irradiated product sample. 9. The irradiator apparatus according to claim 8, wherein:the one or more metallic sheets forming the X-ray filter comprise a material selected from the group consisting of aluminum, tungsten, copper and heavy metals, or combinations thereof. 10. The irradiator apparatus according to claim 8, wherein:the X-ray filter comprises a material composition having a single layer configuration or a multi-layer combination configured to provide a substantially homogeneous and uniform dose distribution in the irradiated product sample. 11. An irradiator apparatus for radiation delivery, comprising:a single radiation source for generating radiation in a radiation beam to irradiate a product sample;a filter to filter the radiation generated by the single radiation source prior to delivery of the generated radiation in the radiation beam to the product sample;a reflector assembly to reflect radiation delivered by the single radiation source to the product sample back to the product sample; anda rotation device associated with a sample holder configured to support the product sample and configured to rotate, flip or orient the product sample to a plurality of positions or orientations for delivery of radiation from the single radiation source to the product sample at each of the plurality of positions or orientations to facilitate a substantially uniform irradiation of the product sample and a substantially uniform radiation exposure delivered to the product sample providing a substantial dose profile uniformity in the irradiated product sample. 12. A method for irradiating a product sample, comprising:positioning a product sample to be irradiated in a product sample container or canister;positioning the product sample container or canister in a sample holder at an initial position or orientation in an irradiator apparatus;irradiating the product sample positioned in the product sample container or canister by generating radiation from a single radiation source in the irradiator apparatus and delivering the generated radiation to the product sample positioned at the initial position or orientation;reflecting the radiation delivered to the product sample at the initial position or orientation back to the product sample by a reflector assembly in the irradiator apparatus;selectively positioning the product sample and the product sample container or canister positioned in the sample holder at one or more other positions or orientations;generating radiation by the single radiation source and delivering the generated radiation to the product sample at the one or more other positions or orientations; andreflecting back to the product sample by the reflector assembly the generated radiation delivered to the product sample positioned at a corresponding position or orientation of the one or more other positions or orientations. 13. The method for irradiating a product sample according to claim 12, further comprising:successively providing a plurality of radiation deliveries to the product sample first at the initial position or orientation and then at each of the one or more other positions or orientations to facilitate a substantially uniform irradiation of the product sample and a substantially uniform radiation exposure delivered to the product sample providing a substantial dose profile uniformity in the irradiated product sample. 14. The method for irradiating a product sample according to claim 12, further comprising:selectively moving at least a part the reflector assembly away from the product sample container or canister to position or orient the product sample container or canister at the initial position or orientation or at the one or more other positions or orientations; androtating or orienting the product sample container or canister including the product sample, after at least the part the reflector assembly has moved away from the product sample container or canister, to position the product sample container or canister including the product sample at the initial position or orientation or at a corresponding position or orientation of the one or more other positions or orientations for irradiation of the product sample. 15. The method for irradiating a product sample according to claim 12, further comprising:filtering the radiation generated by the single radiation source by a filter prior to delivery of the radiation to the product sample to facilitate dose uniformity in the irradiated product sample. 16. The method for irradiating a product sample according to claim 12, wherein:the single radiation source comprises an X-ray source. 17. A method for controlling irradiation of a product sample in an irradiator system, comprising:providing a controlled workflow to control a radiation amount to be delivered to a product sample at each of a plurality of radiation deliveries by a single radiation source in a radiation apparatus to provide a total radiation amount delivered to the product sample for the plurality of radiation deliveries having a substantial dose profile uniformity in the irradiated product sample, the radiation apparatus being configured to position the product sample in the radiation apparatus for radiation delivery to the product sample;determining a beam on-time in the controlled workflow for each of the plurality of radiation deliveries corresponding to a radiation amount to be delivered to the product sample at each of the plurality of radiation deliveries by the single radiation source;determining in the controlled workflow a position or orientation of the product sample for each of the plurality of radiation deliveries;synchronizing in the controlled workflow movements of a sample holder configured to hold the product sample to be irradiated, movements of at least a portion of a reflector assembly, the reflector assembly being positioned in the radiation apparatus and configured to reflect back to the product sample the radiation delivered from the single radiation source, and the determined beam on-time for each of the plurality of radiation deliveries to deliver radiation to the product sample at each corresponding one of the plurality of radiation deliveries; andcontrolling the single radiation source in the controlled workflow to deliver radiation to the product sample for each corresponding determined beam on-time for each of the plurality of radiation deliveries at each corresponding determined position or orientation of the product sample to provide a substantially uniform irradiation of the product sample and a substantially uniform radiation exposure delivered to the product sample to provide a substantial dose profile uniformity in the irradiated product sample. 18. The method for controlling irradiation of a product sample in an irradiator system according to claim 17, further comprising:providing on-time control and synchronization of radiation delivery with the sample holder and the reflector assembly movement. 19. The method for controlling irradiation of a product sample in an irradiator system according to claim 17, further comprising:setting a timer to time each set time corresponding to the beam on-time for radiation delivery at each of the plurality of radiation deliveries by the single radiation source. 20. The method for controlling irradiation of a product sample in an irradiator system according to claim 17, further comprising:transferring data for radiation delivery through a network; andstoring the transferred data corresponding to the radiation delivery in a data storage associated with the network. 21. The method for controlling irradiation of a product sample in an irradiator system according to claim 20, further comprising:reporting the data for the radiation delivery through the network to one or more users of the network. 22. The method for controlling irradiation of a product sample in an irradiator system according to claim 20, further comprising:printing at least a portion of the data corresponding to the radiation delivery on a label; andplacing the label having the printed data in association with the irradiated product sample.
	claims	1. A body composition, comprising:boron in an amount of 21-41 atomic percent (at %);iron in an amount of 25-35 at %;chromium in an amount of 2-4 at %;carbon in an amount of 3-10 at %; anda balance of tungsten of the body composition,wherein at least 95 at % of the Fe is in the form of a boride, an intermetallic boride or an intermetallic carbide. 2. The body composition according to claim 1, wherein at least 95 at % of the Cr is in the form of a boride, an intermetallic boride or an intermetallic carbide. 3. The body composition according to claim 1, wherein less than 5 at % of the Fe is in the form of FeCr. 4. A method of producing a body composition, comprising:providing one or ore powders comprising B, Fe, Cr, C and W;milling the one or more powders with an organic binder to obtain a powder mixture;pressing the milled powder mixture; andsintering the pressed powder mixture to obtain a sintered body,wherein the one or more powders comprising the B, the Fe, the Cr, the C and the W includeboron in an amount of 21-41 atomic percent (at %);iron in an amount of 25-35 at %;chromium in an amount of 2-4 at %;carbon in an amount of 3-10 at %; anda balance of tungsten of the body composition,wherein at least 95 at % of the Fe is in the form of a boride, an intermetallic boride or an intermetallic carbide. 5. The method according to claim 4, wherein the sintering step is a reactive sintering process. 6. The method according to claim 4, wherein the boron is added in the form of B4C. 7. The method according to claim 4, wherein the iron and the chromium are added in the form of FeCr. 8. The method according to claim 4, wherein the W is added in the form of W and optionally WC. 9. The method according to claim 8, wherein the amount of the WC added is less than 5 wt %. 10. A method of manufacturing an object for nuclear shielding in a nuclear reactor, comprising:preparing a body composition comprisingboron in an amount of 21-41 atomic percent (at %);iron in an amount of 25-35 at %;chromium in an amount of 2-4 at %;carbon in an amount of 3-10 at %; anda balance of tungsten of the body composition,wherein at least 95 at % of the Fe is in the form of a boride, an intermetallic boride or an intermetallic carbide. 11. A body composition, comprising:boron in an amount of 21-41 atomic percent (at %);iron in an amount of 25-35 at %;chromium in an amount of 2-4 at %;carbon in an amount of 3-10 at %; anda balance of tungsten of the body composition,wherein at least 95 at % of the Cr is in the form of a boride, an intermetallic boride or an intermetallic carbide. 12. The body composition according to claim 11, wherein at least 95 at % of the Fe is in the form of a boride, an intermetallic boride or an intermetallic carbide. 13. The body composition according to claim 11, wherein less than 5 at % of the Fe is in the form of FeCr. 14. A body composition, comprising:boron in an amount of 21-41 atomic percent (at %);iron in an amount of 25-35 at %;chromium in an amount of 2-4 at %;carbon in an amount of 3-10 at %; anda balance of tungsten of the body composition,wherein less than 5 at % of the Fe is in the form of FeCr. 15. The body composition according to claim 14, wherein at least 95 at % of the Fe is in the form of a boride, an intermetallic boride or an intermetallic carbide. 16. The body composition according to claim 14, wherein at least 95 at % of the Cr is in the form of a boride, an intermetallic boride or an intermetallic carbide.
	claims	1. A fuel rod for a nuclear reactor having a bottom end, a top end, and an intermediate region located between the bottom end and the top end, comprising:a first axial zone positioned proximate to the bottom end;a second axial zone positioned adjacent to the first axial zone in the intermediate region; anda third axial zone positioned proximate to the top end,wherein the first axial zone has an enrichment greater than the second axial zone and the second axial zone has an enrichment greater than or equal to the third axial zone, the first axial zone, the second axial zone, and the third axial zone having different levels of at least one of average enrichment and average gadolinium doping,wherein enrichments for each of the first, second, and third axial zones is uniformly distributed axially within each axial zone. 2. The fuel rod of claim 1 wherein the enrichment of the first axial zone is configured to optimize a local peak power. 3. The fuel rod of claim 2 wherein a zone length of the second axial zone and a zone length of the third axial zone are each dimensioned to optimize an R-factor. 4. The fuel rod of claim 2 wherein the enrichment of the second axial zone and the enrichment of the third axial zones are each selected to optimize for an R-factor. 5. The fuel rod of claim 1, further comprising a bottom end axial zone positioned between the bottom end and the first axial zone, wherein the bottom end axial zone includes at least one of natural uranium and an enrichment about equal to the enrichment of the first axial zone. 6. The fuel rod of claim 5, further comprising a top end axial zone positioned between the third axial zone and the top end, wherein the top end zone includes at least one of natural uranium and an enrichment about equal to the enrichment of the first axial zone. 7. The fuel rod of claim 6 wherein the fuel rod has no more than 5 axial zones. 8. The fuel rod of claim 1 wherein a zone length of the second axial zone and a zone length of the third axial zone are dimensioned for optimizing an R-factor. 9. The fuel rod of claim 1 wherein the fuel rod has no more than 3 axial zones having enrichments so as to be distinct from natural uranium. 10. A fuel assembly for a nuclear reactor comprising a plurality of fuel rods wherein one or more fuel rods include a first axial zone positioned generally at a bottom end, a second axial zone positioned adjacent to the first axial zone in an intermediate region, and a third axial zone positioned generally at a top end, wherein the first axial zone has an enrichment greater than the second axial zone and the second axial zone has an enrichment greater than or equal to the third axial zone, the first axial zone, the second axial zone, and the third axial zone having different levels of at least one of average enrichment and average gadolinium doping,wherein the one or more fuel rods is a substantial portion of the fuel rods in the fuel assembly. 11. The fuel assembly of claim 10 wherein the enrichment of the first axial zone for a substantial portion of the one or more fuel rods is configured to optimize a local peak power. 12. The fuel assembly of claim 11 wherein a zone length of the second axial zone and a zone length of the third axial zone for a substantial portion of the one or more fuel rods are each dimensioned for optimizing an R-factor. 13. The fuel assembly of claim 10 wherein only the first axial zones of fuel rods about the edge of the fuel assembly are optimized for local peak power. 14. The fuel assembly of claim 10 wherein a zone length of the first axial zone for a substantial portion of the one or more fuel rods is dimensioned for optimizing the first axial zone for a local peak power, and a zone length of the second axial zone and a zone length of the third axial zone for the substantial portion of the one or more fuel rods are each dimensioned for optimizing an R-factor. 15. The fuel assembly of claim 10, further comprising a bottom end axial zone positioned between the bottom end and the first axial zone in a portion of the one or more fuel rods and a top end axial zone positioned between the third axial zone and the top end in the portion of the one or more fuel rods, wherein both the bottom end and top end axial zones include at least one of natural uranium and an enrichment about equal to the enrichment of the first axial zone. 16. The fuel assembly of claim 15 wherein a substantial portion of the fuel rods includes the bottom end and top end axial zones and does not include more than 5 axial zones. 17. The fuel assembly of claim 10 wherein the one or more fuel rods includes two or more groups of fuel rods each having a different combination of enrichments and zone lengths for the first axial zone, the second axial zone, and the third axial zone. 18. The fuel assembly of claim 17 wherein each of the different combinations of enrichments and zone lengths for the two or more groups is configured to optimize a local peak power and an R-factor for the fuel assembly. 19. The fuel assembly of claim 17 wherein only the first axial zones of fuel rods about an edge of the fuel assembly are optimized for local peak power and only the second axial zones and third axial zones of the edge fuel rods are optimized for R-factor. 20. The fuel assembly of claim 10 wherein the substantial portion of the fuel rods in the fuel assembly have no more than three enrichment zones each. 21. A fuel assembly for a nuclear reactor comprising a plurality of fuel rods wherein one or more fuel rods includes a first axial zone positioned generally at a bottom end, a second axial zone positioned adjacent to the first axial zone in an intermediate region, and a third axial zone positioned generally at a top end, wherein the first axial zone is configured to optimize a local peak power of the first axial zone and the second and third axial zones are configured to optimize an R-factor for the fuel assembly, the first axial zone, the second axial zone, and the third axial zone having different levels of at least one of average enrichment and average gadolinium doping. 22. The fuel assembly of claim 21 wherein the enrichment of the first axial zone within the fuel assembly is greater than the enrichment of the second axial zone and is greater than the enrichment of the third axial zone. 23. The fuel assembly of claim 22 wherein the enrichment of the second axial zone within the fuel assembly is greater than or equal to the enrichment of the third axial zone.
045227829	description	DETAILED DESCRIPTION FIG. 1 shows a fuel assembly constituted by a bundle of parallel fuel rods 1 whose spacing is maintained by means of spacer grids 2 regularly spaced along the fuel assembly and constituting, as FIG. 2 shows, a square-mesh lattice in which some fuel rod 1 locations are occupied by guide tubes 5. The guide tubes 5 are longer than the rods 1, so that these guide tubes can be fixed on the end plates 3 to assure the rigidity of the assembly. The tubes 5 serve both the keep the assembly rigid and to guide the absorbant rods constituting the control rod associated with the assembly. FIG. 3 shows the end of two guide tubes 6 and 7 of a fuel assembly whose upper end plate, or upper tip, is made of stainless steel and whose spacer grids 9 and 10 are made of a nickel alloy with high elasticity. The guide tube 6 is a stainless steel tube and it is connected to the spacer grids 9 and 10 and to the upper tip 8 by direct welding to these parts. The tube 7 is made of zirconium alloy called "Zircaloy" which has a low neutron capture cross section. This tube 7 is simply engaged in the spacers 9 and 10 and in the upper tip 8, inside openings allowing longitudinal displacement of the tube 7 with respect to these spacer grids and to this upper tip, when the tube 7 lengthens under the action of irradiation and the temperature in the reactor. The whole assembly comprises four guide tubes made of stainless steel such as the tube 6 and twenty tubes made of zirconium alloy such as the tube 7. At their lower part, the tubes 7 are fixed to the lower tip of the assembly, by mechanical connection. It is therefore clear that, in an assembly such as that described with reference to FIG. 3, the mass of material with a high neutron capture cross section is therefore reduced, as twenty out of twenty-four guide tubes are made of zirconium alloy with a low neutron capture cross section. In addition, the assembly is very easily constructed, since the four stainless steel tubes can be directly welded to the end tips and the spacer plates, without the need for intermediate parts. In addition, free sliding of the zirconium tubes in the openings provided in the upper tip and in the spacer grids allows expansion of the zirconium tubes under irradiation to be accommodated, without lengthening of the assembly occurring. FIG. 4 shows two guide tubes 11 and 12 engaged at their upper part in the sleeves 13 and 14 fixed by welding to the upper tip 15 and to the upper spacer grid 16 of the assembly. These stainless steel tips allow connection of the upper spacer grid and the upper tip and thus increase the rigidity and strength of the assembly with respect to axial forces. The stainless steel tube 11 is welded at its upper part to the tip 13 and to each of the spacer grids such as 18 disposed at regular intervals over the height of the assembly. The guide tube 12 made if zirconium alloy with a low neutron capture cross section is simply engaged in openings in the spacer grids such as 16 and 18 and in the upper tip 15. At its lower end, the tube 11 bears a plug 19 made of stainless steel which allows it to be fixed by screw means to the lower end plate 20. The lower end grid 60 is connected to the lower tip 20 by a sleeve 61 to which it is welded. The guide tube 11 is inserted in this sleeve. The sleeve is secured to the end plate when the tube is fixed. The lower part of the zirconium alloy tube 12 is fixed to the lower plate 20 by means of a zirconium alloy plug 23 closing the lower end of the tube 12 and allowing the tube 12 to be fixed by screw means to the lower tip 20. In the complete assembly, four stainless steel tubes similar to the tube 11 are used, fixed in the same way to the end tips and the spacer grids. The other twenty tubes used are zirconium alloy tubes such as the tube 12, only fixed to the lower tip 20. FIG. 5 shows that the stainless steel tip 13 is square-sectioned which allows perfect engagement of this tip in the grid 16 and very strong fixing of the tip to the grid 16 without interfering with the fuel rods thereby. Similarly, the tips 14 are square-sectioned, allowing perfect engagement in the cells of the grid 16. FIG. 6 shows an embodiment of the connections between the stainless steel guide tubes and the tip by direct welding of these tubes to the upper tip 15 and to the upper spacer grid 16, without using a sleeve 13. On the other hand, the Zircaloy tubes 12 are engaged in sleeves 14 identical to those described with reference to FIG. 4, assuring connection between the plate and the upper end and the upper spacer grid. This embodiment of the assembly allows the mass os stainless steel and therefore the absorption of neutrons by the structures of this assembly to be reduced. FIG. 7 shows the upper part of a stainless steel guide tube 24, assuring the rigidity of the assembly, in the case of an easily removable fuel assembly, like that described in French Pat. No. 2,368,785. In such an assembly, the upper end of the guide tubes has widened-out, prismatically shaped part 25 which can be introduced into a recess 26 provided in the lower part of the upper tip 27, to fix the tube to this tip, by means of a hollow bushing 28 including a threaded part 29 which screws into a corresponding screw-threaded part provided on the inner surface of the part 25 of the guide tube. The bushing 28 is engaged in an opening passing through the plate 27 and including a shoulder 31 and an upper part 32 of larger diameter than the lower part in which recesses 33 are provided, for locking the bushing rotationally by expansion of the collar constituting the upper part. In this way, the tube is perfectly locked rotationally and fixed rigidly on the upper tip 27. In the case of such a removable assembly, the provision of four stainless steel tubes 24 like those represented in FIG. 7, welded to the spacer grids such as 35, is all that is necessary to obtain rigid assembling of the guide tubes, spacer grids and tips, while retaining the possibility of removing the upper tip by unscrewing a bushing 28. In an assembly with twenty-four guide tubes, the other twenty tubes are made of zirconium alloy and are only fixed to the lower tip of the assembly by plug and screw as described with reference to FIG. 4. The zirconium alloy tubes are also introduced into openings in the spacer grids and the upper tip, allowing them to move in the event of these tubes expanding. When the upper end plate 27 is removed, it is possible to have access to the fuel rods and to extract these selectively from the assembly, to replace or examine them. In all cases, it is preferable to provide Zircaloy tubes which are sufficiently long for these tubes to open as near as possible to the upper face of the upper tip of the assembly. More efficient retention and guiding of these tubes is thus provided. FIG. 8 shows a variant of the fixing of a stainless steel guide tube, in the case of an easily removable assembly. A cylindrical tip 36 having a prismatically shaped upper end 39 engaged in a correspondingly shaped opening provided in the upper tip 38 is fixed. A bushing 37 can be fixed, as before, inside the tip 36 to fix the upper plate 38 with respect to this tip 36. The tip 36 is also fixed by welding to the spacer grids such as 40 and the stainless steel guide tube 41 is fixed by welding to the end of the tip 36, immediately below the first spacer grid 40. In this embodiment, the connection between the upper plate and the first spacer grid 40 is thus strengthened since it is achieved by means of tips 36, of larger diameter than the guide tubes 41, which can be square-sectioned, as in FIG. 4. As in the case of the apparatus described with reference to FIG. 7, the collar constituting the upper part of the bushing 37 can be deformed so as to enter the recesses 43 by expansion and lock the bushing rotationally. FIG. 9 shows a variant in mounting the zirconium alloy tubes 44 in which these are guided and held at their upper part by a sleeve 45 fixed to the upper plate of the assembly 47 through a bushing 48 which also serves to retain and guide the upper part of the tube 44. The sleeve 45 has an upper part 46 whose outer surface is prismatically shaped to engage in a correspondingly shaped opening in the plate 47. The sleeve 45 is also connected by welding to the upper spacer grid 49 and thus forms the connection between the upper end plate 47 and this spacer grid 49. The tubes 44 are connected at their lower part to the lower end plate of the assembly and engage in openings in the various grids of the assembly so that guiding is assured for them, while retaining the possibility of displacement of the zirconium tubes with respect to these spacer grids, in the longitudinal direction. FIG. 10 shows a variant in mounting the zirconium alloy guide tubes 50 in which these have a widening 51 in their upper part in which the lower part of the sleeve 45, fixed on the upper end plate 47 by means of a bushing 48, as described with reference to FIG. 9, then engages. This type of mounting of the zirconium alloy tubes allows sleeves with a greater thickness to be used, with a given inner diameter of guide tube and a given outer diameter of sleeve 45, since the inner diameter of the sleeves can be reduced, as it is no longer required to pass the guide tubes. FIG. 11 shows a fuel assembly according to a second embodiment of the invention. The spacer grids 65, 66, 67 are made of zirconium alloy (Zircaloy). The guide tubes are also made of the same alloy. The guide tube 68 is connected to the upper tip 15 in a similar way to that represented in FIG. 7, i.e., by using a hollow bushing 74 screwing into a corresponding screw-threaded part provided inside the guide tube. Although it has not been specifically represented, connection of the guide tubes 68 to the tip can be obtained as in all the embodiments represented in FIGS. 7, 8, 9, 10, provided that care is taken to use the materials which allow the necessary welded connections to be made. The freely mounted guide tubes 69 are also made of Zircaloy and are introduced into a blind bore 70 in the lower end plate 20 and into a bore 71 in the upper plate. At its upper end, this bore 71 includes a circular projection 73 for fixing the guide tube. FIG. 12 shows an embodiment of the connection between the grid 65 and the tip 15 in which the guide tube 68 is strengthened in this part and has an outer section which is greater than its normal section over a height between its upper end and the lower level of the end grid 65. This section can be circular or square, identical to that represented in FIG. 5. The guide tubes have the same reinforcement in their lower part, between the end grid and the lower tip. It is clear that the apparatus according to the invention allows simplification of the structure and manufacture of the assembly, less longitudinal deformation of this under irradiation, particularly in the embodiment in which the spacer grids are made of stainless steel and a very strong structure to be obtained, despite use of a small mass of material with a high neutron capture cross section. In addition, having only a small number of connections between the guide tubes and the grids allows an improvement in the thermohydraulic conditions, since the cells adjacent to the freely mounted rods in the grids are very well irrigated. The invention is not limited to the embodiments described; it includes all the variants thereof. Other means of connection, removable or not, between the guide tubes and the end plates and other means of connection between the lower ends of these guide tubes and the lower end plate and thus conceivable. The use of materials other than stainless steel for constituting the guide tubes assuring the rigidity of the assembly and other materials than zirconium alloys for the other guide tubes fixed only to the lower plate of the assembly is also conceivable. Lastly, the fuel assembly according to the invention can be used in all nuclear reactors in which the fuel is in the form of very long rods constituting bundles of parallel rods connected together by spacer grids and end plates.
056087667	description	DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS The preferred embodiment of the present invention is a technique to dope stainless steel surfaces with palladium in situ by injecting a palladium-containing compound into the high-temperature water of a BWR while oxide film is forming on the stainless steel surface. Preferably the palladium compound is injected in the form of a solution or suspension at a point upstream of the feedwater inlet. The high temperatures as well as the gamma and neutron radiation in the reactor core act to decompose the compound, thereby freeing palladium species for incorporation in the oxide film as its grows. As used herein, the term "species" means ions or atoms. One Pd-containing compound successfully used for this purpose is an organometallic compound, palladium acetylacetonate. However, other noble metal compounds of organic, organometallic and inorganic nature can also be used for this purpose. The palladium acetylacetonate compound is dissolved in an ethanol/water mixture or in water alone to form a solution or suspension which is injected into the reactor coolant. The palladium gets incorporated into the stainless steel oxide film via a thermal decomposition process of the organometallic compound. As a result of that decomposition, Pd species become available to replace atoms, e.g., Fe atoms, in the oxide film, thereby producing a Pd-doped oxide film on stainless steel. The method of the present invention involves in situ removal of some or all of the oxide film from the surfaces of wetted reactor component and co-deposition of noble metal during subsequent growth of oxide film on the same wetted surfaces. The result is a noble metal-doped oxide film having a relatively longer catalytic life in the reactor operating environment. Incorporation of palladium into the film provides greatly increased catalytic life as compared to palladium coatings which lie on the oxide surface. In accordance with the broad concept of the present invention, several approaches are possible. In the simplest approach, mechanical cleaning (e.g., by flapper wheel or ultra-high-pressure water jet) is used to remove most or all of the oxide film from the reactor component to be treated. Because the oxide film formed on a reactor component reaches a limiting thickness, some portion of the oxide film must be removed before more oxide film, which forms the matrix for the metal dopant, can be grown. After removal of some oxide film, the appropriate aqueous noble metal compound is added to the reactor water prior to initial heat up. This can be accomplished without the nuclear fuel being present by using the recirculation pumps. As the oxide film reforms, palladium will be incorporated into the film. While it is desirable to use the highest possible palladium concentrations consistent with plant and cost considerations, levels in the preferred range of 5 to 100 ppb Pd should be sufficient. In accordance with the preferred method, after the oxide film has been thinned, noble metal doping of newly formed oxide film can be performed at regular intervals to produce a noble metal concentration which varies cyclically in the thickness direction or can be performed continuously to produce a noble metal concentration which is generally constant in the thickness direction. Since mechanical cleaning is expensive, complex and limited to reactor components that are readily accessible, more attractive approaches for preparing the oxidized alloy surfaces include chemical decontamination (which is periodically performed in many plants to reduce the radioactivity, e.g., of piping from Co.sup.60 and other elements which incorporate into the oxide) and exposure to hydrogen water chemistry, which will thin the existing oxide film. Additions of zinc will also reduce the oxide film thickness. However, it may be desirable to halt the zinc additions during the palladium doping process since zinc appears to densify the film. The formation of ZnO on alloy surfaces has been shown to yield many benefits in BWRs, including reduced incorporation of Co.sup.60 in films (thereby lowering the radiation level, e.g., in piping) and reduced susceptibility to SCC. A further aspect of the present invention is that cycling the temperature during the palladium doping process (e.g., by repeatedly raising the water temperature to 550.degree. F. and then cooling the water to 100.degree. F.) should be beneficial, since the solubility of the metal oxides, film thickness and semiconducting properties of the oxide film change with change in temperature. This may be especially valuable following zinc exposure, since zinc desorbs from the oxide films at lower temperatures, providing more sites for the deposition of palladium and more opportunities for film growth. The advantage of the method of the invention, in which the oxide film on alloy surfaces is removed or thinned before palladium deposition, is that palladium is distributed throughout the oxide film in the thickness direction. In contrast, when pre-oxidized alloy surfaces are treated with, e.g., palladium acetylacetonate, the palladium is deposited only on the surface of the oxide. If this deposited palladium is removed from the surface, e.g., by very high flow rates of the reactor coolant, the catalytic response of the surface coating with palladium is decreased, whereas in the case of co-deposition of palladium during oxide film growth, the catalytic response may be sustained due to the presence of palladium species throughout the thickness of the oxide film. Cylindrical coupons of as-machined Type 304 stainless steel were exposed in 288.degree. C. water containing about 300 ppb O.sub.2 for 16 hr. Thereafter, the coupons were exposed in 288.degree. C. water containing about 300 ppb O.sub.2 and 100 ppb Pd as palladium acetylacetonate for 6-8 hours. This cycle was repeated six times. During palladium doping cycles, palladium acetylacetonate was injected. During oxidizing cycles, palladium acetylacetonate was not injected and the palladium acetylacetonate injected during the doping cycle had been removed by the water cleanup system. During the doping cycle, palladium deposits on the high-temperature oxide film and as this oxide films thickens over time, palladium is incorporated throughout the layer of oxide in the thickness direction. However, the palladium concentration in the thickness direction of the oxide film varies as a function of the amount of palladium in the solution in which the coupon is exposed. During this experiment, the incorporation of palladium was observed by depth profiling the Auger electron spectroscopy of the as-exposed surface. The cyclical variation of the palladium doping in the thickness direction can be seen in FIG. 7. The excellent corrosion potential response of this palladium co-deposited specimen is shown in FIG. 8 by the sharp decrease in corrosion potential at H.sub.2 /O.sub.2 molar ratios in the range of about 1.5-2. The method of the present invention can also be used to dope oxide films on reactor components with corrosion-inhibiting non-noble metal. In accordance with this method, the component or structural material is immersed in a solution or suspension of a compound containing the non-noble metal. The non-noble metal must have the property of increasing the corrosion resistance of the stainless steel or other metal surface when incorporated therein or deposited thereon. The selected compound must have the property that it decomposes under reactor thermal conditions to release species of the selected non-noble metal which incorporate in or deposit on the oxide film formed on the stainless steel or other metal surfaces. The non-noble metals which can be used are selected from the group consisting of zirconium, niobium, yttrium, tungsten, vanadium, titanium, molybdenum, chromium and nickel. The preferred compounds in accordance with the invention are those containing zirconium, e.g., the organometallic compounds zirconium acetylacetonate and inorganic compounds zirconium nitrate and zirconyl nitrate. The present invention offers the advantage that alloy surfaces can be doped with palladium or other metal using an in-situ technique (while the reactor is operating) which is simple in application and also inexpensive. However, this technique can also be implemented for coating ex-situ components. In addition, the technique can be applied to operating BWRs and PWRs and their associated components, such as steam generators. The foregoing method have been disclosed for the purpose of illustration. Variations and modifications of the disclosed method will be readily apparent to practitioners skilled in the art of mitigating stress corrosion cracking. For example, noble metals other than palladium can be applied using this technique. The noble metal can be injected in the form of an organic or inorganic compound in conjunction with injection of small amounts of hydrogen to reduce the potential of stainless steel reactor components. One option is to inject the palladium acetylacetonate solution or suspension via the same port by which dissolved hydrogen is injected. The corrosion-inhibiting non-noble metals can be used even in the absence of hydrogen injection. In addition, the doping technique of the invention is not restricted to use with stainless steel surfaces, but also has application in reducing the ECP of other metals which are susceptible to IGSCC, e.g., nickel-based alloys. All such variations and modifications are intended to be encompassed by the claims set forth hereinafter.
061817600	summary	FIELD OF THE INVENTION The present invention relates generally to nuclear reactors, and more particularly to an electrochemical corrosion potential sensor for sensing the electrochemical corrosion potential of materials exposed to high temperature water. BACKGROUND A nuclear power plant includes a nuclear reactor for heating water to generate steam which is routed to a steam turbine. The steam turbine extracts energy from the steam to power an electrical generator which produces electrical power. The nuclear reactor is typically in the form of a boiling water reactor having nuclear fuel disposed in a reactor pressure vessel in which water is heated. The water and steam are carried through various components and piping which are typically formed of stainless steel, with other materials such as iron based alloys and nickel based alloys being used for various components inside the reactor pressure vessel. It has been found that these materials tend to undergo intergranular stress corrosion cracking depending on the chemistry of the material, the degree of sensitization, the presence of tensile stress, and the chemistry of the reactor water. By controlling one or more of these critical factors, it is possible to control the propensity of a material to undergo intergranular stress corrosion cracking. However, it is conventionally known that intergranular stress corrosion cracking may be controlled or mitigated by controlling a single critical parameter called the electrochemical corrosion potential (ECP) of the material. Thus, considerable efforts have been made in the past decade to measure the electrochemical corrosion potential of the materials of interest during operation of the reactor. This measurement, however, is not a trivial task, because the electrochemical corrosion potential of the material varies depending on the location of the material in the reactor circuit. As an example, a material in the reactor core region is likely to be more susceptible to radiation assisted stress corrosion cracking than the same material exposed to an out-of-core region. The increased susceptibility occurs because the material in the core region is exposed to the highly oxidizing species generated by the radiolysis of water by both gamma and neutron radiation under normal water chemistry conditions in addition to the effect of direct radiation assisted stress corrosion cracking. The oxidizing species increase the electrochemical corrosion potential of the material, which in turn increases its propensity to undergo intergranular stress corrosion cracking or radiation assisted stress corrosion cracking. Thus, a suppression of the oxidizing species is desirable in controlling intergranular stress corrosion cracking. An effective method of suppressing the oxidizing species coming into contact with the material involves the injection of hydrogen into the reactor water via the feedwater system so that recombination of the oxidants with hydrogen occurs within the reactor circuit. The recombination results in an overall reduction in the oxidant concentration present in the reactor which in turn mitigates intergranular stress corrosion cracking of the materials if the oxidant concentration is suppressed to low levels. This method is conventionally called hydrogen water chemistry and is widely practiced for mitigating intergranular stress corrosion cracking of materials in boiling water reactors. When hydrogen water chemistry is practiced in a boiling water reactor, the electrochemical corrosion potential of the stainless steel material typically decreases from a positive value generally in the range of 0.050 to 0.200 V (SHE) under normal water chemistry to a value less than -0.230 V (SHE), where SHE stands for the standard hydrogen electrode. There is considerable evidence that when the electrochemical corrosion potential is below -0.230 V (SHE), intergranular stress corrosion cracking of materials such as stainless steel can be mitigated, and the initiation of intergranular stress corrosion cracking can be largely prevented. Thus, considerable efforts have been made to develop reliable electrochemical corrosion potential sensors to be used as reference electrodes for determining the electrochemical corrosion potential of operating surfaces. These sensors are being used in boiling water reactors worldwide, with a high degree of success, which has enabled the determination of the minimum feedwater hydrogen injection rate required to achieve electrochemical corrosion potentials of reactor internal surfaces and piping below the desired negative value, -0.230 mV (SHE). However, the sensors typically have a limited lifetime, in that some have failed after only a few months of use, while most have shown evidence of successful operation for approximately six to nine months. Only a few sensors have shown successful operation over a period of one fuel cycle, e.g. eighteen months in a U.S. boiling water reactor. Recent experience with boiling water reactors in the United States has shown that the two major modes of failure of the sensor have been cracking and corrosive attack in the ceramic-to-metal braze used at the sensor tip, and the dissolution of the sapphire insulating material used to electrically isolate the sensor tip from the metal conductor cable for platinum or stainless steel type sensors. The electrochemical corrosion potential sensors may be mounted either directly in the reactor core region for directly monitoring electrochemical corrosion potential of in-core surfaces, or may be mounted outside the reactor core to monitor the electrochemical corrosion potential of out-of-core surfaces. However, the typical electrochemical corrosion potential sensor nevertheless experiences a severe operating environment in view of the high temperature of water, typically exceeding 288.degree. C., relatively high flow rates, e.g up to several meters per second (m/s) or more, and the effects of high nuclear radiation in the core region. This environment complicates the design of the sensor, since suitable materials are required for this hostile environment, preferably configured to provide a water-tight assembly for a beneficial useful lifetime. As indicated above, experience with the typical platinum electrochemical corrosion potential sensor has uncovered shortcomings leading to premature failure before expiration of a typical fuel cycle. Accordingly, it is desired to improve the design of electrochemical corrosion potential sensors to increase their useful life, e.g. to at least one fuel cycle. SUMMARY The invention relates to a sensor for a measuring an electrochemical corrosion potential comprising a sensor tip, a conductor electrically connected to the sensor tip, an insulating member which surrounds the conductor, a connecting member which surrounds the conductor, and a sleeve which fits over the sensor tip, the insulating member, and the connecting member, the sleeve having inner threads which engage with corresponding outer threads on at least one of the sensor tip and the connecting member. The invention also relates to a method of making an electrochemical corrosion potential sensor comprising the steps of providing a sensor tip, connecting a conductor to the sensor tip, providing an insulating member around the conductor, providing a connecting member around the conductor, providing a sleeve which fits over the insulating member, a portion of the connecting member, and a portion of the sensor tip, forming inner threads on the sleeve, forming outer threads on at least one of the sensor tip and the connecting member, and engaging the inner threads with the outer threads. The sensor sleeve can be preformed to have a high mechanical strength and high density, which provides excellent protection to the insulating member and braze joints of the sensor in the high temperature water environment. Exemplary embodiments of the sensor typically have a significantly increased lifetime which allows data on electrochemical corrosion potential to be acquired over a complete fuel cycle.
	abstract	A material is exposed to a neutron flux by distributing it in a neutron-diffusing medium surrounding a neutron source. The diffusing medium is transparent to neutrons and so arranged that neutron scattering substantially enhances the neutron flux to which the material is exposed. Such enhanced neutron exposure may be used to produce useful radio-isotopes, in particular for medical applications, from the transmutation of readily-available isotopes included in the exposed material. It may also be used to efficiently transmute long-lived radioactive wastes, such as those recovered from spent nuclear fuel. The use of heavy elements, such as lead and/or bismuth, as the diffusing medium is particularly of interest, since it results in a slowly decreasing scan through the neutron energy spectrum, thereby permitting very efficient resonant neutron captures in the exposed material.
	claims	1. A mechanism for supporting an apertured plate, comprising:a plate holder for holding the apertured plate provided with an aperture through which a beam of charged particle passes;a holder support for supporting the plate holder;grooves extending radially in plural directions about the center axis of said aperture and formed in a supported surface of the plate holder or a supporting surface of the holder support; andconvex portions engaged in the grooves and formed on the other of the supported surface or the supporting surface. 2. A mechanism for supporting an apertured plate as set forth in claim 1, wherein said grooves extend in three directions about the center axis of said aperture. 3. A mechanism for supporting an apertured plate as set forth in claim 1, wherein each of said grooves has a V-shaped cross section. 4. A mechanism for supporting an apertured plate as set forth in claim 1, wherein each of said convex portions is made of a spherical member. 5. A mechanism for supporting an apertured plate as set forth in claim 1, wherein said grooves are formed in the supported surface of the plate holder, while said convex portions are formed on the supporting surface of the holder support. 6. A charged-particle beam instrument having a charged-particle beam source for emitting a beam of charged particles, an apertured plate provided with an aperture through which the beam passes, a plate support mechanism for supporting the apertured plate, and an illumination system for directing the beam passed through the aperture at a substrate,wherein said plate support mechanism has a plate holder for holding the apertured plate and a holder support for supporting the plate holder,wherein grooves extending radially in plural directions about the center axis of said aperture are formed in a supported surface of the plate holder or a supporting surface of the holder support, andwherein convex portions engaged in the grooves are formed on the other of the supported surface or the supporting surface. 7. A charged-particle beam instrument as set forth in claim 6, wherein said grooves in said plate support mechanism extend in three directions about the center axis of said aperture. 8. A charged-particle beam instrument as set forth in claim 6, wherein each of said grooves in said plate support mechanism has a V-shaped cross section. 9. A charged-particle beam instrument as set forth in claim 6, wherein each of said convex portions of said plate support mechanism is made of a spherical member. 10. A charged-particle beam instrument as set forth in claim 6, wherein said grooves in said plate support mechanism are formed in the supported surface of the plate holder, while said convex portions of said plate support mechanism are formed on the supporting surface of the holder support.
	abstract	A modular x-ray lens system for use in directing x-rays comprising a radiation source which generates x-rays and a lens system which directs the x-rays, wherein the x-ray lens system may be configured to focus x-rays to a focal point and vary the intensity of said focal point.
	abstract	A method of producing actinium by using liquefied radium includes producing Ac-225 using Ra-226 of a liquefied state, moving the produced Ac-225 in a liquefied state after Ac-225 is produced, and separating Ac-225 and reusing Ra-226. As a result, a nuclear reaction process of Ac-225 may be performed and loss of Ra-226 may be minimized. Further, such a method may improve safety by including a radon collection unit which is capable of discharging and isolating radon produced from Ra-226, thereby preventing radiation exposure due to radon.
041475917	claims	1. A nuclear reactor fuel assembly comprising, a wrapper tube for flowing coolant therethrough, a plurality of cladding tubes filled with fuel material, said cladding tubes being provided with wound wire spacers and arranged within said wrapper tube, spacing means mounted at upper end portions of said cladding tubes which are adjacent to said wrapper tube for keeping said cladding tubes from contact with said wrapper tube during the reactor operation. 2. A nuclear reactor fuel assembly as claimed in claim 1, wherein said spacing means are formed by thickening said cladding tubes at said upper end portions to a thickness which exceeds the thickness of remaining portions of said cladding tubes. 3. A nuclear reactor fuel assembly as claimed in claim 1, wherein said spacing means comprise ring-shaped spacers. 4. A nuclear reactor fuel assembly as claimed in claim 1, wherein said spacing means are formed by projections at said upper end portions of said cladding tubes adjacent to said wrapper tube, said projections extending outward beyond remaining portions of said cladding tubes. 5. A nuclear reactor fuel assembly as claimed in claim 1, wherein the nuclear reactor comprises a fast breeder reactor. 6. A nuclear reactor fuel assembly as claimed in claim 1, wherein said spacing means are separate from said wire spacers, and are mounted at said upper end portions of all cladding tubes which are adjacent to said wrapper tube. 7. A nuclear reactor fuel assembly as claimed in claim 1, wherein said spaceing means are mounted at said upper end portions of only thos cladding tubes which are adjacent to said wrapper tube. 8. In a fuel assembly for a nuclear reactor of the type comprising a plurality of cladding tubes filled with fuel material, said plurality of cladding tubes being assembled within a wrapper tube, and said plurality of cladding tubes each being of the type which are helically wound with a wire spacer, the improvement comprising separate spacing means provided at upper end portions of those cladding tubes which are adjacent to said wrapper tube for preventing said cladding tubes from coming into direct contact with said wrapper tube. 9. A fuel assembly according to claim 8, wherein said spacing means comprise ring spacers mounted over said upper end portions of said cladding tubes adjacent said wrapper tube. 10. A fuel assembly according to claim 8, wherein said spacing means comprise upper end portion wall thickness of said cladding tubes which are greater than remaining portions of said cladding tubes. 11. A fuel assembly according to claim 8, wherein said spacing means comprise salient projections at said upper end portions of said cladding tubes. 12. A fuel assembly according to claim 8, wherein said nuclear reactor is a fast breeder reactor.
053944461	description	DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS In accordance with a preferred embodiment of the invention (see FIGS. 3 and 4), the uncoupling rod centering gauge 100 comprises a platelike cross handle 102 having a generally square-shaped central portion and four arms 102a-102d extending therefrom. The arms are disposed at right angles relative to each other. The arms are sized such that the gauge cannot be inadvertently left in when reinstailing the CRD after maintenance. The central portion of cross handle 102 further has a pair of concentric rings 104, 106 welded to bottom surface and extending downward. Cross handle 102 sits atop spud 46 with guide ring 104 surrounding the fingers of spud 46 and with centering ring 106 projecting inside the spud. Centering ring 106 has an outer circumferential surface which provides support for the fingers on the inside of the spud by blocking radially inwardly directed flexure of the spud fingers. The annulus between rings 104 and 106 is dimensioned such that the spud fingers reside therein in a flexed state, the return spring force of the spud fingers acting to grip the centering ring 106, whereby the gauge 100 is held securely in place atop the spud. This process ensures that the spud fingers have not been damaged and are concentric. Centering ring 106 further has a circular cylindrical centering bore 108, which continues through cross handle 102. The diameter of centering bore 108 is slightly greater than the diameter of rod 48 of the uncoupling rod. When the uncoupling rod centering gauge is correctly seated atop the spud, free axial movement of rod 48 inside centering bore 108 indicates that the uncoupling rod is correctly installed in the spud. As shown in FIG. 5, the spud 46 has a clover-leaf shaped through-hole consisting of a central hole for receiving the uncoupling rod 48 (which is welded inside tube 43) and three lobes 114, commonly referred to as the "spud flow holes". The spud flow holes are arranged to intersect the central hole at equal angular intervals along the periphery thereof. The radius of each spud flow hole 114 is slightly greater than the radius of center hole. Since the diameter of tube 43 of the uncoupling rod is such that it fits snugly inside the center hole, it is true, albeit undesirable, that the uncoupling rod also fits inside any one of the spud flow holes. Thus, incorrect installation of the uncoupling rod is possible. The gauge of the present invention is designed to ensure correct installation of the uncoupling rod. The gauge is mounted on the spud after the uncoupling rod has been inserted. FIG. 3 shows an uncoupling rod (consisting of rod 48 and tube 43 welded together) which has been correctly inserted in the center hole 112 of spud 46. In this case, the uncoupling rod is free to slide up and down in the centering bore 108 of gauge 100 without binding. In the event that the uncoupling rod had been mistakenly inserted in one of the spud flow holes, binding of rod 48 inside centering bore 108 would occur, indicating incorrect installation. The gauge would then be removed and the uncoupling rod would be withdrawn from the spud flow hole and then re-inserted into the center hole of the spud. After assembly is complete, the CRD must be transported to the under vessel area. In the assembled state, spud 46 protrudes from the outer filter 45, exposing it to possible damage during CRD transport. Therefore, it is desirable to leave the uncoupling rod centering gauge 100 in place to protect the spud during transfer. The uncoupling rod centering gauge in accordance with the preferred embodiment of the invention has a cross handle for ease in handling. The outer perimeter of the gauge as defined by the ends of radially outwardly projecting arms 102a-102d has a radius such that the CRD cannot be installed inside the reactor if the gauge has not been removed from the spud. The preferred embodiment of the uncoupling rod centering gauge has been disclosed for the purpose of illustration. Variations and modifications of the disclosed structure which do not depart from the concept of this invention will be readily apparent to mechanical engineers skilled in the art of tooling. All such variations and modifications are intended to be encompassed by the claims set forth hereinafter.
	claims	1. A specimen box for an electron microscope, comprising:a first substrate, which has a first surface, a second surface, a first concave, and one or more first through holes, wherein the first concave is disposed on the second surface, a first thin film corresponding to the first concave is disposed on the first surface, and the first through hole is disposed around the first concave and penetrates through the first substrate;a second substrate, which has a third surface, a fourth surface, and a second concave, wherein the second concave is disposed on the fourth surface, and a second thin film corresponding to the second concave is disposed on the third surface;a metal adhesion layer, which is disposed between the first substrate and the second substrate; andone or more photoelectric elements, which comprises one or more ends, and the photoelectric element is disposed between the first substrate and the second substrate;wherein a space is formed by the first substrate, the second substrate, and the metal adhesion layer, and the end is disposed in the space. 2. The specimen box for an electron microscope as claimed in claim 1, wherein the first through hole penetrates through the first thin film. 3. The specimen box for an electron microscope as claimed in claim 1, wherein the metal adhesion layer is disposed between the second surface and the fourth surface. 4. The specimen box for an electron microscope as claimed in claim 1, wherein the metal adhesion layer is disposed between the second surface and the second thin film. 5. The specimen box for an electron microscope as claimed in claim 1, wherein the photoelectric element is independently a photo fiber or an electrode. 6. The specimen box for an electron microscope as claimed in claim 1, wherein the second substrate further comprises one or more second through holes, wherein the second through hole is disposed around the second concave and penetrates through the second substrate. 7. The specimen box for an electron microscope as claimed in claim 1, wherein the hole size of the first through hole is 10 μm to 1000 μm. 8. The specimen box for an electron microscope as claimed in claim 6, wherein the hole size of the second through hole is 10 μm to 1000 μm. 9. The specimen box for an electron microscope as claimed in claim 1, wherein the specimen box further comprises one or more plugs assembled into the first through holes. 10. The specimen box for an electron microscope as claimed in claim 6, wherein the specimen box further comprises one or more plugs assembled into the second through holes. 11. The specimen box for an electron microscope as claimed in claim 1, wherein the material of the first thin film and the second thin film is independently silicon dioxide (SiO2), silicon nitride (Si3N4), or a combination thereof. 12. The specimen box for an electron microscope as claimed in claim 1, wherein the thickness of the first thin film and the second thin film is independently 1 nm to 100 nm. 13. The specimen box for an electron microscope as claimed in claim 1, wherein a first protective layer is disposed on the surface of the first thin film. 14. The specimen box for an electron microscope as claimed in claim 13, wherein the material of the first protective film is silicon nitride (Si3N4). 15. The specimen box for an electron microscope as claimed in claim 1, wherein a second protective layer is disposed on the surface of the second thin film. 16. The specimen box for an electron microscope as claimed in claim 15, wherein the material of the second protective film is silicon nitride (Si3N4). 17. The specimen box for an electron microscope as claimed in claim 1, wherein the material of the first substrate and the second substrate is independently silicon substrate, glass substrate, or polymer substrate. 18. The specimen box for an electron microscope as claimed in claim 1, wherein the thickness of the first substrate and the second substrate is independently 10 μm to 1000 μm. 19. The specimen box for an electron microscope as claimed in claim 1, wherein the metal adhesion layer comprises a metal material, wherein the metal material is selected from a group consisting of Ti, Cr, Sn, In, Bi, Cu, Ag, Ni, Zn, Au, and Ti—W alloy. 20. The specimen box for an electron microscope as claimed in claim 1, wherein the material of the plugs is selected from a group consisting of Ni—Ti alloy, copper-base alloy, Cu—Zn alloy, Cu—Al—Mn alloy, Cu—Al—Ni alloy, Cu—Al—Be alloy, Cu—Al—Be—Zr alloy, and Cu—Al—Ni—Be alloy. 21. The specimen box for an electron microscope as claimed in claim 1, wherein the volume of the space is 0.01 mm3 to 100 mm3.
	summary
	summary
061852686	abstract	A main steam pressure disturbance preventing apparatus of a nuclear power plant including, one of a pressure detector and a water level detector provided in one of a steam system from a reactor of the nuclear power plant and a drain system connected to the steam system, an instrumentation pipe connected between one of the pressure detector and the water level detector and one of the steam system and the drain system, and catalyst installed in one of the pressure detector, the water level detector and the instrumentation pipe for recombining hydrogen and oxygen.
	abstract	An object of the present invention is to provide an accumulator including a flow damper which is capable of performing a control so that a vortex may not be formed in a vortex chamber at the time of a large flow injection without requiring huge labors and fabrication costs. The flow damper is configured of a colliding jet controller (a bevel or a projection) for controlling a colliding jet composed of a jet from a large flow pipe and a jet from a small flow pipe flowing into a vortex chamber at the time of a large flow injection so that the colliding jet may proceed directly to an outlet without forming a vortex in the vortex chamber. The colliding jet controller is provided at a junction of an inner surface of the small flow pipe and an inner surface of the vortex chamber.
	summary
	claims	1. A nuclear power plant controlling system, comprising:a first signal receiver configured to receive a first signal, the first signal being acquired at a sampling interval having a shorter time period than that of a calculation time of a second signal and also being compensated by a compensation formula to imitate the second signal such that an error value of a compensated first signal against the second signal is positive;a prospective time deriving unit configured to derive a prospective time for the compensated first signal to reach a full limit, the full limit suspending an automatic control of reactor power with an automatic output controller;a judging unit configured to judge whether a remaining time until the prospective time exceeds a preset value set for a longer time than the calculation time of the second signal, and then to output a request signal for starting calculation of the second signal by a core performance calculator-configured to calculate a power distribution in a reactor core if the remaining time until the prospective time exceeds the preset value;a compensating unit configured to compensate the first signal based on the compensation formula as refined with the calculated second signal calculated in response to the request signal;a first instruction transmitter configured to transmit a first instruction to the automatic output controller to vary a rate factor of the compensated first signal and the second signal by synchronizing with the refined compensation formula so that a power rate of a nuclear reactor is reduced; anda second instruction transmitter configured to transmit a second instruction to the automatic output controller to hold the reactor power after the compensated first signal or the second signal reaches the full limit or a threshold that is just before the full limit,wherein the first signal is outputted from a low power range monitoring (LPRM) detector configured to detect neutrons,wherein the second signal is a maximum linear heat generation ratio or a minimum critical power ratio transmitted from the core performance calculator,wherein the first instruction is a power control instruction, which instructs a control rod operating unit and a recirculation flow operating unit so as to reduce the power rate of the nuclear reactor, andwherein the second instruction is a suspend instruction, which instructs the control rod operating unit and the recirculation flow operating unit so as to suspend the automatic control of reactor power. 2. The nuclear power plant controlling system according to claim 1, further comprising:a rate varying unit configured to vary the rate factor based on another derived prospective time, said another derived prospective time being a time for the second signal to reach the threshold.
	summary
	abstract	The X-ray fluoroscopic imaging system of the present invention comprises: an inspection passage; an electron accelerator; a shielding collimator apparatus comprising a shielding structure, and a first collimator for extracting a low energy planar sector X-ray beam and a second collimator for extracting a high energy planar sector X-ray beam which are disposed within the shielding structure; a low energy detector array for receiving the X-ray beam from the first collimator; and a high energy detector array for receiving the X-ray beam from the second collimator. The first collimator, the low energy detector array and the target point bombarded by the electron beam are located in a first plane; and the second collimator, the high energy detector array and the target point bombarded by the electron beam are located in a second plane.
	claims	1. A method for modulating an energy of a particle beam, comprising:providing a first modulator that is located closer to a particle source than a nozzle;providing a second modulator that is located closer to the nozzle than the particle source;operating the first modulator when an energy of the beam is desired to be decreased; andoperating the second modulator when the energy of the beam is desired to be increased. 2. The method of claim 1, further comprising rotating the nozzle relative to a target region, wherein the operation of the first modulator is synchronized with the rotation of the nozzle. 3. The method of claim 2, wherein the operation of the second modulator is synchronized with the rotation of the nozzle. 4. The method of claim 1, wherein the act of operating the first modulator comprises inserting one or more blocks into a path of the particle beam. 5. The method of claim 4, further comprising:determining an amount of energy to be adjusted for the particle beam; andselecting the one or more blocks based on the determined amount of energy to be adjusted. 6. The method of claim 5, wherein the act of selecting the one or more blocks comprises performing a table lookup based on the determined amount of energy to be adjusted. 7. The method of claim 5, wherein the act of selecting the one or more blocks comprises performing a direction calculation. 8. The method of claim 1, wherein the act of operating the second modulator comprises removing one or more blocks from a path of the particle beam. 9. The method of claim 8, further comprising:determining an amount of energy to be adjusted for the particle beam; andselecting the one or more blocks based on the determined amount of energy to be adjusted. 10. The method of claim 9, wherein the act of selecting the one or more blocks comprises performing a table lookup based on the determined amount of energy to be adjusted. 11. The method of claim 1, further comprising providing a shield for protecting a patient from being irradiated by neutrons generated as a result of the operation of the second modulator. 12. The method of claim 1, further comprising cooling a component of the first modulator. 13. The method of claim 1, further comprising cooling a component of the second modulator. 14. The method of claim 1, wherein the first modulator comprises a plurality of blocks, the plurality of blocks having respective thicknesses that form a logarithmic pattern. 15. The method of claim 1, wherein the second modulator comprises a plurality of blocks, the plurality of blocks having respective thicknesses that form a logarithmic pattern. 16. The method of claim 1, wherein the first modulator comprises a first solid block, and the second modulator comprises a second solid block. 17. An energy modulator for use with a collimator and a particle source that provides a beam of particles, comprises:a first solid block moveable between a first position and a second position, wherein when the first solid block is at the first position, it is out of a path of the beam, and wherein when the first solid block is at the second position, it is in the path of the beam; anda second solid block moveable relative to the first block, wherein the second block and the first block are offset from each other in a direction of the beam;wherein the first block has a first energy absorption characteristic, and the second block has a second energy absorption characteristic that is different from the first energy absorption characteristic; andwherein the first solid block and the second solid block are located upstream from the collimator. 18. The energy modulator of claim 17, wherein the first solid block has a first thickness, and the second solid block has a second thickness that is different from the first thickness. 19. The energy modulator of claim 17, wherein the first solid block has a thickness, and the second solid block has a thickness that is two times the thickness of the first solid block. 20. The energy modulator of claim 17, wherein the first solid block is made from a first material, and the second block is made from a second material that is different from the first material. 21. The energy modulator of claim 17, wherein the first solid block is made from a first material and has a first thickness, and the second solid block is made from a second material and has a second thickness, the second material being different from the first material, and the second thickness being different from the first thickness. 22. The energy modulator of claim 17, wherein the first solid block is made from a material that is at least partially transparent to the beam. 23. The energy modulator of claim 17, further comprising a third solid block, wherein the first, second, and third solid blocks are offset relative to each other in a direction of the beam. 24. The energy modulator of claim 23, wherein the first solid block has a first thickness in a direction of the beam, the second solid block has a second thickness in the direction of the beam, and the third solid block has a third thickness in the direction of the beam; andwherein the first thickness, the second thickness, and the third thickness are different from each other. 25. The energy modulator of claim 24, wherein the third solid block has a thickness that is four times the thickness of the first solid block, and the second solid block has a thickness that is two times the thickness of the first solid block. 26. The energy modulator of claim 24, wherein the first thickness, the second thickness, and the third thickness form a non-linear pattern. 27. The energy modulator of claim 17, further comprising a positioner for moving the first solid block. 28. The energy modulator of claim 17, wherein a surface of the first solid block is perpendicular to the beam. 29. The energy modulator of claim 17, further comprising a mounting structure to which the first and the second solid block are slidably mounted. 30. The energy modulator of claim 29, wherein the mounting structure is mounted to a particle delivery system having the particle source, a particle transport system, and a nozzle. 31. The energy modulator of claim 30, wherein the mounting structure is mounted to the particle delivery system such that the mounting structure is closer to the particle source than the nozzle. 32. The energy modulator of claim 17, wherein the first solid block is made from a first material, the second solid block is made from a second material, and the second material has a Z value that is less then a Z value of the first material. 33. The energy modulator of claim 17, further comprising a cooling system coupled to the first solid block, the second solid block, or both. 34. The energy modulator of claim 17, further comprising an energy sensor, and a control coupled to the energy sensor, wherein the control is configured to adjust a position of the first solid block based on a feedback signal provided by the energy sensor. 35. The energy modulator of claim 17, wherein the particle source comprises a proton source. 36. The energy modulator of claim 17, wherein the first solid block and the second solid block are operable to attenuate the beam before the beam reaches the collimator. 37. The energy modulator of claim 17, further comprising a first end coupled to the particle source, and a second end coupled to a beam transport component. 38. The energy modulator of claim 37, wherein the first end is directly coupled to the particle source. 39. A method for modulating an energy of a particle beam, comprising:determining information regarding a desired particle beam energy;providing a set of at least three solid blocks, wherein the at least three solid blocks in the set are offset from each other in a direction of the beam, and are located upstream from a collimator;selecting one or more of the solid blocks from the set to be placed in a path of the beam based on the determined information; andplacing the selected one or more of the solid blocks in the path of the beam. 40. The method of claim 39, wherein the particle beam comprises a proton beam. 41. The method of claim 39, wherein each of the at least three solid blocks in the set has a thickness that is different from the remaining solid blocks in the set. 42. The method of claim 41, wherein one of the solid blocks in the set has a thickness that is two times a thickness of another one of the solid blocks in the set. 43. The method of claim 39, wherein the solid blocks in the set are at least partially transparent to the beam. 44. The method of claim 39, wherein each of the solid blocks has two surfaces that are parallel to each other. 45. The method of claim 39, wherein one of the solid blocks has a Z value that is different from a Z value of another one of the solid blocks. 46. The method of claim 39, further comprising cooling the selected one or more of the solid blocks. 47. The method of claim 46, wherein one of the solid blocks is cooled using liquid, and another one of the solid blocks is cooled using convection. 48. The method of claim 39, wherein the act of determining the information regarding the desired particle beam energy comprises obtaining the information from a treatment plan. 49. The method of claim 39, wherein the act of determining the information regarding the desired particle beam energy comprises:measuring an energy of a delivered beam; anddetermining a difference between the measured energy and a desired energy, wherein the information comprises the determined difference. 50. The method of claim 39, further comprising using the selected one or more of the solid blocks to attenuate the beam before the beam reaches the collimator. 51. The method of claim 50, wherein the set of at least three solid blocks is a part of an energy modulator having a first end and a second end, the first end coupled to a particle source, the second end coupled to a beam transport component; andwherein the method further comprises passing the beam to the transport component after the selected one or more of the solid blocks has been used to attenuate the beam. 52. The method of claim 51, wherein the first end of the energy modulator is directly coupled to the particle source.
	claims	1. Storage device for the storing and/or transporting of nuclear fuel assemblies, said device comprisinga plurality of adjacent housings, each having a lateral wall, and being able to receive a nuclear fuel assembly therein, said lateral wall being realized using stacking and intercrossing slotted structural assemblies, the structural assemblies oriented lengthwise along a horizontal axis,each structural assembly including at least one plate realized in a first material comprising aluminium, characterized in that each structural assembly further comprises a tubular cross-section element having a hollow interior defined between an upper edge and a lower edge and two lateral flanks, wherein the at least one plate is slidably positioned within the hollow interior along the horizontal axis, the structural assembly having a plurality of vertical slots on its outer surface to allow stacking and intercrossing with an adjacent structural assembly to at least partially form said lateral wall, the at least one plate having a first vertically oriented notch and the tubular cross-section having a second vertically oriented notch, wherein the first notch and the second notch are located to coincide with one another to form the corresponding vertical slot when the at least one plate is positioned within the tubular cross-section element, the tubular cross-section element being realized in a second material selected from steels and titanium or alloys of steels and titanium. 2. Storage device according to claim 1, characterized in that one of said first and second materials are different from one another. 3. Storage device according to claim 2, characterized in that said first material from which is realized each plate comprising aluminium and a neutrophage element(s). 4. Storage device according to claim 3, characterized in that said neutrophage element(s) are selected from the group consisting of the following elements: boron, gadolinium, hafnium, cadmium, indium, samarium and europium. 5. Storage device according to claim 1, characterized in that each housing has a lateral surface delimiting it, said lateral surface being at least partially constituted by said tubular cross-section elements of slotted structural assemblies forming the lateral wall of the housing. 6. Storage device according to claim 1, characterized in that said at least one plate is constituted by a single plate. 7. Storage device according to claim 1, characterized in that said at least one plate is constituted by two plates spaced one in relation to the other. 8. Storage device according to claim 7, characterized in that said two spaced plates respectively face the interior sides of said two lateral flanks of said steel tubular cross-section element. 9. Storage device according to claim 1, characterized in that at least one structural assembly has an upper portion provided with upper slots as well as a lower portion provided with lower slots, upper and lower slots substantially presenting the same height. 10. Storage device according to claim 9, characterized in that each of the upper and lower slots is realized using a notch formed in said tubular cross-section element, and a notch formed in each plate of said structural assembly. 11. Storage device according to claim 1, characterized in that at least one structural assembly has an upper portion provided with upper slots as well as a lower portion provided with lower slots, with the upper slots having a height that is lower than that of lower slots. 12. Storage device according to claim 11, characterized in that each of the upper and lower slots is realized using a notch formed in said tubular cross-section element, and a notch formed in each plate of said structural assembly. 13. Storage device according to claim 11, characterized in that each of the lower slots is realized using a notch formed in said tubular cross-section element, and a notch formed in each plate of said structural assembly, and in that each of the upper slots is realized solely using a notch formed in said tubular cross-section element. 14. Storage device according to claim 1, characterized in that said tubular cross-section element of each structural assembly has a substantially rectangular or square cross-section. 15. Storage device according to claim 1, characterized in that it further includes a plurality of perimeter walls arranged around the stack of structural assemblies. 16. Storage device according to claim 15, characterized in that each structural assembly has two opposite ends each mounted on one of said plurality of perimeter walls, such that a clearance is present between any two directly consecutive structural assemblies of a same lateral wall. 17. Container for the storing and/or transporting of nuclear fuel assemblies, said container including a casing inside of which is housed a storage device having a plurality of adjacent housings, at least one housing being able to receive a nuclear fuel assembly therein, wherein the at least one housing is defined by four lateral walls formed by stacking and intercrossing slotted structural assemblies, each structural assembly including an outer tubular cross-section element being realized in a second material selected from steels and titanium or alloys of steels and titanium, the outer tubular cross-section element having a hollow interior area therein oriented in a horizontal direction, and at least one plate realized in a first material comprising aluminium, wherein the at least one plate is slidably inserted the hollow interior area, the outer tubular element having a vertical first notch and the at least one plate having a vertical second notch, wherein the first notch and the second notch coincide to form an overall slot in the structural assembly. 18. Container according to claim 17, characterized in that said casing includes a base, a lid, as well as a lateral body extending around a longitudinal axis of casing merged with a longitudinal axis of said storage device, said storage device being locked in translation at a distance from said lid in relation to lateral body of casing, along the longitudinal axis of casing, in the direction of said lid. 19. Container according to claim 18, characterized in that the locking in translation is carried out using stop means attached to lateral body of casing and cooperating with perimeter walls of storage device, arranged around the stack of structural assemblies. 20. Storage device according to claim 1, characterized in that at least one of said first and second materials comprises a neutrophage element(s).
	description	This application claims the benefit of U.S. provisional application Ser. No. 60/489,130 filed Jul. 22, 2003, which is incorporated herein by reference. The present invention relates to the diagnostic imaging arts. It particularly relates to computed tomography scanners with a two-dimensional detector arrays, and will be described with particular reference thereto. However, the invention will also find application with other two-dimensional radiation detectors for a variety of imaging and non-imaging applications employing x-rays, radiation from an administered radiopharmaceutical, light, or other types of radiation. Computed tomography (CT) imaging typically employs an x-ray source that generates a fan-beam, wedge-beam, or cone-beam of x-rays that traverse an examination region. A subject arranged in the examination region interacts with and absorbs a portion of the traversing x-rays. A one- or two-dimensional radiation detector including an array of detector elements is arranged opposite the x-ray source to detect and measure intensities of the transmitted x-rays. Typically, the x-ray source and the radiation detector are mounted at opposite sides of a rotating gantry such that the gantry is rotated to obtain an angular range of projection views of the subject. The projection views are reconstructed using filtered backprojection or another reconstruction method to produce a three-dimensional image representation of the subject or of a selected portion thereof. Typically, the reconstruction assumes that the radiation traversed a linear path from the x-ray source directly to the detector. Any scattered radiation that reaches the detector degrades the resultant image. The radiation detector typically includes scintillator crystal arrays, each crystal of which produces bursts of light, called scintillation events, in response to x-rays. Arrays of photodetectors, such as monolithic silicon photodiode arrays, are arranged to view the scintillator crystal arrays and produce analog electrical signals indicative of the spatial location and intensity of the scintillation event. Typically, the detector is focus-centered structure, in which a plurality of scintillator crystal arrays defines a curved detection surface defining a focus that coincides with a focal spot of the x-ray beam. Anti-scatter elements, such as arrays of anti-scatter plates, are mounted in front of the scintillator array, and are precisely aligned with the focus to admit unscattered x-rays and block scattered x-rays, which would otherwise contribute to the measurement as noise. In present anti-scatter elements, plates with heights of between one centimeter and four centimeters are typical. The spacing between the anti-scattering plates defines slits, through which the direct or non-scattered x-rays pass unimpeded. However, scattered x-rays are angularly deviated due to the scattering and strike the anti-scatter plates which absorb the scattered x-rays before they reach the scintillator crystal array. A conventional detector board is assembled starting with a monolithic photodiode array, which is mounted to ceramic support substrates for rigidity. The scintillator crystal arrays are bonded to the monolithic photodiode arrays. Anti-scatter elements are next mounted and aligned with the interface between adjacent scintillation crystals on the detector boards. The detector boards with joined anti-scatter elements are mounted onto a mechanical base plate or support and manually aligned with the focal spot of the x-ray beam. Typically, a test projection image is made and examined to determine which anti-scatter grids are misaligned. The detector boards are shimmed or the anti-scatter elements re-aligned with the detector array. The test image and the adjustment routine are repeated until satisfactory test images are obtained. A common problem in such detector arrays is cumulative alignment or stack-up errors. Typically, the anti-scatter plates are several centimeters long. The thickness of these plates is comparable with the inter-gap spacing between the scintillation crystals and spacing is comparable with the crystals size, e.g., of about 0.5-3.0 mm. The large anti-scatter plates require precise alignment of the anti-scatter elements with the spatial focal point. As detector arrays get larger, e.g., 32 rows of detectors, 64 rows of detectors, etc., it becomes progressively more difficult to maintain every anti-scatter plate accurately positioned between the scintillation crystals over their entire length. Slight misalignment, deflection along the plate's length, or wobble of a plate due to vibration or rotation, can cause the plate to shadow the adjoining scintillation crystals. Shadowing, in turn, leads to reduced x-ray intensities, which signify more dense material along the x-ray path. This leads to image artifacts, which generally manifest as rings in the image reconstruction. Spatially non-uniform shadowing also leads to spectral differences in the detected x-rays and non-linear detector array characteristics. Further, the replacement of defective detector electronics requires removal of the entire detector module including anti-scatter grid. When the new parts are installed the alignment process is repeated, making a field replacement of the defective detector expensive and time-consuming. The present invention contemplates an improved apparatus and method that overcomes the aforementioned limitations and others. According to one aspect of the invention, a two-dimensional radiation detector for a radiographic scanner is disclosed. A first aligning means aligns an anti-scatter module, disposed on a support frame, with a spatial focus. A second aligning means aligns the anti-scatter module with a detector subassembly module and a radiation absorbing mask. Each radiation subassembly module includes a substrate and an array of detector elements arranged on a substrate to detect radiation. The radiation absorbing mask is formed as a grid and arranged between the array of detector elements and the anti-scatter module. According to another aspect of the invention, a computed tomography scanner is disclosed. An x-ray source is mounted to rotate about an examination region. The x-ray source emits a cone shaped x-ray beam from a radiation focal point that traverses the examination region. A two-dimensional radiation detector receives the cone beam of radiation that has traversed the examination region. The radiation detector includes a plurality of detector modules. Each detector module includes an anti-scatter module, a detector subassembly module, and a radiation absorbing mask that are aligned with each other. Each detector subassembly module includes a substrate and an array of detector elements arranged on the substrate to detect radiation. The radiation absorbing mask is positioned between the anti-scatter module and the detector elements. A reconstruction processor reconstructs signals from the detector elements into a volumetric image. According to yet another aspect of the invention, a method is provided for manufacturing a radiation detector for a computed tomography scanner. An anti-scatter module is aligned with a detector subassembly module and a radiation absorbing mask. The anti-scatter modules are disposed on a support frame. The detector subassembly module includes a substrate and an array of detector elements arranged on the substrate to detect radiation. The radiation absorbing mask is disposed between the anti-scatter module and the array of the detector elements. One advantage of the present invention resides in the improved uniformity in x-ray sensitivity among the individual scintillation crystals. The scintillation crystals are generally larger than the openings in the radiation absorbing mask, hence the spatial resolution of the detector is established by the radiation absorbing mask. Another advantage of the present invention resides in a simplified process for assembling a detector module for computed tomography imaging. Numerous additional advantages and benefits of the present invention will become apparent to those of ordinary skill in the art upon reading the following detailed description of the preferred embodiment. With reference to FIG. 1, a computed tomography (CT) imaging apparatus or CT scanner 10 includes a stationery gantry 12. An x-ray source 14 and a source collimator 16 cooperate to produce a fan-shaped, cone-shaped, wedge-shaped, or otherwise-shaped x-ray beam directed into an examination region 18 which contains a subject (not shown) such as a patient arranged on a subject support 20. The subject support 20 is linearly movable in a Z-direction while the x-ray source 14 on a rotating gantry 22 rotates around the Z-axis. In an exemplary helical imaging mode, the rotating gantry 22 rotates simultaneously with linear advancement of the subject support 20 to produce a generally helical trajectory of the x-ray source 14 and collimator 16 about the examination region 18. However, other imaging modes can also be employed, such as a multi-slice imaging mode in which the gantry 22 rotates as the subject support 20 remains stationary to produce a generally circular trajectory of the x-ray source 14 over which transverse parallel slice images are acquired. After the parallel slice images are acquired, the subject support 20 optionally steps a pre-determined distance in the Z-direction and the image acquisition is repeated to acquire a larger volumetric data set in discrete steps along the Z-direction. A radiation detector assembly 30 is arranged on the gantry 22 across from the x-ray source 14. In the exemplary CT scanner 10, the radiation detector assembly 30 spans a selected angular range that preferably comports with a fan angle of the x-ray beam. The radiation detector assembly 30 includes a plurality of modules 32 for acquiring imaging data along a portion of the Z-direction in each projection view. The radiation detector assembly 30 is arranged on the rotating gantry 22 opposite to the x-ray source 14 and rotates therewith so that the radiation detector assembly 30 receives x-rays that traverse the examination region 18 as the gantry 22 rotates. With continuing reference to FIG. 1, the rotating gantry 22 and the subject support 20 cooperate to obtain selected projection views of the subject along a helical trajectory or other trajectory of the x-ray source 14 relative to the subject. The path of the x-ray source 14 preferably provides sufficient angular coverage for each voxel of the imaged region of interest to prevent undersampling. Projection data collected by the radiation detector assembly 30 are communicated to a digital data memory buffer 40 for storage. A reconstruction processor 42 reconstructs the acquired projection data, using filtered backprojection, an n-PI reconstruction method, or other reconstruction method, to generate a three-dimensional image representation of the subject or of a selected portion thereof which is stored in a volumetric image memory 44. The image representation is rendered or otherwise manipulated by a video processor 46 to produce a human-viewable image that is displayed on a graphical user interface 48 or another display device, printing device, or the like for viewing by an operator. Preferably, the graphical user interface 48 is programmed to interface a human operator with the CT scanner 10 to allow the operator to initialize, execute, and control CT imaging sessions. The graphical user interface 48 is optionally interfaced with a communication network such as a hospital or clinic information network via which image reconstructions are transmitted to medical personnel, a patient information database is accessed, or the like. With continuing reference to FIG. 1 and with further reference to FIGS. 2 and 5, the radiation detector assembly includes a pair of alignment plates 60, positioned about support frame 62. A plurality of alignment openings 70 disposed on the alignment plates 60 for mounting and aligning of the detector modules 32. As shown in FIG. 2, the alignment openings 70 are arranged in pairs along radial lines 72 that converge at a spatial focal spot 74, which coincides with the focal spot of the x-ray source 14. In FIG. 2, a few exemplary radial lines 72 are shown to indicate the alignment of pairs of the alignment openings 70 with the spatial focal point 74. Further additional alignment openings 76 in the alignment plates 60 are included for mounting the plates 60 in the CT scanner 10. With reference to FIGS. 3-5, each detector module 32 has an anti-scatter module 78, each including a plurality of anti-scatter plates or vanes 80 arranged generally in conformity with the rays or planes 72 and separated by radio translucent spacer plates 82. The spacer plates 82 taper to define a selected spacing and convergence angle between anti-scatter plates 80. The non-scattered radiation is directed parallel to the anti-scatter plates 80 and pass between any two of them, while scattered radiation angularly deviates from parallel with the anti-scatter plates 80 and is typically absorbed by the anti-scatter plates 80. Although the anti-scatter plates or vanes 80 are generally parallel to one another, those skilled in the art will recognize that precisely parallel plates do not exactly align with the spatial focal point 74. That is, precisely parallel planes do not contain any points in common, and hence cannot contain the spatial focal point 74 in common. Preferably, the generally parallel anti-scatter plates or vanes 80 are each aligned with a plane 72 that intersects the spatial focal point 74. Such planes are close to, but not exactly, parallel over a length L of the anti-scatter plate 80 since L is short, compared a distance between the anti-scatter module 78 and the spatial focal point 74. The anti-scatter plates or vanes 80 are preferably formed of a material with a high atomic number that is highly absorbing for radiation produced by the x-ray source 14, such as tantalum, tungsten, lead, or the like. The spacer plates 82 are formed of a material that is substantially translucent to radiation produced by the x-ray source 14, and are suitably formed of a plastic material. In a preferred embodiment, the spacer plates 82 are substantially hollow molded plastic frames, rather than full molded plastic slabs, to further reduce radiation absorption in the spacer plates 82. The arrangement of generally parallel anti-scatter plates 80 and spacer plates 82 is secured at the sides by two end caps 84. Each end cap 84 includes alignment pins or other alignment protrusions 86 that are received in openings 70 to align the plates with the radial lines or planes 72, as best seen in FIG. 6. With reference to FIG. 7, each detector subassembly module 100 includes a substrate 102, such as a circuit board, on which a detector array 104 and associated electronics 106 are mounted. In the preferred embodiment, the detector array 104 includes a scintillator crystal array 108 including individual scintillation crystals 110, each optically coupled to a photodetector 112 of a photodetector array 114. The scintillator crystal array 108 converts x-rays to light that is detectable by the photodetector array 114. The photodetector array 114 is preferably a monolithic array of silicon photodiodes, amorphous silicon, charge-coupled devices, or other semiconductor photodetectors. Other detector arrays such as CZT detectors that convert x-rays directly into electrical signals are also contemplated. With continuing reference to FIG. 7 and further reference to FIGS. 8, 9 and 10A, an x-ray radiation absorbing mask 120 is arranged between the base of each anti-scatter module 78 and the scintillator crystal array 108. The radiation absorbing mask 120 has the form of a grid and surrounds a periphery of the radiation receiving face of each of the scintillation crystals 110. A width of a Z-direction grid strip 122 of the radiation absorbing mask 120, which is mounted parallel to the anti-scatter plates 80 is wider than the width of one anti-scatter plate 80 and equal to or greater than a width of an intra-element gap 124 between detector elements. The Z-direction strips 122 of the radiation absorbing mask 120 are wider than the projection of the distal end of the anti-scatter plate 80. Preferably, the Z-direction strips 122 are generously wide to compensate for any allowed tolerance, vibrational motion, and cumulative alignment stack-up errors leading to mispositioning of anti-scatter plates 80 that could cast shadows on crystals 110 of array 108. Preferably, the Z-direction strips 122 of the radiation absorbing mask 120 are sufficient to absorb 90% or more of the x-rays that are incident on it. Preferably, the radiation absorbing mask 120 is a high-density absorber and constructed of a material with a high atomic number that is highly absorbing for radiation produced by the x-ray source 14, such as tungsten or any other material that may be etched or manufactured in another precise way. The masks can be stacked to provide 0.5-2 mm thickness to increase the radiation attenuation. Typically, the tungsten sheets are available in 0.125 mm thick sheets. Consequently, three or four 0.125 mm thick radiation absorbing masks 120 are preferred for attenuating the x-ray beam. Thicker masks do not need to be stacked and can be made by other suitable methods. With continuing reference to FIG. 110A, apertures 126 in the radiation absorbing masks 120 are photochemically etched and accurately and repetitiously controlled. Typically, the spatial resolution of the scanner is chosen by the width of the crystal 110 of the array 108 and is controlled by the size of the aperture or the solid angle of radiation that can measured by radiation detector assembly 30. The spatial resolution is controlled by the radiation absorbing mask 120 that is a precise part providing accurate apertures 126 that determine the spatial resolution. Further, the scintillation crystals 110 in large arrays tend to have a lack of precise uniformity in the surface area that is exposed to x-ray radiation, due to the cutting or etching tolerances, uneven reflective coating layers, or the like. The use of the grid that defines apertures 126 precisely fixes the cross section of the radiation beam that each scintillation crystal 110 receives and causes each to have exact same sampling window. Alternatively, the sampling window and the apertures can have a precisely controlled non-uniform size to provide different resolution across a detector module. Analogously, the aperture sizes can change from detector module to detector module to provide non-uniform resolution across the detector assembly. The apertures 126 are precisely referenced to alignment openings 128 in the same grid and are made to high tolerances as afforded by the photochemical etching process or other precise methods of manufacturing. With continuing reference to FIG. 10A, the radiation absorbing masks 120 are preferably configured to include thin bridges or circumferential strips 130, which provide mechanical stability by reducing the free aperture length. Preferably, strips 130 are comparable with the width of an intra-element gap between detector elements running in a circumferential direction. With continuing reference to FIG. 10A and further reference to FIG. 10B, the radiation absorbing masks 120 are constructed to have an interleaving edges geometry to prevent the radiation from passing through a gap between any two individual masks. Each edge of the radiation absorbing mask 120 along its width is etched or cut approximately half way, creating a stepped edge 134. Steps on opposite sides of the mask are provided on the opposite surfaces of the mask to permit alternate stacking of the radiation absorbing masks 120 across the length of the radiation detector assembly 30 without inter-module gaps. Alternately, the stepped edges can be etched or cut into the same surface for simplicity of manufacture. Alternate masks are turned over to engage their neighbors. With reference again to FIGS. 7, 9 and 10A, alignment pins 160 of each anti-scatter module 78 precisely mate with the precision openings 128 of radiation absorbing mask 120 and precision alignment openings 162 of the corresponding detector subassembly module 100 to provide alignment of the scintillator crystal array 108 with the mask 120 and the vanes 80 of the anti-scatter module 78. Because both the alignment openings 162 and the pins 160 are defined with precision, the parts are precisely aligned upon insertion. No adjustment in the alignment is necessary. As best seen in FIG. 7, several radiation absorbing masks 120 can be stacked on the pins 160 without any loss of accuracy. The alignment of the scintillator crystal array 108 to the anti-scatter module 78 arranges the scintillation crystals 110 in the gaps between the anti-scatter plates 80 as best seen in FIGS. 8 and 11. The radiation absorbing mask 120 is arranged between the anti-scatter plates 80 and the scintillation crystals 110. The scintillation crystals 110 view the x-ray source through the grid apertures 126 and between the anti-scatter plates 80, such that the radiation, which does not angularly deviate from the unscattered radiation, reaches the individual elements of the detector array 104. With reference to FIG. 12, the radiation absorbing mask 120 made by one of the alternative methods of manufacturing the radiation absorbing mask is depicted. Slots or grooves 164 are provided in a matrix 166 of a low radiation absorbing material such as BeAl Alloy, plastics, epoxy resin, or any other material that is relatively transparent to the x-ray radiation. A tungsten powder or any other high atomic powder material is then deposited by casting or injection molding in the provided slots and grooves. Yet the radiation absorbing mask might be made by using ECOMASS™ high atomic number powder materials in a plastic matrix available from PolyOne Corp. These materials can be formed in different shapes by standard manufacturing techniques such as extrusion. The radiation absorbing mask might be fabricated by a use of a very precise tooling. With continuing reference to FIGS. 1-12 and with further reference to FIG. 13, a preferred method 170 for assembling the radiation detector assembly 30 is described. In a step 172, the anti-scatter modules 78 are aligned with the alignment openings 70 by coupling the alignment pins 86 with the alignment openings 70 of the alignment plates 60. In a step 174, radiation absorbing masks 120 are stacked on the pins 160 of the anti-scatter modules 78. In a step 176, each detector subassembly module 100 is mounted to each corresponding fixed anti-scatter module 78 using the mating alignment pins 160 and openings 162. Although the radiation detector assembly 30 has been described with reference to a computed tomography imaging scanner, it is readily modified for use in other imaging systems. For example, a gamma camera for nuclear medical imaging typically includes detector arrays substantially similar to the photodetector array 104 with scintillators suitable for converting radiation produced by an administered radiopharmaceutical to light detectable by the detector array. Analogously, these techniques can be applied to conventional x-ray, digital x-ray, fluoroscopy, and the like. The invention has been described with reference to the preferred embodiments. Modifications and alterations will occur to others upon a reading and understanding of the preceding detailed description. It is intended that the invention be construed as including all such modifications and alterations insofar as they come within the scope of the appended claims or the equivalents thereof.
	abstract	In a reactor vessel thermal load reducing system near the surface level of a coolant, the present invention is characterized in that a heat conductive member is installed not contacting the reactor vessel wall in an area above and below the coolant liquid surface, and the heat conducting member is attached to a guard vessel, the heat conducting member being made of material of good heat conductivity.
	abstract	A method and system for thermal-dynamic modeling and performance evaluation of a nuclear Boiling Water Reactor (BWR) core design is presented. Based on a predetermined generic transient bias and uncertainty distribution in the change in critical power ratio (xcex94CPR/ICPR), shifted histograms of fuel rod critical power ratio (CPR) are generated. Ultimately, the operating limit minimum critical power ratio (OLMCPR) of the reactor is directly evaluated from a shifted histogram of fuel rod CPRs for a particular set of initial conditions which result in a probability calculation representing the number of fuel rods subject to a boiling transition (NRSBT) during the transient condition being equal to a predetermined value.
062332988	summary	BACKGROUND OF THE INVENTION 1. Field of the Invention (Technical Field) The present invention relates to transmutation of nuclear waste. 2. Background Art Geologic storage of nuclear waste has been for the past thirty years the primary plan for permanent storage of nuclear waste worldwide. However, concerns about the effectiveness of confinement, the loss of the large energy content of the plutonium waste, the accumulation of thousands of tons of weapons-useful plutonium in the stored waste, and the possibility of recriticality of the waste in permanent storage have delayed the implementation of geologic storage worldwide. At present, no nation has Identified a permanent site for geologic storage of high-level nuclear waste and implementation of geologic waste storage anywhere is at least a decade away. Several nations have attempted to address this problem by destroying the waste using technologies such as mixed oxide (MOX) waste burning in conventional light water reactors (LWRs) or in fast breeder reactors converted for waste burning. While some gains are possible using these approaches, the impact on the waste problem is either minor or the time scale for making a significant impact is much longer than a human generation. Therefore there is no consensus that destruction of waste using conventional nuclear technology is practical for improving geologic storage significantly. Over the last decade, scientists have proposed schemes for improved waste destruction using accelerator-driven reactor-like systems and some of these concepts have been patented. Typically, such systems offer gains by using an accelerator to supplement the number of neutrons beyond those available in an ordinary reactor and thereby to obtain more complete destruction or burn-up of the waste. An accelerator also may be used to allow operation outside of technical constraints imposed by the criticality requirement of normal reactors. The following patents disclose technology related to reduction and production of nuclear matter. U.S. Pat. No. 5,774,514, entitled "Energy Amplifier for Nuclear Energy Production Driven by a Particle Beam Accelerator," to Rubbia, issued Jun. 30, 1998. This patent discloses a method for producing energy from a nuclear fuel material contained in an enclosure. A high energy particle beam is directed into the enclosure for interacting with a heavy nuclei target to produce high energy spallation neutrons, such target comprising bismuth and/or lead, wherein the bismuth and/or lead are in a molten state. The spallation neutrons are multiplied in steady subcritical fission conditions. This patent also discloses the use of a plurality of fuel bodies each encapsulated in a shell of a solid-phase moderator, such moderator comprising graphite. U.S. Pat. No. 5,768,329, entitled "Apparatus for Accelerator Production of Tritium," to Berwald, issued Jun. 16, 1998. This patent discloses a process for preparing or breeding tritium gas from dense molten lithium alloy, such as an eutectic lead lithium alloy. The molten lithium alloy serves as a target material for a high energy particle beam whereby the beam's high energy protons interact with the target to generate a neutron flux. The molten state lead lithium alloy circulates past the beam impact area and through a heat exchanger to recover thermal energy. U.S. Pat. No. 5,545,797, entitled "Method of Immobilizing Weapons Plutonium to Provide a Durable Disposable Waste Product," to Ewing et al., issued Aug. 13, 1996. This patent discloses a method of fixation and immobilization of plutonium whereby the plutonium is fixed in the form of either PuO.sub.2 or Pu(NO.sub.3).sub.4 and is mixed with ZrO.sub.2 and SiO.sub.2. U.S. Pat. No. 5,513,226, entitled "Destruction of Plutonium," to Baxter et al., issued Apr. 30, 1996. This patent discloses a method of using plutonium in a manner so as to render it no longer suitable for employment in a device to create nuclear detonation. The first three steps of the ten step method comprise forming plutonium oxide spheroids, coating the spheroids with a multi-layer fission-product retentive coating and disposing the coated spheroids in a plurality of graphite block elements. Fissioning of the fissle plutonium nuclides occurs through neutrons primarily in the thermal range. U.S. Pat. No. 5,499,276, entitled "Method for Minor Actinide Nuclides Incineration," to Wakabayashi, issued Mar. 12, 1996. This patent discloses a method of minor actinide incineration by adding neptunium of minor actinide nuclides separated from spent fuel to a reactor core fuel of a fast reactor and adding americium of the separated minor actinide nuclides and rare earth elements to radial and/or axial blankets and/or shield of the fast reactor. U.S. Pat. No. 5,160,696, entitled "Apparatus for Nuclear Transmutation and Power Production using an Intense Accelerator-generated Thermal Neutron Flux," to Bowman, issued Nov. 3, 1992, expired Nov. 6, 1996 due to failure to pay maintenance fees. This patent discloses an apparatus using a high energy proton beam and a spallation target to generate high thermal neutron fluxes wherein the target comprises a high Z-material such as a liquid lead-bismuth eutectic mixture. The high thermal neutron fluxes are used to burn-up higher actinide nuclear waste and rapid burn-up of fission product waste. U.S. Pat. No. 4,721,596, entitled "Method for Net Decrease of Hazardous Waste Materials," to Marriott et al., issued Jan. 26, 1988. This patent discloses a method for decreasing reactor waste materials through use of a thermal neutron flux whereby neutrons for transmutation are produced from a fission or non-fission, e.g., fusion, source. U.S. Pat. No. 4,309,249, entitled "Neutron Source, Linear-Accelerator Fuel Enricher and Regenerator and Associated Methods," to Steinberg et al., issued Jan. 5, 1982. This patent discloses an apparatus for producing fissle material using a high energy particle beam, nuclear fuel elements and a liquid metal target material. For example, a proton accelerator produces high energy protons that interact with a liquid lead-bismuth metal that surrounds LWR fuel elements placed in pressure tubes whereby the interaction of the beam and liquid metal produces neutrons. The neutrons are absorbed by the nuclides in the fuel elements and transformed to fissle material. U.S. Pat. No. 3,349,001, entitled "Molten Metal Proton Target Assembly," to Stanton, issued Oct. 24, 1967. This patent discloses a thermal nuclear apparatus having a molten metal high nergy proton target, e.g., molten lead, surrounded by a blanket of fertile material and a recirculating coolant. To date, none of the accelerator-driven systems have received the funding necessary to construct and operate a system demonstrating enhanced waste burning capabilities. One reason seems to be that the addition of an expensive accelerator and extensive chemical separations increase the cost of these transmutation concepts well beyond that of conventional reactor technology which is itself too expensive for deployment for waste burning. Since geologic storage adds only incrementally to nuclear electric power costs, transmutation should also add only incrementally to the cost. Therefore, if an accelerator must be added, other components or operations normally required for nuclear power should be eliminated to keep costs under control. SUMMARY OF THE INVENTION (DISCLOSURE OF THE INVENTION) A subcritical reactor-like apparatus for treating actinide wastes, the apparatus comprising a vessel having a shell and an internal volume. In the preferred embodiment, the shell comprises a metal, for example, HASTELLOY.RTM., a federal trademark owned by Haynes International, Inc. (Kokomo, Ind.), while the apparatus typically has means for removing volatile material from the vessel. The internal volume of the vessel houses graphite, preferably a solid graphite matrix. The graphite matrix in certain embodiments comprises greater than 80% of the internal volume of the vessel. The apparatus preferably also has a fluid medium comprising molten salts and plutonium and minor actinide waste and/or fission products. The molten salts are typically fluoride salts and/or chloride salts. When fluoride salts are used, NaF and/or ZrF.sub.4 are suitable. The apparatus also introduces the fluid medium into the internal volume continuously and/or periodically. The apparatus introduces neutrons into the internal volume wherein absorption of the neutrons after thermalization forms a processed fluid medium through fission chain events averaging approximately 10 fission events to approximately 100 fission events. This is preferably accomplished using at least one high energy particle beam and at least one target material, where the target material is preferably lead and/or bismuth. The apparatus of the present invention also removes the processed fluid medium from the internal volume continuously and/or periodically. In addition, the processed medium typically comprises less than approximately 0.1 mole fraction percent of .sup.242m am and/or an isotopic composition of plutonium of less than that necessary for production of nuclear weapons. Fluid medium introduction and removal processes transfer fluid medium and processed fluid medium at rates sufficient to maintain an isotopic equilibrium and constant average fission chain length within the vessel wherein the fission event chains average approximately fission events to approximately 100 fission events and preferably from approximately 20 fission events to approximately 40 fission events. In a preferred embodiment of the present invention, the apparatus additionally separates zirconium cladding and uranium from actinide wastes. In another preferred embodiment, the apparatus additionally recovers energy released by fission events, such as by at least one heat exchanger contained within the internal volume of the vessel, the heat exchanger(s) preferably have metal heat exchange surface for heat transfer and fission product deposition. In yet another preferred embodiment, the recovered energy provides energy for introducing neutrons. The present invention also comprises a method of treating nuclear material comprising: providing a fluid medium comprising molten salts and at least one member selected from the group consisting of plutonium and minor actinide waste and fission products; introducing the fluid medium into a vessel, the vessel containing graphite; introducing neutrons into the vessel wherein absorption of the thermalized neutrons forms a processed fluid medium through fission chain events averaging from approximately 10 fission events to approximately 100 fission events and preferably averaging from approximately 20 fission events to approximately 40 fission events; and removing the processed medium from said vessel. The method of present invention also includes a method of introducing processed fluid medium into the vessel in place of unprocessed fluid medium. In this particular method of introducing, the processed fluid medium from one apparatus of the present invention is useable as a start-up fluid medium for another apparatus of the present invention. Therefore, start-up is facilitated after the first such apparatus becomes operational. Likewise, all apparatuses of the present invention produce processed fluid media suitable for start-up of additional apparatuses. Operation in such a manner contributes to the efficiency of the present invention. A primary object of the present invention is to transmutate nuclear waste. Other objectives include, but are not limited to: Destruction of essentially all of the weapons-useful material in commercial reactor spent fuel; Recovery of nearly all of the fission energy of the plutonium and other fissioning nuclides; Elimination of the possibility of waste recriticality in permanent storage; Reduction of the long-term waste radioactivity; Decoupling of nuclear power production from an associated large weapon-useful waste inventory; and Recovery of full costs by the generation and sale of electric power. A primary advantage of the present invention is a synergy derived through recovery of fission energy whereby operation costs are minimized. This advantage and other advantages are derived in part from the following features: Replacement of reprocessing with more modest front-end chemistry; Elimination of separation of a pure stream of plutonium; Elimination of back-end chemistry; Elimination of fuel fabrication and refabrication; and Elimination of the need for fast spectrum reactors for fuel destruction. Other objects, advantages and novel features, and further scope of applicability of the present invention will be set forth in part in the detailed description to follow, taken in conjunction with the accompanying drawings, and in part will become apparent to those skilled in the art upon examination of the following, or may be learned by practice of the invention. The objects and advantages of the invention may be realized and attained by means of the instrumentalities and combinations particularly pointed out in the appended claims.
	summary
	abstract	A substrate inspection method includes: generating an electron beam and irradiating the electron beam as a primary electron beam to a substrate as a specimen; inducing at least any of a secondary electron, a reflected electron and a backscattering electron which are emitted from the substrate receiving the primary electron beam, and magnifying and projecting the induced electron as a secondary electron beam so as to form an image of the secondary electron beam; a trajectory of the primary electron beam and a trajectory of the secondary electron beam having an overlapping space and space charge effect of the secondary electron beam occurring in the overlapping space, detecting the image of the secondary electron beam to output a signal representing a state of the substrate; and suppressing aberration caused by the space charge effect in the overlapping space.
051757559	abstract	An X-ray lithography device which utilizes a Kumakhov lens is disclosed. This device is capable of using both small area sources and synchrotron sources. This device provides improved X-ray control, precision and accuracy. Also provided is a method of X-ray lithograph which incorporates a Kumakhov lens.
	abstract	In a particle therapy treatment planning system for creating treatment plan data, the movement of a target (patient's affected area) is extracted from plural tomography images of the target, and the direction of scanning is determined by projecting the extracted movement on a scanning plane scanned by scanning magnets. Irradiation positions are arranged on straight lines parallel with the scanning direction making it possible to calculate a scanning path for causing scanning to be made mainly along the direction of movement of the target. The treatment planning system can thereby realize dose distribution with improved uniformity.
	description	In accordance with a preferred embodiment of the present invention, FIG. 1 depicts a system 10 for the x-ray fluorescence analysis of a sample of interest. An x-ray source 20 emits a field of x-ray radiation 12 directed at a reflective optic 22. The reflective optic 22 may be used for collimating or monochromatizing the x-ray radiation 12. Alternatively, the system 10 may operate without the reflective optic 22. As shown, however, the field of x-ray radiation 12 impinges upon a sample of interest 24, such as a silicon wafer that needs to be analyzed to determine chemical impurities. Due to a known physical reaction between the field of x-ray radiation 12 and the sample 24, a field of fluorescent radiation 14 is emitted from the sample. The field of fluorescent radiation 14 contains information in the form of radiation emission lines about the type of atomic or molecular elements present in the sample 24. The field of fluorescent radiation 14 is selectively reflected from the multilayer structure 26 of the present invention, creating a reflected fluorescent radiation field 36. The reflected fluorescent radiation field 36 is subsequently received and analyzed by a detector 28 that is adapted to interpret qualitative and quantitative aspects of the reflected fluorescent radiation field 36. Radiation is selectively reflected from the multilayer structure 26 in accordance with Bragg""s equation, Equation 1 above, where a distance d is schematically referred to in FIG. 2 as reference numeral 18. As shown in FIG. 2, incident radiation 16 that impinges upon a surface at an angle xcex8 is reflected at intervals that correspond to the d-spacing 18. Constructive interference between a predetermined number of layers 18 creates a uniform field of reflected radiation 17. FIG. 3 depicts a multilayer structure 26 in accordance with a preferred embodiment of the present invention. The multilayer structure 26 generally includes a substrate 34, upon which a series of triadic layers 30 may be periodically formed. As shown, the substrate 34 is planar in nature. However, in alternative embodiments, the substrate 34 may be formed into a curved member. For example, the substrate 34 may be formed into an ellipsoid, a paraboloid, or a spheroid as necessary to accomplish a particular objective. A series of triadic layers 30 is periodically formed on the substrate 34 to create the multilayer structure 26 of the present invention. Each triadic layer 30 includes a triad of layers 32a, 32b, 32c, which are sequentially deposited upon the substrate 34 to create the necessary periodicity. The multilayer structure 26 is composed of between 1 and 100 triadic layers 30, or between 3 and 300 individual layers 32a, 32b, 32c. In a preferred embodiment, the multilayer structure 26 is composed of between 30 and 60 triadic layers 30, and each triadic layer 30 is between 5 and 60 nanometers in thickness. This thickness is otherwise referred to as the d-spacing of the multilayer structure 26. As noted, each triadic layer 30 is composed of a triad of layers 32a, 32b, 32c including a first layer 32a, a second layer 32b, and a third layer 32c. Preferably, the first layer 32a is composed of one member from a first group, where the first group includes lanthanum (La), lanthanum oxide (La2O3), or a lanthanum-based alloy. The second layer 32b is preferably composed of one member from a second group, where the second group includes carbon (C), boron (B), silicon (Si), boron carbide (B4C), or silicon carbide (SiC). The third layer 32c is preferably composed of one member from a third group, where the third group includes boron (B) or boron carbide (B4C). As depicted in FIG. 3, the second layer 32b is preferably disposed between the first layer 32a and the third layer 32c. In a second preferred embodiment of the present invention, shown in FIG. 4, the base period 31 of the multilayer structure 26 includes at least one quartet of layers 36a, 36b, 36c, 36d. A series of quartic layers 31 is periodically formed on the substrate 34 to create the multilayer structure 26 of the present embodiment. Each quartic layer 31 includes a quartet of layers 36a, 36b, 36c, 36d which are sequentially deposited upon the substrate 34 to create the necessary periodicity. The multilayer structure 26 is composed of between 1 and 100 quartic layers 31, or between 4 and 400 individual layers 36a, 36b, 36c, 36d. In a preferred embodiment, the multilayer structure 26 is composed of between 30 and 60 quartic layers 31, and each quartic layer 30 is between 5 and 60 nanometers in thickness. This thickness is otherwise referred to as the d-spacing of the multilayer structure 26. As noted, each quartic layer 31 is composed of a quartet of layers 36a, 36b, 36c, 36d including a first layer 36a, a second layer 36b, a third layer 36c, and a fourth layer 36d. Preferably, the first layer 36a is composed of one member from a first group, where the first group includes lanthanum (La), lanthanum oxide (La2O3), or a lanthanum-based alloy. The second layer 36b is preferably composed of one member from a second group, where the second group includes carbon (C), boron (B), silicon (Si), boron carbide (B4C), or silicon carbide (SiC). The third layer 36c is preferably composed of one member from a third group, where the third group includes boron (B) or boron carbide (B4C). The fourth layer 36d is preferably composed of one member from a fourth group, where the fourth group includes carbon (C), boron (B), silicon (Si), boron carbide (B4C), or silicon carbide (SiC). In a preferred embodiment, the second layer 36b and the fourth layer 36d are chemically identical, although their respective geometrical characteristics will preferably be non-identical. As depicted in FIG. 4, the second layer 36b is preferably disposed between the first layer 36a and the third layer 36c, and the third layer 36c is preferably disposed between the second layer 36b and the fourth layer 36d. It is a feature of the present invention that the multilayer structure 26 may be shaped or otherwise tailored to maximize the performance of the system 10. For example, the multilayer structure 26 shown in FIGS. 3 and 4 may be shaped into a conic section, such as an ellipsoid, paraboloid, or spheroid in order to regulate the magnitude of the angle of incidence xcex8 at different points on the surface of the multilayer structure 26. By shaping the surface of the multilayer structure 26, the field of fluorescent radiation 14 can be conditioned in a particular manner such that the reflected field of fluorescent radiation 36 is focused upon the detector 28 in a preferred fashion. Additionally, the d-spacing of the multilayer structure 26 shown in FIGS. 3 and 4, i.e. the thickness of the triadic layer 30 or the quartic layer 31, may be varied along the depth of the multilayer structure 26, or alternatively, along a lateral axis of the multilayer structure 26. The latter manipulations are known as depth graded d-spacing and laterally graded d-spacing, respectively. The present invention as described in its preferred embodiments thus improves the procedure of x-ray fluorescent spectroscopy by providing a durable multilayer structure with improved spectral resolution, in particular with respect to the fluorescent radiation of boron. In particular, the formation of the multilayer structure composed of triadic or quartic periods greatly increases the overall performance of an x-ray fluorescence spectroscopy system. Both the triadic and quartic periods increase the longevity of the multilayer optic by adding structural integrity to the system, as well as dramatically improved resistance to water. It should be apparent to those skilled in the art that the above-described embodiments are merely illustrative of but a few of the many possible specific embodiments of the present invention. Numerous and various other arrangements can be readily devised by those skilled in the art without departing from the spirit and scope of the invention as defined in the following claims.
052176797	description	DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT Referring to the drawings, it is seen in FIG. 2 that the invention is generally indicated by the numeral 10. Device 10 is generally comprised of cylindrical sleeve 12 having upper, middle, and lower sections 14, 16, 18 respectively. The outer diameter of sleeve 12 is sized to be received through an existing bore in the lower core support plate of a nuclear reactor. The inner diameter is sized to receive a thimble tube during cold conditions in the nuclear reactor. Upper section 14 is formed from any material suitable for such use inside a nuclear reactor. 316L stainless steel is used in the preferred embodiment partly because a hard roll expansion of the invention in this area will be used to lock it in position inside the reactor internals. Over-expanded area 20 is provided slightly below the top end of upper section 14. The outer diameter of area 20 is sized to be slightly larger than the internal diameter of the bore in the lower core support plate into which device 10 is installed. This feature prevents inadvertent misalignment of device 10 inside the lower internals instrument guide tube. The expanded area also helps with accurate positioning of the device by preventing it from being installed at an improper elevation. The return to the normal internal diameter above expanded area 20 provides better coolant flow excitation dynamics for the thimble tube(s). Flow testing has confirmed that a flared top end results in a disruption of coolant flow and causes more thimble tube vibration than the illustrated design. Upper section 14 is welded to middle section 16 as indicated at numeral 22. Middle section 16 is comprised of outer section 24 and inner section 26. Outer and inner sections 24, 26 are welded to each other at their top ends at weld 22 and at their lower ends at each of several plug welds 28. Outer and inner sections 24, 26 are formed from materials having different coefficients of thermal expansion. The materials used are selected and attached to each other such that inner section 26 is in a first normal relaxed state during cold conditions in the reactor but during hot conditions in the reactor (normal operating temperature) inner section 26 moves from its first relaxed state to a second flexed state, thus providing middle section 16 with a variable inner diameter. When inner section 26 is in the first relaxed state, the internal diameter of middle section 16 will readily receive a thimble tube. When inner section 26 is in the second flexed state, the internal diameter of middle section 16 is effectively reduced by inner section 26. This results in a preload or clamping action by inner section 26 on the thimble tube therein. This delivers a significant dampening effect to the thimble tube in preventing vibration caused by coolant flow. As seen in phantom view in FIG. 2, the greatest area of internal diameter reduction is in the area of the center of inner section 26 due to the attachment points being only at the ends. In the preferred embodiment, as best seen in FIG. 1, inner section 26 is comprised of three separate bars equally spaced around the inner circumference of outer section 24. Although a different number of bars can be used for inner section 26, three are used in the preferred embodiment to obtain the clamping action on three opposing sides of the thimble tube. In the preferred embodiment, inner section 26 is formed from cold worked 316L stainless steel and outer section 24 is formed from inconel.RTM.. Lower section 18 may be integral with middle section 16. In the preferred embodiment, lower section 18 is formed from the same, continuous material used to form middle section 16. Lower section 18 may also be a separate part attached to middle section 16 at weld 28. Lower section 18 is provided with baffle 30 along its inner diameter. Baffle 30 sets up effective areas of coolant flow restriction. This reduces coolant flow velocity and pressure upstream of baffle 30, that is, through middle and upper section 16, 14. This maintains cooling requirements while reducing the potential for causing thimble tube vibration. Baffle 30 is formed from a plurality of alternating rings 32, 34. Rings 32 have an outer diameter slightly larger than the interior diameter of lower section 18 and are received in circumferential grooves on the inner surface of lower section 18. Rings 34 have an outer diameter substantially equal to that of lower section 18. The inner diameter of rings 34 is less than that of rings 32 but still large enough to allow a thimble tube therethrough. The changing inner and outer diameter between rings 32, 34 serves to reduce coolant flow and pressure. In operation, device 10 is installed in the lower core support plate of a nuclear reactor during a scheduled outage. Over-expanded area 20 prevents device 10 from being positioned too low in the reactor. A thimble tube is received through device 10 while the reactor is in a cold condition and inner section 26 in its first normal relaxed state. When reactor operations are started, the temperature rise causes inner section 26 to move to its second flexed state in which it provides a clamping action on the thimble tube. Baffle 30 reduces the velocity and pressure of coolant flow through device 10 to further aid in preventing vibration and excessive wear of the thimble tube. Because many varying and differing embodiments may be made within the scope of the inventive concept herein taught and because many modifications may be made in the embodiment herein detailed in accordance with the descriptive requirement of the law, it is to be understood that the details herein are to be interpreted as illustrative and not in a limiting sense.
062326791	abstract	A combined heat and electricity generating unit is suitable for use in class 8 trucks and the like. When it is operating it provides heat for keeping the engine and cabin warm and electricity for use by the electricity consuming devices in the cabin. It is a closed cycle system that includes a radial inlet turbine driven by a low pressure vaporized and superheated working fluid, an oil fired heater for vaporizing the working fluid, a pump for pumping condensed working fluid, a heat exchanger for heating engine coolant with heat from the condensing working fluid and an electric generator for converting energy produced by the turbine to electricity.
	abstract	A method for producing a radionuclide is provided that produces molybdenum trioxide 99 (Mo-99.O3) and technetium oxide 99m (Tc-99m2.O7) by emitting an electron beam accelerated by an electron linear accelerator to a molybdenum trioxide 100 (Mo-100.O3) powder sample, and which separates and purifies technetium oxide 99m from both the molybdenum trioxide 99 and the technetium oxide 99m by using a radionuclide separation/purification unit. The method for producing a radionuclide supplies temperature-regulated gas to the molybdenum trioxide 100 powder sample during an irradiation period during which the electron beam is emitted to the molybdenum trioxide 100 powder sample.
	description	The present application is a divisional of U.S. patent application Ser. No. 14/771,018 filed Aug. 27, 2015, which is a U.S. national stage application under 35 U.S.C. § 371 of International Patent Application No. PCT/US2014/019042 filed Feb. 27, 2014, which claims the benefit of U.S. Provisional Patent Application Ser. No. U.S. 61/770,213 filed Feb. 27, 2013; the entireties of which are incorporated herein by reference. The present invention relates to nuclear reactor vessels, and more particularly to a nuclear reactor shroud surrounding the fuel core. Many nuclear reactor designs are of circulatory type wherein the water heated in the reactor fuel core region must be separated from the cooler water outside of it. Such a nuclear reactor may be typically equipped with a cylindrical shroud around the fuel core. The shroud serves to separate the internal space in the reactor vessel between an “up-flow” (e.g. riser) region in which primary coolant heated by the core flows inside the shroud and the “downcomer” region in which colder primary coolant returned to the reactor vessel from the Rankine cycle steam generating system flows outside the shroud. It is desirable to minimize heat transfer from the heated hot reactor water inside the riser region of the shroud to the colder downcomer water outside the shroud which is deleterious to the thermodynamic performance of the reactor. The standard practice in shroud design has typically consisted of hermetically enclosing a fibrous or ceramic insulation in a stainless steel (or another corrosion resistant alloy) enclosure. Such a shroud works well until a leak in the enclosure develops, usually caused by the thermal stresses and strains that are inherent to any structure operating under a temperature differential. Concerns regarding failure of the shroud and subsequent dismembering of the insulation have been a source of significant and expensive ameliorative modification efforts in many operating reactors. The present disclosure provides a reactor shroud which minimizes heat transfer between the hot reactor riser water and cold downcomer water in a manner which eliminates drawbacks of the foregoing insulated enclosure designs. In an embodiment of the present invention, the shroud may be comprised of a series of concentric cylindrical shells separated by a small radial clearance. The top and bottom extremities of the shells are each welded to common top and bottom annular plates (“closure plates”) to create an essentially isolated set of narrow & tall annular cavities. Each cavity is connected to its neighbor by one or more small drain holes such that submerging the multi-shell body in water (e.g. demineralized primary coolant in a reactor vessel) would fill all of the internal cavities with water and expel virtually all entrapped air, thereby creating water-filled annular cavities. In one non-limiting embodiment, the thin walled concentric shells may be buttressed against each other with a prescribed gap by small fusion welds made by a suitable process such as spot, plug, or TIG welding. In such a welding process, a small piece of metal (e.g. spacer) equal in thickness to the radial gap or clearance in the cavity serves to enable a fusion nugget to be created between the two shell walls. The number of such nuggets is variable, but preferably is sufficient to prevent flow induced vibration of the shroud weldment during reactor operation. One principal advantage of the multi-shell closed cavity embodiment described herein is that it is entirely made of materials native to the reactor's internal space, namely demineralized water (e.g. primary coolant) disposed within the radial gaps between the concentric shells and metal such as stainless steel. No special insulation material of any kind is used in the reactor shroud (which may degrade and fail over time). Advantageously, the present shroud design provides the desired heat transfer minimization between the hot reactor water inside the riser region of the shroud to the colder downcomer water outside the shroud without insulation, thereby preserving the thermodynamic performance of the reactor. According to one exemplary embodiment, a nuclear reactor vessel includes an elongated cylindrical body defining an internal cavity containing primary coolant water; a nuclear fuel core disposed in the internal cavity; an elongated shroud disposed in the internal cavity, the shroud comprising an inner shell, an outer shell, and a plurality of intermediate shells disposed between the inner and outer shells; and a plurality of annular cavities formed between the inner and outer shells, the annular cavities being filled with the primary coolant water. In one embodiment, the annular cavities are fluidly interconnected by a plurality of drain holes allowing the primary coolant to flow into and fill the cavities from the reactor vessel. According to another embodiment, a shroud segment for a nuclear reactor vessel includes an elongated inner shell; an elongated outer shell; a plurality of elongated intermediate shells disposed between the inner and outer shells; the inner shell, outer shell, and intermediate shells being radially spaced apart forming a plurality of annular cavities for holding water; a top closure plate attached to the top of the shroud segment; and a bottom closure plate attached to the bottom of the shroud segment, wherein the top and bottom closure plates are configured for coupling to adjoining shroud segments to form a stacked array of shroud segments. A method for assembling a shroud for a nuclear reactor vessel is provided. The method includes: providing a first shroud segment and a second shroud segment, each shroud segment including a top closure plate and a bottom closure plate; abutting the top closure plate of the second shroud segment against the bottom closure plate of the first shroud segment; axially aligning a first mounting lug on the first shroud segment with a second mounting lug on the second shroud; and locking the first mounting lug to the second mounting lug to couple the first and second shroud segments together. In one embodiment, the locking step is preceded by pivoting a mounting clamp attached to the first shroud segment from an unlocked open position to a locked closed position. All drawings are schematic and not necessarily to scale. Parts given a reference numerical designation in one figure may be considered to be the same parts where they appear in other figures without a numerical designation for brevity unless specifically labeled with a different part number and described herein. The features and benefits of the invention are illustrated and described herein by reference to exemplary embodiments. This description of exemplary embodiments is intended to be read in connection with the accompanying drawings, which are to be considered part of the entire written description. Accordingly, the disclosure expressly should not be limited to such exemplary embodiments illustrating some possible non-limiting combination of features that may exist alone or in other combinations of features. In the description of embodiments disclosed herein, any reference to direction or orientation is merely intended for convenience of description and is not intended in any way to limit the scope of the present invention. Relative terms such as “lower,” “upper,” “horizontal,” “vertical,”, “above,” “below,” “up,” “down,” “top” and “bottom” as well as derivative thereof (e.g., “horizontally,” “downwardly,” “upwardly,” etc.) should be construed to refer to the orientation as then described or as shown in the drawing under discussion. These relative terms are for convenience of description only and do not require that the apparatus be constructed or operated in a particular orientation. Terms such as “attached,” “affixed,” “connected,” “coupled,” “interconnected,” and similar refer to a relationship wherein structures are secured or attached to one another either directly or indirectly through intervening structures, as well as both movable or rigid attachments or relationships, unless expressly described otherwise. Referring to FIG. 1, a reactor vessel 20 includes a vertically elongated cylindrical body defining a longitudinal axis LA and having a top 21, closed bottom 22, and a circumferentially extending sidewall 24 extending between the top and bottom. Sidewall 24 defines an internal cavity 25 configured for holding a nuclear fuel core 26. Internal cavity extends axially along the longitudinal axis from the top 21 to the bottom 22 of the reactor vessel 20 in one embodiment. The bottom 22 may be closed by a lower head 23, which may be without limitation dished or hemispherical in configuration. In one embodiment, the internal cavity 25 may be filled with a liquid such as primary coolant which may be demineralized water. The reactor vessel 20 may be made of any suitable metal, including without limitation coated steel or stainless steel for corrosion resistance. Referring to FIGS. 1-3 and 7, a vertically elongated shroud 30 is provided which is disposed in the internal cavity 25 of the reactor vessel 20. Shroud 30 may be cylindrical in shape with a circular annular cross-section; however, other suitable shapes may be used. Shroud 30 is coaxially aligned with the reactor vessel 20 along the longitudinal axis LA. The fuel core 26 may be located inside the shroud 30, and in one non-limiting embodiment nearer to the bottom 22 of the reactor vessel 20. Shroud 30 includes a top 34 and bottom 35 which may be spaced vertically apart from the bottom 22 of reactor vessel 20 to provide a flow passage into the shroud 30 at the bottom of the reactor vessel 20 (see, e.g. directional flow arrows FIGS. 1 and 8). In one embodiment as best shown in FIG. 8, the bottom 35 of the shroud 30 may be spaced apart from bottom 22 of reactor vessel 20 and supported by a plurality of radially oriented and circumferentially spaced apart support plates 42. Support plates 42 are configured to engage the reactor vessel bottom 22 at one extremity and bottom 35 of shroud 30 at another extremity. In one embodiment, support plates 42 may include one or more flow holes 41 to allow primary coolant to flow and circulate through the plates at the bottom of the reactor vessel. In other embodiments, the holes may be omitted. The shroud 30 divides the internal cavity 25 of reactor vessel 20 into an outer annular space which defines a vertical downcomer region 28 (i.e. down-flow region) and an inner space which defines a vertical riser region 27 (up-flow region). Primary coolant flows downwards in reactor vessel 20 through the annular downcomer region 28, reverses direction and enters the bottom 35 of the shroud 30, and flows upwards through riser region 27 through the fuel core 26 where the primary coolant is heated for generating steam in an external steam generator. In one embodiment, the shroud 30 may comprise an elongated outer shell 31, an inner shell 32, and a plurality of intermediate shells 33 disposed between the outer and inner shells. Shells 31-33 are cylindrically shaped in one embodiment. Shells 31-33 are concentrically aligned with respect to each other and spaced radially apart forming an array comprised of a plurality of relatively thin concentric annular cavities 40 between the outer and inner shell 31, 32. In one embodiment, the cavities 40 are fluid-filled with primary coolant, as further described herein. Annular cavities 40 extend longitudinally from the top 34 to bottom 35 of shroud 30. Accordingly, the annular cavities 40 have a length or height substantially coextensive with the length of the shells 31-33. The shells 31-33 may be formed of a suitable corrosion resistant metal, such as coated or stainless steel for example. In one exemplary embodiment, the number of intermediate shells 33 may be at least two to provide at least three annular cavities 40. In non-limiting preferred embodiments, at least six or more intermediate shells 33 (divider shells) may be provided to divide the space between the inner and outer shells 32 and 31 into at least seven annular cavities 40. In one representative embodiment, without limitation, eight intermediate shells 33 are provided to create nine intermediate shells 33. The number of water-filled annular cavities 40 selected correlates to the insulating effect and heat transfer reduction from the inner shell 32 through the shroud to the outer shell 31. The number of intermediate shells 33 will be one less than the number of water-filled annular cavities 40 to be created. In order to provide inter-shell connectivity and maintain the radial gap of annular cavities 40 between intermediate shells 33 and between the innermost and outermost intermediate shells and inner shell 32 and outer shell 31 respectively, spacers 80 may be provided as shown in FIG. 2. Spacers 80 are disposed in annular cavities 40 between the shells 31-33 and have a radial thickness sufficient to provide the desired radial width of each annular cavity. Each annular cavity 40 preferably includes spacers 80 in an exemplary embodiment. To retain the spacers 80 in their desired vertical position, the spacers may be rigidly attached to a shell 31-33 by any suitable means such as fusion welding in an exemplary embodiment. In one embodiment, a spot weld 81 may be used to attach spacer 80 to a shell 31-33 as shown. The spot welds 81 may have any suitable diameter, such as without limitation about 1 inch as a representative example. The number of spot welds 81 (spot nuggets) needed for joining neighboring shells 31-33 together may be estimated by the following empirical formula: Number=(shroud diameter times height (in inches)/100). Preferably, the spot welds 81 and spacers 80 should be spaced as uniformly as possible. In one embodiment, the spacers 80 may be radially staggered such that the spacers between adjacent shells 31-33 do not lie on the same radial axis (see, e.g. FIG. 2 showing a set of spacers aligned radially only in every other annular cavity 40). Other suitable arrangements of spacers 80 may be used. Spacers 80 may have any suitable shape, including circular or polygonal configurations. Preferably, spacers 80 may be formed of metal such as steel or other. Referring to FIGS. 2, 3, and 7, each annular cavity 40 may be connected to its adjoining cavities by one or more small fluid drain holes 90. Drain holes 90 are configured and arranged to hydraulically or fluidly interconnect all of the annular cavities 40. The outer shell 31 includes drain holes 90 which fluidly connect the outermost annular cavity 40 in shroud 30 to the annular downcomer region 28 in reactor vessel 20. This allows the primary coolant water to enter the outermost cavity 40 and then flow inwards successively through the plurality of drains holes in intermediate shells 33 for filling all the annular cavities with the fluid. Submerging the multi-shell shroud 30 body in the water-filled reactor vessel (e.g. demineralized primary coolant) will fill all of the internal annular cavities 40 with water and expel virtually all entrapped air, thereby creating water-filled annular cavities. In one arrangement, the drains holes 90 may be radially staggered as best shown in FIG. 7 so that the holes in one shell 31 or 33 do not radially align with holes in its neighboring shells. This forms a staggered flow path through the shroud 30. The inner shell 32 may not have drain holes 90 and is solid in one embodiment. Preferably, a plurality of drain holes 90 are spaced both circumferentially and longitudinally apart along the entire height or length of the shroud 30 in each shroud segment 30A-C Referring to FIG. 2, the inner and outer shells 32 and 31 may have thicknesses T2 and T1 respectively which are larger than the intermediate shells 33 in one embodiment to stiffen and strengthen the shroud 30. For example, in one representative example without limitation inner and outer shells 32 and 31 may have a plate thickness (T1 and T2) of about ¼ inch and intermediate shells 33 may have a thickness T3 of about ⅛ inch. Each annular cavity 40 has a depth D2 (measured in the radial direction transverse to longitudinal axis LA) which is less than the total depth D1 between the inner and outer shells 32 and 31. In one embodiment, the water-filled annular cavities 40 may have a depth D2 that is less than the thickness T1-T3 of the shells 31-33. In one representative example without limitation, the depth of cavity 40 may be about 3/16 inch. This arrangement provides a plurality of thin water films or chambers comprised of primary coolant sandwiched between the inner and outer shells 32 and 31 in the multi-shell weldment (MSW) shroud wall construction. The thin water films have an insulating effect for shroud 30 which minimizes heat transfer between the hot riser region 27 and colder downcomer region 28 (see FIG. 1). Advantageously, the water films eliminate the need for traditional insulation materials in the shroud which may be wetted or otherwise damaged. In one embodiment, inner shell 32, outer shell 31, and intermediate shells 33 may have vertical heights or lengths which are substantially coextensive. According to one aspect of the invention, the shroud 30 may comprise a plurality of vertically stacked and coupled shroud sections or segments 30A, 30B, and 30C. Referring to FIGS. 1 and 3, each shroud segment 30A-C includes an upper end 48, lower end 49, an annular top closure plate 36 attached to upper end 48, and an annular bottom closure plate 37 attached to lower end 49. The top closure plate 36 and bottom closure plate 37 may be formed of a suitable metal such as steel. Corrosion resistant closure plates 36, 37 formed of coated or stainless steel may be used. Within each shroud segment 30A-C, the annular cavities 40 and shells 31-33 extend longitudinally between the top and bottom closure plates 36 and 37, and may have coextensive lengths or heights. The outer shell 31, inner shell 32, and intermediate shells 33 in each segment 30A-C may be rigidly attached to the top and bottom closure plates, such as via a rigid connection formed by welding for structural strength. In one embodiment, the shells 31-33 may be hermetically seal joined to the top and bottom closure plates such as with full circumferential seal welds. This forms a water-tight joint between the shells 31-33 and the top and bottom closure plates 36 and 37, respectively. Each shroud segment 30A-C is a self-supporting structure which may be transported, raised, and lowered individually for ease of maneuvering and assembly to adjoining segments during fabrication of the shroud 30. To facilitate handling the shroud segments 30A-C individually, the top closure plates 36 may include radially extending lifting lugs 38 which include a rigging hole 39 for attachment of lifting slings or hoists. A suitable number of lifting lugs 38 circumferentially spaced apart at appropriate intervals are provided to properly and safely hoist the shroud segments 30A-C. The weight of each shroud segment 30A-C may be vertically supported by the shroud segment immediately below with the weight being transferred through the top and bottom closure plates 36 and 37, respectively. Accordingly, in some embodiments, the entire weight of the shroud segments 30A-C may be supported by support plates 42 (see, e.g. FIGS. 1 and 8). In one embodiment, adjoining shroud segments 30A-C may be coupled together at joints 43 between segments via a plurality connectors 76 such as of clamps 50. Referring to FIGS. 1 and 3-5, clamps 50 are configured to detachably join and engage the bottom closure plate 37 of one shroud segment (e.g. 30B) to top closure plate 36 of the adjoining lower shroud segment (e.g. 30C). Clamps 50 each include a U-shaped body 51 defining a recess 52 configured to receive a mounting lug 55 formed on bottom closure plate 37 and a mating mounting lug 56 formed on top closure plate 36 as shown. Mounting lugs 55 and 56 are radially extending and circumferentially spaced apart on bottom and top closure plates 37 and 36, respectively. Each mounting lug 55 is arranged in a pair and coaxially aligned along the longitudinal axis LA with a corresponding mounting lug 56. In one embodiment, the mounting lugs 55 and 56 are integrally formed with and a unitary structural part of the bottom and top closure plates 37, 36. Accordingly, the mounting lugs 55, 56 may preferably be formed of metal similarly to bottom and top closure plates 37, 36 for structural strength. In one arrangement, clamps 50 may each be pivotably connected to a mounting lug 55 on the bottom closure plate 37 by a pivot pin 54 which defines a pivot axis. Pivot pins 54 are oriented parallel to longitudinal axis LA so that the clamp 50 may be pivotably swung or moved transversely to the longitudinal axis LA between a closed locked position (see, e.g. FIG. 4) and open unlocked position (see, e.g. FIG. 5). In one embodiment, pivot pin 54 is disposed proximate to one end 58 of the clamp body 51 and the opposing end 57 is open to receive mounting lug 56 of a top closure plate 36 into recess 52. Pivot pin 54 extends axially through the mounting lug 55 and the bottom and top flanges 59, 60 of clamp 50. To secure the clamp 50 in the closed locked position shown in FIG. 4, a locking fastener such as set screw 53 may be provided which is configured and arranged to engage a top surface of mounting flange 55. Set screw 53 may be threadably engaged in threaded bore 61 formed in top flange 60 of clamp 50. The bore 61 extends completely through top flange 60 to allow the bottom end of the set screw shaft to be projected into or withdrawn from clamp recess 51 for engaging or disengaging mounting flange 55. Raising or lowering the set screw 53 alternatingly disengages or engages the set screw with the mounting flange 55. Set screw 53 is preferably withdrawn from. A method for assembling shroud 30 comprised of segments 30A-C using clamps 50 will now be described. For brevity, assembly of shroud segment 30B onto segment 30C will be described; however, additional shroud segments may be mounted in a similar manner Referring to FIG. 3, a pair of shroud segments 30B and 30C are provided each configured as shown. Clamps 50 are in the open unlocked position (see, e.g. FIG. 5). Shroud segment 30B is first axially aligned along longitudinal axis LA with segment 30C. Segment 30B may then be rotated as needed to axially align mounting flanges 55 on bottom closure plate 37 with mounting flanges 56 on top closure plate 36 of segment 30C. Each pair of mounting flanges 55 and 56 may be brought into abutting relationship. In the process, bottom closure plate 37 is brought into abutting contact with top closure plate 36 forming the joint 43 between segments 30B and 30C. Clamp 50 is then pivoted about pivot pin 54. Mounting flanges 55 and 56 are inserted into recess 51 of clamp 50 between flanges 59 and 60 (see, e.g. FIG. 5). The set screw 53 is then tightened to secure the clamp 50 in the closed locked position shown in FIG. 5. It will be appreciated that the order of performing the steps of the fore steps may be varied. In addition, numerous variations of the foregoing assembly process are possible. Referring to FIG. 2, a sealing gasket 44 may be provided in between each pairing of a top closure plate 36 and bottom closure plate 37 to seal the interface at joint 43 therebetween. In one embodiment, the gasket 44 may be metallic formed of steel, aluminum, or another seal material suitable for the environment within a reactor vessel 20. The gasket 44 may be situated in an annular groove 45 formed in the bottom closure plate 37 as shown, or alternatively in the top closure plate 36 (not shown), to seal water seepage at the interface of joint 43 and also provide a certain level of verticality alignment capability during installation and joining of shroud segments 30A-C. In one embodiment, gasket 44 may be circular in transverse cross-section prior to the joint 43 being closed which will compress and deform the gasket. According to another aspect of the invention, a plurality of lateral seismic restraints such as restraint springs 70 may be provided to horizontally support and protect the structural integrity of the shroud 30 inside reactor vessel 20 during a seismic event. In one embodiment as shown in FIGS. 4 and 5, a dual purpose connector 76 (fastener or coupler for joints 43 between shroud segments 30A-C and lateral restraint) may be provided which combine the clamps 50 and seismic springs 70 into a single assembly. Referring to FIGS. 1 and 3-5, seismic springs 70 are disposed between and engage shroud 30 and the interior surface 74 of the reactor vessel 20. A plurality of seismic springs 70 are provided which are circumferentially spaced apart on the outer shell 28 of the shroud 30. In one embodiment, the seismic springs 70 may be spaced apart at equal intervals. Seismic springs 70 are elastically deformable to absorb lateral movement of the shroud 30. In one embodiment, each spring 70 may be in the form of an arcuate leaf spring comprised of a plurality of individual leaves 75 joined together to function as a unit. The leaves 75 may be made of suitable metal such as spring steel having an elastic memory. Other appropriate materials however may be used. The thickness and number of leaves 75 may be varied to adjust the desired spring force K of the spring 70. Seismic springs are arranged with the concave side facing outwards away from shroud 30 and towards reactor vessel 20 when in the fully mounted and active operating position. Opposing ends 72 and 73 of each seismic spring 70 are arranged to engage the interior surface 74 of reactor vessel 20. In one embodiment, seismic springs 70 may be rigidly attached to shroud 30 to provide a stable mounting for proper operation and deflection of the spring to absorb energy during a seismic event. In one possible arrangement, seismic springs 70 may be rigidly attached to clamps 50 via a fastener 71 or another suitable mounting mechanism. Spring 70 may be fastened to clamp 50 at the midpoint between ends 72 and 73 in one embodiment. Accordingly, seismic springs 70 may be pivotably movable with clamps 50 in the manner already described herein. In FIG. 1, for example, the seismic spring 70 and clamp 50 shown between shroud segments 30A and 30B is in the open unlocked position. In this same figure, seismic springs 70 shown between shroud segments 30B and 30C are in the pivoted closed locked position in which the seismic springs 70 are in the active operating position with ends 72 and 73 engaged with the reactor vessel 20. During a seismic event when the shroud 30 may shift laterally/horizontally in one or more directions, the seismic springs 70 will deform and deflect assuming a more flattened configuration until the seismic load is removed, thereby returning the spring elastically to its original more arcuately-shaped configuration shown. In one embodiment, each joint 43 between shroud segments 30A, 30B, and 30C may include seismic springs 70 to horizontal support the shroud 30 intermittently along its entire height. Underlying Operating Principle of the Shroud The multi-shell weldment (MSW) design for shroud 30 described herein is based on the principle in applied heat transfer which holds that an infinitely tall and infinitesimally thin closed end cavity filled with water would approximate the thru-wall thermal resistance equal to that of the metal walls and the water layer conductances. The governing dimensionless quantity that provides the measure of departure from the ideal (conduction only) is Rayleigh number defined as the product of the Prandtl number (Pr) and the Grashof number (Gr). Heat transfer in a differentially heated vertical channel of height H and gap L is characterized by Nusselt number correlation as a function of Rayleigh number as follows:Nu=0.039Ra1/3 Where: Nu is Nusselt Number (=hL/k) h is heat transfer coefficient k is conductivity of water Ra is Rayleigh number (=gβΔTL3ρ2/μ2)*Pr g is gravitational acceleration β is coefficient of thermal expansion of water ΔT is hot-to-cold face temperature difference ρ is density of water μ is water viscosityAs Rayleigh number defined above exhibits an L3 scaling it follows that gap reduction substantially affects Ra number. For example a factor of 2 gap reduction cuts down Ra number by a factor of 8 (almost by an order of magnitude). Thus engineering the shroud with small gaps has the desired effect of minimizing heat transfer. To further restrict heat transfer a multiple array of gaps are engineered in the shroud lateral space to have the effect of resistances in series. An example case is defined and described below to illustrate the concept. A Small Modular Reactor (SMR), such as the SMR-160 available from SMR, LLC of Jupiter, Fla., may have a particularly long shroud (e.g. over 70 feet). In such a case, the principal design concerns are: ease of installation, removal, verticality of the installed structure, mitigation of thermal expansion effects and protection from flow induced vibration of the multi-wall shell. The design features, described below to address the above concerns for such an SMR, can be applied to any shroud design. A. Narrow cavity geometry: The height of each shroud (e.g. shroud segments 30A-C) is approximately three times its nominal diameter. The innermost and outer most shells (e.g. shells 32 and 31) are relatively thick compared to the intermediate (inner) shells (e.g. shells 33). The water cavity is less than 0.1% of the shroud's height. The table below provides representative dimensions for demonstrating the concept: Dimensions of a typical shroud in SMR-160:Inner diameter 71⅛ inch Height 71 ft.(Shroud built in four stacked sections (segments), 3 × 20 ft.(lower) and 1 × 11 ft. (top))Number of water annuli (cavities) 9Thickness of inner most shell ¼ inch Thickness of outermost shell ¼ inch Thickness of interior shell walls ⅛ inch Thickness of water cavities 3/16 inch B. Inter-shell connectivity: The number of spot nuggets (approximately 1 inch diameter) joining neighboring shells should be estimated by the following empirical formula: Number=(shroud diameter times height (in inches)/100). The spot welds should be spaced as uniformly as possible. C. Handling: The top plate 36 of each shroud segment 30A-C is equipped with lift lugs 38 for handling and installation. Typically six lift lug locations, evenly spaced in the circumferential direction, will suffice. D. Stacked construction: The multi-shell weldments (MSW) of shroud segments 30A-C are stacked on top of each other as shown in FIGS. 1 and 2. One or more round metallic gaskets 44 as described above are provided at the interface between the annular top and bottom closure plates 36, 37 of successive stacks of shroud segments 30A-C. The gaskets 44 situated in the annular grooves 45 in the bottom closure plate 37 serve to seal water seepage at the interface of joint 43 and also provide a certain level of verticality alignment capability. E. Thermal expansion: The axial thermal expansion of a tall stack of shroud segments 30A-C will cause severe stresses in adjoining structures such as the return piping that delivers the reactor coolant from the steam generator to the reactor's outer annulus (downcomer). To mitigate the thermal stresses, the upper region of the shroud may be equipped with a multi-ply bellows type expansion joint. F. Seismic restraints: The junctions or joints 43 of the MSW shroud segments 30A-C provide the “hard” locations to join them and to secure them against lateral movement during earthquakes. The dual purpose connector 76 (fastener and lateral restraint) design concept shown in FIGS. 3-5 comprising the clamps 50 and seismic springs 70 as described herein provide the joining and lateral restraint functionality. This dual purpose connector 76 has the following capabilities: (i) The two interfacing closure plates 36 and 37 are prevented from significant rotation or separation from each other during earthquakes. (ii) The connector 76 is amenable to remote installation and removal. (iii) The connector 76 is equipped with the seismic springs 70 (e.g. leaf springs) to enable it to establish a soft contact or a small clearance with the reactor's inside wall under operating condition (hot). A set of three connectors 76, equipment-spaced in the circumferential direction at each closure plate 36, 36 elevation, is deemed to be adequate for the SMR described above. Additional connectors may be employed in other reactor applications at the designer's option. Performance assessment: The efficacy of the MSW design is demonstrated by the case of the SMR-160 described above. Calculations show that the decrease in the hot leg temperature (primary coolant inside shroud 30) using water-filled annular cavities 40 due to heat loss across the shroud is merely 0.355 deg. F. As a point of reference, the idealized temperature loss would be 0.092 deg. F. if the water layers were instead omitted and “solid,” i.e., heat transferred only by conduction through the shroud. It can be seen that the Rayleigh effect, responsible for the movement of water in closed cavities, has been largely suppressed by the MSW design of shroud 30. Extension to vessels and conduits: The concept of establishing a thin water layer inside pipes (hereafter called “water lining”) carrying heated water is proposed to be employed at the various locations in the power plant where minimizing heat loss from the pipe is desired. For example, the lines carrying hot and cooled reactor coolant are water lined to limit heat loss. Water lining is achieved by the following generic construction: (i) An inner thin walled (liner) pipe that is nominally concentric with the main pipe. The liner pipe has a few small holes to make the narrow annulus communicate with the main flow space. (ii) The small gap between the main and liner pipes is held in place by small spacer nuggets attached to the outside surface of the liner pipe. (iii) In piping runs subject to in-service inspection of pressure boundary welds, the liner pipe is discontinued at the location of such welds. The foregoing water lining approach is also proposed to be used to reduce thermal shock to pressure retaining vessel/nozzle junctions (locations of gross structural discontinuity) where large secondary stresses from pressure exist. This is true of penetrations in the reactor vessel, steam generator as well as the superheater. Water lined pressure boundaries will experience significantly reduced fatigue inducing cyclic stresses which will help extend the service life of the owner plant. While the foregoing description and drawings represent exemplary embodiments of the present disclosure, it will be understood that various additions, modifications and substitutions may be made therein without departing from the spirit and scope and range of equivalents of the accompanying claims. In particular, it will be clear to those skilled in the art that the present invention may be embodied in other forms, structures, arrangements, proportions, sizes, and with other elements, materials, and components, without departing from the spirit or essential characteristics thereof. In addition, numerous variations in the methods/processes described herein may be made within the scope of the present disclosure. One skilled in the art will further appreciate that the embodiments may be used with many modifications of structure, arrangement, proportions, sizes, materials, and components and otherwise, used in the practice of the disclosure, which are particularly adapted to specific environments and operative requirements without departing from the principles described herein. The presently disclosed embodiments are therefore to be considered in all respects as illustrative and not restrictive. The appended claims should be construed broadly, to include other variants and embodiments of the disclosure, which may be made by those skilled in the art without departing from the scope and range of equivalents.
	claims	1. Device (1) for generating voltage dips in an electrical power generator (2) with a control cabinet (4) at the output of the generator and an output transformer (3) to an electricity network (30) characterised in that it comprises:a) a circuit located between the control cabinet (4) of the machine and the output transformer (3) to the network (30) with a transformer (31), a plurality of impedances (11, 14, 17; 12, 15, 18; 13, 16, 19) in-series for each phase, having a first group of switches (24, 25, 26) associated respectively with each phase and a second group of switches (20, 21, 22) associated with the in-series impedances, and a switch (27) for connecting the circuit to ground;b) mechanisms for generating short circuits by activating the first group of switches (24, 25, 26) in order to select a single phase, two-phase or three-phase voltage dip, by activating the second group of switches (20, 21, 22) in order to select the depth of the voltage dip and by activating the third switch (27) in order to generate a short circuit to ground; andc) mechanisms for protecting the network (30) during voltage dip generation. 2. Device (1) for generating voltage dips in accordance with claim 1, characterised in that the mechanisms for protecting the network (30) comprise an inductor (23) to obtain a minimum level of disturbance to the network during voltage dip generation. 3. Device (1) for generating voltage dips in accordance with claim 1, characterised in that the electricity generation machine (2) is a wind turbine. 4. Device (1) for generating voltage dips in accordance with claim 3 characterised in that the power of the wind turbine (2) is between 850 kW and 2 MW. 5. Device (1) for generating voltage dips in accordance with claim 3 characterised in that it also includes bridging mechanisms enabling the wind turbine (2) to be directly connected to the network (30). 6. Device (1) for generating voltage dips in accordance with claim 3 characterised in that it also includes switching mechanisms for connecting/disconnecting it to/from the transformer (3).
	summary
	abstract	A correlation tolerance limit setting system using repetitive cross-validation includes: a variable extraction unit randomly classifying data of an initial DB set into training set data and validation set data at a specific rate and then extracting variables for determining a DNBR limit by optimizing coefficients of a selected correlation; a normality test unit testing normality for a variable extraction result; a DNBR limit unit determining whether data sets have a same population or not depending on normality result and determining DNBR limit from a distribution of 95/95 DNBR; and a controller.
	summary
	description	The present application claims the benefit of priority to U.S. Provisional Application Ser. No. 62/280,569 filed Jan. 19, 2016, of Jeffrey R. Ridgway and Scott H. Bloom, also the benefit of priority to U.S. Provisional Application Ser. No. 62/348,331 filed Jun. 10, 2016, of Jeffrey R. Ridgway, Scott H. Bloom and Daniel Adam Prelewicz, and the benefit of priority to U.S. Provisional Ser. No. 62/384,501 filed Sep. 7, 2016, all priority applications being incorporated by reference in their entirety. The invention disclosed in U.S. Provisional Application Ser. No. 62/280,569 filed Jan. 19, 2016, was made with partial government support under Contract DOE Grant No. DE-SC0013729 awarded by the U.S. Department of Energy, Nuclear Energy NEET. The government may have certain rights in this invention. The present invention generally relates to the technical field of measuring fluid levels in a tank using a gravity meter and, in particular, and by way of example to a gravity-based, non-invasive system applicable to nuclear reactor systems as a method for monitoring the level of coolant in a reactor pressure vessel and also of reporting a gain or loss of coolant fluid in a nuclear reactor module. Known sensors for measurement of coolant/fluid levels in a nuclear reactor module typically are invasive sensors, for example, differential pressure gauges, that must reside inside pressurized modules and must penetrate the module walls, for example, of concentric cylindrical vessels or a single cylindrical vessel, each having, for example, an outer vertical wall comprising, for example, an outer containment pressure vessel (CPV), sometimes referred to as simply a containment vessel or (CNV) and an inner reactor pressure vessel (RPV). Typical pressure ranges in a reactor pressure vessel are on the order of 1800 to 2250 PSI and may be within a larger range of 1000 to 3000 PSI. For example, reactor steam pressure may be typically measured by one or more pressure gauges of an RPV. It is generally necessary that a nuclear reactor system be capable of providing measurements of the fluid level inside a reactor pressure vessel to an accuracy of a few centimeters, for example, for normal regulation of fluid level within a nuclear reactor module. An invasive fluid measurement system is typically used so that fluid level may be controlled, for example, via a Chemical Volume and Control System (CVCS), alternatively referred to as a Makeup/Letdown or a Makeup and Purification System. In the present patent application and claims, Chemical Volume and Control System shall be used generically to refer to all such systems. The CVCS circulates continuously to control both water chemistry and coolant fluid level. There are typically separate inlet and outlet lines for the CVCS. The CVCS typically utilizes fluid flow rate sensors to measure inlet and outlet flow rates. One purpose of the CVCS is to provide a means of regulating fluid level in a reactor so that the fluid level remains within a normal operating band. It is normal for a CVCS to add fluid or remove fluid, for example, from a reactor pressure vessel in order to maintain fluid level in a vessel within the operating band. On the other hand, a leak from a vessel, such as a reactor pressure vessel (RPV), to an external vessel, such as a containment vessel may be indicated by a drop in coolant level below the operating band lower limit. In such a situation, a lowering of the center of mass of the fluid contained within the reactor may indicate a leak, rather than normal fluid level regulation. The loss of fluid sometimes referred to as a LOCA (loss of coolant accident) may require immediate intervention, for example, initiation of an emergency core cooling system, remedial action and possible reactor shut-down and repair. In a NuScale small modular reactor (SMR), for example, the CVCS lines penetrate both the RPV and the CPV, but the lines are not used to measure coolant level; (invasive level measurement systems introduced above may be used). Two types of known invasive sensors used are differential pressure gauges for measuring pressure levels in a pressurized module and heated thermocouples. Another type utilizes a guided wave radar device that penetrates the top of the reactor module to measure the water level in the reactor module. Typically, for installation of either known invasive sensor type (or the radar type), the pressurized modules are pre-drilled to allow the invasive sensors to operate within the reactor module, which is typically cylindrical, and report the measurement result to systems outside the reactor module. The pre-drilling may increase the risk of the undesirable result of a leakage through the reactor module wall(s). A leak may take a long time to develop and so measurements are often taken as a time series. A slow loss of fluid may simply be a sign of normal fluid regulation by the CVCS. So factors in distinguishing normal operation from abnormal operation may include, but not be limited to, the rate of loss of fluid from the reactor module or the fluid level reaching a low fluid level below the band lower limit or the fluid level reaching a fluid level that does not compare with a computer processor estimated fluid level considering other systems such as flow rate, temperature and pressure systems other than the CVCS. The pressure within the pressure vessel may precipitate a vessel failure caused by the invasive nature of any invasive sensor installation through a wall of the reactor module. One may recall the partial meltdown that occurred at the Three Mile Island nuclear reactor in Pennsylvania in 1979. It has been reported that the accident at Three Mile Island was caused by operator error based upon an erroneous estimate of coolant fluid levels in the reactor core. A coolant fluid level regulation problem may be a serious problem that requires a more efficient and accurate solution than an invasive sensor solution to the problem of monitoring fluid level in a reactor system. Consequently, there is a need for an alternative method and apparatus and system for sensing fluid levels, for example in nuclear reactors, that is more accurate and more efficient than known invasive methods and systems for measuring fluid levels in a tank, such as are used in nuclear reactors. This summary is provided to introduce a selection of concepts. These concepts are further described below in the Brief Description and the Detailed Description. This summary is not intended to identify key features or essential features of the claimed subject matter, nor is this summary intended as an aid in determining the scope of the claimed subject matter. The present invention meets the above-identified needs and solves the fluid level measurement problem by, for example, providing non-invasive measurement of a fluid level in a tank. One may mount at least one gravity meter (also referred to herein as a gravimeter or gravimetric mass sensor) outside the tank, preferably as close to the tank as possible, and derive the center of mass of fluid contained within the tank, eliminating noise caused by surrounding effects such as tidal influences, groundwater levels and the like. In one embodiment, one gravity meter may be located as close to the estimated center of mass of the fluid in the tank as possible so any detected gravity data is not impacted by measurements made of levels in tanks nearby the tank under analysis. The present non-invasive method of measuring fluid level may be utilized in a modular system architecture, for example, a pressurized system comprising a CPV and an RPV of a small modular reactor (SMR) or similar modules of other reactor architectures including the architecture of a large nuclear reactor. A module as used in the specification and claims refers to a container for a volume of fluid having a center of mass. Most nuclear reactors involve a tall cylindrical container or module which employs a reactor core to heat water to create steam under pressure which in turn drives a turbine to generate, for example, electricity. Small Modular Reactors (SMRs) offer the promise of nuclear power that is potentially safer and less expensive than large nuclear plants. A known manufacturer and developer of SMR's is NuScale Power, LLC of Corvallis, Oreg. Other manufacturers of SMR's include Babcock and Wilcox (who, with Bechtel, formed Generation mPower LLC with research facilities in Bedford County, Va.) which is represented in the SMR market by their mPower SMR and Westinghouse Electric Company LLC near Pittsburgh, Pa., which is represented in the SMR market by their IRIS under development. Internationally, SMR's are being developed and produced, for example, in China, Korea and Argentina, among other countries, for providing electricity to the public electric grid, as well as for private industry applications in their respective countries and for export. In a known NuScale architecture, several SMR modules, each comprising a concentric outer CPV and inner RPV described briefly above, may be situated side by side or in an array within a common coolant pool, while other SMR designs of other manufacturers may use a cooling air flow to cool the SMR modules. On the other hand, all SMR's and any cylindrical tank containing fluid would find useful a method and system of measuring coolant fluid level that is non-invasive. Moreover, besides determining level, the method and system may determine whether a given loss of fluid relates to normal operation or, possibly, a LOCA. For example, in the petro-chemical field, it is well known that a petro-chemical leak of certain petro-chemicals from a tank may require immediate remedial action and be quite expensive to cure and it may be also useful to determine the location of an abnormal leak so that the leak may be repaired. Measurement of coolant fluid levels within nuclear reactors, as indicated above, is required for reliable and efficient operation and for control of fluid levels within a band of normal reactor operation. In accordance with the present invention, one embodiment of a gravity meter may be located on a platform proximate the center of mass of a tank (of a nuclear reactor module) containing fluid, typically comprising water and added chemicals, and pressurized steam for generating electricity. In order to accomplish the measurement in a coolant pool of water environment, the gravity meter may be placed, for example, in a dry, temperature controlled housing close to the pressurized vessels and/or on a stable platform that may be fixed to the ground level and/or the gravity meter may be fixed to the outside of the outer pressurized vessel or just outside or be attached to the reactor module itself. In order to avoid detecting data from another gravity meter of one reactor vessel proximate another reactor vessel, it is preferred that the gravity meter be mounted on the side of the module and, when one gravity meter is used, as close to an estimated location of a center of mass of the fluid in the module as possible. Hence, an object of an embodiment of the present invention is to use one or more sensitive gravity meters to monitor nuclear reactor coolant fluid levels noninvasively, in particular, for an SMR design architecture wherein a module is typically mounted vertically, and the module is typically tall and relatively thin in diameter. According to the present invention, gravity meters may be used to measure a lowering of the center of mass of a fluid in a tank, for example, attached to an external vessel of a nuclear reactor module. One known gravity meter is the CG-5 AUTOGRAV™ gravity meter manufactured by Scintrex of Concord, Ontario, Canada. Another is the Micro-g Lacoste gPhone gravimeter available from Micro-g LaCoste of Lafayette, Colo., United States. Six Micro-G LaCoste gPhone Gravity Meters, in July 2008, captured the 12 May 2008 Sichuan, China earthquake with extremely high correlation. These gravity meters, however, may not be as sensitive as a superconducting type of gravity meter. The superconducting type of gravity meter may operate by suspending a helium-cooled diamagnetic superconducting niobium sphere in a stable magnetic field. An example of a superconducting type of gravity meter is an iGrav meter manufactured by GWR Instruments Inc., of San Diego, Calif., United States. The iGrav meter has been measured to be more sensitive than the MicroG-Lacoste gPhone meter or the Scintrex CG-5 meter. Other gravity meters may be developed and offered commercially that may be even more sensitive and require less space. For example, the iGrav meter requires a separate system for super-cooling. An object of measuring the changes of and a lowering of the center of mass of a fluid in a tank, for example, a fluid level in a nuclear reactor, may be achieved by applying at least one gravity meter, for example, one of the superconducting type or other gravity meter which accurately measures a fluid level in a given tank to within a few centimeters when a raising or lowering of the center of mass of the fluid in the tank occurs. The gravity meter monitors coolant fluid level in the tank when the gravity meter is placed proximate but outside of the tank, for example, a pressurized SMR reactor module. Known algorithms may be used to minimize noise, for example, ground water influences as will be further discussed herein. Some nuclear reactors rest within a coolant pool while others rest in an air-cooled environment. For example, in the NuScale SMR architecture, a gravity meter may be mounted in a vertical open-ended cylinder on a stable mount in the benign environment of a coolant pool for one of a plurality of SMR modules. In other SMR architectures, the environment may be air and a gravity meter may be mounted to a side of a module not facing other modules so as to minimize noise that may be received by those other modules also having coolant fluid levels to be measured. In one embodiment, a single gravity meter is mounted on a stable mount and may be attached to a reactor module at approximately one half to three quarters of the height of coolant fluid in the reactor module. The output data is output as a time series, for example, of gravimetric data at approximately one second intervals, not intended to exclude other data sampling intervals between less than a second and every minute or every hour, the point being that a fluid loss from any tank or an abnormal regulation of fluid level may occur at any time and be a slow or a fast leak. External gravity signals, such as earth tides, humidity changes in the weather and local ground water levels (soil moisture levels) may be sensed or estimated and separated from the gravimetric signal and excluded from the analysis of the fluid level in a tank. To this end, known software may be used with a computer data processing system to sense and to subtract the influences on gravity meters as noise from the data signals that sense the fluid center of mass in a tank and so monitor and regulate fluid level and/or detect a leak or gain of fluid over time. Whenever fluid level falls below a given level for a given reactor module, it may be necessary to actuate remedial measures. We have determined that multiple gravimeters spaced vertically beside the vertical pressurized vessel(s) of a nuclear reactor module improve the measurement process, for example, a dual-gravity meter configuration is preferable over a single gravity meter. The expected signal from, for example, two gravity meters providing gravity data used for regulating the level of coolant in a module or two gravity meters may detect the occurrence of a loss or gain of coolant fluid in an actual reactor may produce a large difference data signal over a short period of time between the two gravity meters which may indicate (considering pressure, flow rate, temperature and other data taken for the reactor over the same time series) that the normal coolant regulation process is not operating properly or that a LOCA or other event has occurred. Perhaps, some remedial measures must be taken by reactor personnel to restart normal coolant regulation or minimize the impact of a LOCA. This two gravity meter configuration has been demonstrated to allow a resolution of water level monitoring using superconducting type gravity meters to about 3 cm, given typical noise levels and a desired signal-to-noise ratio (SNR) of 2-X (6 dB). A gravity meter system of one, two, three or more gravity meters may monitor fluid level and detect different types of fluid mass loss or normal or abnormal changes of fluid levels. For a LOCA scenario where coolant is leaked outside the facility, for example, via a break in the Chemical Volume and Control System (CVCS) line outside the reactor module (for example, outside a NuScale SMR CPV), the usage of gravity meter measurements may differentiate that scenario versus a contained-leak LOCA where the water leaks, for example, in a NuScale SMR, from the pressurized RPV to the CPV and also may differentiate normal versus abnormal fluid level regulation. Also, the gravity meter measurements may incorporate input data from CVCS fluid flow sensors, temperature sensors and/or reactor pressure sensors as a time series of additional data which may help to indicate a normal or abnormal reduction or addition of fluid to the reactor containment vessel. In a NuScale SMR or other reactor module systems such as those of Westinghouse or Babcock and Wilcox, the non-invasive gravimetric meter system (referred to as a gravity meter in the specification and claims) of the present invention may determine coolant fluid level for regulating the CVCS line delivery and normal removal of coolant fluid as well as determine whether a LOCA has occurred. Thus, a gravity meter deployment may comprise first or first and second gravity meters mounted above one another as close to a module as possible, preferably, at a location as far away from another module as possible. A third or third and fourth gravity meter may also be used at different locations vertically or horizontally to further reduce noise influences in a measurement system according to the present invention and to improve the accuracy of the measurement. In an SMR having a coolant pool, the coolant measurement system may comprise a dry cylinder for holding one, two, or more gravity meters at specified heights on stable mounts and may be open at one end for, for example, power, refrigeration (for example, for super-cooling) and data cables, if necessary. Each non-invasive gravity meter or cylinder containing a gravity meter may be placed as closely as possible or even mounted to an outer vertical wall of the pressure vessel or, in a NuScale SMR, near or attached to the CPV, for example, in a coolant pool in a common tank of a NuScale SMR. The concepts and principles of using at least one gravity meter, associated software systems and other known systems such as flow-rate, pressure and temperature sensing to non-invasively probe in-reactor water levels will be described by the drawings and their brief and detailed descriptions below. In an alternative preferred embodiment, then, at least two gravity meters may be mounted, for example, vertically in a cooling chamber of air. In a further alternative embodiment at least two gravity meters may be mounted in a dry cylinder fixed to the bottom of a coolant pool in at least a partially underground tank surrounding an SMR such that the top of the cylinder is open-ended. Data, refrigeration and power cables, (if needed) may exit the open upper end of the dry cylinder. The data cables may transmit time series gravity signal data, for example, once per second to a data processing system and signal a change in coolant fluid level via a display or other reporting or alerting system such as a sounder. Pressure, temperature and flow rate measurements may be taken at this same once per second time series. In this embodiment, one gravity meter is placed on a stable mount proximate the bottom of the fluid tank whose coolant fluid level is to be measured and, in an SMR, the other gravity meter may be placed on a stable mount at a point, for example, one half to three quarters of the height of the fluid in the tank, proximate to its expected center of mass. In another embodiment, a computer processor may rely on the many sensor systems available in a reactor setting for pressure, temperature, fluid level and fluid in-flow and out-flow to estimate the current fluid level in a reactor module for comparison with the fluid level determined by the one or more gravity meters. Other than cylindrical containment of the two or more gravity meters is also possible such as a container that is rectangular in shape but a cylindrical shape or any shape directly attachable to a module is preferred. One may prefer a shape, for example, that hugs and may be fixed to the outer surface of a cylindrical reactor module and has a curved outer wall and thus provides a means of achieving greater proximity of the gravity meters and related equipment to the nuclear reactor module and center of mass of the coolant fluid. All of the embodiments and further embodiments under development comprise systems for measuring coolant levels in nuclear reactors and may be referred to herein as a GraviSense™ system available from Information Systems Laboratories, Inc. Further features and advantages of the present invention, as well as the structure and operation of various aspects of the present invention, are described in detail below with reference to the accompanying drawings. These drawings will now be described in some detail in the following DETAILED DESCRIPTION wherein similar reference numerals may be used to denote similar elements. The present invention is generally directed to a method and apparatus for monitoring and automatically regulating the fluid level as well as possibly detecting a leak of fluid from a module, the module typically oriented vertically with respect to a gravitational field of the earth, via at least one gravity meter located next to or attached to the module (such as an SMR module comprising a tall cylindrical structure located in a coolant pool or air-filled structure). An SMR module is used by way of example of a nuclear reactor, but the present invention has application in large nuclear reactors (up to over 1000 MWe) as well as applications, for example, in monitoring fluid level and quantifying petro-chemical spills or, generally, leaks from any tanks or other chemical leaks from containers. While the module may be a fuel tank, petro-chemical tank or other tank, for example, on a tanker ship, the specific vessel/module investigated and for which an embodiment of the present invention may be utilized is exemplified by a small modular reactor (SMR) module. One such small modular reactor (SMR) is described, for example, in some detail in “Highly Reliable Nuclear Power for Mission-Critical Applications,” by J. Doyle et al. in Proceedings of ICAPP 2016, Apr. 17-20, 2016, San Francisco, Calif., incorporated by reference as to its entire contents using a NuScale SMR architecture by way of example. The Doyle article submitted by NuScale Power, LLC of Corvallis, Oreg. includes but is not limited to inclusion of a NuScale SMR plant design overview, analysis and a Table I showing calculated initiating LOCA event frequencies and Tables II-V showing, for example, how modules may be swapped in and out of an SMR reactor by number of modules and by given power levels in three cases of available power and module maintenance. Doyle et al. also discusses the percentage of time the plant may operate with an indicated number of SMR modules producing electric power. Another NuScale SMR example is shown in FIG. 1 of the present application and described in priority provisional patent application U.S. Ser. No. 62/348,331 filed Jun. 10, 2016 of Jeffrey R. Ridgway, Scott H. Bloom and Daniel Adam Prelewicz, a priority application incorporated by reference in its entirety. Westinghouse Electric Company (coordinating a team of research companies) and Babcock and Wilcox (in cooperation with Bechtel forming Generation mPower LLC) and international manufacturers of SMR's may design their reactor modules differently from NuScale. However, the great majority, if not all SMR modules, are presently similarly constructed as tall, relatively thin, cylindrical modules having their reactor core at the bottom and steam collection for power generation near the top. A flow of fluid passes the reactor core directly or indirectly (using a steam generator) to create pressurized steam to drive an electric power generator. As discussed above, typically the level of fluid in the system is controlled, for example, via a Chemical Volume and Control System (CVCS) system (which is not, by itself, capable of measuring the level of the water). The CVCS requires a fluid level measurement that may be provided by the noninvasive gravity meter(s) of the present GraviSense™ system and method in order to control the level within a desired band. This task is typically accomplished by invasive sensors such as differential pressure sensors or heated thermocouples known in the art. According to the present invention, changes in or leakage of fluid from the SMR module (used as an example) may be exhibited as a shift of the center of mass of the fluid in the module up or down. Calculation of the center of mass noninvasively with at least one gravity meter, after eliminating noise influences, according to the present invention will assist to quantify a gain in fluid level or loss of fluid/coolant in the SMR module, and, if the loss exceeds a predetermined value for the SMR module (excluding normal CVCS operation to replenish fluid and excluding the pressurizer operating within a predetermined pressure range for the nuclear reactor), that a LOCA event or abnormal regulation of fluid level by a CVCS outside of a band has occurred and remedial action may be required. A NuScale SMR is shown schematically by way of example only and is not intended to be limiting to the present invention. Similar reference numerals are used in the figures to represent similar elements, for example, a cooling pool shown as reference 120 in FIG. 1 is likewise shown in FIG. 2 and the first number of a reference number such as the 1 in 120 indicates the figure number where the element first appears. Another example is upper gravity-meter 220 in FIG. 2 not seen in FIG. 1 but first shown in FIG. 2 such that in subsequent figures such as FIG. 3A and FIG. 3B, upper gravity-meter 220 is seen but is shown to the right of the SMR module aligned vertically with lower gravity-meter 225 in the cooling pool 120 tank. Referring to FIG. 1, a nuclear reactor including a number of modules is shown (for example, having an output of approximately 300 MWe or less) when compared to a large nuclear reactor operated by a major power company which are typically designed to produce 1,000 MWe or higher. These smaller, modular reactors 110-1, 110-2, 110-3 . . . 110-n are housed in a SMR building or structure 100 and may be partially or wholly below the ground level (earth) 130. The SMR modules typically may share a large tank or cooling pool 120 (or air cooling structure, not shown) shown in cross-section, and the coolant pool 120 filled to cover the modules 110. In some SMR reactors, the coolant pool may be replaced by an air conditioned room (not shown). This coolant pool 120 yields a fairly benign environment for an external-to-a-module gravity meter to operate in, as opposed to sensors that are invasive and must connect directly to inside the hot, pressurized and radioactive pressure vessels and are known from the prior art. FIG. 1 also shows room for additional modules than the six shown. The modules are typically tall on the order of 50-80 feet and wide, approximately 6 to 12 feet wide, and, as more clearly seen in FIG. 2, are wholly covered within the cooling pool 120 or air-conditioned room by coolant fluid, for example, water or air, and the modules may be separated from one another by concrete walls. These smaller, modular SMR designs have an advantage of being fabricated at a factory and may be transported to a site by truck or rail where they are installed in an SMR building or structure 100 and actuated once a common cooling pool is filled or the structure air-conditioned (not shown). FIG. 2 will now be described in some detail. FIG. 2 is a schematic picture of a single SMR module in a cooling tank of a water-filled coolant pool 120 and one embodiment of the present invention wherein an upper gravity meter 220 is placed in the vicinity of steam generator headers 240 within an RPV 235. A lower gravity meter may be located at the bottom, for example, within a dry cylinder 210 mounted to the bottom of the water-filled pool 120 common to all modules as seen in FIG. 1; (in SMR modules of other manufacturers, air may surround the modules rather than water). The air cylinder 210, if used, and not to exclude other shaped volumes that may be used for stable sensor platforms, is preferably wide enough to enable maintenance of the gravity meters with, for example, access from the top. Access from the bottom is also possible through the ground 130 or the sides invasively, but the cabling then may be subjected to coolant pool water. Power and data cables, not shown, will likewise exit the dry cylinder 210 top end. The SMR module comprises a reactor pressure vessel (RPV) 235 having its own set of fuel rods 250, control rods 245, and steam pipes 240. Steam is shown differentiated at the top of RPV 235 as white in the gray scale drawing. If the RPV leaks steam, then, the steam may condense against the relatively cool walls of the CPV and collect in the bottom of the CPV tank 230. (A drop in steam pressure and temperature may be sensed as well). Thus, in this scenario #1, a steady internal leak, the upper water level will gradually drop and the CPV will gradually fill up; (see FIG. 3B). In a second scenario #2, a CVCS line may break outside the CPV and allow the RPV 235 to leak externally, such that there is a loss of water from the RPV 235 but no gain in water by the CPV 230. Still referring to FIG. 2, the narrow cylindrical geometry of the SMR module 110 lends itself well to placing a gravity meter (or meters) immediately adjacent to the CPV 235 in the cooling pool 120, for example, on stable mount(s) in a dry cylinder 210, if used. The gravity meter senses the change in gravimetric pull of the fluid mass (in the CPV/RPV module) as it transfers from top to bottom as indicated by, for example, the movement of the center of mass in FIG. 3B. The closer the gravity meter may be placed to the CPV wall, the better, so the distance to a location of a center of mass of fluid may be minimized. Compared with the prior art invasive sensor, the gravity meter will not penetrate the CPV or RPV wall in order to work properly. The gravity meter must have sufficient sensitivity to measure the changes of fluid level in an RPV and also the raising of fluid level in a CPV versus a normal fluid level regulation condition over time (scenario #1). Also, the gravity meter should be sufficiently sensitive to determine when the fluid level falls below a lower limit of a regulation band. It should be capable of determining a lowering of water level in an RPV with no water mass added to the CPV (scenario #2). Scenario #1 will be further explained with reference to FIG. 3A. The impact of one gravity meter measuring gravity time series data of a single module is related as one divided by the distance squared to the center of mass of the fluid in a container. Consequently, the closer an upper or lower gravity meter is to the center of the fluid mass contained in the module, the more accurate the gravity measurement. Conversely, the influence of another, more distant second (or third or fourth) module in a nuclear reactor facility is minimized by this same distance squared factor. Moreover, any change in fluid level of a distant, for example, second module may be measured by a proximate gravity meter (or gravity meter pair) and so subtracted as noise from the gravity meter measurements taken by a gravity meter proximate the first module. As already suggested above, temperature, pressure, in-flow and out-flow fluid rates may be utilized to help differentiate normal fluid regulation from abnormal fluid level. FIG. 3A shows gravity meter placement 220, 225 on a right side of the CPV 230 proximate the CPV wall on stable mounts (not shown) above and below a center of mass of fluid within the RPV 235 where the white cap again represents pressurizer steam. Steam flow to the turbine is generated in the steam generator with headers 240 and is used for generating electric power as seen in FIG. 2. FIG. 3A represents an SMR power module in a cooling pool 120 under normal operating conditions. FIG. 3B, on the other hand, may represent a coolant leak scenario #1 for an SMR module in a coolant tank having common pool 120. In this scenario #1, as opposed to the CVCS line break scenario #2, pressurized water may leak from the RPV 235 to the CPV 230. The water may escape from the RPV 235 (inner cylinder) as steam and may condense as water, filling up the CPV 230 (outer containment cylindrical). The two gravity meters 220, 225 at the two different locations are affected differently by the change in location of a center of mass 310 to new center of mass 320 after the LOCA coolant leak or abnormal CVCS regulation. Consequently, the original center of mass 310 before a loss of fluid lowers its level to a new center of mass 320 after the coolant fluid leak, to a location closer to the lower gravity meter 225. There will now be water in the CPV 230. In a scenario #2 where a CVCS line breaks outside the CPV, there is no coolant fluid fed to the CPV from the RPV 235. The center of mass still lowers but is only a change in center of mass of the coolant fluid in just the RPV 235, and the resultant center of mass is not a combined center of mass including the mass of coolant lost to the CPV 230 as in scenario #1. Introduction to Gravitational Measurement The Earth is a predominantly spherical mass that exerts a large, mean gravitational attraction at any location on its surface above or below sea level (influenced by the tides). However, all bodies of mass likewise exert gravitational force on each other according to the classic equation of gravitational force, which is proportional to the product of the two masses and inversely proportional to their distance from each other. Generally, the force caused by a local mass is described in terms of a “field,” which is force per unit mass that would be felt by a unit test mass. The most commonly used unit of gravitational acceleration is the Gal (named for Galileo) where 1 Gal=1 cm/sec2. The Earth's mean gravitational field is approximately 980 Gal's. Calculations of expected fields resulting from water mass changes due to leaks in an RPV range, for example, from 1 to 26 μGal in range. Consequently, a very accurate gravity meter is needed. As mentioned above, there are a number of gravity meters which are commercially available. A gravity meter of the superconducting type may be preferable over other accurate gravity meters of different types such as those manufactured by Scintrex and Micro-g LaCoste. On the other hand, a gravity meter of the superconducting type has a disadvantage of requiring a nearby, connected super-cooling system to achieve super-conduction. Moreover, even more sensitive gravity meters may be commercially introduced in the future. An iGrav gravity meter instrument manufactured by GWR Instruments of San Diego, Calif., USA may be useful for simulating applicability in the present invention because it is small, has great sensitivity, can be placed and fit in, for example, a tall cylinder or be located as proximate as possible to an SMR module along with its super-cooling system, has small drift and may compensate for external gravity forces on the SMR module automatically (via using two or more such gravity meters, one above the other or at different locations, vertical or horizontal, and mounted on stable platforms). The external gravity forces may comprise, for example, earth tides, humidity changes, ground water levels, and local changes in soil moisture which are separated from the gravimetric signal(s) and some of these noise influences excluded from the analysis if two or more vertically spaced gravity meters are used and a difference signal over time calculated. Considering the external noise signals in measuring a gravitational field because of tides, ground water, vibrations due to rotating equipment, movement of modules within the reactor facility, humidity and other such influences, a differential gravity measurement at different heights may be automatically provided by using two gravity meters, one above the other and subtracting their signals or by using a gravity gradiometer. Gravity gradiometers measure the spatial gradient of gravity using a closely-spaced pair of sensors. An embodiment as seen in FIG. 2 may utilize gravity meters placed, for example, fifteen to thirty feet apart, vertically, depending on the height of the SMR module and expected center of mass of the fluid in the module calculated by a computer processor as will be explained herein. The tidal effect can exceed 250 μGal and is most easily removed using a known sophisticated tidal calculation algorithm, for example, as taught by V. P. Dehant et al. in “Tides for a Convective Earth,” Journal of Geophysics Research, vol. 104, No. B1, pp. 1035-1058, 1999 and incorporated by reference as to its entirety. However, minor residual effects from tidal removal may remain and increase the level of long-period noise in the output gravity data for water level measurement thus decreasing signal-to-noise ratio. Again, these effects may be avoided by differencing the signal in a dual gravity meter or gravity gradiometer setup. Other time-varying effects such as humidity, ground water level, soil moisture and related effects beneath a gravimeter installation are likewise common to two gravitationally aligned gravity meters and may be removed as described by subtraction of one gravity meter data signal from the other. Thus, a multiple-gravity meter setup yields the lowest level of common-mode noise and produces a large signal-to-noise ratio for resolving changes in the level of reactor coolant fluid center of mass. Referring to scenario #1 in FIG. 3A and FIG. 3B, if the total water mass inside the SMR module example of CPV and RPV is considered, then, as a leak progresses, the center of mass of the water moves downward over time. A single gravity meter mounted inside the coolant tank at the bottom of the CPV container and as close as possible to the CPV will experience a negative change in gravity over time as a leak progresses. This is because water mass is transferred from far above the single gravity meter to immediately above it and next to it laterally. Similarly, an increase in fluid level will move the center of mass upward and the single gravity meter will experience a positive change in gravity, for example, during automatic regulation of fluid level. Now, the single gravity meter model will be described in greater detail with reference to FIG. 4. FIG. 4 provides a graph showing the gravity meter response for a leak scenario for a single gravity meter 225 located proximate the bottom of the CPV 235 and slightly outside its cylindrical outer wall (for example, mounted as the lower gravity meter 225 shown in FIG. 2). The gravitational change is plotted in μGal versus the fractional loss of the coolant from inside the RPV 235. The starting point is arbitrarily set to 0 μGal in this graph. At 60% leakage fraction, the change in gravity is measured by the single sensor 225 as minus twelve μGal i.e. the dashed line shows a curve to a value of minus twelve μGal and to 0.6 (or 60%) over a time series of data measurements, for example, a measurement every second. It may be seen that a single gravity meter may easily detect a gain or loss of fluid over time but will have some difficulty detecting which scenario has caused the gain or loss of fluid to occur and may alert for a LOCA or a lower level of coolant fluid under normal operation of a CVCS system but not identify what scenario caused a decrease in fluid level to occur, for example, to a level below the lower level of the regulation band. Temperature, pressure and in-flow and out-flow fluid rate sensors may assist in the differentiating normal from abnormal fluid regulation as well as determining if a fluid level falls below a pre-determined level of a fluid level regulation band. For example, if the determined level of fluid left in the module falls below the lower limit of the band or if the fluid level falls below a computer estimated level of fluid that may be out of the ordinary, remedial action or further testing may be required. Now, a two gravity meter differentiated in height and vertically one above the other proximate the SMR module as seen in FIG. 2 will be discussed for both scenarios with reference to FIG. 5. Referring to FIG. 5, there is shown a graphical plot of data collected from both an upper gravity meter 220 and a lower gravity meter 225 over time where the solid line graph represents the upper gravity meter data and the dashed line graph represents the lower gravity meter data signal over time. FIG. 5 thus provides graphs in solid line and dashed line for an upper gravity meter 220 and a lower gravity meter 225 located, for example, as per FIG. 2. Gravity is shown as a starting measurement point of 0 μGal for both the solid and dashed line graphs over time. A composite signal is formed on a gravity meter display and gravity data is output to a computer processing system, for example, as a time series once per second. The composite signal is formed by an upper gravity meter signal in solid line where the upper gravity meter 220 may be placed approximately halfway up the RPV water height and a lower gravity meter signal in dashed line output by the lower sensor, for example, at a, RPV/CPV module base level directly below the upper gravity meter 220. The differential signal amplitude (between the solid line and the dashed line) runs from 0 at normal to 26 μGal in magnitude, which flattens after a fractional loss of approximately 55% (0.55) coolant is reached. The differential signal clearly demonstrates that a loss of coolant fluid has occurred based on the measured location of the center of mass of the fluid (typically water mixed with chemicals) in the SMR module (for example, an RPV). The graph of FIG. 5 without further investigation could be interpreted as a scenario #1 or a scenario #2 LOCA event or could be normal regulation of coolant level depending on the rapidity of the change in location of the center of mass, for example, or whether the differential level of fluid measured by FIG. 5 suggests that the level of fluid has fallen below the lower limit of the fluid regulation band. Hence, a discussion of LOCA scenario #2 follows in a discussion of FIG. 6 showing LOCA Scenario #2 with reference to the SMR module and use of dual gravity meters 220, 225 with assistance of temperature, pressure and in-flow and out-flow rate sensors (not shown). FIG. 6 is a schematic diagram of an alternative loss of coolant accident (LOCA) leak scenario #2 showing how a leak may be determined by the apparatus of FIG. 2 using one gravity meter 220 mounted on a stable platform above the other lower gravity meter 225, both meters preferably mounted so as to be directly adjacent or even mounted to the exterior wall of the module. In this scenario #2, water may be lost from inside the RPV 235 and may exit the CPV/RPV RMS module via a break in a CVCS line (not shown) connected to the RPV 235 that bypasses the CPV and exits the module. The water level drops due to the outflow of coolant fluid through the CVCS line (not shown) and does not leak, for example, into the CPV 230. Less mass may be left in the system than in scenario #1 because no mass is in the CPV directly next to the lower gravity meter 225. The resulting center of mass is shown as CM″ 410 and should be compared with CM 320 of FIG. 3B to understand that the mass of water is greater in FIG. 3B and scenario #1 than the mass of water remaining after a loss of fluid recorded for scenario #2 as seen in FIG. 6. This scenario as shown in FIG. 6 is identical to a deliberate removal of coolant via the CVCS line for coolant regulation purposes. In the regulated removal of coolant, however, the outward flow of water mass is a known quantity (for example, from an out-flow rate sensor appropriately located) and its gravity signal can be calculated and compared with the measured gravity data. In all further discussions, scenario #2 can be used interchangeably for a specific type of LOCA accident or a deliberate coolant removal via the CVCS for regulation. Now, FIG. 7 will be discussed to demonstrate the differential graphs of scenario #1 versus scenario #2. FIG. 7 provides graphs for both scenario #1, a loss of fluid from RPV to CPV (Gravity Response: Leak Contained), and a leak for scenario #2, (Gravity Response: Leak External), a leak in the CVCS line that connects to the RPV 235 and bypasses the CPV. In scenario #1 (upper graph), the upper gravity meter 220 gravity data in solid line exhibits a strong increase in gravity and the lower gravity meter 225 demonstrates a fairly strong decrease as discussed above with reference to FIG. 5. In scenario #2 (lower graph), however, the upper gravity meter 220 gravity data signal shown in solid line is weakly positive over time and the lower gravity meter data signal shows a very weak positive increase. Using two gravity meters, the two LOCA scenarios may be easily differentiated from one another by the time series of the graphs of the gravity data signals from the two gravity meters and the assistance of other sensor data. But using just one gravity meter data signal over time, either upper or lower, it would be difficult to differentiate the two scenarios and identify or alert as to which LOCA scenario (leak contained or external) has occurred. Now, a data memory and data processing system will be discussed for detecting a gain or loss of fluid, the quantity of leakage or gain in fluid and differentiating between a scenario #1 or #2 with reference to FIG. 8. FIG. 8 shows a flow of data as seen in FIG. 3B for a contained leak scenario #2. The flow of data over the two lines shown from two vertically fixed gravity meters are processed by a computer (data processing) to determine if a gain or loss of fluid has occurred, whether it is a significant gain or loss and the quantity or magnitude of the leak over the time series of measurements. In the case of a loss of fluid or a lowering of the center of mass, the lower fluid level is determined and may be estimated in part by a computer processor from in-flow, out-flow, temperature and pressure data, and, if the fluid level is below a lower limit of a regulation band, then, remedial action may be required. Moreover, one scenario—contained—may be differentiated from another—external—and appropriate action taken. FIG. 8 also provides a further schematic diagram similar to FIG. 2 showing an RPV vessel 235 contained within and concentric within a CPV vessel 230, the upper and lower gravity meters 220 and 225 respectively, and a LOCA leak scenario according to scenario #1 where water has leaked from the RPV 235 to the CPV 230 and the resultant lower than normal center of mass 320. What is added to FIG. 8 are two data lines (power, refrigeration and the like leads, if needed, not shown) transmitting data, for example, up the dry cylinder 210 shown in FIG. 2, (but not shown in FIG. 8) and to data memory or storage 820 as a time series of data for the upper and lower gravity meters 220, 225 respectively (and other sensor data such as temperature, pressure, in-flow rate and out-flow rate data). The gravitational data from gravimeters are typically stored as a time series of gravity measurements in μGal. A data processing system comprising a programmed computer processor receives the time series of gravity data for each gravity meter and from in-flow and outflow rate sensors, temperature and pressure sensors and subtracts out gravitational data from noise such as tidal, groundwater level and humidity and the earth location of the SMR by subtracting one gravity meter signal from the other gravity meter signal at each time period, for example, each second. The computer processor may calculate what a fluid level should be with the measured fluid level and may determine if the fluid level likely has fallen below a predetermined lower limit of the fluid regulation band. As will be explained further herein, there may be three or more software routines working together to analyze the time series of gravity data and other data to distinguish a loss of fluid Scenario #1 from a loss of fluid Scenario #2 or a normal regulation of fluid level in the module, and thus allow personnel to take appropriate remedial action, if necessary, shown as a Leak Detect/Quantity box 840. The placement of the gravimeter(s) can be rearranged and the present invention will still determine the location of the center of mass of a fluid containing module. The gravimeter(s) need to be proximate to the SMR module or other fluid containing module such as a combination of nuclear reactor pressure vessel (RPV) and containment vessel (CPV), but their exact location can be changed (e.g., the gravity meters) could be placed directly on top of the SMR module or other structure, or underneath, for example, in a below-ground cavity) and the method of measuring CM of a fluid in a container will still effectively monitor fluid level and distribution. Software Discussion for a GRAVISENSE™ Fluid Level Measurement System of One or Multiple Gravity Meters In one embodiment of the present invention, three or more software routines are utilized. Data inputs may be provided from one or more gravimeters, plus detectors of, for example, the flow rates provided for in-flow and out-flow of a CVCS for a nuclear reactor, pressure in the module and temperature of the fluid at various vertical fluid levels. The data inputs may be provided to the software routines from a pressure sensor and from a temperature sensor to allow compensation for fluid density changes and permit the computer processor to estimate fluid level at a given point in time. Firstly, the natural time-varying changes in gravitational magnitude (such as tidal effects) are calculated and removed from the data time series in order to determine whether changes in the gravity signal of importance to the reactor are observed (Routine #1). This is, for example, a subroutine to calculate and remove natural gravitational signals from the time series of measured gravity data, such as tides or local groundwater effects. Routine #1 is described by well-known scientific literature including but not limited to: Pertsev, B. P. “Tidal corrections to gravity measurements,” in Izvestiya, Physics of the Solid Earth, July 2007, Volume 43, Issue 7, pp. 547-553 and see also: Dehant, V., P. Defraigne and J. Wahr, “Tides for a Convective Earth,” Journ. Geophys. Res., Vol. 104, No. B1, pp. 1035-1058, 1999, incorporated by reference in their entirety. If changes of importance in the gravity signal are observed, they are compared to a computed signal that is modeled from a calculation software subroutine (Routine #2), assuming an estimate of the change in coolant level within the pressure vessel. The signal thus calculated, when compared to the measured data, indicates whether or not a coolant loss has occurred, by a significant positive correlation between modeled and measured signal, and whether a significant loss of fluid above and beyond a normal loss of fluid (the measured level being within a regulation band) has been detected. (Routine #3). The amount of coolant fluid loss or gain can be quantified by the coolant software calculation (Routine #2). Method of Calculation of Gravity Signal due to Change in Nuclear Coolant Levels First, a method of calculation of a gravity signal due to a change in fluid levels will be discussed (referred to above as routine #2). The gravity signal due to a change in nuclear coolant levels may be calculated using a finite-element treatment of Newton's gravitational equation, utilizing knowledge of the geometry of the coolant fluid inside the reactor, the density of the fluid, and the position of the gravitational data collecting gravity meter with respect to the expected center of mass of the fluid. The gravitational equation is: F → 12 = Gm 1 ⁢ m 2  r → 12  2 ⁢ r ^ 12 ( 1 ) where {right arrow over (F)}12 is the force exerted on the masses due to their mutual gravitational attraction, G is the gravitational constant, m1 and m2 are the mass of the objects, {right arrow over (r)}12 is the distance vector between them, and {right arrow over (r)}12 is the unit direction. Generally, we may describe the force caused by a local mass m2 in terms of a gravity field E12 at an observation point 1, which is force per unit mass that is felt by a test mass, and so Eqn. 1 above is modified so that only the source mass m2 is left in the formula: E → 12 = Gm 2  r → 12  2 ⁢ r ^ 12 ( 2 ) The gravitational field strength has identical units to acceleration. For gravitational studies, we use the units of “Gal” (after Galileo), where 1 Gal is 1 cm/sec2. The magnitude of amplitudes that are measured in nuclear monitoring applications, lie between 0 and 26 micro-Gal (μGal), which is an extremely small quantity. The direction of the field is along the line between the observation point 1 and m2. In general, gravimeters can only measure the vertical component of the field to the required level of sensitivity. The vertical component of Eqn. (2) is easily calculated from knowing the direction of {circumflex over (r)}12. This component of the gravity field can be either positive or negative, depending upon the position of the source mass versus the gravity meter position. In order to evaluate Eqn. (2) for an actual large source mass, we define m2 as a point mass and integrate Eqn. (2) over the source mass's entire geometry. This is well approximated using a finite element approach whereby the source mass is broken into small cubes (voxels), each voxel is assigned a mass m2, the gravity field effect of the voxel's mass is approximated as a point mass using Eqn. (2), and all of the voxels that make up the volume of fluid are summed. The first usage that may be made to the gravity calculation may be to approximate the nuclear reactor and its containment vessel as two concentric cylindrical structures, which hold coolant fluid. This is a simple geometric case and the voxels are created to make use of the cylindrical coordinate system to easily break the body into finite elements without resorting to cubicle voxelization programs (FIG. 9). In this figure, r, radial distance, becomes R in the below equation. The finite element at (R, θ, and Z) has a differential volume given by:dV=RdRdθdZ (3)The finite element mass (m2 in Eqn. 2) is simply the density p times change in volume dV: m2=ρdV. The gravitational effect of this finite element at an arbitrary position is itself a complicated formula, due to its odd shape. However, we can make use of the fact that, given an element of small volume and a distance to the calculation point that is many times greater than its characteristic size, the gravity effect of a differential cylindrical element is very well approximated by the gravity of a point mass of equivalent mass; (see Waldman, C., “Comparison of Gravimeter & Gradiometer Fields,” ISL internal report for Contract N00024-04-C-4179, 2008, incorporated herein by reference as to its entire contents), which is calculated using the formula Eqn. (2). Thus, we may calculate the volume of the finite element and its center position, and compute the gravitational attraction due to an equivalent point mass, the calculation being simplified and rapid. The total gravitational effect of the cylinder is the sum over all finite elements that constitute the cylinder. The algorithm can compute the gravity field of the entire cylinder at any point and in any direction. In the program, the user sets the granularity of the finite elements in R, θ, and Z by setting the number of sections over the radial, circumferential and axial directions. FIG. 10 is an example plot of the finite elements from the cylindrical gravity calculation program setup. (The center of each finite element is plotted and the observation points shown.) The program was tested by setting an observation point on the cylinder axis and comparing to an analytical calculation from a Nettleton formula (1976); see Nettleton, L. L., “Gravity and Magnetics in Oil Prospecting,” McGraw-Hill, New York, N.Y., 464 pp, 1976, incorporated by reference in its entirety. This calculation agrees with an analytical calculation to within 99.9%. Method of Determining from Gravity Data Whether Coolant Levels Changed, and by How Much A method of determining from time series gravity data whether coolant levels have changed, for example, by how much, and, for example, by a predetermined amount signifying a significant gain or loss of fluid may have occurred, i.e. a LOCA or regulation event (a fluid level above an upper limit or below a lower limit of a fluid regulation band has occurred, is now described in some detail (routine #3). Referring briefly again to FIGS. 3A and 3B, this software module may have as its input the gravity data from two gravity meters, for example, in a vertical configuration as shown (gravity meters 220, 225) and are represented by black dots in FIGS. 3A and 3B). A third gravity meter can be used as a “remote” sensor to remove common-mode noise (not shown), which may be placed, for example, 25 to 100 feet away horizontally from the other two gravity meters on their respective upper and lower stable mounts (FIG. 2), for example, in a dry cylinder 210 or on a stable platform. In an alternative embodiment, a third gravity meter may be vertically aligned between the first two gravity meters or at a location above the upper gravity meter. For example, there may be third, fourth, fifth or more gravity meters for further improving accuracy, noise elimination and efficiency. Temperature, pressure, CVCS in-flow and out-flow rates and other data may be weighted and incorporated into the algorithm to determine normal versus abnormal level regulation in a known manner. The data will be processed in the form of time series (for example, gravity data plotted vs. time). The data in each time series first have tides removed (see discussion of routine #1 for tide removal) and are low-pass filtered to remove high-frequency noise prior to the next step. If a horizontally remote gravity meter is used, then its time series is subtracted from the data of the two or more gravity meters proximate to the mass of fluid whose level is to be measured, in order to remove common-mode noise. In a situation where there is no leak in fluid, the residual time series (after tide removal and common-mode noise removal) should look flat over time. We can determine this by examining the slope of the time series of both gravity meters over a set time period (for example, fifteen minutes or one hour at one second intervals) and their slope values should be significantly less than the drift rate and/or root mean square (rms) noise level of an individual gravity meter. If the slope of each gravity meter's time series (#1 and #2) is indeed less than the specified drift rate, then no leak detection can be asserted. If the slope of the time series from both gravity meters exceeds the specified drift rate, then a gain or loss of fluid can be inferred and differentiated from a normal regulation of fluid in the module, and the amount of the gain or loss of fluid can be calculated by using routine #2 and by comparing the calculation to the data. Inherent in the invention is the fact that the data signal so derived can be used for normal coolant level regulation. If the level is decreasing toward the lower level of the control band then the flow into the RPV from the CVCS can be increased (and the in-flow rate should be reflected by a reading of an in-flow rate sensor). Similarly, if the level is increasing toward the upper limit of the control band then the flow from the CVCS can be decreased to a net output of fluid from the RPV. FIG. 11 is a flowchart of the above-described process of a method of determining a loss or gain of fluid from first and second concentric container modules consistent with the above-described apparatus. The process starts at step 1100 with the acquisition of gravity signal data periodically over time. At step 1101, a first gravity sensor mounted, for example, above an expected center of mass of a fluid in a first concentric container module measures first gravity signals. The first gravity sensor is checked for a center of mass reading for a center of mass where fluid is entirely contained in an inner first concentric container. The object of the process is to discover if fluid has leaked from the inner first concentric container to a second outer concentric container or from the second concentric container module to outside the second concentric container module. At step 1102, a second gravity meter is mounted below the expected center of mass of fluid inside the first concentric container module for a second series of gravity signals. At step 1103, a third gravity meter may be mounted horizontal and distant from the first and second gravity meters. Step 1105 relates to calculating gravity drift rate over time and a predetermined value for the expected center of mass (when no leaks are occurring). Step 1106 is a step of removing common mode noise by subtracting the first and second gravity meter signals from one another or from the third gravity meter. Step 1108 discusses low pass filtering the gravity data to remove high frequency noise. Step 1109 calculates the center of mass (CM) of fluid over time at the location of the first and second gravity meters, and step 1110 evaluates when these start to show a lowering of the expected center of mass of fluid (by comparing them to the step 1109 CM calculations), for example, due to a leak from the first concentric container module to the next. Meanwhile, step 1104 determines CVCS, flow rate, temperature, pressure and any other sensors at a given time. Next step 1107 calculates the changes in gravity signals expected from changes in flow rate, temperature, pressure and any other sensors. At step 1111, one may determine if gravity falls outside a predetermined value using a computer processor. If Yes, then go to 1113; if No, go to 1112 and to 1100 to acquire more data over time. If the gravity signals fall outside of the predetermined value, then, step 1113 differentiates, by comparing the signals to the expected center of mass calculations, between: a) an internal loss of fluid occurrence, b) an external loss of fluid occurrence, and c) no loss of fluid. If Yes a) at 1114, then, call it an internal loss of fluid (from one concentric container module to the other at 1117) and if Yes b) call it an external loss of fluid (to outside of both concentric container modules) at 1116, and if Yes c) (normal fluid levels) go to 1115 and return to 1100 to acquire more data over time. While various aspects of the present invention have been described above, it should be understood that they have been presented by way of example and not limitation. It will be apparent to persons skilled in the relevant art(s) that various changes in form and detail can be made therein without departing from the spirit and scope of the present invention. Thus, the present invention should not be limited by any of the above described exemplary aspects, but should be defined only in accordance with the following claims and their equivalents. In addition, it should be understood that the figures in the attachments, which highlight the structure, methodology, functionality and advantages of the present invention, are presented for example purposes only. The present invention is sufficiently flexible and configurable, such that it may be implemented in ways other than that shown in the accompanying figures. Further, the purpose of the foregoing Abstract is to enable the U.S. Patent and Trademark Office and the public generally and especially the scientists, engineers and practitioners in the relevant art(s) who are not familiar with patent or legal terms or phraseology, to determine quickly from a cursory inspection the nature and essence of this technical disclosure. The Abstract is not intended to be limiting as to the scope of the present invention in any way.
048448613	claims	1. A fuel assembly for a nuclear reactor comprising a top end piece, a bottom end piece, a plurality of elongate elements extending between and interconnecting said top end piece and bottom end piece, a bundle of parallel fuel elements located between said top and bottom end pieces, and a plurality of grids spaced apart along said elongate elements and forming cells for retaining said fuel elements at the nodes of a regular lattice, said grids including a plurality of top grids, a plurality of bottom grids and a plurality of median grids, wherein: said median grids are constructed and arranged to withstand lateral shocks and provided with turbulence creating fins, said bottom grids are all located between said median grids and bottom end piece, are devoid of fins and are arranged fro cross-bracing said fuel elements, and said top grids are all located between said median grids and top end piece, are provided with turbulence creating fins and are arranged for cross-bracing said fuel elements, each of said bottom and top grids being constructed to impress to a coolant flow along and within said fuel assembly a head loss smaller than any one of said median grids. said median grids are constructed and arranged to withstand lateral shocks and provided with turbulence creating fins, said bottom grids are devoid of fins and are arranged for cross-bracing said fuel elements, and said top grids are provided with turbulence creating fins and are arranged for cross-bracing said fuel elements, each of said bottom and top grids being constructed to impress to a coolant flow along and within said fuel assembly a head loss smaller than any one of said median grids, wherein said top end piece comprises a first plate having wide openings for passage of the coolant, fixed to said elongate elements and a second plate fastened to said first plate and formed with a checker board pattern having nodes coinciding with the axis of the fuel elements and forming means for limiting longitudinal movement of the latter. 2. The fuel assembly as claimed in claim 1, wherein each said median grid has a height greater than that of the top and bottom grids and comprises cell defining walls and wherein each of said walls is formed with bosses contacting said fuel elements, each of said walls being provided with two bosses offset in the longitudinal direction of the fuel element received in the cell, said bosses being the only fuel element contacting means in said median grids. 3. The assembly as claimed in claim 1, wherein each of said median grids is provided with tongues projecting upstream with respect to the direction of coolant flow, each formed with a stiffening rib transverse to the direction of coolant flow. 4. The assembly as claimed in claim 1, wherein the grids are spaced apart at progressively decreasing intervals in the flow direction. 5. The assembly as claimed in claim 1, further comprising an additional grid devoid of fins, located between said top grids and said top end piece and of a material having a greater neutron absorption than said top, bottom and median grids. 6. The assembly as claimed in claim 5, wherein said additional grid is secured to some at least of said elongate elements and is provided with bosses for supporting shoulders formed on plugs of the fuel elements and the upper end piece is provided with means limiting the upward movement of the elements so that these latter can only extend downwardly by sliding in the other grids. 7. The assembly as claimed in claim 6, wherein said top end piece comprises a first plate having wide openings for passage of the coolant, fixed to said elongate elements and a second plate fastened to said first plate and formed by a checker board pattern having nodes coinciding with the axis of the fuel elements, and forming means for limiting the movement of these latter. 8. The assembly as claimed in claim 7, wherein said elongate elements consist of only some of a plurality of guide tubes and constitute a framework for said fuel assembly with said bottom end piece and a top end piece. 9. The assembly as claimed in claim 1, further comprising a grating located immediately above the lower of said bottom end pieces, said grating being arranged for guiding extensions of plugs of the elements. 10. The assembly as claimed in claim 9, wherein the grating is fixed at its periphery to a skirt belonging to the bottom end piece. 11. The assembly as claimed in claim 9, wherein said extensions of plugs of the fuel elements have a sliding fit in said grating, some at least of said plugs being formed, at their end, with a bulge for retaining the grating. 12. A fuel assembly for a nuclear reactor comprising a top end piece, a bottom end piece, a plurality of elongate elements extending between and interconnecting said top end piece and bottom end piece, a bundle of parallel fuel elements located between said top and bottom end pieces, and a plurality of grids spaced apart along said elongate elements and forming cells for retaining said fuel elements at the nodes of a regular lattice, said grids including top grids, bottom grids and median grids, wherein:
	description	This is a non-provisional application claiming benefit of U.S. Provisional Application Ser. No. 60/998,504 filed Oct. 11, 2007, and entitled Passive Actinide Self-burner, incorporated herein by reference. Embodiments of the invention relate to the field of nuclear waste (radioactive waste materials) disposal and methods. More particularly, embodiments of the invention relate to the disposal and accelerated destruction of transuranic actinide materials that are the residual products of the chemical dissolution of spent nuclear fuel and other components containing fissile materials. Actinide and/or transuranic material destruction occurs naturally along well-understood decay chains to eventually become a stable non-radioactive element, lead. However, in natural decay, some of these elements remain dangerous to man over hundreds of thousands or millions of years. Actinides are separated from spent nuclear fuel, and other components containing fissile materials, by chemical means. Fission and actinide products are left over from the splitting of atoms to make power and are the principal residual material in spent nuclear fuel. The fission products are short-lived by comparison to the actinide products, and are destroyed by the natural radioactive decay process in about one thousand years. Most actinide products, on the other hand, remain dangerous for many centuries. It is this extremely long decay process that results in target isolation objectives of the US Department of Energy (DOE) of 1,000,000 years. In destruction by the natural decay process, the radioactive elements arrive at a stable state by the spontaneous emission of radioactive particles, including alpha particles and neutrons. By increasing the rate at which actinides decay to a stable state, the technological challenge and cost burden of establishing long-lived containment systems for the disposal of spent fuel may be significantly reduced. The potential benefits of effectively reducing the half life of actinide elements has resulted in extensive experimental and operational programs having the sole objective of hastening the transmutation and decay of these products to a stable, and inherently safe, state. Most applications directed to the destruction of actinides by transmutation or fission rely upon high neutron flux rates, similar to those found in operating reactors. To that end, many applications propose the inclusion of waste actinides in some portion of an active fuel assembly so that the material is transmuted or destroyed using the high neutron flux of the operating reactor. While these methods do accelerate transmutation of the actinides to more stable forms thereby reducing the actinide waste quantity requiring disposal, they have the disadvantage of requiring special packaging, high-energy neutron sources and handling of the waste forms, which increases personnel radiation exposure and the risk of accident. The nuclear reactor or particle accelerator operations required by these methods are both complex and expensive. This invention operates by converting the abundant alpha particles emitted by the actinides into neutrons via an alpha—neutron (alpha, n) reaction that is a property of Beryllium and some other elements such as oxygen. This raises the neutron flux of the container to about one ten thousandth of the level present in a nuclear reactor intended to burn the actinides in one or two years. The invention makes use of its passive nature and the 10,000-year minimum period required by regulations to accomplish the same level of destruction of the actinide waste after emplacement in a geological repository. By utilizing the invention, a million year waste isolation period is no longer required, and the shorter 10,000-year waste isolation period is much less complex to analyze and regulate. As described herein, the preferred embodiment of the invention is intended for use with the current series of DOE Standard Canisters (herein after “Canisters” or “Canister”) designed for geologic disposal. An apparatus that provides for the passive destruction of the actinides meets the waste actinide destruction and disposal needs described above. The invention relies upon the neutrons generated within the waste actinides to achieve accelerated destruction, by reflecting those neutrons back into the actinides for efficiency. No neutrons external to the system are introduced into the apparatus. The invention builds upon the current seal source technology that uses the same principles to generate neutrons for industrial testing and well logging purposes, but significantly extends that art to meet a need for accelerating the decay of very long-lived waste actinide isotopes. The invention provides for the confinement of actinide or transuranic radioactive wastes, alloyed with beryllium, inside a graphite disk to cause accelerated destruction (burning) of actinide wastes. Actinides, including plutonium, neptunium, americium, and curium, emit alpha particles by radioactive decay. The alpha particles are converted into neutrons by the beryllium through an alpha-neutron (also called an alpha, n) reaction. The neutrons created in this reaction are absorbed by the actinides causing them to transmute to a heavier actinide isotope with a shorter half-life. An outer layer of graphite is provided to moderate and reflect neutrons back into the actinide zone to improve the efficiency of actinide burning. The outer layer could consist of beryllium metal as the neutron moderator and reflector, but its use results in a prohibitively expensive configuration. The process is passive because the alpha particles that initiate the actinide destruction by radioactive decay are an intrinsic physical property of the actinides. The decay process is accelerated because neutrons that would escape the confinement system are reflected back into the actinide waste where they are captured, reducing the stability of those wastes. The use of a neutron moderator and reflector such as beryllium or graphite to initiate self-burning of actinides allows the quantity of actinides to be reduced much more rapidly than if decay where to occur naturally in accordance with the half-life of the material. Using this method, the actinides can be destroyed in the repository design period of 10,000-year instead of requiring 100,000 years to one million years to attain the same waste reduction by natural radioactive decay alone. The waste actinides may be either individual actinide elements, or a mixture of actinides, which may also be mixed with beryllium metal, in any convenient solid metal, glass, resin, powder or other stable form. The waste actinide product, in addition to the beryllium, may contain binders or other compounds consistent with the stabilization method used. In the preferred embodiment, the waste actinide and beryllium are melted together to obtain a substantially mixed alloy. Intimate mixing of the waste actinides and beryllium improves alpha,n conversion efficiency. The ideal configuration for the waste actinide/beryllium mixture or alloy is a sphere, which would be encased by a spherical shell of beryllium. This configuration presents the best volume to surface ratio for heat rejection. However, it also has the highest cost and is not considered suitable for routine use. The preferred embodiment for the waste actinide/beryllium alloy is as a cylinder sized for use with existing Standard Source capsules. This configuration is consistent with current methods of solidifying or stabilizing actinides, and with current storage and disposal package construction and fabrication practice. More importantly, it is consistent with the configuration of DOE Standard Sources capsules that are manufactured using actinides such as americium and curium, which have been separated from recycled spent nuclear fuel and refined. These sources, which emit neutrons by design, are widely used in industry. Each DOE Standard Source is a cylindrical double-sealed capsule, 0.75 inches in diameter and 2.00 inches in length. The source configuration consists of an inner capsule with domed ends, an outer capsule with domed ends, and the metallic actinide source inside the inner capsule. Double encapsulation is provided to prevent the leakage of radioactive material from the source capsule. The capsules are fabricated from stainless steel. The typical Standard Source contains 3.0 Curies of americium/beryllium in a metallic alloy form with an atomic ratio of beryllium to americium of 13:1. In the preferred embodiment described for this invention, the atomic ratio remains the same, but the actinide is comprised of unrefined actinides consisting primarily of americium, neptunium and curium, but also including a number of related isotopes, and is referred to herein as “waste actinide.” The waste actinides in the source produce alpha particles, many of which are converted to neutrons by collision with the beryllium atoms. The neutrons, in turn, are captured by the isotopes of the waste actinide, destroying a portion of the waste actinide and transmuting a larger portion to heavier isotopes. These heavier isotopes are much less stable than the original isotopes and decay more rapidly than the original isotopes. Thus the quantity of waste actinide remaining at any given future time, is less than if the Am and Cm were disposed of in another form to undergo natural decay. In the preferred embodiment as described herein, individual sealed capsules are placed in wells drilled into a graphite disk. The graphite acts as a moderator for the neutrons created in the alpha, n reaction, allowing the slowed (or thermalized) neutron to transmute other actinide isotopes. Graphite is a cost-effective substitute for beryllium. To provide for structural integrity and facilitate heat rejection, the exposed graphite surface, including the surface of the wells, is faced with stainless steel or aluminum. Once sealed capsules are inserted in the wells of the graphite disk, a closure plate of the same material, acting as a cover, is welded in place over the capsules. Based on the planned use of the Canister, the graphite would have the external shape of a disk, between 18 and 24 inches in diameter, and approximately 3 inches thick. One or more loaded and sealed graphite disks could be stacked within the Canister. The number of capsules installed in each disk, and the number of disks placed in each Canister, is limited primarily by the heat of radioactive decay associated with each capsule. Such limit is determined by appropriate analysis to ensure that heat rejection limits of the repository package are not exceeded. In the preferred embodiment, the number of capsules in the 18-inch diameter graphite disk is 21. FIG. 1 depicts the cylindrical shape of the waste actinide/beryllium mixture (100). In the preferred embodiment, the mixture is an alloy of waste actinide(s) and beryllium. The passive operation of the actinide self-burner relies upon the close proximity of the waste actinide material to beryllium metal. Close proximity is necessary because of the short mean free path of the alpha particles within the mixture, since in the most effective operation, the alpha particle must encounter a beryllium atom in order to generate the neutron that will hasten actinide destruction. Consequently, the actinide, or combination of actinides, and beryllium are substantially mixed. Once mixed, the material may be handled as a dry stable powder or alloy. The cylindrical volume of the mixture is approximately 0.5 inches in diameter (D) and 1.75 inches in length (L). The radioactive contents would be approximately 3 curies of waste actinide material. FIG. 2 depicts the cylindrical capsule that contains the waste actinide/beryllium alloy. In the preferred embodiment, each first (outer) capsule (200) is a cylindrical stainless steel vessel, being 0.75 inches in external diameter (D2) and 2.00 inches in external length (L2). Within this first capsule is a second (inner) capsule (210) with the necessary reduced external dimensions. Each capsule has domed upper ends such that the domed ends can be welded to the cylindrical body of the capsule. Double encapsulation is provided to prevent the leakage of radioactive material from the source capsule. The capsules are fabricated from stainless steel and are individually welded shut by a suitable process and may be leaked tested to verify closure. It is intended that DOE Standard Source capsules, or Standard capsules, be used for the purpose of confining the waste actinide/beryllium material. This allows the capsules to be prepared, filled, sealed and handled in accordance with DOE procedures. Further, the capsules are procured in accordance with the applicable DOE specifications. However, use of the Standard Source capsule is not required to achieve the desired results. Use of the Standard capsules takes advantage of the familiarity of that design to those operators who normally load, handle and maintain these devices. It further ensures the material, welding, testing and handling controls required by the applicable standards for fabrication and use. In the embodiment shown in FIG. 3, structure 300 depicts a graphite disk and illustrates typical placement of a capsule (200) within it. Note that when welded closed, Item 200 contains within it the second sealed capsule (210) and the actinide/beryllium material (100). Structure 300 consists of a 2.5-inch-thick cylindrical graphite reflector disk into which the capsules are inserted. The cylindrical graphite reflector is 17 inches in diameter for the 18″ Canister or 22.75 inches in diameter for the 24″ Canister. The graphite disk is perforated by a number of cylindrical holes or wells 310 (typical) to allow the insertion of the capsules. The disk bottom surface and side is covered with ⅛-inch thick 316 L stainless steel or aluminum (320). Each well is lined with a 3/16-inch thick 316 L stainless steel tube. Once all of the capsule positions are loaded, the graphite disk top surface is closed with a welded cover (330) of the same thickness. The stainless steel or aluminum provides structural support and heat conduction for the capsules. The purpose of the graphite reflector is to slow down the fast neutrons created by the alpha particle collisions with beryllium so that the neutrons can transmute the actinide isotopes. Crumpled aluminum foil or other similar filler material may also inserted into the top and bottom of the cylindrical well that holds each capsule to provide axial support for the capsule and to aid in heat transfer to the stainless steel or aluminum cover of the graphite reflector assembly. The graphite reflector assembly, loaded with source capsules, is seal welded close to provide a third barrier against the release of radioactive material during handling operations. The heat conduction disks are used as separators for the graphite reflector assemblies when the graphite assemblies are stacked within a Canister. The heat conduction disks are fabricated from copper or aluminum and have the same outside diameter as the graphite reflector assembly. The thickness is nominally ¼ inch for copper and 3/16 inches for aluminum. The disks provide a conduction path to the Canister cavity surface for decay heat generated in the capsules. The sealed reflector disks (300), loaded with source capsules (200), may be stacked into a 10-foot or 15-foot long Canister. Each loaded reflector disk assembly is separated from the one below it by a copper or aluminum heat conduction disk. Extra space at the end of the Canister may be filled, e.g. by aluminum foil, to prevent axial motion of the loaded reflector disks. While a Canister is assumed for geologic disposal of such waste, there is no requirement to use the Canister. Any container qualified for long-term storage or testing could be used. The preferred embodiment assumes that the actinide waste material is packaged for geologic disposal. Consequently, the graphite disk(s) are sized to fit the Canister, the DOE disposal package. However, any size graphite disk may be used, including a disk that is designed to hold a single capsule. Such configurations may be required for long-term storage that anticipates future disposal, or for testing. FIG. 4 depicts an alternate graphite disk embodiment in which an outer most 1-inch ring of graphite is replaced by a ring of beryllium and the individual capsule wells are lined with a one half-inch ring of Ricorad™ material. In all other respects, this embodiment is the same as that depicted in FIG. 3. The inclusion of the outer ring of beryllium significantly improves the effectiveness of the apparatus of FIG. 3 by increasing the number of neutrons that are reflected, while the Ricorad™ liner material improves neutron thermalization. These effects combine to cause additional decay events within the waste actinide material. While the beryllium ring and Ricorad™ improve neutron production by a factor of approximately 2.2, the inclusion of these materials significantly increase the cost of this embodiment. FIG. 5 depicts an alternate spherical embodiment of the invention. This embodiment consists of a beryllium sphere (500) approximately 6.0 inches in diameter, having an interior cavity (510) approximately 2 inches in diameter. In this embodiment, the interior cavity is filled with a waste actinide/beryllium substantially mixed powder (100). The fill hole is then closed and sealed. In this embodiment, the beryllium sphere acts in the same manner as the graphite disk previously described. The sphere both moderates the fast neutrons developed by the alpha-n reaction and reflects those slowed neutrons back into the waste actinides, thereby accelerating the destruction of the waste actinides. While this alternate embodiment accomplishes the same objective as the preferred embodiment, the cost of the beryllium shell makes its use unattractive for other than experimental purposes. An embodiment shown by the diagram in FIG. 6 depicts a flowchart of the typical steps of a method of making a sealed capsule and closing the graphite disk. The preferred embodiment includes the steps of measuring 602 by volume and material curie content, a first amount of a waste actinide powder, and a second amount of a beryllium metal powder material; blending 604 the powders to form a uniform first mixture; loading 606 the first mixture into a suitable beryllium crucible and heating the crucible and its first mixture contents to 1375° F. in a vacuum chamber to form a first alloy material; shaping 608 the alloy material in a suitable die; loading 610 the alloy ingot into an inner capsule; sealing, inspecting and testing 612 the inner and an outer capsules; loading 614 the sealed capsules into a prepared graphite disk; sealing 616 the graphite disk by welding a top cover plate; and, loading 618 one or more sealed graphite disks into a Canister for disposal or other container. The measuring step 602 considers that the interior space of the Standard Source capsule volumetrically limits the quantity of the first mixture, but the quantity may also be limited by the curie content of the actinide material. The curie content determines the heat output of each capsule, which must be considered in the managing the total heat load within the Canister. The actinide powder may be comprised of unrefined individual actinide isotopes; or, a mixture of such waste actinides consisting primarily of americium, neptunium and curium, but also including a number of related isotopes. In addition, the quantities of the powders are controlled to achieve a preferred 1:13 atomic ratio of waste actinide to beryllium. The atomic ratio achieved is determined by weighing powders, which may result in some minor deviations from the target ratio. In addition, other ratios can be used, such as 1:167. The blending step 604 is important since the efficiency of the alpha, n reaction requires that the waste actinide material and the beryllium be in close proximity within the mixture. Consequently, substantial mixing of the two powders is required. The process step 606 forms the waste actinide/beryllium alloy by loading the mixed the powders into a beryllium oxide crucible. The crucible is heated to 1375° F. in a vacuum oven until the mixture is a molten alloy. A vacuum is applied to draw off any fluorine gases, F2, which might be present from the chemical processing of the waste actinide(s). The actual temperature applied is somewhat less important than achieving the molten state necessary to form the alloy. However, a temperature of 1375° F. is typically used, and is considered to be an acceptable target value for this purpose. The alloy is cooled in a suitable die 608 to establish the form of the ingot that allows it to be inserted in the inner capsule. The cooled ingot is inspected and polished if necessary. The alloy ingot is inserted 610 into the inner capsule. As previously described, the capsule assembly consists of an inner and outer capsule, which conforms to the design and specifications of the DOE Standard Source capsule. This capsule assembly, and arrangement, are identical to those already supplied to industry as neutron sources, and there are no unique procurement, process or handling steps associated with the use of the Standard Source capsule. While use of the Standard Source capsule is anticipated, its use is not required. Close 612 the inner and outer capsules welding, and inspect and test the completed capsule for leakage. The post seal welding inspection and test activities are preformed individually on the inner and outer capsules. Once sealed, the capsules are loaded 614 into wells in the graphite disk assembly. Within each well, the individual capsules may be supported in the axial direction by crumpled aluminum, or other similar filler material, to provide axial support and to aid in the transfer of decay heat away from the capsule. There are no unique processing activities associated with the graphite disk except to cut or trim the disk to the appropriate dimensions and then to enclose the disk with a stainless steel or aluminum covering to create a disk assembly. The covering lends structural integrity to the graphite disk, protects the graphite from handling damage, facilitates the transfer of decay heat to the walls of the Canister, and allows of the welding of a closure plate to the top of the disk. A top cover is welded 616 to the disk assembly to seal the assembly and to retain the individual capsules in the wells that are incorporated into the disk to receive the capsules. The closure weld is inspected and tested to verify sealing. One or more sealed graphite disk assemblies may be loaded 618 into a designed Canister or other container. Individual loaded disk assemblies may be separated by placing copper or aluminum disks between the disk assemblies to assist in transferring decay heat from the capsules to the interior walls of the Canister or container. The diameter of both the disk assembly and heat transfer disks must be determined based on the interior dimensions of the DOE waste disposal Canister or container intended for use. The number of loaded disk assemblies that may be placed inside a Canister in a stacked array may be limited by either the total decay heat load allowed by the Canister specifications or by the available stacking height within the Canister. Since this configuration is not a requirement of the embodiment, its representation is not provided. However, where stacking of graphite disks is used, separating cooper or aluminum disks to improve decay heat transfer should also be use. The description provided is a preferred embodiment that utilizes a known process and configuration for encapsulating the waste actinide/beryllium alloy. While there are significant benefits in using this encapsulating method, the success of the technique described herein is not dependent upon the use of that encapsulating method. Consequently, in other embodiments, the diameter and length of the alloy ingot and its method of confinement (i.e., such as alloy ingots stacked in a long tube closed at each end) may altered to conform to the intended handling, long-term storage or disposal requirements of given container. The foregoing description of preferred embodiments for this invention have been presented for purposes of illustration and description. They are not intended to be exhaustive or to limit the invention to the precise form disclosed. Obvious modifications or variations are possible in light of the above teachings. The embodiments are chosen and described in an effort to provide the best illustrations of the principles of the invention and its practical application, and to thereby enable one of ordinary skill in the art to utilize the invention in various embodiments and with various modifications as are suited to the particular use contemplated. All such modifications and variations are within the scope of the invention as determined by the appended claims when interpreted in accordance with the breadth to which they are fairly, legally, and equitably entitled.
	abstract	An active sensor 10 is positioned on an outside of a pipe 60 so as to detect a thickness of the pipe. The active sensor comprises: an oscillator 15 capable of inputting oscillatory waves into the pipe and sweeping a frequency of the oscillatory waves within a desired range; and an optical fiber sensor mounted on the pipe, the optical fiber sensor detecting the oscillatory waves generated in the pipe.
	summary
	summary
051280957	claims	1. Grab for lifting and displacing a object between a position submerged in a hot liquid and a position emerged in a gaseous atmosphere, said grab comprising a tubular body for axial sliding movement connected to a lifting means for gripping said object which said tubular body as well as means for gripping said object which are mounted within said tubular body and which are radially movable between a position for picking up and a position for releasing said object, radial displacement of said gripping means being obtained as a result of said axial movement of said slide within said tubular body, wherein (a) said tubular body comprises a bell-shaped open end part, within which said slide is connected to an inner surface of said tubular body over an entire periphery of said tubular body by means of an axially deformable gas-tight elastic member; and (b) said gripping means comprise pawls mounted pivotably on a support fastened to an inside of said bell-shaped open end part, said pawls being arranged within the bell-shaped open end part between said elastic member ensuring its closure and its open end, so that, during introduction of a lower part of said tubular body into said hot liquid in order to carry out pick-up and displacement of said object, gas is trapped inside said bell-shaped open end part and forms a reservoir in which said pawls are located during said pick-up and displacement of said object below the level of hot liquid. 2. Grab according to claim 1, wherein an annular space is formed between the inner surface of said bell-shaped open end and an end part of said slide located within said bell-shaped open end part, said annular space having a first end delimited by said deformable sealing member, said pawls of said grab being arranged in said annular space. 3. Grab according to claim 1, wherein said slide is tubular to allow cooling of an object on which said grab is engaged during handling of said object above the level of said hot liquid. 4. Grab according to claim 1, wherein said deformable sealing member comprises a bellows having corrugations succeeding one another in an axial direction. 5. Grab according to claim 4, wherein said bellows comprises two cylindrical and coaxial envelopes. 6. Grab according to claim 1, wherein said deformable sealing member has a first end fastened to an annular inwardly projecting part of an inner surface of said bellow-shaped open end part and a second end fastened to an annular outwardly projecting part of an outer surface of said slide. 7. Grab according to claim 1, wherein said slide comprises at least two actuating ramps inclined relative to an axis of said slide and of said tubular body and coming into contact with actuating surfaces of said pawls during axial displacement of said slide, in order to ensure pivoting and radial outward and inward displacement of said pawls, so as to carry out pick-up and release of said object.
047388218	abstract	In a reconstitutable nuclear fuel assembly, a top-nozzle-to-control-rod-guide-thimble attachment system employing a reusable locking tube having dimples. The guide thimble has longitudinal slots defining fingers therebetween. The fingers have rim (bulge) portions which engage a groove in the control rod passageway of the top nozzle adaptor plate for top-nozzle-to-guide-thimble attachment. The finger rim portions are prevented from moving radially inward, and thus leaving the groove, by a locking tube placed in the guide thimble. The locking tube has dimples at two elevations. The dimples interact with the other components of the attachment system to assure proper seating of the locking tube during locking tube installation and to assure against unintentional longitudinal movement of the locking tube during fuel assembly handling. An improperly seated locking tube complicates the underwater fuel assembly reconstitution operation.
	description	FIG. 1 is a perspective view of a related art nuclear boiling water reactor (BWR) jet pump assembly 8. The major components of the jet pump assembly 8 include a riser pipe 3 and two inlet mixers 4 that insert into respective diffusers 2. Jet pump restrainer brackets are used to stabilize movement of the inlet mixers 4 and reduce movement of and leakage at slip joint 6 that exists at the interface between inlet mixers 4 and diffusers 2. One type of movement is Flow Induced Vibration, or FIV, that causes slip joint leakage due to high-velocity flows in and around assembly 8. Restrainer brackets minimize relative movement between inlet mixers 4 and restrainer brackets to minimize leakage or damage around slip joint 6. FIG. 2 is a detailed view of related art slip joint 6 that exists between inlet mixer 4 and diffuser 2 of a BWR jet pump assembly. Bottom portion 4a of the inlet mixer 4 inserts into upper crown 2a of diffuser 2. A top edge of diffuser 2 includes one or more guide ears 2b to allow tolerances and easier connection between inlet mixer 4 and diffuser 2. The interface or mating between inlet mixer 4 and diffuser 2 is referred to as slip joint 6. FIG. 3 is a cross-sectional view of related art slip joint 6 between inlet mixer 4 and diffuser 2 of a BWR jet pump assembly, showing internal relationships between components. Lowest distal end 4b of inlet mixer 4 rests in upper crown 2a of diffuser 2, to form slip joint 6. Inlet mixer FIV may occur in the slip joint 6 when tolerances between distal end 4b of inlet mixer 4 and upper crown 2a of diffuser 2 do not exactly match due to wear or improper machining. Leakage may occur at this interface due to both a poor fit and FIV, as fluid coolant leaks between lowest distal end 4b of inlet mixer 4 and upper crown 2a of diffuser 2 and out of the slip joint 6. Example embodiments include slip joint clamps that can vertically join to a diffuser end and laterally push or drive an inlet mixer to stabilize and prevent vibration and leakage in a slip joint between the diffuser and inlet mixer. Example clamps may include clamp arms that are moveable with respect to each other to allow expansion and closing around a slip joint to seat on the diffuser, such as ring halves joined about a clevis pin for example. Example clamps further include structures that push against the inlet mixer, like a lateral drive that transversely pushes the inlet mixer against the clamp. For example, the lateral drive may include a leaf spring that can be biased through a driving bolt and transmission to allow biasing and preloading internal to the clamp from the accessible driving bolt at an exterior surface of the clamp. Example clamps include an axial mount that attach and secure the clamp to a diffuser exterior, such as guide ear clamps that extend around a guide ear common on a diffuser terminus, for example. This axial mounting may permit example clamps to seat on a diffuser end and fill the slip joint around the diffuser end without requiring disassembly or loading on the inlet mixer. The lateral loading may thus independently compress the inlet mixer against an interior of the clamp to prevent vibration in and leakage through the slip joint. Because this is a patent document, general broad rules of construction should be applied when reading and understanding it. Everything described and shown in this document is an example of subject matter falling within the scope of the appended claims. Any specific structural and functional details disclosed herein are merely for purposes of describing how to make and use example embodiments or methods. Several different embodiments not specifically disclosed herein fall within the claim scope; as such, the claims may be embodied in many alternate forms and should not be construed as limited to only example embodiments set forth herein. It will be understood that, although the terms first, second, etc. may be used herein to describe various elements, these elements should not be limited by these terms. These terms are only used to distinguish one element from another. For example, a first element could be termed a second element, and, similarly, a second element could be termed a first element, without departing from the scope of example embodiments. As used herein, the term “and/or” includes any and all combinations of one or more of the associated listed items. It will be understood that when an element is referred to as being “connected,” “coupled,” “mated,” “attached,” or “fixed” to another element, it can be directly connected or coupled to the other element or intervening elements may be present. In contrast, when an element is referred to as being “directly connected” or “directly coupled” to another element, there are no intervening elements present. Other words used to describe the relationship between elements should be interpreted in a like fashion (e.g., “between” versus “directly between”, “adjacent” versus “directly adjacent”, etc.). Similarly, a term such as “communicatively connected” includes all variations of information exchange routes between two devices, including intermediary devices, networks, etc., connected wirelessly or not. As used herein, the singular forms “a”, “an” and “the” are intended to include both the singular and plural forms, unless the language explicitly indicates otherwise with words like “only,” “single,” and/or “one.” It will be further understood that the terms “comprises”, “comprising,”, “includes” and/or “including”, when used herein, specify the presence of stated features, steps, operations, elements, ideas, and/or components, but do not themselves preclude the presence or addition of one or more other features, steps, operations, elements, components, ideas, and/or groups thereof. It should also be noted that the structures and operations discussed below may occur out of the order described and/or noted in the figures. For example, two operations and/or figures shown in succession may in fact be executed concurrently or may sometimes be executed in the reverse order, depending upon the functionality/acts involved. Similarly, individual operations within example methods described below may be executed repetitively, individually or sequentially, so as to provide looping or other series of operations aside from the single operations described below. It should be presumed that any embodiment having features and functionality described below, in any workable combination, falls within the scope of example embodiments. The Inventors have newly recognized that slip joints in nuclear reactor jet pumps often have worn interfaces between diffusers and inlet mixers at the slip joint. The wear may be ¼-inch of depleted metal or other material due to FIV around a perimeter of the slip joint, which can both worsen leakage through the slip joint and render existing slip joint clamps and FIV solutions inoperable without adequate material to seal. Conventional repairs for worn slip joint interfaces may involve disassembly of the inlet mixer, requiring substantial downtime and repair resources. The Inventors have newly recognized a need for slip joint repair without significant disassembly or dependence on pristine slip joint structures that still reduces leakage and FIV in the slip joint. Example embodiments described below uniquely enable solutions to these and other problems discovered by the Inventors. The present invention is clamps that are useable with slip joints in nuclear reactor jet pumps to preload the same. In contrast to the present invention, the few example embodiments and example methods discussed below illustrate just a subset of the variety of different configurations that can be used as and/or in connection with the present invention. FIG. 4 is an illustration of an example embodiment slip joint clamp 100. As seen in FIG. 4, slip joint clamp 100 is shaped to match an inlet mixer and diffuser interface at a slip joint, such as an annular shape. Slip joint clamp 100 is shaped to seat axially onto a diffuser (such as diffuser 2 in FIG. 2) while surrounding an exterior of an inlet mixer (such as inlet mixer 4 in FIG. 2) at a slip joint. Slip joint clamp 100 may completely or partially surround and/or fill a slip joint by seating on a diffuser and surrounding an inlet mixer of the slip joint. Example embodiment slip joint clamp 100 may include two ring halves 120 that are joined to form an annular shape or other shape to match a slip joint shape. Ring halves 120 may be moveable with respect to one another if joined by a hinge or socket or any other relative joining mechanism, including a clevis pin 130, for example. Clevis pin 130 may permit ring halves 120 to expand/be separated in a transverse or radial direction without fully disconnecting or moving relatively in an axial position, allowing clamp 100 to adjust to and move over diffuser and/or inlet mixer structures. In this way, example embodiment slip joint clamp 100 can be installed about a slip joint without disassembly of any inlet mixer or diffuser, because clamp 100 can open halves 120 to fit around such structures and close halves 120 when in place on a diffuser, for example. Example embodiment slip joint clamp 100 may include a fastening element to ensure it may be expanded in a transverse direction so as to be removably installed about a slip joint. For example, a collar bolt 135 may be used to engage and draw together ring halves 120, such as by screwing into one half 120 while being mounted in another half 120, to form a substantially annular shape of clamp 100 with no relative movement of ring halves 120 when collar bolt 135 is engaged between the two. Collar bolt 135 may not apply additional tension or shaping beyond a point when ring halves 120 are fully mated; that is, collar bolt 135 may rigidly yet removably join ring halves 120 into a configuration that mounts on a diffuser without potential for additional clamping from collar bolt 135 when so joined. In this way, example embodiment clamp 100 may remain reliably closed without significantly transversely loading a diffuser on which it seats. Slip joint clamp 100 includes an inner surface 121 that is shaped to seat against an inlet mixer and extend down along an inner surface of a diffuser at a slip joint. Inner surface 121 may be formed by ring halves 120, for example, being brought together about clevis pin 130 and closed into a ring shape. Inner surface 121 may be substantially annular at higher axial positions to match an outer surface of a cylindrical inlet mixer. Inner surface 121 may further include a flange or thinner ring element at a lower axial position that matches an interface between an outer surface of a cylindrical diffuser and an inner surface of a cylindrical diffuser at the slip joint. In this way, example embodiment slip joint clamp 100 may be shaped and sized like a sleeve that internally fits against a diffuser while externally seating on a top of the diffuser and fitting externally against an inlet mixer. Slip joint clamp 100 may include axial joints or anchors that retain clamp 100 on an upper end, such as a crown, of a diffuser at a slip joint. For example, ear clamp 181 may be shaped and sized to clamp around a guide ear of a diffuser (such as ear 2b in FIG. 2) in order to axially hold clamp 100 at a top end of a diffuser. A draw bolt 182 may be paired with ear clamp 181 to allow axial movement of ear clamp 181 and thus clamping against a lower side of the ear. Further, a ratchet surface 190 or other locking mechanism can permit one-way movement or tensioned securing of draw bolt 182 when paired with a matching ratchet surface of draw bolt 182 in example embodiment clamp 100. When draw bolt 182 is turned, ear clamp 181 may be drawn upward against an ear or other surface, axially clamping clamp 100, and ratchet surface 190 may prevent reversing of draw bolt 182 and thus loosening. Multiple sets of ear clamp 181, draw bolt 182, and ratchet 190 may be positioned about clamp 100. In this way, clamp 100 may be axially secured to and tightened against each guide ear at multiple radial positions, ensuring clamp 100 remains stationary and secure while exerting axial clamping forces only against a top end of a diffuser. When halves 120 and inner surface 121 are shaped to substantially fill a slip joint between a diffuser and inlet mixer, axial securing of clamp 100 may prevent fluid from escaping the slip joint. Because guide ears are less likely to become worn through FIV and other jet pump operations, they may be used for axially clamping and anchoring clamp 100 without regard for wear or other damage that may have occurred inside a diffuser or inlet mixer at the slip joint. Example embodiment slip joint clamp includes a lateral-loading drive that can independently push or bias an inlet mixer at a slip joint to a desired preloading condition. Such lateral loading may secure the inlet mixer against inner surface 121 and further prevent FIV and leakage. The lateral-loading drive provides at least up to 750 pounds-force of lateral preload against an inlet mixer. For example, a leaf spring 140 may be laterally driven by a lateral driving bolt 160 mounted in a top plate 150. A ratchet surface 170 may allow tightening or one-way movement of driving bolt 160 until 750 or more pounds of force are exerted by leaf spring 140. Additional operational examples of driving bolt 160 and leaf spring 140 are described below in connection with FIG. 6. FIG. 5 is an illustration of example embodiment slip joint clamp 100 as installed at a slip joint between inlet mixer 4 and diffuser 2, such as related art structures of FIGS. 1-3 in existing or future nuclear reactor jet pumps. Inlet mixer 4 and/or diffuser 2 may have been damaged or subject to extensive FIV during operation or otherwise, causing wear and damage to their surfaces where they mate at the slip joint. As such, there may be significant fluid leakage and relative movement between inlet mixer 4 and diffuser 2 without slip joint clamp 100 being installed. Or, inlet mixer 4 may require axial adjustment with respect to diffuser 2 without disassembly or removal of the entire slip joint when installing an example embodiment clamp. As shown in FIG. 5, example embodiment clamp 100 can axially seat at a top terminus, or crown, of diffuser 2 about a slip joint with inlet mixer 4. This positioning may be achieved without movement of diffuser 2 or inlet mixer 4, because of how clamp 100 can be opened and closed or otherwise fit around these structures. For example, clamp 100 may be opened in lateral or radial direction 95 to fit around inlet mixer 4 without disassembly of the same, and then clamp 100 may be closed to fit about a top end of diffuser 2 by radially closing halves or other portions of clamp 100, potentially about a clevis pin or other fastener. Collar bolt 135 may secure clamp 100 in a closed, continuous shape in radial direction 95 on diffuser 2. Example embodiment clamp 100 may further be axially secured to diffuser 2 in order to prevent relative movement between clamp 100 and diffuser 2. For example, clamp 100 may be rotated in radial direction 95 until each ear clamp 181 is positioned axially under a corresponding guide ear 2b of diffuser 2. Draw bolt 182 may be tightened to move ear clamp 181 upward in axial direction 181, such as through threads or another connection. Because both ear clamp 181 and draw bolt 182 may be seated in clamp 100, this axial movement may cause a net axial downward force on clamp 100, securing clamp 100 to diffuser 2 in an axial direction. Ratchet surface 190 may prevent loosening of draw bolt 182 in order to maintain the secured positioning. Example embodiment clamp 100 can be axially secured to diffuser 2 despite potential wear or damage to terminal or inner surfaces of diffuser 2. Moreover, example embodiment clamp 100 can be axially secured as seen in FIG. 5 without removing or requiring movement of inlet mixer 4 in axial direction 90, because clamp 100 can be axially secured through guide ears 2b. Thus, inlet mixer 4 can still be axially adjusted and repositioned with respect to a slip joint during installation of example embodiment clamp 100. Securing clamp 100 via guide ears 2b may also prevent relative movement of example embodiment clamp 100 in a radial direction 95. FIG. 6 is a cross-section of the Detail A region of FIG. 5. As seen in FIG. 6, inner surface 121 of clamp 100 may seat down into diffuser 2, such that clamp 100 is flush against an inner perimeter of the same. For example, inner surface 121 may have a lower flange or other fitted section that narrows to fit within slip joint components. A bottom outer portion of inlet mixer 4 also seats against inner surface 121 of clamp 100. In this way, a narrowing portion of example embodiment clamp 100 and/or an otherwise specially shaped inner surface 121 may fit down into and seal a slip joint between diffuser 2 and inlet mixer 4, regardless of wear, damage, or non-fit among ends of those pieces and without requiring disassembly of those pieces for installation. As further seen in FIG. 6 a lateral loading structure is useable in example embodiment clamp 100. Lateral driving bolt 160 may be extended through top plate 150 and extend down into a chamber 167 inside of example embodiment clamp 100 (FIG. 4). Wedge 154 may be secured to or a part of driving bolt 160 and captured by chamber 167, except where a wedged or angled surface seats against a surface of leaf spring 140 in inner surface 121 of clamp 100 (FIG. 4). Leaf spring 140 may additionally be axially restrained adjacent to chamber 167 by top plate 150. Because top plate 150 and chamber 167 may be bolted to, in integral within, example embodiment clamp 100, driving bolt 160 may be axially driven upward or downward relative to clamp 100 when seated in threads in top plate 150. Such axial movement in wedge 154 translates to transverse or radial compression of leaf spring 140 due to the angled surfaces and otherwise captured nature of leaf spring 140 and wedge 154 in chamber 167. A desired axial displacement or resultant force may be sustained through ratchet surface 170 or another lock that prevents drive bolt 160 from further moving after being set at a desired axial position. Thus, when clamp 100 is axially secured, such as to diffuser 2, a transverse load may be applied internal to clamp 100 without axial and/or diffuser involvement. FIG. 7 is an illustration of an example leaf spring 140 illustrating its shape and compression. Leaf spring 140 may extend some distance along a perimeter of inner surface 121 (as seed in FIG. 4) and be shaped such that under compression leaf spring 140 compresses against, and distributes force along, some length of an outer surface of inlet mixer 4. For example, leaf spring may extend along an eighth or more of a perimeter of inlet mixer 4. Leaf spring 140 may define a central void 145 where a post from top plate 150 (FIG. 6) may extend through to retain leaf spring 140 in chamber 167 in contact with wedge 154 (FIG. 6). As seen in FIG. 7, force may be applied in transverse direction 99 when leaf spring 140 is installed in an example embodiment clamp at a slip joint and in contact with inlet mixer 4. Such force may come from, for example, wedge 154 being drawn up in cavity 167 by driving bolt 160 in FIG. 6. As the force in direction 99 approaches a desired preload force, such as 750-lbs, outer contact pads 141a and 141c of leaf spring 140 may extend angularly (direction 99 of FIG. 5) along inlet mixer 4, distributing such force. Finally, central contact pad 141b may come into contact with inlet mixer at the desired preload force, essentially distributing a large static force in a radial direction against inlet mixer 4. This force may compress inlet mixer 4 against an opposite inner surface 121 in clamp 100 (FIG. 4). Under a sufficiently large preload force, such as 750 pounds-force or more, inlet mixer 4 may be prevented from moving relative to a diffuser or undergoing FIV via this contact with clamp 100. As seen, example embodiment slip joint clamp 100 can be axially secured to a diffuser and independently bias an inlet mixer. Installation on the diffuser may require attachment only to guide ears or other external structures without movement or involvement with an inlet mixer. Subsequent to installation on an end of the diffuser about a slip joint, example embodiment clamp 100 may be laterally biased via a lateral drive. This lateral biasing may exclusively preload the inlet mixer with up to or exceeding 750 pounds force in a lateral or radial direction to seat the inlet mixer against the clamp perimeter without involvement of the diffuser. This independent axial attachment to the diffuser and lateral preload of the inlet mixer may permit installation of example embodiment clamps on a variety of slip joint types and in varying conditions, reduce leakage through such slip joints, and prevent FIV in and damage between slip joint components. Example embodiment clamp 100 may be fabricated of any materials that are compatible with an operating nuclear reactor environment, including materials that maintain their physical characteristics when exposed to high-temperature fluids and radiation. For example, metals such as stainless steels and iron alloys, nickel alloys, zirconium alloys, etc. are useable in example embodiment clamp 100. For example, leaf spring 140 may be X750 inconel of approximately 1-inch radial depth/thickness to provide a spring constant the yields up to 750 lbf preload force when compressed across its thickness. Bolts, clamp body, and connectors may be fabricated of stainless steels and other compatible materials to prevent fouling or metal-on-metal reactions. Example embodiments and methods thus being described, it will be appreciated by one skilled in the art that example embodiments may be varied and substituted through routine experimentation while still falling within the scope of the following claims. For example, a generally annular slip joint connection has been shown in connection with an example; however, other configurations and shapes of slip joints, and diffusers and inlet mixers therein, are compatible with example embodiments and methods simply through proper dimensioning and placement—and fall within the scope of the claims. Such variations are not to be regarded as departure from the scope of these claims.
	abstract	A chamber for exposing a workpiece to charged particles includes a charged particle source for generating a stream of charged particles, a collimator configured to collimate and direct the stream of charged particles from the charged particle source along an axis, a beam digitizer downstream of the collimator configured to create a digital beam including groups of at least one charged particle by adjusting longitudinal spacing between the charged particles along the axis, a deflector downstream of the beam digitizer including a series of deflection stages disposed longitudinally along the axis to deflect the digital beams, and a workpiece stage downstream of the deflector configured to hold the workpiece.
	summary
	abstract	The invention relates to a smoke analysis characterization cell employing optical spectroscopy, which comprises: a reaction chamber, an inlet orifice for injecting smoke into the reaction chamber; an outlet orifice for discharging the smoke from the reaction chamber; and an analysis window for the entry of a laser beam intended to form the plasma inside the reaction chamber, which cell is characterized in that the system further includes a blower for blowing an inert gas close to the analysis window; and a shielding gas injector for the shielded injection of the smoke into the reaction chamber, the shielding being provided by a jet of inert gas around the smoke.
042082498	claims	1. In combination with a fuel assembly for a nuclear reactor of the type having a plurality of long and slender fuel rods, each of said rods having two ends, said rods being spaced from each other and generally parallel to each other in the lengthwise direction, the improvement comprising a cellular grill engaging one of the ends of each of the fuel rods, all of said grill engaged ends being adjacent to each other at the lengthwise rod ends, a hollow guide post coupled on one end to the grill for movement therewith, said guide post having at least one lengthwise slot-formed therein, said slot extending from the end of the guide post that is spaced from the grill through a portion of the guide post length, a plate mounted within the guide post for movement therewith, a pad disposed generally transverse to the guide post slot, said pad having an aperture formed therein to receive the guide post and to accommodate movement of the guide post in a lengthwise direction relative to said pad, and a plunger within the guide post and secured to the pad, said plunger progressively blocking and unblocking the guide post slot in response to relative motion between the plunger and the guide post. 2. A fuel assembly according to claim 1 further comprising spring means circumscribing said guide post, said spring means being interposed between said grill and said pad. 3. A fuel assembly according to claim 1 wherein said plate further comprises an orifice formed therein to establish fluid communication through said plate in said lengthwise direction. 4. A fuel assembly according to claim 2 wherein said plate further comprises an orifice formed therein to establish fluid communication through the plate. 5. In combination with a fuel assembly for a nuclear reactor of the type having a plurality of long and slender fuel rods, each of said rods having two ends, said rods being spaced from each other and generally parallel to each other in the lengthwise direction, the improvement comprising a cellular grill engaging one of the ends of each of the fuel rods, all of said grill engaged ends being adjacent to each other at the lengthwise rod ends, a hollow guide post coupled on one end to the grill for movement therewith, said guide post having at least one lengthwise slot-formed therein, said slot extending from the end of the guide post that is spaced from the grill through a portion of the guide post length, a pad disposed generally transverse to the guide post slot, said pad having an aperture formed therein to receive the guide post and to accommodate movement of the guide in a lengthwise direction relative to the pad, and a plunger within the guide post and secured to the pad, said plunger progressively blocking and unblocking the guide post slot in response to relative motion between the plunger and the guide post. 6. A fuel assembly according to claim 5 further comprising a plate mounted within the guide posts having an orifice to establish fluid communication through the plate.
	summary
045490830	summary	BACKGROUND OF THE INVENTION An X-ray is an electromagnetic wave having a high energy which has a power to penetrate through many materials, e.g.- a human body, packed materials, industrial products and the like, and a transmitted X-ray has a space distribution in its intensity due to the difference in the absorption coefficients in the materials. If our eyes had the ability to see X-rays, we would be able to see the space distribution of the transmitted X-rays, i.e.- the inside of the materials. Unfortunately, our eyes cannot detect X-rays. To observe the inside of the materials, a conversion from X-rays to another media, which can be observed with our eyes, is absolutely necessary. However, there is no available three dimensional image detector, and a transmitted X-ray is usually projected on a two dimensional plane (e.g.- a screen), which detects X-rays, so as to observe the inside of the materials. A typical X-ray detecting screen is an X-ray film which is coated with an X-ray sensitive emulsion of a suspended fine powder of silver halides. When the X-ray film has been exposed to X-rays, a latent X-ray image is formed in the X-ray film and the exposed film is chemically developed in a darkroom to semi-permanently fix the X-ray image on the X-ray film. The developed X-ray film has different reflections (or transmittances) with respect to light, e.g.- dark portions correspond to exposed areas and light portions correspond to unexposed areas. Thus, doctors in hospitals can diagnose the inside of a body by observing the developed X-ray film under a light or by transmitted light, and the inspectors in industrial processes and security areas can nondestructively inspect the inside of materials of industrial products and packages. Thus, the X-ray images are essential for enabling doctors in hospitals and clinics to diagnose patients, and for enabling inspectors to inspect the products in quality control areas of industry, and for enabling security guards to inspect packages in security areas. Because X-ray film has a poor sensitivity to X-ray exposure, a large dosage of X-rays to a patient's body is needed for the formation of a proper X-ray image on the X-ray film for use in diagnostic purposes by doctors. The large dosage of X-rays to a body is hazardous and sometimes causes severe damage to the health of patients. The reduction of the dosage of the exposure of X-rays to the human body is a primary subject for radiologists. A successful approach is an application of a fluorescent intensifier screen which converts X-rays to a fluorescent light in the wavelengths of the maximum senstivity of X-ray film, noting that X-ray film is more sensitive in the near ultraviolet light range in comparison to X-rays. If the fluorescent intensifier screen is attached to the X-ray film, then the X-ray film absorbs both X-rays and fluorescent light; naturally, the formation of the X-ray image on the X-ray film is significantly improved. The fluorescent intensifier screen is coated with phosphor crystals which have a high absorption coefficient of X-rays and a high conversion efficiency of luminescence. However, the application of the fluorescence intensifier screen does not eliminate the essential difficulties of the use of X-ray film. They are: the need for a chemical process in a darkroom; the need for a treatment of the image information to enhance the objective images on the X-ray film; and storage of the developed X-ray films. The X-ray image intensifier tube, i.e.- a special cathode ray tube consisting of a layer that emits photoelectrons as X-rays are absorbed, and electrodes for accelerating the photoelectrons, and electric lenses and a phosphor screen which emits cathodoluminescence when the accelerated photoelectrons hit the phosphor screen, has been developed to eliminate the need for a process in a darkroom. The X-ray image intensifier tube detects a small limited area of the body, and the X-ray image on the phosphor screen is only observed during the exposure of X-rays to the body. This increases the X-ray exposure dosage on the body which is needed for diagnosis by doctors. The permanent or semi-permanent recording of the X-ray image on a screen may therefore reduce the X-ray exposure dosage on the body. There has been an attempt to record the X-ray image on the phosphor screen, that is, a phosphor screen utilizing thermoluminescence. Some phosphors emit thermoluminescence if the phosphors which are exposed to X-ray radiation are heated to an elevated temperature by either irradiation by an infrared light or by a laser beam which is focused on the exposed screen. However, the sensitivity to X-rays and intensity of thermoluminescence are not high enough to obtain a sharp image. A more practical device utilizes an electrostatic image formed on the dielectric layer. The dielectric layer is charged as the electric field has been applied, and the charged layer is discharged as X-rays irradiate the charged layer. The sensitivity of such a device is not high enough for a wide application to radiology, limiting the application of such a device. SUMMARY OF THE INVENTION The object of this invention is to provide a cathode ray tube which can display X-ray images on a phosphor screen. In more detail, the present invention relates to a phosphor screen in a cathode ray tube which has no cathodoluminescence if the phosphor crystals in the phosphor screen are persistently polarized, and in which the phosphor screen emits steady cathodoluminescence if the phosphor crystals in the screen are depolarized. In the cathode ray tube of the present invention, the phosphor crystals in the phosphor screen are persistently polarized by an external electric field which has been applied across the phosphor screen, and the persistently polarized crystals in the screen are then depolarized by the irradiation of X-rays. Hence, when X-rays which are trasmitted through a human body or materials irradiate the persistently polarized phosphor crystals arranged on the phosphor screen in the cathode ray tube of the present invention, the phosphor screen displays an X-ray image on the screen with a steady cathodoluminescence. The cathode ray picture tube of the present invention essentially consists of: (a) -an envelope which keeps the inside of the tube at a high vacuum; (b) -an electrically conductive and optically transparent face plate which has a low absorption coefficient with respect to X-ray radiation; (c) -a phosphor screen formed on the inside of the face plate; (d) -a reading gun which supplies a sharply focused high energy electron beam; (e) -flood guns which steadily supply low evergy electrons, showering them uniformly throughout the phosphor screen; and (f) -electrodes for collecting both electrons repulsed from the phosphor screen and secondary electrons emitted from the phosphor crystals. In the picture tube of the present invention, the penetration of the low energy electrons from the flood guns into phosphor crystals, giving rise to the cathodoluminescence from the phosphor screen, is controlled by means of the persistent polarization and depolarization of the phosphor crystals.
054855004	description	DESCRIPTION OF THE PREFERRED EMBODIMENTS Embodiments of the present invention will be described in detail with reference to the accompanying drawings. FIG. 1 is a schematic diagram showing the structure of a chest X-ray imaging system according to an embodiment of the present invention. The chest X-ray imaging system of this embodiment includes components, such as an imaging sequence controller 1, an X-ray tube high voltage generator 2, an X-ray tube 3, an X-ray iris and filter 4, an X-ray slit 5, an X-ray grid 6, an X-ray image intensifier 7 (hereinafter abbreviated as X-ray II), an optical lens and mirror unit 8, an iris 9, an ultra high precision television camera 10, an image acquisition and processing unit 11, an image display unit 12, a data storage unit 13, a position change mechanism 14 for controlling the position of an X-ray detection unit 16 enclosed by a broken line, and a position change controller 15. The imaging posture of a subject is a standing position, or a dorsal or ventral position on a bed. For the standing position, an X-ray incident plane of the X-ray detection unit is positioned generally upright on the floor (like shown in FIG. 1). For the dorsal or ventral position, X-ray beams are applied to a subject from the higher or lower position and the detector is positioned under or above the bed. In the example shown in FIG. 1, the X-ray detection unit 16 moves perpendicular to the surface of the drawing sheet (parallel to the floor) for imaging a subject a plurality of times. The television camera 10 uses as its image pickup device a high resolution image pickup tube 20. The X-ray II 7 has an effective input view field diameter of about 14 inches, corresponding to a field of view covering a single lung and mediastinum. The X-ray incident plane of the X-ray detection unit 16 is set vertical to the floor and to the center line inter-connecting the X-ray tube and the center of the X-ray incident plane. The function of each component will be summarized in the following. The imaging sequence controller 1 defines the imaging sequence of the X-ray detection unit 16 for X-ray imaging a single lung (right or left lung) and mediastinum at two predetermined positions, namely it defines the width and period of two X-ray pulses, the amplitude of a high voltage (X-ray tube voltage), the value of a tube current, and a position change of the X-ray detection unit 16. The position change controller 15 controls the position change mechanism 14 to maintain the X-ray detection unit 16 still at the two predetermined positions during X-ray imaging and move it during the period between the two X-ray pulses. In this manner, X-ray imaging can be performed two times consecutively at the two positions. The X-ray slit 5 has a predetermined width through which an X-ray is transmitted, and is adapted to move in one direction and change its position with a changed imaging region. The high voltage generator 2 generates a voltage and current and the X-ray tube 3 generates an X-ray, in accordance with an imaging sequence. An X-ray transmitted through the subject 17 is applied to the X-ray II 7 via the X-ray grid 6 at which the X-ray is attenuated by shielding its scattered X-ray. An X-ray image projected to an input phosphor screen 18 of the X-ray II 7 is converted into a visual image on an output phosphor screen 19. The optical lens and mirror unit 8 focusses the visual image onto the television camera 10 which converts the image into a video signal to be inputted to the image acquisition and processing unit 11. At the image acquisition and processing unit 11, the video signal is A/D converted and stored in an internal frame memory. Geometric distortion and image level (density) shading of the two digital X-ray images inherent to the X-ray detection unit are corrected. Two corrected images are joined together so as to align the subject regions commonly contained in the two images, image-processed, displayed on the image display unit 12, and stored in the data storage unit 13. The data display unit 12 displays either a single lung image, a joined image, or a composite image of right and left lung juxtaposed images. FIG. 2 illustrates an example of a position change sequence in the imaging sequence. Both the position (position A) of the X-ray detection unit 16 covering the region of one lung and mediastinum and the position (position B) covering the region of the other lung and mediastinum are determined beforehand. Typically, positions A and B are located at the right and left sides of a subject, 85 mm from the center line of it. First, the position of the X-ray detection unit 16 is set to position A. In the typical operation mode of the television camera shown in FIG. 2, a video signal is read 30 frames per second and 1000 scan lines per frame. The read time is 33 ms, and the X-ray irradiation time is 5 ms. The irradiation time may be set to, for example, about 30 ms. At the frame when an X-ray is irradiated, an image-read-out scan is stopped by the frame blanking operation of the television camera to record an X-ray image on an image pickup screen. At the next frame, the frame blanking operation is stopped and the image-read-out scan is executed to read first image data and store it in the image acquisition and processing unit. Next, the X-ray detection unit 16 is moved to position B. An average motion speed is typically about 50 cm/s, and it takes 330 ms to move the unit 16. After the unit 16 is moved to position B, second image data is acquired in the similar manner to the case of position A. FIG. 3 illustrates an example of an image joining algorithm. This algorithm includes an image correction process for correcting the sensitivity and geometric distortion of a plurality of digital X-ray images and an image joining process for joining the plurality of digital X-ray images by aligning the common subject regions contained in the images, and for correcting the image densities. The image correction process 320 is divided into a pre-process 300 for generating correction tables prior to X-ray imaging a chest and a post-process 310 for correcting image data. The pre-process 300 generates a non-uniformity correction factor table (302) for correcting image sensitivity non-uniformity (303-1, 303-2, . . . ) and a corresponding position relation table (304) for correcting image geometrical distortions (305-1, 305-2, . . . ). In order to generate the non-uniformity correction factor table (302), a standard chart having a uniform X-ray transmittance is imaged (301). In order to generate the corresponding position relation table (304), a standard chart having elements (such as holes) at known positions is imaged (303). In order to generate the non-uniformity correction factor table, the standard chart with the uniform X-ray transmittance is imaged to measure an image-level (density) data of each pixel detected by the detector. The measured data itself, or the measured data with high frequency noises removed therefrom by smoothing it, is stored in the non-uniformity correction factor table. The image data D(x, y) of a subject is given by the following equation (1) in the x-y coordinate system of the X-ray detector: EQU D(x, y)={T(x, y)+S(x, y)}* H(x, y) * A(x, y) (1) where A(x, y) represents an X-ray emission intensity, T(x, y) represents an X-ray transmittance distribution of the subject, and S(x, y) represents a scattered X-ray distribution. The image data B(x, y) of the chart with a uniform X-ray transmittance is given by the following equation (2): EQU B(x, y)=kH(x, y)A(x, y) (2) where k is a constant. For example, for the non-uniformity factor correction, the left side of the following equation (3) is calculated: EQU lnD(x, y)-lnB(x, y)=ln {T(x, y)+S(x, y){-lnk (3) The correction result is the right side of the equation (3). In this case, the scattered X-ray distribution S(x, y) is left uncorrected. As a method of correcting the scattered X-ray distribution, there is a method of substituting the scattered X-ray distribution actually measured into the equation (3). Alternatively, as will be later described, the scattered X-ray distribution may be corrected by correcting the density difference of two images as the joining point when a plurality of images are joined together. In order to generate the corresponding position relation table, the standard chart having elements at known positions is imaged. The image of the standard chart is analyzed so that the direction and magnitude of a distortion at each element position can be known. The corresponding position relation between a subject image and the standard chart image is stored in the corresponding position relation table, using the coordinate values of each element as parameters. This table is generated using as parameters the coordinate values of all pixels of the standard chart image in the following manner. Since pixels of a distortion-corrected image are not all in correspondence with the discrete positions of the measured standard chart image, the corresponding coordinates values are estimated from adjacent known discrete positions, by interpolation and extrapolation calculations. As the interpolation and extrapolation calculations, various interpolation methods may be used such as a nearest neighborhood method, a Lagrange interpolation method, a sampling function interpolation method, and a spline interpolation method. FIG. 4 shows an example of the standard chart. This standard chart has holes at a plurality of lattice points 140. In this case, the coordinate values of the lattice points are stored as parameters in the corresponding position table. Other coordinate values at positions other than the lattice points to be stored in the table are calculated by interpolation and extrapolation. In the case where the Lagrange interpolation method using four lattice points is used, as shown in FIG. 4, the coordinate values of pixels in an oblique line area 144 surrounded by four lattice points within a circle view field 42 are calculated by interpolation, and the coordinate values of pixels in another oblique line area 146 near the inside and outside of the view field 42 are calculated by extrapolation. At the post-process (310) shown in FIG. 3, the images (308-1, 308-2, . . . ) obtained through radiography of a subject (307-1, 307-2, . . . ) undergo the sensitivity non-uniform correction process (303-1, 303-2, . . . ). In this correction process, the non-uniformity correction factor for each pixel of a subject image is obtained from the non-uniformity correction factor table. The pixel value of the subject is divided by the correction factor. Next, the geometric distortion correction process (305-1, 305-2, . . . ) is executed. In this correction process, the coordinate value of each pixel of the distortion-corrected image is related to the corresponding coordinate value of a pixel of the measured image, by using the corresponding position relation table. The value of the pixel at the corresponding coordinate value is used as the image density. As described previously, since pixels of a distortion-corrected image are not all in correspondence with the discrete positions of the measured image, the corresponding coordinates values are estimated from adjacent known discrete positions, by interpolation and extrapolation calculations. As the interpolation and extrapolation calculations, various interpolation methods may be used such as a nearest neighborhood method, a Lagrange interpolation method, a sampling function interpolation method, and a spline interpolation method. After the image correction process (320), an image joining process (330) is executed. At the image joining process 330, a joining point for image joining is determined (331-1, 331-2, . . . ) and thereafter the coordinate systems are unified so as to make the coordinate values of joining points of images coincide with each other (332-1, 332-2, . . . ). The images are then joined together (335), and the image level (density) of the subject image near the joining point is corrected (336). With the procedure of the algorithm illustrated in FIG. 3, it is possible to obtain a joined image 337 of a plurality of subject images while performing various corrections. FIGS. 5A and 5B illustrate how a joining point is determined. In FIGS. 5A and 5B, there is shown the position relation between fields of view obtained by two radiographic exposures. Two circles 42 represent the fields of view of the X-ray detection unit at different imaging positions. FIG. 5A illustrates a method of using as a reference point the characteristic point (such as an edge of a particular bone) of a subject common to two X-ray images. In FIG. 5A, the first to fifth thoracic vertebrae 43-1 to 43-5 and ribs 44 are shown. One edge point on these bones, e.g., the vertebra, in this example, an edge point on the second vertebra indicated by a solid black circle, is used as a marker 40. FIG. 5B illustrates a method of using, as reference points, the points of markers (40-1, 40-2) common to two X-ray images. In this example, the markers made of circular metal pieces are attached to a subject when imaging it. FIG. 5C illustrates a method of imaging a subject 47 by placing a marker chart 45 having a known transmittance and known positions of lattice points usable as joining points, and using as reference points some lattice points, such as points (140-3, 140-4) indicated by black solid circles. With this method, it is possible both to determine reference points and to correct geometric distortions by referring to the positions of lattice points on the subject image. It is preferable that a reference point for determining the joining point is contained in a plurality of X-ray images. The joining method used in this case will be later described with reference to Fig. 6. Even if a reference point is contained in only one X-ray image, a plurality of X-ray images can be joined together by using known reference point coordinate values in the subject space. For example, in the case of the marker chart 45 shown in FIG. 5C, if the coordinate values of four lattice points (46-1 to 46-4) indicated by white empty circles are known, the images can be joined together. As a method of joining a plurality of digital images, a method disclosed, for example, in JP-A-2-264372 may be used. According to this method, two points commonly imaged on a plurality of digital images are sampled, the images are joined together by using a line passing through the two points as a border line (joining line) for two images, and the average density distribution of each image is calculated to correct the density of each pixel. FIG. 6 illustrates an example of an image joining method. In FIG. 6, two reference points A and B are commonly contained in two X-ray images R and L. If the joining points on the image L are points A1 and B1 and the joining points on the image R are points A2 and B2, points A1 and A2 and points B1 and B2 correspond to reference points A and B at the same positions of the subject. The coordinate systems of the two images R and L are unified by making points A1 and A2 coincide with each other at point A so that points A1 and A2 are shown at the same point A in Fig. 6 and points B1 and B2 coincide with each other at point B so that points B1 and B2 are shown at the same point B in Fig. 6. Affine transformation is used for the coordinate transformation, and the corresponding position relation table obtained through interpolation and extrapolation is used. After the coordinate system transformation, two images are superposed one upon the other so as to make corresponding joining points coincide with each other. A polygonal line CABD is used as a joining line, where A is the coincident point of the joining points A1 and A2, B is the coincident point of the joining points B1 and B2, C and D are cross points of the outer peripheries of the two fields of view. The left image L to the left of the joining line and the right image R to the right of the joining line are joined together. Generally, a polygonal line coupling coincident joining points and cross points of the outer peripheries of fields of view is used as the joining line to join images together. After the image joining, the density (image value) of the superposed area of the images L and R is corrected in order to correct the discontinuity of the density at the joining points. FIGS. 7A to 7E illustrate an example of the density correction. The pixel value of the image L is represented by 1, and that of the image R is represented by r. In this invention, the pixel values in the area only contained in one of two images are kept unchanged, and the pixel values in the area contained on both the two images are corrected by using a density correction factor calculated by the algorithm to be described later. In FIG. 7A, R and L represent two images, and a bold polygonal line is a joining line (CABD) (refer to FIG. 6). The density correction for a line G (coincident with X-axis) is illustrated in FIG. 7B. RD and LD are density curves of the images R and L on the line G. The pixel value on the joining line takes an arithmetic mean of pixel values of the images R and L on the line. The pixel values near the joining line are multiplied by correction factors, so as to make the change in the pixel values of the images R and L on the line continuous. As shown in FIG. 7C, correction factors (represented by a broken line RC1 and a solid line LC1 for the images R and L) are linearly changed so as to make the pixel values of the images R and L continuous on the joining line. The correction factors are calculated for each line (in the X-direction). An example of determining correction factors will be described, The pixel values on the joining line are set to an arithmetic means of the pixel value (1) of the image L and the pixel value (r) of the image R. The correction amount m (refer to FIG. 7B) on the joining line is calculated from the following equation (4). The position of the line G is changed in the Y-direction to obtain an arithmetic means on a new line from the equation (4). EQU m=(L-r)/2 . . . (4) The correction amount m changes discontinuously in the vertical direction (Y-direction). Namely, as shown in FIG. 7D, correction factors (represented by a solid line LC2 and broken line RC2) calculated from the correction amount m on the joining line at each point become discontinuous in the vertical direction. As a result, if the density correction is performed by using these correction factors, the density of the corrected image becomes discontinuous in the vertical direction although it is continuous in the horizontal direction (X-direction). In order to obtain correction factors continuous also in the vertical direction, the correction amount m' on the joining line given by the following equation (5) is used: ##EQU1## where i represent the pixel position in the X-direction, and j represents the pixel position in the Y-direction. The correction amount m' is an arithmetic mean of correction amounts m given by the equation (4) in a window defined by p pixels in the horizontal direction (X-direction) and q pixels in the vertical direction (Y-direction). As shown in FIG. 7E, correction factors (a solid line LC3 and broken line RC3) calculated so that the correction amounts is equal to m' on the joining line m' become continuous also in the vertical direction. If the density correction is performed using these correction factors the density of the corrected image becomes continuous both in the horizontal direction (X-direction) and in the vertical direction (Y-direction) at each line. In the above embodiment, the input field of view of the X-ray II covers one lung and mediastinum, and the whole chest can be covered by two radiographic exposures, minimizing the time required for imaging the subject and processing image data. The shape of the view field of a joined image obtained by a chest X-ray imaging system is preferably a square or a shape like a square. The imaging time is preferably 2 seconds or less in order to minimize the motion of the chest while the breathing is stopped. In this embodiment, the input field of view of the X-ray II is a circle having a diameter of about 30 to 45 cm. The relative motion distance of the X-ray II between two radiographic exposures is about 1/2 of the diameter of the input field of view. The shape of a joined image is a square having a side length of about .sqroot.3/2 times the diameter of the input field of view, which provides a sufficient field of view of a 35 cm square required for the chest diagnosis. The imaging time is about 0.5 sec or less, which is proper in practical use. In the above embodiment, the position change mechanism moves the X-ray detection unit horizontally between consecutive radiographic exposures. This is suitable for imaging the chest of a subject in a standing posture, and for moving the X-ray detection unit having a large weight by using a simple structure of the mechanism. As a trajectory of the X-ray detection unit between two radiographic exposures, a circular trajectory 80 indicated by a bold line in FIG. 8 as well as a straight trajectory may also be used. In FIG. 8, reference numerals 85 represent two imaging centers. The position change mechanism for a straight trajectory involves reversed motion directions. The position change mechanism for a circular trajectory does not require reversed motion directions so that the structure of the mechanism can be simplified. If a small X-ray II of about 12 inches is used, the number of chest radiographic exposures is preferably four. In this case, it is preferable to move the imaging center of the X-ray detection unit along the circular trajectory indicated by the bold line in FIG. 9 so that a field of view like a square can be obtained. FIG. 10 is a plan view showing the structure of an X-ray imaging system according to the second embodiment of the present invention. In the X-ray imaging system of this embodiment, the positions of an X-ray source 3 and a subject 17 are fixed. When imaging the subject 17, the X-ray source 3 positions on a center line coupling the X-ray source 3 and the center of the X-ray input plane of an X-ray detection unit 16, and the X-ray detection unit 16 moves while always maintaining its X-ray input plane perpendicular to the center line. As a result, the X-ray detection unit 16 moves along the arc line about the center of the X-ray source 3. Instead of moving the X-ray detection unit along the arc line in the above manner, another position change mechanism may also be used wherein a linear position change system for linearly moving the center of the X-ray input plane of the X-ray detection unit 16 to predetermined imaging positions, and a rotary system for rotating the X-ray detection unit 16 placed on a stage rotating about its rotation axis and facing the unit 16 towards the X-ray source 3. In this case, the predominant motion is a linear motion so that the X-ray detection unit 16 can be moved at high speed. In the above embodiment, the X-ray source 3 and the subject 17 are fixed. Instead, the X-ray source 3 is faced to the center of the X-ray detection unit 16 at each exposure time. In this case, the X-ray intensity distribution can be maintained constant, allowing to make the focal size small and the resolution high. In changing the relative position between an X-ray detection unit and a subject, the subject 17 may be moved instead of moving the X-ray detection unit 16. In this case, as shown in FIG. 11, the X-ray source 3 and X-ray detection unit 16 are fixed on a center line passing through the X-ray source and the center of the X-ray detection unit 16, while maintaining the X-ray input plane of the X-ray detection unit perpendicular to the center line. When imaging the X-ray source 3, the subject 17 is located at a predetermined position on a center line passing through the X-ray source and the center of the subject 17, while the subject 17 maintaining its posture perpendicular to the center line. The method of moving only the subject does not give any direct vibration to the X-ray detection unit, eliminating the adverse effect of vibration noises. In addition to the above methods of imaging a plurality of divided target imaging regions by changing the relative position between the X-ray detection unit and subject, both the X-ray detection unit and subject may be moved at the same time or time sequentially. The method of moving both the X-ray detection unit and subject provides an advantage of shortening the time required for moving them. If the relative position between the X-ray detection unit and subject is changed even during X-ray exposure, the position change mechanism can be simplified. In a ultra high precision camera using an image pickup tube, the resolution in the horizontal scanning line direction is generally lower than that in the vertical direction. Therefore, if an X-ray absorption plate constituting the X-ray grid is vertical to the horizontal scanning line direction of the image pickup tube, an image of the X-ray absorption plate, i.e., a grid pattern, becomes unsharp on the subject image, so that the visual diagnosis of the subject image will not be obstructed. In the above embodiments, an X-ray II-TV system using an X-ray II and a television camera of an image pickup tube has been used as the X-ray detection unit. This system has the most advanced technical development and is now widely used as the real time X-ray imaging system. A ultra high precision camera with 2100 or 4200 scanning lines may be used with this system, allowing a high resolution X-ray image to be imaged relatively easily. Also with this system, X-ray fluoroscopy is possible, realizing easy imaging position alignment. Since an image can be read in real time, the next imaging can be started immediately after the X-ray detection unit 16 is moved to the next imaging position, and also the continuous imaging can be performed. If the horizontal scanning line direction of an image pickup tube used as the image pickup element of a television camera is made parallel to the motion of the X-ray detection unit taking a plurality of subject images, then the position alignment of a plurality of subject images can be achieved only by parallel position changes between a plurality of subject images. Therefore, calculation for generating a joined image can be simplified, and the time required for obtaining the joined image from a plurality of subject images can be shortened. Instead of the image pickup tube, a CCD image pickup device maybe used for a television camera, in which the direction of arrays of CCD elements is made parallel to the direction of moving the X-ray detection unit taking a plurality of subject images. If a camera using an image pickup tube is mechanically vibrated during read-out scanning, noises may be generated by the characteristic oscillation of the electrodes of the image pickup tube, resulting in a possible error. After the mechanical vibration is removed, it takes a certain time thereafter to attenuate noises. In the case of a CCD element which is a solid element, the noises are rarely generated by mechanical vibration. It is therefore possible to perform reading an image and moving the X-ray detection unit at the same time after X-ray exposure, and to perform next imaging immediately after moving the X-ray detection unit. As a result, a plurality of X-ray images used for forming a joined image can be obtained at high speed, reducing the influence of a motion of a subject during the period while the X-ray detection unit is moved. Not only a combination of an X-ray II and a television camera but also a combination of a phosphor plate and a television camera may be used as the X-ray detection unit. In the latter case, the geometrical distortion and non-uniformity of sensitivity at low spatial frequencies of an image on the phosphor plate are smaller than an X-ray II, allowing images to be corrected and joined more easily. However, the sensitivity of a high resolution phosphor plate is lower than an X-ray II, so that used as a television camera is a high sensitivity imaging device such as a ultra high sensitivity image pickup tube using avalanche phenomenon or a ultra low noise imaging device such as a cooled CCD device. The application fields of the present invention are not limited to chest radiography, but are applicable to other large internal organs such as the colon of a subject with a large body structure. An angiogram of a whole body may be formed by smoothly coupling joining regions of a plurality of images, for the proper diagnosis of blood vessels near the joining regions. The digital X-ray imaging system of this invention can use various imaging methods available to conventional X-ray imaging systems, extending the application fields of the present invention. For example, an image of a wider field of view and more precise than a conventional system can be obtained by using a tomography imaging method, an enlarged imaging methods or a stereo imaging method.
	description	The nuclear fuel cycle is the series of industrial processes used to produce electricity from uranium in a nuclear reactor. The nuclear fuel cycle can be described as having three major parts: (1) the “front end” where uranium is mined and processed into fuel for use in a nuclear reactor, (2) the use of the fuel in a reactor, and (3) the “back end” where spent fuel is stored and eventually disposed or reprocessed (if the spent fuel is reprocessed, remaining wastes would be temporarily stored and eventually disposed). The nuclear fuel cycle begins with the extraction of uranium from ores or other natural sources. Uranium provides the basic fissile material or “fuel” for nearly all nuclear reactors. Extracted uranium consists almost entirely of two isotopes of uranium atoms, mostly uranium-238 (U-238) (99.3%) together with a much smaller fraction (0.7%) of the fissionable isotope uranium-235 or “U-235.” In its natural state, mined uranium is only weakly radioactive meaning that it can be handled without the need for radiation shielding. Before it can be used in a commercial reactor, natural uranium must be purified and enriched to boost the amount of fissionable U-235 present in the fuel. Most of the commercial nuclear power plants in operation today use fuel enriched to a U-235 concentration of anywhere from 3 to 5 percent—a typical figure for fuel used in commercial U.S. reactors is 4 percent. The enriched uranium is cast into hard pellets and stacked inside long metal tubes or “cladding” to form nuclear fuel rods. The uranium in the pellets is not pure elemental uranium but rather uranium oxide. The fuel rods are bundled into nuclear fuel rod assemblies that are typically about 12 to about 14 feet long. The core of a typical light-water commercial nuclear power reactor in the U.S. contains roughly about 200 to about 500 nuclear fuel rod assemblies, totaling approximately 100 metric tons of uranium oxide. Inside the reactor, the enriched uranium sustains a series of controlled nuclear reactions that collectively liberate substantial quantities of energy. The energy is converted to steam and used to drive turbines that generate electricity. Meanwhile, the fission process inside the reactor creates new elements or “fission products,” and gives rise to some heavier elements, collectively known as “transuranics,” which may take part in further reactions (among the most important is plutonium-239). The preponderant reactor type currently used in the majority of commercial nuclear power plants is the light water reactor (LWR). There are also several other reactor types in commercial use such as the heavy water reactor (HWR), gas cooled reactor (GCR), boiling water cooled graphite moderated pressure tube type of reactor (RBMK), etc. Nuclear fuel remains in a commercial power reactor for about four to six years, after which it can no longer efficiently produce energy and is considered used or spent. The spent fuel removed from a reactor is thermally hot and emits a great deal of radiation. Upon removal from the reactor, each spent nuclear fuel rod assembly emits enough radiation to deliver a fatal radiation dose in minutes to someone in the immediate vicinity who is not adequately shielded. The spent nuclear fuel rod assemblies are transferred to a deep, water-filled pool and stored in a rack. Wet storage keeps the spent fuel cool and protects the workers from the radiation. Ideally, spent fuel is kept in the pool for at least five years, although spent fuel at many U.S. reactor sites has been in pool storage for several decades. After the fuel has cooled sufficiently in wet storage, it can be transferred to dry storage. Dry storage systems generally consist of multiple nuclear fuel rod assemblies positioned in a fuel storage grid that is placed in a steel inner container and a concrete and steel outer container. The hazards posed by spent fuel makes it difficult to transport. For this reason, government regulators require spent fuel to be shipped in containers or casks that shield and contain the radioactivity and dissipate the heat. In the U.S., spent fuel has typically been transported via truck or rail although other nations also use ships for spent fuel transport. Spent fuel can be reprocessed to produce additional nuclear fuel. Even after commercial fuel is considered “spent,” it still contains unused uranium along with other re-usable elements such as plutonium which is generated within the fuel while it is in the reactor and fission products. Current reprocessing technologies separate the spent fuel into three components: uranium, plutonium (or a plutonium-uranium mix), and waste, which contains fission products and transuranic elements that are produced within the fuel. The plutonium is mixed with uranium and fabricated into new fuel while the fission products and other waste elements are packaged into a new form for disposal. Regardless of whether spent fuel is reprocessed or directly disposed of, every approach to the nuclear fuel cycle requires disposal of spent fuel that assures the very long-term isolation of radioactive wastes from the environment. Many nations, including those engaged in reprocessing, are working to develop disposal facilities for spent fuel and/or high level waste, but no such facility has yet been put into operation. Every nation that is developing disposal capacity plans to use a deep, mined geologic repository for this purpose. The lack of operational disposal facilities makes storage that much more important. Storage in some form, for some period of time, is an inevitable part of the nuclear fuel cycle. In the early days of the nuclear energy industry it was assumed that storage times for spent fuel would be relatively short—on the order of several years to a decade or two at most—before spent fuel would be sent either for reprocessing or final disposal. The current reality is much different. Storage is not only playing a more prominent and protracted role in the nuclear fuel cycle than once expected, it is the only element of the back end of the fuel cycle that is currently being deployed on an operational scale in the U.S. In fact, much larger quantities of spent fuel are being stored for much longer periods of time than policy-makers envisioned or utility companies planned for when most of the current fleet of reactors were built. The dominant form of storage for spent fuel at operating reactor sites is wet storage in pools. In some countries, pools are even used at consolidated storage facilities that are distant from the reactor sites. Pools are the de facto storage solution because they are essential to operating a nuclear power plant given the need to cool newly discharged spent fuel close to the reactor core. Once spent fuel is in the pool, it is easy and inexpensive to leave it there for long periods of time. Storing spent fuel in pools presents a number of problems. One problem is the limited capacity of the pools. Over the years, the pools fill up until there is no more additional capacity. The operator of the reactor must then transfer some of the spent fuel to dry storage, which is an expensive and difficult operation. Another problem is that spent fuel stored in pools is susceptible to natural disasters such as earthquakes. The earthquake may cause the pool to lose water and the spent nuclear fuel to meltdown. The Fukushima disaster in Japan is an example of a cooling pool losing water causing the spent fuel to overheat and meltdown. Another problem is that spent fuel can begin to lose its structural integrity when stored for long periods of time in a pool. Once this happens, the structural integrity of the spent fuel must be restored, a process that requires considerable time and resources. After an initial period of cooling in wet storage (generally at least five years), dry storage (in casks or vaults) is the preferred option for extended periods of storage (i.e., multiple decades up to 100 years or possibly more). Unlike wet storage systems, dry systems are cooled by the natural circulation of air and are less vulnerable to system failures and natural disasters. The most common type of dry storage system is shown in FIG. 1. The system includes a canister 12 that encloses multiple spent nuclear fuel rod assemblies 10. The canister 12 is positioned inside a concrete structure or cask 14. The canister 12 is formed of ½ inch to ⅝ inch thick stainless steel or concrete and serves as the primary boundary to confine radioactive material. The canister 12 can be oriented vertically or horizontally inside the cask 14. The cask 14 is a reinforced concrete structure that provides shielding from radiation and protects the canister 12. The cask 14 can be positioned in a vault 16 for long term storage as shown in FIG. 2. Casks can be designed and licensed as single-purpose casks (storage only), dual-purpose casks (storage and transport), and multi-purpose casks (storage, transport, and disposal). Typically, the more uses the casks are licensed for, the more they cost. Conventional cask systems present a number of problems. One is that many nuclear power plants require expensive and time-consuming upgrades to make it possible to handle and maneuver the casks while loading them with spent nuclear fuel assemblies. For example, many of these plants have to retrofit the pool area with a larger overhead crane to handle the tremendous load of the casks. These improvements can cost tens of millions of dollars, which tends to deter plant operators from moving spent fuel from wet storage to dry storage. Another problem with conventional cask systems is that unless the cask is a multi-purpose cask, there is a good chance that the bare fuel assemblies will need to be handled again in order to transport and/or eventually dispose of the spent fuel. Handling bare fuel greatly increases the difficulty and cost required to transport and/or dispose of the spent fuel. The current management strategy for spent nuclear fuel relies on dry storage to provide adequate capacity to allow continued operation of commercial nuclear plants. Utilities meet their interim storage needs on an individual basis with large-capacity, dry storage casks that are focused on meeting existing storage and transportation requirements because disposal requirements are not available. The problem with this is that disposal of the large canisters currently used by the commercial nuclear power industry represents many significant engineering and scientific challenges. Additionally, the expanded use of high-burnup (>45 GWd/MTU) fuel increases licensing uncertainty associated with transporting existing spent nuclear fuel. The problem is exacerbated by the uncertainty surrounding the requirements for the geological disposal repository. For example, several repositories under consideration are formed of materials (i.e., clay/shale, salt, and crystalline rock) that require limited canister/cask sizes due to thermal or physical constraints. This combined with the above discussion indicates that the canister/casks that end up satisfying the as yet unknown disposal requirements will likely be significantly different than what is being used for dry storage today. This difference means that the existing canisters/casks will likely need to be repackaged in canisters/casks that satisfy future transportation and disposal requirements. Repackaging the spent nuclear fuel for the purpose of transportation and/or disposal, particularly following an extended storage period, creates radiological, operational, and financial liabilities and uncertainties and should be avoided or minimized. Given the current status, the most imminent service needed worldwide for spent fuel management is the supply of sufficient and prolonged storage capacity that solves one or more of the problems identified above for the future spent fuel inventory arising from both the continued operation of nuclear power plants and from the removal of fuel in preparation for plant decommissioning. A number of representative embodiments are provided to illustrate the various features, characteristics, and advantages of the disclosed subject matter. It should be understood, however, that other embodiments can be created by combining individual features and/or components of the explicitly disclosed embodiments. For example, the features, characteristics, advantages, etc., of one embodiment can be used alone or in various combinations and sub-combinations with one another to form other embodiments. A system and method for managing spent nuclear fuel includes managing the spent fuel from the time it is discharged from the reactor to the time it is disposed of in a geological repository. The system and method is not limited to managing spent nuclear fuel that comes straight from the reactor. It can also accommodate spent nuclear fuel regardless of what stage it is at in the back end of the nuclear fuel cycle. For example, spent fuel stored in pools or in dry storage can be incorporated into the system. The system includes a small capacity canister that encloses or encapsulates up to six spent nuclear fuel rod assemblies. In the preferred embodiment, the canister is sized and configured to enclose a single spent nuclear fuel rod assembly. Individually enclosing the spent nuclear fuel rod assemblies maximizes the advantages of the system. It should be appreciated, however, that canisters that enclose more than one spent nuclear fuel rod assemblies can still realize many of the benefits of the system. The canister is engineered to satisfy safety related criteria while minimizing reliance on other systems and components that are difficult to monitor or examine such as the cladding integrity of the fuel rods. The canister is also engineered to be versatile. It can be used in connection with multiple disposal paths. The canister also provides flexibility for meeting existing and future licensing objectives and requirements. The canister is configured to receive and enclose a spent nuclear fuel rod assembly in an air tight fashion. In effect, the spent nuclear fuel rod assembly is sealed inside the canister. Multiple canisters are loaded into a cask for interim storage and/or transport. The canister can eventually be disposed of in the geological repository. A method for enclosing a spent nuclear fuel rod assembly in the canister includes positioning a single spent nuclear fuel rod assembly in the canister and closing or sealing the canister to make it air tight. Alternatively, two to six spent nuclear fuel rod assemblies can be sealed in a single canister. In one embodiment, the canister is lowered over a spent nuclear fuel rod assembly positioned in a staging rack in a pool such as the spent nuclear fuel pool (cooling pool) that is part of a commercial nuclear power station. In other embodiments, the spent nuclear fuel rod assembly may be lowered into the top of a stationary canister. Loading the canister preferably takes place in a pool, but can also take place outside of a pool, such as, for example, at an interim dry storage location. The staging rack can include multiple holding areas each of which is configured to receive and support a spent nuclear fuel rod assembly. A retaining member is positioned at the bottom of each holding area underneath the spent nuclear fuel rod assembly. The canister moves over the spent nuclear fuel rod assembly until it reaches the bottom and engages the retaining member. The retaining member is coupled to the bottom of the canister in such a way that the spent nuclear fuel rod assembly is held in the can as the canister is lifted out of the pool. The canister is lifted out of the pool and water is allowed to drain out the bottom through openings in and around the retaining member. The interior of the canister is actively or passively dried and a cover is secured to the bottom of the canister to make it air tight. The canister is filled with an inert gas up to a pressure of about 1 psi to 3 psi. The canister includes a coupler or fitting through which the inert gas can pass into the interior of the canister. Once the coupler is no longer needed, a cap is sealed over it to make the top of the canister air tight. With the cap in place, there is no way for gas to escape from the canister. In one embodiment, the cap, bottom cover and other components of the canister are welded together. The welds are inspected radiographically to make sure that it is completely sealed and meets all applicable standards. The use of radiographic testing is advantageous because it can eliminate the need to provide double containment such as, for example, a secondary enclosure. Alternatively, the canister can be enclosed in a second canister that is slightly larger to provide double containment. The sealed canister is put back in the staging rack in the pool. The canister can remain in the pool indefinitely or can be transferred to a cask for dry storage. Alternatively, the canister can be transferred directly to a cask without being put back in the pool. It should be noted that storing sealed canisters in the pool is preferable to storing bare spent nuclear fuel rod assemblies. For example, if the water level unexpectedly drops, the spent nuclear fuel is much less likely to produce a radioactive event because it is enclosed in the canister. The Fukushima disaster in 2011 is a good example. If the spent nuclear fuel rod assemblies had been enclosed in canisters, then they would have been much less likely to have released harmful radiation to the environment. It may be desirable to periodically transfer groups of canisters from the pool to storage casks. In one embodiment, a transfer platform is placed on the staging rack directly above a group of canisters. A cask is positioned above the transfer platform and the group of canisters are lifted into the cask. The canisters are maintained in a fixed, spaced apart relationship to each other in the cask to facilitate criticality safety. The cask can be stored in a vault at the interim storage area at the reactor site or in a vault at a consolidated storage area that is not part of the reactor site. In one embodiment, the cask is configured to be put directly in the vault without removing the canisters from the cask. In another embodiment, the cask is a transfer cask and the canisters are transferred from the transfer cask to a storage cask, which can be placed in the vault. The vault can have a modular construction so that the capacity of the vault can be expanded on an as-needed basis instead of as a large, one-time capital expenditure. The vault can include a plurality of panels, preferably made of concrete and/or steel, that can be coupled together to form one or more chambers each of which is configured to receive and hold a cask. The system includes the following main components: a staging rack, a canister, a canning module, a transfer rack, a cask, and a vault. The canister includes an elongated tubular member having a top and a bottom, a first end cover coupled to the top of the tubular member and a second end cover coupled to the bottom of the tubular member. The staging rack and the transfer rack are positioned in the pool and used to facilitate enclosing the spent nuclear fuel rod assemblies in the canister and moving them out of the pool. The canning module is positioned out of and adjacent to the pool and is used to enclose or seal the canister with the spent nuclear fuel rod assembly inside. The cask can be used to transport and/or store multiple canisters on-site or off-site (e.g., an intermediate waste transfer station). It should be appreciated that one cask can be used to remove the canisters from the pool and another cask can be used to store the spent nuclear fuel rod assemblies in a vault. The vault holds the casks and provides shielding and passive cooling. The casks can be licensed for on-site and/or off-site usage. For example, one cask can be designed and licensed for on-site transport and/or storage. Another cask can be designed and licensed for off-site transport and/or storage (dual-use cask). Yet another cask can be designed and licensed for off-site transport and/or storage as well as final disposal (multi-purpose cask). The Summary is provided to introduce a selection of concepts in a simplified form that are further described below in the Detailed Description. The Summary and the Background are not intended to identify key concepts or essential aspects of the disclosed subject matter, nor should they be used to constrict or limit the scope of the claims. For example, the scope of the claims should not be limited based on whether the recited subject matter includes any or all aspects noted in the Summary and/or addresses any of the issues noted in the Background. A system is disclosed for flexibly and safely managing the entire back end of the nuclear fuel cycle. The spent nuclear fuel is managed from the time it is discharged from the reactor to the time it is disposed of in a geological repository. The system is also capable of managing legacy spent nuclear fuel that is stored in dry storage. The system includes a small capacity canister 20 that is preferably configured to enclose or encapsulate up to six spent nuclear fuel rod assemblies 22. Preferably, the canister 20 is sized and configured to enclose a single spent nuclear fuel rod assembly 22. However, in other embodiments, the canister 20 is sized to enclose two, three, four, five, or six spent nuclear fuel rod assemblies 22. The canister 20 is engineered to satisfy various safety related criteria associated with storing and transporting spent nuclear fuel. The canister 20 is configured to provide a sealed containment enclosure for the nuclear fuel rod assembly 22. If the cladding on the spent fuel rods deteriorates, it will still be contained inside the canister 20. The canister 20 is also versatile. For example, the canister 20 can be used in connection with multiple storage and disposal paths. The canister 20 can be loaded with a spent fuel assembly 22 and then stored in a pool or in a dry storage vault. Once the disposal criteria has been established, the canister 20 can be transferred to an appropriate disposal cask or directly disposed without the need to handle and expose bare fuel, especially bare fuel that has been in storage for decades. Conventional systems enclose large numbers of spent fuel assemblies 22 in large canisters and casks. Enclosing individual or small groups of the spent fuel assemblies 22 in a single canister 20 provides a number of advantages over conventional systems. One advantage is that expensive upgrades to the reactor site are not required. Conventional canisters and casks are so large that most reactor sites must be retrofitted with expensive upgrades just to lift and move the canisters and casks. The canister 20 and associated components are small enough that they can be handled using the existing reactor site infrastructure. For example, the overhead crane present at most spent fuel pools can be used to handle the canister 20 and associated components although it is too small to handle the enormous size of conventional canisters and casks. The use of the canister 20 provides the ability to enclose the spent fuel assemblies 22 immediately or shortly after exiting the reactor core, which significantly increases the safety of the system. If the pool loses water like it did in Fukushima Japan, the spent fuel assemblies 22 will still be contained in the canisters 20. This will prevent a large scale release of radioactive particles into the environment. Individually enclosing the spent fuel assemblies 22 in the canisters 20 makes them much easier to handle and transport, both now and in the future, because they are always contained. Once the fuel assemblies 22 are sealed in the canisters 20, there is no need to handle bare spent fuel again. If the canisters 20 need to be transferred to a different cask or system for interim storage of final disposal, then they can without exposing the bare fuel. One of the reasons the Yucca Mountain disposal site is so complex and expensive is because it is designed to handle bare fuel assemblies 22. If this was no longer required, then it would significantly reduce the complexity and cost of the geologic disposal site regardless whether it is at Yucca Mountain or somewhere else. The same considerations apply to regional interim storage sites. The canister 20 provides structural support and integrity to the spent fuel assemblies 22. One of the problems with storing spent fuel assemblies for long periods of time is that they lose their structural integrity, e.g., the cladding on the spent fuel rods can crack or break. Once this happens, it becomes much more difficult and expensive to handle the spent fuel assemblies 22. Enclosing the spent fuel assemblies 22 in the canister 20 prevents this from happening Enclosing the spent fuel assemblies 22 in the canisters 20 allows for passive cooling of the spent fuel assemblies 22 during dry storage. Air can enter the bottom of the vault or cask, travel upward past the canisters 20, and exit through openings in the top. Turning to the Figs., they show the canister 20 sized and configured to enclose a single spent fuel assembly 22. It should be appreciated, however, that the canister 20 can be designed to hold up to six spent fuel assemblies 22 as mentioned above. The canister 20 can include a framework that holds the spent fuel assemblies 22 in a fixed, spaced apart relationship to each other. The framework can be configured to hold the spent fuel assemblies 22 in the most compact way possible. For example, if the canister includes four spent fuel assemblies 22, then they may be arranged in a 2×2 matrix. Also, if the canister includes six spent fuel assemblies 22, then they can be arranged in a 2×3 matrix. Numerous other configurations are possible. Turning to FIG. 3, a staging rack 24 is shown positioned at the bottom of a spent nuclear fuel pool (also referred to as a cooling pool). The staging rack 24 includes a plurality of holding areas 26 (also referred to as holding bays) each of which is sized to receive and securely hold the canisters 20 and the spent fuel assemblies 22 in an upright position. The staging rack 24 includes the bare spent fuel assemblies 22 in the left rear area and the loaded canisters 20 in the right rear area. The canisters 20 have lifting members 28 (also referred to as handles) on the top and the bare spent fuel assemblies 22 do not. The canisters 20 and bare spent fuel assemblies 22 positioned along the front of the staging rack 24 illustrate the process of enclosing the spent fuel assemblies 22, which is discussed in greater detail later. The term “spent nuclear fuel rod assembly” and corresponding terms such as spent fuel assembly” shall mean the bundle or cluster of nuclear fuel rods held together in a fixed relationship to each other by a framework. This is a discrete assembly of nuclear fuel rods that is positioned inside a nuclear reactor. The spent fuel assembly 22 can have a variety of sizes and configurations. For example, the spent fuel assembly 22 can have any suitable length and cross-sectional shape. The spent fuel assembly 22 can be 1 m to 15 m long and have a rectangular, circular, hexagonal, or other cross-sectional shape. The configuration of the spent fuel assembly 22 largely depends on the type of reactor and characteristics of the fuel rods. The preponderant fuel type currently used for the majority of commercial nuclear power today is that required for the LWR. However, there are other fuel types in commercial use such those used in HWR reactors, GCR reactors, RBMK reactors, etc. Table 1 below shows some of the main characteristics of these fuel types and their respective associated fuel cycle post-operation disposition. TABLE 1Fuel types in commercial use in the worldReactorTypeDesignPhysical Specs.NotesLWRPWRSquare/hexagonal cross-Usually stored intactsectionFuel rods areBWR4 m to 5 m longconsolidated inWWER200 kg to 500 kg perfuel assembliesassemblyPHWRCANDUØ 10 cm × 50 cmHandled in tray/20 kg per bundlebasketGCRMagoxØ 3 cm × 1.1 m long slug;Usually reprocessedAGR24 cm diameter, 1 m longDry storage possibleassemblyOthersRBMKØ 8 cm × 10 m long assemblySized to half length(2 sections)for storagePBMRØ 6 cm spherical fuel elementCanned for storage The nuclear fuel rods in the spent fuel assemblies 22 can have any suitable configuration. In one embodiment, the nuclear fuel rods include a plurality of nuclear fuel pellets clad in a sleeve or rod of zirconium oxide. The pellets are stacked up, enclosed, and sealed in a zirconium alloy tube to form a single nuclear fuel rod. Before describing the process of loading the canisters 20 with the spent fuel assemblies 22, the construction of the canisters 20 is described with reference to FIGS. 4-14. FIG. 4 shows a perspective view of the canister 20. FIG. 5 shows a cross-sectional perspective view along a cross-sectional plane that extends longitudinally the length of the canister 20. The canister 20 includes an elongated tubular member 30 (also referred to as a tubular body or main body), a first end cover 36 coupled to the top end 32 (also referred to as a first end) of the tubular member 30 and a second end cover 38 coupled to the bottom end 34 (also referred to as a second end) of the tubular member 30. The covers 36, 38 close the ends 32, 34 of the tubular member 30. The second end cover 38 seals the bottom end 34 of the tubular member 30 so that it is air tight—i.e., so that gases cannot enter or escape. It should be noted that for purposes of this disclosure, the term “coupled” means the joining of two members directly or indirectly to one another. Such joining may be stationary in nature or movable in nature. Such joining may be achieved with the two members or the two members and any additional intermediate members being integrally formed as a single unitary body with one another or with the two members or the two members and any additional intermediate member being attached to one another. Such joining may be permanent in nature or alternatively may be removable or releasable in nature. The covers 36, 38 can be coupled to the tubular member 30 in any way so long as it produces an air tight seal. In one embodiment, the covers 36, 38 are welded to the tubular member 30. The welds can be inspected using radiographic testing to ensure that there are no flaws that could allow gas to escape through the welds. Radiographic testing can be used to ensure compliance with ASME standards so that it is not necessary to use a double containment system, e.g., two canisters 20 enclosing a single spent fuel assembly 22. It should be appreciated that the above techniques can be used to couple together any of the components described in this document. Other fasteners and fastening techniques can also be used depending on the situation. For example, bolts, screws, adhesives, and the so forth, can be used to couple the various components together. The top end 32 of the canister 20 is shown in greater detail in FIG. 6. The first end cover 36 is welded to the tubular member 30 as explained above. The lifting member 28 is coupled to the top of the first end cover 36 using fasteners 40 that engage corresponding holes 42 in the first end cover 36. FIG. 7 shows that the holes 42 do not extend all the way through the first end cover 36. The lifting member 28 provides a convenient way for a crane or other lifting device to engage and lift the canister 20. The lifting member 28 in FIG. 6 is a removable bail that includes a loop that extends from one corner of the first end cover 36 upward and then back down to the opposite corner of the first end cover 36. It should be appreciated that the lifting member 28 can have any suitable configuration so long as it is capable of being used to lift the canister 20. The lifting member 28 can also be coupled to other components of the canister 20 such as the tubular member 30. In an alternative embodiment, a threaded lifting member is provided to enable lifting with a suitable remotely operated lifting device. An example of a threaded lifting member and corresponding remotely operated lifting device is a Zip Lift available from FastTorq, New Caney, Tex. Referring to FIG. 7, the first end cover 36 includes a hole 44 through which fluids can pass into the interior of the canister 20. A coupler 46 is coupled to the first end cover 36 over the hole 44. The coupler 46 is shown in FIG. 8. The coupler 46 defines a passageway through the front end cover 36 and into the interior of the canister 20. The coupler 46 includes a sleeve 48 and a quick release fitting 50. The sleeve 48 is coupled to the top of the first end cover 36. In the embodiment shown in FIG. 6, the sleeve 48 is welded to the first end cover 36. The sleeve 48 provides a secure base to which the quick release fitting 50 can be coupled. Returning to FIG. 8, the quick release fitting 50 is coupled to the sleeve 48 using a threaded engagement. The quick release fitting 50 is a female type fitting and is configured to receive a corresponding male quick release fitting 50. The quick release fitting 50 includes a valve assembly that is closed when the corresponding male quick release fitting 50 is not present and is open when it is present and securely coupled to the quick release fitting 50. The coupler 46 can be attached to a vacuum pump to remove residual moisture from the canister 20. The coupler 46 can also be used to supply gases such as air, inert gases (noble gases), heated gases, and so forth to the interior of the canister 20. For example, the coupler 46 can be used to supply heated air to dry the interior of the canister 20 including the spent fuel assembly 22. The coupler 46 can also be used to charge the loaded canister 20 with inert gases for long term storage and/or disposal of the spent fuel assembly 22. It should be appreciated that the configuration of the coupler 46 shown in the Figs. is but one example of numerous other configurations it can have. For example, the coupler 46 can be positioned at other locations on the canister 20 such as the tubular member 30 or the second end cover 38. Also, the coupler 46 can be provided with or without a valve that closes the passageway into the canister 20. Referring to FIG. 6, the canister 20 includes a cap member 52 positioned over the coupler 46 and coupled to the first end cover 36. The cap member 52 encloses the coupler 46 and seals the passageway so that it is air tight. The cap member 52 is welded to the top of the first end cover 36 in the embodiment shown in FIG. 6. The weld 54 is shown separately to depict that it can be a v-groove fillet type weld. The cap member 52 includes a tubular body 56 capped with a circular end plate 58. The tubular body 56 is sized to fit over the coupler 46. The circular end plate 58 is welded to the tubular body 56 to seal the two components together in an air tight manner. FIGS. 9-10 show another configuration for the top end 32 of the canister 20. In this embodiment, the cap member 60 includes the lifting member 62. The cap member 60 includes a tubular body 66 and circular end plate 68 welded to the top of the tubular body 66 with weld 64. The lifting member 62 is a threaded rod that is coupled to and extends upward from the circular end plate 58. In this embodiment, the lifting member 62 is both welded (weld 65) and threaded to the circular end plate 58 but it should be appreciated that these components could be coupled together in other ways. The cap member 60 includes threads that engage corresponding threads on the sleeve 48 of the coupler 46. This allows the cap member 60 to be screwed on to the coupler 46 and then welded in place for a strong and secure connection. In another embodiment, the cap member 56 can be used to lift the canister 20 without a separate lifting member. For example, the circular end plate 58 forms a lip that could be engaged by lifting equipment such as cranes. In this situation, the cap member 56 doubles as a lifting pintle. The construction of the bottom end 34 of the canister 20 is shown in FIGS. 11-14. FIG. 11 shows an exploded view of the bottom end 34 that includes the bottom end of the tubular member 30, a retaining member 70, and the second end cover 38. The retaining member 70 fits inside the tubular member 30 and the second end cover 38 is coupled to the bottom end of the tubular member 30 to seal the bottom end 34 of the canister 20 closed. The second end cover 38 is coupled to the tubular member 30 in any suitable manner. In one embodiment, the second end cover 38 is welded to the tubular member 30 in a similar manner as the first end cover 36. It should be appreciated, however, that any of the other fastening techniques described in this document could be used as well. The retaining member 70 includes a support plate 72 (also referred to as a support member) and support posts 74 positioned underneath the support plate 72. The support plate 72 includes a plurality of holes 76 that extend through the support plate 72 and are arranged in a regular pattern. The holes 76 are provided to allow water to drain out the bottom of the canister 20 through the retaining member 70. It should be appreciated that the configuration of the retaining member 70 shown in the Figs. is but one example of a suitable configuration. The retaining member 70 can have a variety of additional configurations. For example, the retaining member 70 can be configured to allow water through gaps between the edges of the support plate 72 and the walls of the tubular member 30 instead of or in addition to the holes 76. Also, the support plate 72 can have a concave or convex shape instead of a flat plate shape. Numerous other configurations are possible. The support plate 72 includes recesses 78 that are configured to engage retaining latches 80 (also referred to as tabs) coupled to the interior walls of the tubular member 30. The retaining latches 80 are biased outward from the interior walls of the tubular member 30. As the support plate 72 enters the bottom of the tubular member 30, the recesses 78 contact the latches 80 and bias them toward the interior walls of the tubular member 30 until the recesses 78 reach a corresponding recess 82 in the latches 80. At this point, the latches 80 bias outward from the interior walls of the tubular member 30 and the recesses 78, 82 engaged each other holding the retaining member 70 in place. This arrangement allows the tubular member 30 to be coupled to the retaining member 70 by lowering the tubular member 30 on to the retaining member 70. As the tubular member 30 is lowered, the latches 80 contact the support plate 72 and hold the retaining member 70 in the position shown in FIG. 14. The retaining member 70 is configured to support the weight of the spent fuel assembly 22. Before the second end cover 38 is put in place, the weight is supported entirely by the latches 80. Once the second end cover 38 is put in place, the support posts 74 rest on the inside surface of the second end cover 38 and transfer the weight load from the support plate 72 to the second end cover 38. The second end cover 38 includes recesses 84 that correspond to the support posts 74 to keep the support posts 74 in an upright position over the long term and through numerous moves. The canister 20 and any of its components can be made of any suitable material. In one embodiment, the canister 20, including the tubular member 30 and the covers 36, 38 are made of stainless steel that is at least 3 mm thick (e.g., 3 mm to 7 mm). It should be appreciated that other materials can be used as well such as composites, carbon steel, various alloys, and the like. The exterior of the canister 20 can have a smooth finish (e.g., 2B finish for stainless steel) to facilitate decontamination. Criticality control can be provided using a variety of different techniques. In one embodiment, the canister 20 does not include a borated neutron absorber. Criticality control is provided by soluble boron credit, geometric spacing and moderator exclusion. In another embodiment, the canister 20 includes a borated neutron absorber surrounds the spent fuel assembly 22. The canister 20 can also be any suitable size. In one embodiment, the canister 20 is sized to at least roughly correspond to the size of an individual spent fuel assembly 22. For example, if the spent fuel assembly 22 is square like those in the Figs., then the canister 20 is square and slightly larger to enable it to receive the spent fuel assembly 22. If the spent fuel assembly 22 is hexagonal, then the canister 20 would also be hexagonal and so forth. In one embodiment, the canister 20 has cross-sectional dimensions of approximately 24 cm×24 cm. The canister 20 can be any suitable height such as 1 m to 35 m, 2 m to 30 m, and so forth. Referring back to FIG. 3, one embodiment of a process for loading the canisters 20 with spent fuel assemblies 22 is shown. The process is represented by the canisters 20/spent fuel assemblies 22 shown in the first row of the staging rack 24. The process proceeds from right to left. The first step in the process is to position a retaining member 72 at the bottom of each holding area 26. A bare spent fuel assembly 22 is positioned in the holding area 26 on top of the retaining member 72 as shown by the bare spent fuel assembly 22 positioned on the right side of the front row of the staging rack 24. The spent fuel assembly 22 is shown as it is being lowered down on to the retaining member 72. The canister 20 is then lowered over the spent fuel assembly 22 as depicted in the middle right position of the front row of the staging rack 24. The canister 20 has been lowered most of the way down but has not yet reached the retaining member 72. Note that the coupler 46 on the canister 20 has not been enclosed by the cap member 56. The canister 20 is lowered until it reaches and is coupled to the retaining member 72 in the manner described above. The retaining member 72 is coupled to the bottom end 34 of the canister 20 and is configured to support the weight of the spent fuel assembly 22. The canister 20 is lifted out of the pool and water drains out the bottom end 34 through the holes 76 in the retaining member 72. The canister 20 being lifted out of the pool is depicted in the middle left position of the front row of the staging rack 24. While out of the pool, the interior of the canister 20 is dried, charged with an inert gas, and then the canister 20 is sealed air tight. The details of this process are described in greater detail as follows. The second end cover 38 and the cap member 52 are coupled to the canister 20 to seal it closed and make it air tight. The canister 20 is returned to the staging rack 24 in the pool as shown by the left position of the front row of the staging rack 24. Alternatively, the canister 20 could be placed directly in a transfer cask or storage cask for dry storage instead of being returned to the pool. It should be noted that the canister 20 on the far left includes both the second end cover 38 and the cap member 52. FIGS. 15-18 show one embodiment of a process for sealing the canister 20 using a canning module 90 that is positioned out of the pool. The process of sealing the canister 20 is designed to be controlled remotely so that personnel are not exposed to harmful radiation (e.g., ionizing particle and electromagnetic radiation). For example, the process can be controlled in a control room 86 such as that shown in FIG. 18. The canning module 90 includes a lifting mechanism 92 that lifts the canister 20 out of the pool. In the embodiment shown in FIG. 15, the lifting mechanism 92 includes a winch 93, cable 95, and a hook 97 on the end of the cable 95 (FIG. 16). The hook 97 engages the lifting member 28 at the top of the canister 20. It should be appreciated that the lifting mechanism 92 can include any suitable mechanism in any configuration as long as it is capable of lifting the canister 20 into the canning module 90. The canning module 90 includes lifting members 97 on the top that are configured to be coupled to a lifting mechanism such as a crane. The lifting members 97 allow the canning module 90 to be suspended above the pool while loading and unloading the canisters 20. The canning module 90 includes an elongated, shielded chamber 94 that is sized to receive the canister 20. The canister 20 is lifted into the chamber 94 through an access door 96 at the bottom of the canning module 90. The chamber 94 is open at the top and the bottom to allow remote operations to be performed on the canister 20 such as drying the interior and sealing it air tight. The top and bottom of the chamber 94 are referred to as top chamber 100 and bottom chamber 102 even though they are part of chamber 94. Alternatively, the chambers 100, 102 can be separate from the elongated chamber 94. The canning module 90 includes multiple layers of shielding to protect against harmful radiation. The shielding can be provided by a variety of materials such as layers of concrete, lead, and so forth. The shielding is provided to prevent or reduce exposure to harmful electromagnetic radiation. The access door 96 on the bottom of the canning module 90 can be closed by a door mechanism 98 (FIG. 17) to prevent exposure to harmful particle radiation. The door mechanism 98 includes one or more electric, hydraulic, or pneumatic actuators that close the access door 96 by, for example, sliding it closed. The top chamber 100 includes components that allow the interior of the canister 20 to be remotely dried and facilitate putting the cap member 52 in place. For example, the top chamber 100 includes a robotic arm 104, video camera 106, and drying and inerting apparatus 108. The video camera 106 can be used to remotely monitor the process from the control panel 86. The canister 20 undergoes the following operations in the canning module 90. The interior of the canister 20 is dried using the apparatus 108. In one embodiment, the apparatus 108 is configured to vacuum dry the interior of the canister 20. In another embodiment, the apparatus 108 is configured to blow air through the canister 20 to dry it. It should be appreciated that the interior of the canister 20 can be dried before or after the second end cover 38 is attached. The drying and inerting apparatus 108 is configured to engage the coupler 46 on the top of the canister 20. The robotic arm 104 can be used to engage and/or disengage the apparatus 108 and the coupler 46. Once the interior of the canister 20 is dry, it is charged with an inert gas. In one embodiment, the inert gas is a noble gas such as helium. The inert atmosphere prevents the spent fuel assembly 22 from oxidizing and/or otherwise decomposing during long periods of storage and/or after disposal. Alternatively, the interior of the canister 20 can be placed under a vacuum. It should be appreciated that the second end cover 38 should be put in place before the canister 20 is charged with inert gas. The apparatus 108 can have any of a variety of configurations. In one embodiment, the apparatus 108 is replaced by two separate apparatuses. One apparatus is configured to dry the canister 20 and the other apparatus is configured to charge it with an inert gas. The disadvantage of this configuration is that it can require connecting and disconnecting the apparatuses from the coupler 46 multiple times. Once the canister 20 is charged with inert gas, the cap member 52 is positioned over the coupler 46 and coupled to the canister 20 in the manner described above. A robotic welder can be used to weld the cap member 52 to the canister 20. In one embodiment, the robotic welder is mounted on a turntable to allow it to rotate all the way around the cap member 52. In another embodiment, the robotic arm 104 includes the robotic welder. The bottom chamber 102 includes components used to couple the second end cover 38 to the tubular member 30. For example, the bottom chamber 102 can include a video camera 110, robotic welder 112, and a radiographic testing device 114. The video camera 112 can be used to remotely monitor the process from the control room 86. The second end cover 38 is positioned on a staging platform 116 that can move vertically and horizontally. Once the canister 20 is in position, the staging platform 116 moves horizontally underneath the bottom end 34 of the canister 20. The staging platform then moves vertically until the second end cover 38 is positioned adjacent to or in contact with the bottom of the tubular member 30. The second end cover 38 is now in position to be welded to the tubular member 30. The robotic welder 112 welds the second end cover 38 to the tubular member 30. In one embodiment, the robotic welder 112 is coupled to a turntable 118 that rotates around the exterior of the canister 20. The video camera 110 and the radiographic testing device 114 can also be coupled to the turntable 118. This allows a full 360 degree view of the welding operation. The radiographic testing device 114 is used to inspect the welds to ensure that they meet applicable standards and do not contain any defects. If the welds are defective, then the robotic welder can be used to weld the area again and fix the defects. It should be appreciated that the canister 20 can be sealed shut using any of a number of other methods and devices. For example, the process can be modified to seal the canister 20 in the pool while still drying and charging it with inert gas (e.g., an air lock can be used to remove the water from the canister 20). Numerous other modifications are also possible. FIGS. 19-21 show one embodiment of a process for moving the loaded canisters 20 from the pool to dry storage. The first step in the process is to place a transfer platform 120 on top of the staging rack 24 as shown in FIG. 19. The transfer platform 120 is configured to support a transfer cask 122 placed on top of the transfer platform 120. It should be appreciated that the transfer platform 120 is in the pool and all or a portion of the transfer cask 122 is also in the pool. The transfer platform 120 is divided into nine sections 124, each of which corresponds to a group of canisters 20 in the staging rack 24 that will be loaded into the transfer cask 122. In the embodiment shown in FIG. 19, each section 124 corresponds to a 3×3 group of nine canisters 20. This is the number of canisters 20 that are loaded into the transfer cask 122. In another embodiment, a 4×4 group of sixteen canisters 20 are loaded into the transfer cask 122. It should be appreciated that the transfer platform 120 and the transfer cask 122 can be configured to handle any number and/or size of canisters 20. For example, the transfer platform 120 and the transfer cask 122 can be configured to handle BWR or other types of spent fuel that have different shapes and cross-sectional sizes. The transfer cask 122 can be formed of any material that is capable of providing the desired amount of structural strength and radiation shielding. In one embodiment, the transfer cask 122 is made of concrete, metal (e.g., stainless steel), or a combination of both. The transfer cask 122 includes trunnions 126 that are used to lift and handle the transfer cask 122. The trunnions are capable of supporting the weight of the loaded cask 122. The canisters 20 are loaded into the transfer cask 122 as a group with a lifting assembly 128. The lifting assembly includes a lifting cable 130 and hook 132 for each of the canisters 20. The hooks 132 are configured to engage the lifting members 28 at the top of each canister 20. Once engaged, the lifting cables 130 lift the canisters 20 into the transfer cask 122. Alternatively, each canister 20 can be lifted separately into the transfer cask 122. FIG. 22 shows one embodiment of the lifting assembly 128 that includes a support member 134 (also referred to as a support plate), nine lifting cables 130 coupled to and extending downward from the support member 134, nine hooks 132 coupled to the end of the lifting cables 130 and an alignment member 136 (also referred to as an alignment plate) positioned just above the hooks 132. The alignment member 136 holds the lifting cables 130 and hooks 132 in a fixed spatial relationship to each other to make it easier for the hooks 132 to engage the lifting members 28 on the canisters 20. The alignment member 136 includes slots 138 that engage a corresponding section on the hooks 132 to prevent the hooks 132 from rotating. The hooks 132 are configured to all face the same direction to make it easier to engage the lifting members 28. When the hooks 132 reach the lifting members 28, the lifting members 28 hit the underside of the hooks 132 and deflect the hooks 132 to one side until the lifting members 28 have cleared the opening of the hooks 132. At this point, the lifting members 28 move back the opposite direction until the open part of each hook 132 is directly below the corresponding lifting members 28. The hooks 132 are raised and engage the lifting members 28 and lift the canisters 20. It should be noted that the alignment member 136 causes the hooks 132 move as a single body and makes it impossible for them to twist or change the direction they face. The lifting assembly 128 includes a plurality of cables 140 that extend from the top of the support member 134 upwards to a lifting ring 142. The support member 134 is configured to be positioned outside the transfer cask 122 while the alignment member 136 is positioned inside with the lifting cables 130 extending through openings in the top. A crane or other lifting device can be coupled to the lifting ring 142 to lift the canisters 20 into the transfer cask 122. The opening on the underside of the transfer cask 122 through which the canisters 20 passed is closed before the transfer cask 122 is moved beyond the pool area. The exterior components of the lifting assembly 128 are kept inside the transfer cask 122 until it reaches its destination and the canisters 20 are placed in a storage cask and/or storage vault. FIG. 23 shows another embodiment of the lifting assembly 128. This embodiment is similar to the one shown in FIG. 22 except that the alignment member 136 has been replaced by a plurality of hook actuator assemblies 144. Each hook actuator assembly 144 includes a housing 146, a hook actuator 148, and a hook 132. The housing 146 is sized to receive the lifting members 28 inside the housing 146 and to maintain the desired spacing between adjacent hook actuator assemblies 144. The size and configuration of the housing 146 can help maintain the hook actuator assemblies 144 in the proper orientation to allow them to drop down over the corresponding lifting members 28. The operation of the hook actuator assemblies 144 is shown in FIGS. 24-26. The hook actuator assembly 144 is lowered until it reaches the lifting member 28 as shown in FIG. 24. The hook actuator 148 moves the hook 132 to a retracted position where the lifting member 28 can move past the hook 132 as shown in FIG. 25. The hook actuator assembly 144 is then lowered until the lifting member 28 is above the opening in the hook 132. The hook actuator 148 moves the hook 132 forward until the hook 132 securely engages the lifting member 28 as shown in FIG. 26. The canisters 20 are ready to be lifted into the transfer cask 122. The hook actuator assemblies 144 can also be used to release the canisters 20 when they are lowered out of the transfer cask 122 and placed in a storage cask or the like. It should be appreciated that the hook actuators 148 can include any suitable hydraulic, electric, or pneumatic actuator. In one embodiment, the hook actuators 148 are operated electrically. FIGS. 27-31 show one embodiment of a method to space the canisters 20 apart in the transfer cask 122. In this embodiment, the transfer cask 122 includes a plurality of spacers 150 that are actuated using a corresponding plurality of drive mechanisms 152. Each drive mechanism 152 includes a motor 154 drivingly connected to a screw 156. The spacers 150 can be used to stabilize and hold the canisters 20 while the transfer cask 122 is in motion. The spacers 150 can also be used to provide criticality control by keeping the canisters 20 spaced apart from each other a safe distance. The drive mechanisms 152 can also be configured to decontaminate and/or clean the exterior surface of the canisters 20. The contaminants accumulate on the exterior of the canisters 20 during storage in the pool. In one embodiment, the spacers 150 include cleaning equipment such as spray headers 158 and/or cleaning pads (e.g., Scotch-Brite cleaning pads). As the spacers 150 move up and down, the spray headers and cleaning pads move along the exterior of the canisters 20 to remove contaminants. As shown in FIG. 27, the spacers 150 are raised while the transfer cask 122 is being loaded with canisters 20. This keeps the spacers 150 out of the way while the canisters 20 are raised from the staging rack 24. Once the canisters 20 are in the transfer cask 122, the spacers 150 are lowered to different heights using the drive mechanisms 152. The process of lowering the spacers 150 and the final heights of the spacers 150 are shown in FIGS. 27-29. FIG. 30 shows the spacers 150 and drive mechanisms 152 in greater detail. It should be appreciated that each screw 156 is configured to move a single spacer 150 even though the screw 156 is configured to extend through all four spacers 150. This is accomplished by configuring the screw 156 and spacers 150 so that the screw 156 only engages a single spacer 150 having a corresponding set of threads and passes freely through the other three spacers 150. FIG. 31 shows that the bottom spacer 150 includes sprayers 158 and cleaning pads 159 that surround all of the canisters 20. As the bottom spacer 150 moves downward, the sprayers 158 and cleaning pads 159 remove contaminants on all sides of the canisters 20. Turning to FIGS. 32-33, the transfer cask 122 can be moved from the pool to a dry storage area using a truck 160, cask transporter 162, or any other suitable mode of transportation. In one embodiment, the transfer cask 122 is moved to an independent spent fuel storage installation located on or near the reactor site and the canisters 20 are moved to dry storage. FIGS. 34-38 show one embodiment of a dry storage system that includes a modular vault 200 that encloses one or more storage casks 202. The storage casks 202 are configured to receive the canisters 20 from the transfer cask 122. FIGS. 34-35 show a perspective view and an exploded view, respectively, of the modular vault 200. The vault 200 includes shielded openings 204 that allow passive ventilation by natural convection to dissipate the decay heat of the spent fuel. The air circulates from near ground level up through the interior of the vault 200 and escapes out the top. The circulating air passively cools the canisters 20 inside the vault 200. The vault 200 is modular in that it includes functionally separate expandable units configured to hold additional canisters 20. The vault 200 can be expanded on an as-needed basis so that capital improvement costs are spread out evenly over a longer time period. The savings reaped from minimizing idle vault capacity can be substantial depending on the facility and time span of the implementation. FIG. 38 shows one embodiment of the vault 200 after it has been expanded multiple times. The vault 200 can be made of any suitable material that is capable of shielding the surrounding area from harmful radiation and providing passive cooling to the canisters 20. In one embodiment, the vault 200 is made of reinforced concrete. Such concrete components are sized to facilitate manufacture offsite and transport to the site for assembly. The concrete can be preformed panels that are coupled together on-site. The concrete can also be poured on-site. Preferably, preformed concrete is used so it can be easily disassembled to expand the vault 200. Referring to FIG. 35, the vault 200 includes four storage casks 202. The storage casks 202 include a thick outer shell 206, a metal liner 208, an interior framework 210, and a lid or top 212. The shell 206 is made of a thick, solid material such as reinforced concrete that is capable of shielding harmful radiation. The metal liner 208 provides a thermal radiation shield to reduce concrete temperatures and a loose contamination barrier. The metal liner 208 is made of a corrosion resistant material such as stainless steel or galvanized steel. The interior framework 210 is likewise made of metal (e.g., stainless steel) and is configured to hold the canisters 20 in a spaced apart relationship that provides criticality control. Heat resistant ceramic plates 214 can be positioned at the bottom of each holding area in the framework 210 to minimize heat damage to the underlying material and to mitigate galvanic corrosion (FIG. 36). The lid 212 allows access to the top of the storage cask 202 for loading and unloading operations. FIG. 37 shows the canisters 20 being moved from the transfer cask 122 and loaded into the storage cask 202. The first step is to remove the top panel of the modular vault 200 that covers the storage cask 202. The lid 212 of the storage cask 202 then slides outward to expose the interior framework 210 while maintaining radiation shielding. The transfer cask 122 is lifted over the storage cask 202 and the canisters 20 are aligned with the holding areas of the framework 210. The canisters 20 are lowered into the storage cask 202 using the lifting assembly 128. The lifting assembly 128 disengages the canisters 20 in the manner described above and the lid 212 is moved back into place before the transfer cask 122 is moved away from the vault 200 to shield radiation. The lid 212 of the storage cask is put back into position and the top panel of the vault 200 is put in place. It should be noted that the aspect ratio and dimensions of the vault 200 are configured to provide a stable structure to resist earthquake loads without being anchored to the basemat. It should be appreciated that one advantage of this system is the reduction of the need to handle bare spent fuel assemblies 22 for transfer operations between the different steps of spent fuel management. This reduces the potential for radiation exposure and human error. It also reduces the need for specialized transfer facilities and equipment and the concomitant safety risks and costs. It also eliminates the need to open, repackage, and rehandle the fuel as is currently the case with large conventional canisters. It also facilitates operations involved in the interface operations between different steps of the spent fuel management down to disposal, including safeguards inspections. It should also be appreciated that the casks 122, 202 can be single-purpose, dual-purpose, or multi-purpose casks. For example, the cask 122 can be licensed as a multi-purpose cask so that the canisters 20 can be loaded into it once, stored, and disposed of without further handling. Illustrative Embodiments Reference is made in the following to a number of illustrative embodiments of the disclosed subject matter. The following embodiments illustrate only a few selected embodiments that may include one or more of the various features, characteristics, and advantages of the disclosed subject matter. Accordingly, the following embodiments should not be considered as being comprehensive of all of the possible embodiments. In one embodiment, a method for enclosing a spent nuclear fuel rod assembly in an air tight canister comprises positioning a single spent nuclear fuel rod assembly in the canister and closing the canister to make it air tight. The canister is configured to only enclose the single spent nuclear fuel rod assembly. Positioning the spent nuclear fuel rod assembly in the canister can include lowering the canister over the spent nuclear fuel rod assembly. Positioning the spent nuclear fuel rod assembly in the canister can take place in a pool. The method can comprise positioning the spent nuclear fuel rod assembly in a staging rack before positioning the spent nuclear fuel rod assembly in the canister. The staging rack can include a plurality of holding areas each of which is configured to receive a spent nuclear fuel rod assembly. The staging rack can include a retaining member positioned at the bottom of each of the plurality of holding areas where the retaining members are configured to couple to the canister. Multiple storage racks can be used to store caniserized fuel in the pool or until removal to dry storage or transport. The method can comprise coupling the canister to a retaining member positioned below the spent nuclear fuel rod assembly. The method can comprise lifting the canister with the spent nuclear fuel rod assembly in it out of a pool before closing the canister to make it air tight. The method can comprise drying the interior of the canister and the spent nuclear fuel rod assembly. The method can comprise filling the canister with inert gas before closing the canister to make it air tight. Closing the canister can include welding a cover over any opening that provides access to the spent nuclear fuel rod assembly in the interior of the canister. The method can comprise positioning the spent nuclear fuel rod assembly in a staging rack after closing the canister to make it air tight. The staging rack can be in a pool. The method can comprise positioning a plurality of the canisters in a cask. The method can comprise positioning a plurality of the casks in a storage vault. In another embodiment, a canister for enclosing a spent nuclear fuel rod assembly comprises a single spent nuclear fuel rod assembly positioned in the canister. The canister can enclose the spent nuclear fuel rod assembly. The canister can be air tight. The spent nuclear fuel rod assembly can be enclosed in a gaseous atmosphere. The spent nuclear fuel rod assembly can be enclosed in an inert atmosphere. The spent nuclear fuel rod assembly can include a framework and a plurality of spent nuclear fuel rods held together in a fixed spatial relationship to each other by the framework. The interior of the canister can have the same general shape as the exterior of the spent nuclear fuel rod assembly. The canister can comprise a lifting member at the top of the canister. The canister can comprise a coupler that provides a passageway into the canister to the spent nuclear fuel rod assembly and a cap member that covers the coupler and prevents gas from escaping from the canister. The coupler can be configured to connect to a source of compressed gas. The canister can comprise a tubular member having an elongated shape and a top and a bottom, a first end cover coupled to the top of the tubular member, and a second end cover coupled to the bottom of the tubular member. The interior of the tubular member can have the same general shape as the exterior of the spent nuclear fuel rod assembly. The canister can comprise a top, a bottom, and a retaining member. The retaining member can be located at the bottom of the canister and can support the spent nuclear fuel rod assembly. The retaining member can include openings through which water can flow. In another embodiment, a system comprises a staging rack and the canister positioned in the staging rack. The staging rack can be positioned in a pool. A system can comprise a cask and a plurality of the canisters recited in claim INDEP positioned in the cask. The cask can include at least three of the canisters. A system can comprise a storage vault and a plurality of the casks positioned in the storage vault. The storage vault can be modular. The concepts and aspects of one embodiment may apply equally to one or more other embodiments or may be used in combination with any of the concepts and aspects from the other embodiments. Any combination of any of the disclosed subject matter is contemplated. The terms recited in the claims should be given their ordinary and customary meaning as determined by reference to relevant entries in widely used general dictionaries and/or relevant technical dictionaries, commonly understood meanings by those in the art, etc., with the understanding that the broadest meaning imparted by any one or combination of these sources should be given to the claim terms (e.g., two or more relevant dictionary entries should be combined to provide the broadest meaning of the combination of entries, etc.) subject only to the following exceptions: (a) if a term is used in a manner that is more expansive than its ordinary and customary meaning, the term should be given its ordinary and customary meaning plus the additional expansive meaning, or (b) if a term has been explicitly defined to have a different meaning by reciting the term followed by the phrase “as used herein shall mean” or similar language (e.g., “herein this term means,” “as defined herein,” “for the purposes of this disclosure the term shall mean,” etc.). References to specific examples, use of “i.e.,” use of the word “invention,” etc., are not meant to invoke exception (b) or otherwise restrict the scope of the recited claim terms. Other than situations where exception (b) applies, nothing contained herein should be considered a disclaimer or disavowal of claim scope. The subject matter recited in the claims is not coextensive with and should not be interpreted to be coextensive with any particular embodiment, feature, or combination of features shown herein. This is true even if only a single embodiment of the particular feature or combination of features is illustrated and described herein. Thus, the appended claims should be given their broadest interpretation in view of the prior art and the meaning of the claim terms. As used herein, spatial or directional terms, such as “left,” “right,” “front,” “back,” and the like, relate to the subject matter as it is shown in the drawings. However, it is to be understood that the described subject matter may assume various alternative orientations and, accordingly, such terms are not to be considered as limiting. Articles such as “the,” “a,” and “an” can connote the singular or plural. Also, the word “or” when used without a preceding “either” (or other similar language indicating that “or” is unequivocally meant to be exclusive—e.g., only one of x or y, etc.) shall be interpreted to be inclusive (e.g., “x or y” means one or both x or y). The term “and/or” shall also be interpreted to be inclusive (e.g., “x and/or y” means one or both x or y). In situations where “and/or” or “or” are used as a conjunction for a group of three or more items, the group should be interpreted to include one item alone, all of the items together, or any combination or number of the items. Moreover, terms used in the specification and claims such as have, having, include, and including should be construed to be synonymous with the terms comprise and comprising. Unless otherwise indicated, all numbers or expressions, such as those expressing dimensions, physical characteristics, etc. used in the specification (other than the claims) are understood as modified in all instances by the term “approximately.” At the very least, and not as an attempt to limit the application of the doctrine of equivalents to the claims, each numerical parameter recited in the specification or claims which is modified by the term “approximately” should at least be construed in light of the number of recited significant digits and by applying ordinary rounding techniques. All ranges disclosed herein are to be understood to encompass and provide support for claims that recite any and all subranges or any and all individual values subsumed therein. For example, a stated range of 1 to 10 should be considered to include and provide support for claims that recite any and all subranges or individual values that are between and/or inclusive of the minimum value of 1 and the maximum value of 10; that is, all subranges beginning with a minimum value of 1 or more and ending with a maximum value of 10 or less (e.g., 5.5 to 10, 2.34 to 3.56, and so forth) or any values from 1 to 10 (e.g., 3, 5.8, 9.9994, and so forth).
	abstract	Described herein are embodiments of a method to control energy dose output from a laser-produced plasma extreme ultraviolet light system by adjusting timing of fired laser beam pulses. During stroboscopic firing, pulses are timed to lase droplets until a dose target of EUV has been achieved. Once accumulated EUV reaches the dose target, pulses are timed so as to not lase droplets during the remainder of the packet, and thereby prevent additional EUV light generation during those portions of the packet. In a continuous burst mode, pulses are timed to irradiate droplets until accumulated burst error meets or exceeds a threshold burst error. If accumulated burst error meets or exceeds the threshold burst error, a next pulse is timed to not irradiate a next droplet. Thus, the embodiments described herein manipulate pulse timing to obtain a constant desired dose target that can more precisely match downstream dosing requirements.
	summary
	summary
	description	1. Field of the Invention The present invention relates to a method for safely and securely storing radiocontaminated wastes, such as soil, sludge generated from treatment of waste water and sewage, boiler ashes, rubbles from manmade and/or natural disasters, farmed and/or forest mushrooms, and leaves and also relates to the container for the above method. 2. Description of the Related Art According to the U.S. classification system, nuclear waste is classified into high-level waste (HLW), transuranic (TRU) waste, uranium mill tailings, and low-level waste (LLW). Generally, not less than 99% of the total radioactivity in nuclear waste is contained in HLW, while LLW takes up the biggest share, about 85% or more of the entire weight of nuclear waste generated. Of the above, soil, sludge, boiler ashes, rubbles, forest mushrooms, fallen leaves, and the like (hereinafter referred to as “radiocontaminated waste matter”) are belonged to LLW. The general practice for decontaminating radiocontaminated agricultural lands, roads, school lands, and the like resulted from a nuclear accident is to remove the contaminated surface soil and for private houses and public buildings is to wash their roofs using jet water. The said removed radiocontaminated waste matter is then shielded by storing in, for example, flexible container bags, sandbags or the like and then isolated. The Japanese Environment Ministry has proposed temporary storage sites for smoothly storing and managing a large amount of radiocontaminated waste matter (Patent document 1). In those temporary storage sites, flexible container bags, sandbags, and the like, each filled with radiocontaminated waste matter, are stacked on a water-barrier sheet laid on the ground. These flexible container bags, sandbags are, in turn, shielded by placing filling soil and sandbags upon them. However, since the said shielding filling soil and sandbags are susceptible to damage from rain, snow, earthquakes, and other causes, and because the radiocontaminated wastes-containing flexible container bags and sandbags have low densities and thickness, moisture from the surrounding ground may permeate the water-barrier sheet, and hence, radiocontaminated waste matter may, in turn, contaminate the surrounding environment (air, groundwater, and the like) of the temporary storage site. In order to improve the temporary storage site of the type described above, storage facilities for radiocontaminated materials disclosed in the Patent document 1 have been proposed as alternative facilities. In the proposed storage facilities, radiocontaminated materials are shielded by a dome-shaped structure made of corrugated steel sheets and buried in healthy soil. For these storage facilities, the thickness of the corrugated steel sheet is 0.6 cm and the shallowest depth of the healthy soil 30 cm. It is calculated that these values allow the radiation dosage to be reduced by 90%. In this calculation, the half value layers (HVL) of iron and soil are assumed to be 1.5 cm and 5 cm, respectively, and these HVL's are applied only to 137Cs. Since the radiation dosage from radiocontaminated materials is due to 134Cs, 137Cs, 60Co, and the like in the gamma ray, the accuracy of the said calculation remains uncertain. In a storage structure for shielding radioactive substances-containing materials disclosed in the Patent document 2, bags filled with contaminated soil are placed on a bottom water barrier layer (sheet), and a radiation shielding wall structure is constructed composed of stacked sandbags or retaining walls around its periphery. The topmost surfaces of the bags are covered with a covering body made of a radioactive cesium absorbing powder containing one or more of zeolite, lead, tungsten, barium sulfate, and the like, and a resin or a rubber blended therewith. In order to keep out water, the storage structure is covered with a tent roof. However, the shielding effect of the covering body of this storage structure against gamma rays and the specifications thereof are not disclosed. In both the storage facilities for radiocontaminated materials (Patent document 1) and the storage structure for radioactive substances-containing materials (Patent document 2), radiocontaminated soil is filled into a flexible container bag and/or a sandbag, neither of which can be shielded against gamma rays. However, in a radiation shielding building disclosed in the following Patent document 3, a mixture of highly functional ceramic concrete and construction materials (such as wood, iron, and concrete) is used without incorporating ordinary lead, lead alloy, antimony-containing material, or the like. In one embodiment, when radiocontaminated soil was filled into a first box made of a highly functional ceramic concrete having a thickness of 5 cm, the radiation dose decreased from 147 μSv/h to 7.5 μSv/h, and a shielding rate of 94.9% was obtained, corresponding to a half value layer of 1.165 cm of the highly functional ceramic concrete. When a second box made of a highly functional ceramic concrete having a thickness of 5 cm was provided around the first box, the radiation dose further decreased from 7.5 μSv/h to 2.0 μSv/h, and a shielding rate of 73.3% was obtained, corresponding to a half value layer of 2.622 cm. Furthermore, when a third box made of a highly functional ceramic concrete having a thickness of 10 cm was provided around the second box, the radiation dose decreased from 2 μSv/h to 0.9 μSv/h, and a shielding rate of 55% was obtained, corresponding to a half value layer of 8.681 cm. The final radiation dose of 0.9 μSv/h is still higher than the 0.065 to 0.072 μSv/h radiation dosage in Fujinomiya city, Shizuoka prefecture, at a straight-line distance of approximately 330 km from the earthquake- and tsunami-damaged nuclear power plant in Fukushima Prefecture. Since the half value layer of the highly functional ceramic concrete changed with varied radiation dosages, it seems likely that the density was uneven and/or the shielding efficiency decreased at low radiation dosages. The following equation 1 was used to calculate the half value layer (Patent document 1).1−Shielding rate=1/[e(thickness of shielding structure+half value layer of shielding structure×ln2)] [Equation 1] Additionally, a radiation shielding material disclosed in the Patent document 4 is made by granulating or molding a water slurry containing magnesium oxide and debris of a lead-containing glass, such as discarded cathode-ray tube glass, followed by drying. Although lead-containing glass can be formed as a plate having a thickness of 1 to 10 cm, the performance (half value layer), density, Mohs hardness, and the like thereof are not disclosed. This applicant carried out a decontamination field test for a radiocontaminated rice paddy using a paper sludge-derived sintered carbonized porous grains and obtained the results indicating that radioactive substances could be removed from radiocontaminated agricultural soil by this method. Furthermore, it was found that the polished rice harvested from the improved soil contained a total of 30 Bq/kg of 134Csc and 137Cs, which is lower than the new Japanese standard limits of 100 Bq/kg for radiocesium in foods. Details are disclosed in the Patent document 5 below. The said paper sludge-derived sintered carbonized porous grains are formed by sintering and carbonization of paper sludge discharged from paper manufacturing mills which use either waste paper or wood chip or both waste paper and wood chip and the composition thereof is as described below. (1) Paper sludge discharged from paper manufacturing mills which use either waste paper or wood chip or both waste paper and wood chip is processed by sintering/carbonization to form a paper sludge-derived sintered carbonized porous grains which have a pH of 8 or more and preferably 10 or more; an alkalinity equivalent value of 1.0 to 4.0 meq/g (as NaOH) and preferably 1.5 to 2.5 meq/g (as NaOH); a cation exchange capacity of 1.0 to 4.0 meq/100 g (as NH4+) and preferably 1.5 to 3.0 meq/100 g (as NH4+); an electric conductivity of 70 to 150 μS/cm; a sodium content of 0.0003% or more; and a potassium content of 0.0003% or more, and the paper sludge-derived sintered carbonized porous grains thus obtained is dispersed on or mixed with radiocontaminated soil to remove radioactive substances therefrom. (2) In the manufacturing process of the said paper sludge-derived sintered carbonized porous grains, the impregnation of the paper sludge with either potassium iodide (KI) alone or ethylenediaminetetraacetic acid (EDTA) alone or a combination of KI and EDTA was not incorporated. (3) The radiocontaminated soil contains radioactive 134Cs and 137Cs at a total dosage of 800 Bq/kg or above. (4) The dosage of the said paper sludge-derived sintered carbonized porous grains spread on or mixed with the radiocontaminated soil is 0.1 to 6 kg/m2 (0.5 to 50 kg/m3) (0.1 to 6 percent by weight of dry soil) and preferably 1.0 to 3.5 kg/m2 (8 to 30 kg/m3) (0.9 to 3.3 percent by weight of dry soil). (5) The paper sludge has a moisture content of 50% to 85%, and after being pelletized and dried, this paper sludge is pyrolyzed in a reducing carbonization sintering furnace at a temperature of 500° C. to 1,300° C., preferably 700° C. to 1,200° C. Furthermore, carbonization is preferably carried out at 800° C. to 1,100° C. (6) The said paper sludge-derived sintered carbonized porous grains contain, on oven-dry weight basis, 15% to 25% of combustibles (including carbon), 0.5% to 3.0% of TiO2, 0.0001% to 0.0005% of Na2O, 0.0001% to 0.0005% of K2O, 15% to 35% of SiO2, 8% to 20% of Al2O3, 5% to 15% of Fe2O3, 15% to 30% of CaO, 1% to 8% of MgO, and a balance of 0.5% to 3.0% (including impurities), the total of these being 100%; and has a water absorption rate of 100% to 160% in accordance with JIS C2141, a specific surface area of 80 to 150 m2/g in accordance with the BET adsorption method, and an interconnected cell structure. (7) The said paper sludge-derived sintered carbonized porous grains are to have a porosity volume of not less than 70%, a porosity volume of not less than 1,000 mm3/g, an average pore radius of 20 to 60 μm, and pores with radius of not less than 1 μm constitute not less than 70% of the total porosity volume, and are a mixture of various forms such as spherical, oval, or cylindrical or the like forms with each having an axis length of 1 to 10 mm, and a black color. Patent document 1: Japanese Unexamined Patent Application Publication No. 2013-134226 Patent document 2: Japanese Unexamined Patent Application Publication No. 2013-130403 Patent document 3: Japanese Unexamined Patent Application Publication No. 2013-195416 Patent document 4: Japanese Unexamined Patent Application Publication No. 2013-210342 Patent document 5: Japanese Unexamined Patent Application Publication No. 2013-068459 As described above, because treatment facilities for radiocontaminated waste matter are not established yet, the most suitable decontamination methods for radioactive substances are currently not available. In Fukushima prefecture, where radioactive substances from the Nuclear Power Plant disaster on Mar. 11, 2011, are detected in some soil areas, the amount of radiocontaminated waste matter is at least 250,000 ton, which, in turn, is bagged into one tone-sized blue vinyl bags. These bags are piled on each other and stored atop manmade plateaus built on nearby mountains and around people's homes and rice fields. There are currently 30 of such locations around the prefecture. The blue bags are temporary and designed to withstand the environment for 5 years. (www.foreignpolicy.com/articles/2014/02/20/250000_tons_of_radioactive_soil_in_fukushima_japan). Additionally, the blue bags can only partially shield the gamma radiation from the inside radiocontaminated waste matter and their usable life is limited. The said temporary manmade plateaus are therefore liable to be contaminated by the radiocontaminated waste matter. The objective of the present invention is to provide a method for shielding radiocontaminated waste matter and a container therefor. Specifically, the invented method would comprehensively satisfy the requirements of cost, practicality, safety, and security, and that it would shield gamma radiation emitted from radiocontaminated waste matter, as well as it can make the spatial radiation dosage at the storage site similar to that of a place at a straight-line distance of 330 km from the Fukushima Nuclear Power Plant where the nuclear disaster took place. In the present invention, it is assumed that the location of the present applicant at a straight-line distance 330 km from the abovementioned Nuclear Power Station is a place (hereinafter referred to as a “blank spatial radiation dosage”) not influenced by the gamma radiation emitted by the Fukushima Daiichi Nuclear Power Plant disaster. The method and the container for shielding radiocontaminated waste matter according to the present invention can comprehensively satisfy requirements for cost, practicality, safety, and security, can shield the gamma radiation from radiocontaminated waste matter, and can also make the spatial radiation dosage at the storage site similar to that of a place at a straight-line distance of 330 km from the Fukushima Daiichi Nuclear Power Plant where the nuclear disaster occurred. In order to temporarily or permanently store radiocontaminated waste matter, building a tank is advantageous in terms of cost, location, and practicality. According to the method and the container for shielding radiocontaminated waste matter of the present invention, when radiocontaminated waste matter is partially replaced with the said paper sludge-derived sintered carbonized porous grains or potassium chloride-impregnated paper sludge-derived sintered carbonized porous grains, the spatial gamma radiation dosage around a receiving or storage tank would be similar to the blank spatial gamma radiation dosage, and thus the advantage of the present invention is that the safety and security for the environment and health can be maintained. When radiocontaminated waste matter alone or a mixture of it with the said paper sludge-derived sintered carbonized porous grains are ashed, and the ashes thus obtained are mixed again with the said paper sludge-derived sintered carbonized porous grains, the weight, volume, and radioactive 134Cs and 137Cs are all decreased, and the gamma radiation level around the storage site is equal to the blank spatial gamma radiation dosage; therefore, a large amount of radiocontaminated waste matter and paper sludge-derived sintered carbonized porous grains can be charged into a container/tank made of, for example, steel sheet, concrete, or concrete containing paper sludge-derived sintered carbonized porous grains, without causing any problem regarding spatial gamma radiation dosage in the environment around the container/tank and as such its long-term retention and storage can be done safely and securely. Hereinafter, embodiments of the present invention will be described. However, the present invention is not limited to the following embodiments. In the method of shielding radiocontaminated waste matter according to an embodiment of the present invention, paper sludge-derived sintered carbonized porous grains (hereinafter referred to as PSC)) are mixed with radiocontaminated waste matter, and the mixture thus obtained is charged into and stored in a tank 1 functioning as a container. The tank 1 has a lid 2 and a bottom base plate 4, and the materials of the tank 1, the lid 2 and the bottom base plate 4 are the same. According to the Occupational Safety and Health Administration of the U.S. Department of Labor, radioactive isotopes 37Cs and 60Co have adverse effect on human health. The present invention examined the efficacy of iron (density: 7.86 g/cm3) and concrete (density: 2.35 g/cm3) in shielding 60Co and 137Cs. The half value layers of both iron and concrete in shielding 60Co are 20 mm and 61 mm, respectively (International Commission on Radiological Protection, ICRP Pub. 21). These values are higher than those of iron (1.5 cm) and concrete (4.9 cm) in shielding 137Cs. Compared to 60Co, 137Cs has a longer half-life (30.17 years vs. 5.27 years) and a lower energy (0.6616 MeV vs. 1.3325 MeV). Therefore, the half value layer of a material for gamma radiation shielding depends on the density of the material and the energy rather than on the half-life of the radioisotopes to be shielded. The tank 1 does not require a retaining wall structure. The tank 1 in which radiocontaminated waste matter is to be filled is made of steel sheet, concrete or other material that shields gamma radiation emitted from radiocontaminated waste matter, and is set up with the bottom base plate 4 resting on the ground. The tank 1 is, for example, made in the shape of a regular polygonal cylinder having at least four corners or a circular cylinder in accordance with the shape of the packaged radiocontaminated waste matter. In order to cope with cases in which a large amount of radiocontaminated waste matter is stored, the volume of the tank 1 is required to be at least 1,000 m3. As such a tank farm would be formed. When the shape of the tank 1 is circular cylindrical, its diameter must be equal to its height so that its volume would be maximum. In this case, the correlation between the tank volume (y) and the tank diameter (x) is expressed by y=0.7854x3 (R2=1) (Equation 2), and when the tank volume is 1,000 m3, both the tank diameter and height would be 10.838513 m. When circular cylindrical vinyl bags (diameter=height=1.0838513 m) each with 1 m3 volume to hold one ton of radiocontaminated waste matter are loaded in a circular cylindrical storage tank with a volume of 1,000 m3, the filling rate is only 78.5% because even though the cylindrical vinyl bag has a circular shape it would occupy a square area. The tank 1 therefore preferably has a cubic shape (10 m×10 m×10 m). Accordingly, the diameter of the cubic tank 1 is smaller than that of the tank 1 of circular cylindrical shape, and when a square vinyl bag having a volume of 1 m3 is used, the filling rate of the tank 1 would be approximately 100%. In order to improve the filling rate when a circular cylindrical storage tank with a volume of 1,000 m3 is used, radiocontaminated waste matter is loaded directly into the storage tank in bulk without using a circular cylindrical or square vinyl bag of 1 m3 volume. When steel sheet is used as the material of the tank 1, a thickness of approximately 10 to 15 mm and a density of approximately 7.0 to 7.8 g/cm3 are preferable; when concrete without PSC is used, a thickness of approximately 60 to 65 mm, and a density of approximately 2.0 to 2.4 g/cm3 are preferable; and when PSC-containing concrete PSC is employed, a thickness of approximately 60 to 65 mm and a density of approximately 1.8 to 2.4 g/cm3 are preferable. When the material of the tank 1 is steel sheet, the lid 2 and the bottom base plate 4 of the tank 1 are preferably made of material similar to that of the tank 1 and of about the same thickness. When the material of the tank 1 is concrete, and the lid 2 and the bottom base plate 4 are also made from concrete, their thicknesses are made substantially equal to that of the tank 1. On the other hand, when the material of the tank 1 is concrete, and the lid 2 and the bottom base plate 4 are made from steel sheet, their thicknesses are made about equal, in the range of approximately 10 to 15 mm. The concrete is composed of cement, sand and gravel, and 15% to 35% of the gravel content can be replaced with PSC. The spatial gamma radiation dosage around the tank 1 made of concrete containing PSC becomes similar to the blank spatial gamma radiation dosage. When 20% (by weight) of radiocontaminated waste matter is replaced with PSC, followed by mixing, and the resulting mixture is loaded into the tank 1, the spatial gamma radiation dosage around the tank 1 becomes similar to the blank spatial gamma radiation dosage. As the decontamination of radioactive wastes contaminated by 134Cs and 137Cs is concerned, the decontamination degree of the PSC with impregnated potassium chloride is approximately two times higher than that of the PSC without impregnated potassium chloride. Thus, the mixing ratio of PSC with radiocontaminated waste matter can be decreased from 20% to 10%. Accordingly, the loading rate of radiocontaminated waste matter into the tank 1 will be increased. When radiocontaminated waste matter and/or a mixture of radiocontaminated waste matter and PSC which replaces 20% thereof is ashed, and PSC in an amount corresponding to 20% of this ash is added thereto and then mixed therewith to form an ash mixture, the gamma radiation dosage of this ash mixture is similar to that of the blank spatial gamma radiation, and radioactive 134Cs and 137Cs are also decreased. As a result, this ash mixture can be safely and securely loaded and stored in the tank 1. More specifically, when radiocontaminated waste matter and a mixture of radiocontaminated waste matter and PSC which replaces 20% thereof are each ashed at a temperature of 850° C. for 90 minutes, the weight of the radiocontaminated waste matter and that of the mixture of the radiocontaminated waste matter and PSC which replaces 20% thereof are each decreased by 10% to 15%. Thus, it is expected that the volume thereof can also be decreased by about the same percentage. Since ashing lowers the gamma radiation dosage of radiocontaminated waste matter to a value close to the blank spatial gamma radiation dosage, it is estimated that a large amount of ashes can be loaded into the tank 1, and that the spatial gamma radiation dosage around the tank 1 is to be about the same as the blank spatial gamma radiation dosage of 0.065 to 0.072 μSv/h. When radiocontaminated waste matter or a mixture of radiocontaminated waste matter and PSC which replaces 20% thereof is ashed, and PSC in an amount corresponding to 20% of this ash thus obtained is again added thereto and mixed therewith, the weight, volume, and radioactive 134Cs and 137Cs of the obtained ash mixture are all decreased, and the surrounding gamma radiation dosage becomes similar to the blank spatial gamma radiation dosage. Hence, the filling rate of the obtained ashes in the tank 1 would be improved, the spatial gamma radiation dosage in the surrounding environment of the tank 1 would not cause any problem, and a safe and secure long-term retention and storage is thus possible. Below are the examples of the present invention. In order to confirm the shielding efficiency of radiocontaminated waste matter by a tank made of steel sheet and a tank made of concrete, a rectangular steel sheet tank, a rectangular concrete tank, and a rectangular concrete tank in which 15% of gravel was replaced with PSC were constructed, and the shielding tests were performed. As the sample of radiocontaminated waste matter, radiocontaminated soil from a paddy in Iitate village, Fukushima prefecture was collected at the beginning of September 2013 (soil in some areas of Fukushima prefecture contained radioactive matters because of the Nuclear Power Plant disaster on Mar. 11, 2011). The gamma radiation dosage was measured using a Hitachi-Aloka pocket survey meter PDR-111. 134Cs and 137Cs were determined using a Canberra coaxial germanium detector in accordance with “the manual for radiation measurement of foods in emergencies” issued by the Japanese Ministry of Health, Labor and Welfare and “Gamma-ray spectrometry with germanium semiconductor detectors”, issued by the Japanese Ministry of Education, Culture, Sports, Science and Technology. The gamma radiation dosage of the radiocontaminated paddy soil was 1.763 μSv/h, and the total of 134Cs and 137Cs thereof was 26,914 Bq/kg (30,277 Bq/kg oven-dry weight). The steel sheet tank was a cold rolled square steel pipe for construction purposes (BCR). It had a thickness of 6.05 mm, an inside width of 8.78 cm, a height of 30.11 cm, and a density of 7.10 g/cm3. The lid and the bottom base plate were steel sheets having a width of 22.9 cm and a thickness of 8.90 mm. After the steel sheet tank was fixed on the bottom base plate, 1,310 g oven-dry weight (1,853.6 g air-dried weight) of the radiocontaminated paddy soil was loaded into the tank. The tank was covered with the lid and left to stand. The gamma radiation dosage was measured on the lid and on the ground side of the bottom base plate for five minutes per measurement. On Day 15, an outside steel sheet tank, namely, a cold rolled rectangular steel pipe having a thickness of 8.01 mm, an inside width of 18.35 cm, a height of 30.11 cm, and a density of 7.18 g/cm3, was installed to enclose the first-mentioned steel sheet tank. PSC was charged into the 4.18 cm space between the outside and the inside steel sheet tanks, and the tanks were again covered with the lid and left to stand. Subsequently, the gamma radiation dosage was measured. TABLE 1<Shielding Efficiency of a RadiocontaminatedPaddy Soil by a Rectangular Steel Pipe>Spatial Gamma Radiation Dosage (μSv/h)On Upper LidGround at Bottom BaseDayMeanσMeanσDay 10.1040.0100.1110.012Day 80.0970.0130.1040.009Day 150.0970.0140.0990.005Day 22 0.0970.0090.0950.011Day 29 0.0950.0080.0900.010* After PSC was charged in the space between the outside and inside rectangular steel pipes. As shown in Table 1, the gamma radiation dosage decreased from an initial value of 1.763 μSv/h to 0.097 μSv/h measured on Day 15, and a shielding rate of 94.5% was obtained. The last two gamma radiation dosage measurements were made after installing the outside steel sheet tank and charging the PSC. The final value (0.090 to 0.095 μSv/h) was slightly higher than the blank spatial gamma radiation dosage of 0.065 to 0.072 μSv/h. In a test in which 20% of the radiocontaminated paddy soil was replaced with PSC, a mixture of the radiocontaminated paddy soil (1,048 g oven-dry weight, 1,482.9 g air-dried weight) and PSC (262 g oven-dry weight, 266.6 g air-dried weight) was loaded into the same steel sheet tank of Reference Example 1, and the experiment was performed in accordance with the method described in Reference Example 1 but without the outside steel sheet tank. Compared to the mixture of the radiocontaminated paddy soil and PSC, the spatial gamma radiation dosage of the original radiocontaminated paddy soil was decreased by 72.1% from the initial value of 1.763 μSv/h to 0.491 μSv/h, and the total of 134Cs and 137Cs decreased by 27.9%, that is, from 30,227 Bq/kg oven-dry weight to 21,788 Bq/kg oven-dry weight. These results suggest that there were other components in the gamma-ray that were easier to be shielded under the presence of PSC than 134Cs and 137Cs. In this example, the oven-dry weight is the weight obtained when the moisture content of the paddy soil is 0%, i.e. the paddy soil is dried at 105° C. until the weight thereof is constant. TABLE 2<Shielding Efficiency of a Mixture of RadiocontaminatedPaddy Soil and PSC by a Rectangular Steel Pipe>Spatial Gamma Radiation Dosage (μSv/h)On Upper LidGround at Bottom BaseDayMeanσMeanσDay 10.1010.0060.0870.006Day 20.1030.0080.0850.008Day 80.0960.0060.0840.007Day 210.0800.0100.0800.005Day 280.0780.0080.0800.006 A comparison of Tables 1 and 2 shows that the spatial gamma radiation dosage on Day 1 of the mixture of radiocontaminated paddy soil and PSC (Table 2) was slightly lower than that of the radiocontaminated paddy soil alone (Table 1), and that the value on Day 28 was approximately equal to the blank spatial gamma radiation dosage of 0.065 to 0.072 μSv/h. It is believed that this is due to the ion exchange between PSC and radioactive substance in the radiocontaminated paddy soil. The tank made of PSC-containing concrete was constructed from cement, sand, gravel, and PSC without using any reinforcing steel, and the mixing rate was 12% of cement, 24% of sand, 51% of gravel, and 20.3% of PSC with respect to the gravel. The main body, the lid, and the bottom base plate of the concrete tank were made of the same raw materials and mixing ratios. The specifications of the tank made of PSC-containing concrete were: a thickness of 61.02 mm, an inside width of 86.65 mm, a height of 30.56 cm, and a density of 1.817 g/cm3. The specifications of the lid were: a width of 34.85 cm and a thickness of 32.28 mm. The specifications of the bottom base plate were: a width of 40.5 cm and a thickness of 32.73 mm. The tank made of PSC-containing concrete was used to carry out the same shielding test as that performed with the steel sheet tank. The concrete tank was enclosed by an outside concrete tank on Day 25, and the shielding test was continued afterward. Of the outside concrete tank, the raw materials and their mixing ratios were similar to those of the inside concrete tank but the thickness was 61.94 mm, the outside width 34.52 cm, and the height 30.55 cm. The space between the outside and inside concrete tanks was approximately 12 mm. TABLE 3<Shielding Efficiency of Radiocontaminated Paddy Soilby PSC-containing Concrete Rectangular Tank>Spatial Gamma Radiation Dosage (μSv/h)On Upper LidGround at Bottom BaseDayMeanσMeanσDay 10.1040.0100.1150.012Day 100.1010.0110.1040.009Day 250.0970.0100.0990.005Day 40 0.0900.0100.0950.011Day 51 0.0900.0170.0900.010* After the inside rectangular tank was enclosed by the outside rectangular tank. The shielding test was carried out for 51 days using a double shielding structure composed of the outside and inside concrete tanks. The final value (0.090 μSv/h) of the spatial gamma-ray dosage was slightly higher than the blank spatial gamma radiation dosage of 0.065 to 0.072 μSv/h (Table 3) and was substantially the same as the result of the shielding test using the steel sheet tank (Table 1). Hence, a test for confirming the effect of the addition of PSC to the radioactively contaminated paddy soil was performed. Similar to the shielding test using the steel sheet tank, a mixture of the radiocontaminated paddy soil and PSC at a oven-dry weight ratio of 4:1 was loaded into an inside concrete tank, and the test was performed according to the method of Reference Example 2 but without the outside concrete tank. As in the case of Example 1 of the steel sheet tank, the value on Day 51 was almost equivalent to the blank spatial gamma radiation dosage of 0.065 to 0.072 μSv/h. The results are shown in Table 4. TABLE 4<Shielding Efficiency of Mixture of Radiocontaminated PaddySoil and PSC by PSC-containing Concrete Rectangular Tank>Spatial Gamma Radiation Dosage (μSv/h)On Upper LidGround at Bottom BaseDayAverageσAverageσDay 10.0970.0100.1100.012Day 80.0900.0100.1040.010Day 200.0830.0090.0920.008Day 360.0800.0100.0840.011Day 510.0770.0070.0800.008 In order to improve the filling rate of the radiocontaminated paddy soil in the storage tank, the said paddy soil was ashed in an electric furnace at 850° C. for 90 minutes. The ash contents of the radiocontaminated paddy soil alone and the mixture of PSC and radiocontaminated paddy soil were 89.67% and 86.03%, respectively. Thus, the ashing lowered the weight of the radiocontaminated paddy soil by 10% to 15%, and it is expected that the volume thereof is also reduced by similar percentages. As indicated in Table 5 below, because the ashes of the radiocontaminated paddy soil alone and the mixture of PSC and radiocontaminated paddy soil showed spatial gamma radiation dosages close to the blank spatial gamma radiation dosage, it is believed that if the ashes were to be loaded and stored in a tank, such as a steel sheet tank, a concrete tank, or a PSC-containing concrete tank, the spatial gamma radiation dosage of the environment around the tank would be decreased to a value similar to the blank spatial gamma radiation dosage of 0.065 to 0.072 μSv/h. Compared to the radiocontaminated paddy soil, the total of 134Cs and 137Cs of the mixture of PSC and the radiocontaminated paddy soil decreased by 27.9% (from 30,227 Bq/kg oven-dry weight to 21,788 Bq/kg oven-dry weight) as in the case of Reference Example 1. However, after the radiocontaminated paddy soil itself and the mixture of PSC and the radioactively contaminated paddy soil were ashed, the results for 134Cs and 137Cs were approximately the same as those for the samples before the ashing. Thus, the spatial gamma radiation dosage was decreased by ashing but no change in 134Cs and 137Cs was observed. TABLE 5<Effect of Ashing on Radiocontaminated Paddy Soil alone aswell as Mixture of Radiocontaminated Paddy Soil and PSC>ConstituentsRadio-Mixture of radio-contaminatedcontaminatedpaddy soilpaddy soil andalone20% PSCAsh content (%)Mean89.6786.03σ0.791.00Gamma ray radiation dosage (μSv/h)ContaminatedMean1.2810.148paddy soilσ0.1240.015AshMean0.0820.078σ0.0100.007Radiocesium (Bq/kg oven-dry weight)Contaminated134Cs8,3096,025paddy soil137Cs21,91815,763Total30,22721,788Ash134Cs8,2336,573137Cs21,07417,299Total29,30723,871 Since the gamma ray radiation dosage of the ash in {circle around (11)}Example 3 was slightly higher than the blank spatial gamma radiation dosage of 0.065 to 0.072 μSv/h, this ash was added with PSC in an amount corresponding to 20% thereof (percentage to the weight of the ash), followed by mixing. After the mixture had stood for 3 days, the gamma radiation dosage of the mixture of the ash and PSC was measured. As shown in Table 6, when PSC was added to the ash of the radiocontaminated paddy soil alone, there was no change in the gamma radiation dosage. However, when PSC was added to the ash of the mixture of PSC and radiocontaminated paddy soil, the gamma radiation dosage was approximately the same as the blank spatial gamma radiation dosage of 0.065 to 0.072 μSv/h. Hence, the suitable method to securely and safely retain and store radiocontaminated paddy soil is as follows. First, the radiocontaminated paddy soil is mixed with PSC and then ashed. The ash thus obtained is again mixed with PSC and then loaded and stored in a tank, such as a steel sheet tank, a concrete tank, or a PSC-containing concrete tank. After PSC was added to the ash of the radiocontaminated paddy soil, the spatial gamma radiation dosage was slightly higher than the blank spatial gamma radiation dosage of 0.065 to 0.072 μSv/h but similar to the value prior to addition of PSC. On the other hand, the total of radiocesiums was decreased by 24% (from 29,307 Bq/kg oven-dry weight to 22,354 Bq/kg oven-dry weight). Accordingly, when the mixture thus obtained is loaded and stored in a tank, such as a steel sheet tank, a concrete tank, or a PSC-containing concrete tank, safe and secure long-term storage is possible. When PSC was added to the ash of the mixture of PSC and radiocontaminated paddy soil, the resulting spatial gamma radiation dosage was approximately the same as the blank spatial gamma radiation dosage of 0.065 to 0.072 μSv/h, and the radiocesiums were decreased by approximately 40% (23,871 Bq/kg oven-dry weight to 14,878 Bq/kg oven-dry weight) as compared to those before the addition of PSC. Accordingly, when the mixture thus obtained is loaded and stored in a tank, such as a steel sheet tank, a concrete tank, or a PSC-containing concrete tank, safe and secure storage is possible for a long-period of time. TABLE 6<Effect of Addition of PSC to the Ashes of RadiocontaminatedPaddy Soil Alone and Mixture thereof with PSC>ConstituentsAsh of radio-Ash of mixturecontaminatedof radio-paddy soilcontaminatedalone +paddy soil and20% PSCPSC + 20% PSCGamma radiation dosage (μSv/h)Before addition of PSCMean0.0820.078σ0.0100.007After addition of PSCMean0.0830.071σ0.0070.006Radiocesium (Bq/kg oven-dry weight)Before addition of PSC134Cs8,2336,573137Cs21,07417,299Total29,30723,871After addition of PSC134Cs6,2874,218137Cs16,06710,660Total22,35414,878 After radiocontaminated waste matter, such as radioactively contaminated rubble, soil, soil slurry, farmed mushrooms, and leaves generated by the decontamination works carried out in regions contaminated radioactively by a nuclear power plant accident; and/or radiocontaminated sludge and boiler ashes generated from treatment facilities for radiocontaminated waste water; and/or radiocontaminated mushrooms, leaves, and the like in radiocontaminated forests are mixed with PSC and then ashed, PSC is further added to the ash thus formed and then mixed therewith to form a mixture, and the mixture thus obtained is loaded and stored in steel sheet, concrete, or PSC-containing concrete tank 1 having a regular polygonal cylindrical shape with at least four corners, a circular cylindrical shape, or other suitable shape. Accordingly, the spatial gamma radiation dosage of the environment around the tank 1 is to be similar to the spatial gamma radiation dosage at a place which receives no fall-out radioactive substances from the nuclear power plant accident, and thus safe and secure retention and storage is possible. In addition, the lid 2 and the bottom base plate 4 of the tank 1 are made of the same raw material as that of the main body of the tank 1, and the lid 2 is secured by hooks 3. Since the tank 1 can be set up on the ground, it can be advantageously built from the cost and technical aspects. Although embodiments of the present invention were described in detail in the foregoing, the present invention is not limited to the embodiments described above. Additionally, various changes in design may be performed without departing from the scope disclosed in claims of the present invention.

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