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claims | 1. An apparatus for reducing contamination during ion implantation, the apparatus comprising:a platen to hold a workpiece for ion implantation by an ion beam;a mask, located in front of the platen and downstream from a corrector magnet, to block the ion beam and at least a portion of contamination ions having a charge state different than a desired charge state from reaching a first portion of the workpiece during ion implantation of a second portion of the workpiece, wherein a position of the mask relative to the ion beam is chosen based on a likelihood of impact by the contamination ions; anda control mechanism, coupled to the platen, to reposition the workpiece to expose the first portion of the workpiece for ion implantation. 2. The apparatus according to claim 1, wherein the second portion comprises the remaining portion of the workpiece. 3. The apparatus according to claim 1, wherein the first portion comprises one half of the workpiece, and wherein the second portion comprises the other half of the workpiece. 4. The apparatus according to claim 3, wherein the mask has a semi-circular shape. 5. The apparatus according to claim 1, wherein the position of the mask remains in a fixed relative position with respect to the ion beam during the ion implantation of the first portion and the second portion of the workpiece. 6. The apparatus according to claim 1, wherein the mask is made from one or more materials selected from a group consisting of: silicon, carbon, and silicon carbide. 7. The apparatus according to claim 1, wherein the control mechanism repositions the workpiece by rotating the workpiece by a predetermined angle. 8. The apparatus according to claim 1, wherein the ion implantation on the first portion of the workpiece is based on a recipe different from the ion implantation on the second portion of the workpiece. 9. The apparatus according to claim 1, wherein the ion beam is a ribbon beam, and wherein the ion implantation of the first portion and the second portion of the workpiece is performed by translating the workpiece and the mask relative to the ribbon beam. 10. A method for reducing contamination during ion implantation, the method comprising the steps of:positioning a mask in front of a workpiece and downstream from a corrector magnet to block an ion beam and at least a portion of contamination ions having a charge state different than a desired charge state from reaching a first portion of the workpiece during ion implantation of a second portion of the workpiece, wherein a position of the mask relative to the ion beam is chosen based on a likelihood of impact by the contamination ions; andrepositioning the workpiece, after the ion implantation of the second portion, to expose the first portion of the workpiece for ion implantation. 11. The method according to claim 10, wherein the position of the mask remains in a fixed relative position with respect to the ion beam during the ion implantation of the first portion and the second portion of the workpiece. 12. The method according to claim 10, wherein the second portion comprises remaining portion of the workpiece. 13. The method according to claim 10, wherein the first portion comprises one half of the workpiece, and wherein the second portion comprises the other half of the workpiece. 14. The method according to claim 13, wherein the mask has a semi-circular shape. 15. The method according to claim 13, wherein the mask has a rectangular shape. 16. The method according to claim 10, wherein the step of repositioning the workpiece comprises a step of rotating the workpiece by a predetermined angle. 17. The method according to claim 10, wherein the ion implantation on the first portion of the workpiece is based on a recipe different from the ion implantation on the second portion of the workpiece. 18. The method according to claim 10, wherein the ion beam is a ribbon beam, and wherein the ion implantation of the first portion and the second of the workpiece is performed by translating the workpiece and the mask relative to the ribbon beam. 19. The apparatus according to claim 1, wherein the charge state is a singly charged state and the desired charge state is a doubly charged state. 20. The method according to claim 10, wherein the charge state is a singly charged state and the desired charge state is a doubly charged state. |
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055263879 | claims | 1. A spacer for use with a substantially vertically oriented fuel bundle in a nuclear reactor, the spacer comprising: a matrix of ferrules for surrounding individual fuel rods within a bundle; a substantially horizontally oriented band surrounding said matrix and defining a peripheral wall of the spacer, said band having an upper edge; a plurality of laterally spaced flow tabs extending upwardly from said upper edge, each flow tab having a substantially vertical portion and an inclined portion; and first means on said inclined portion for imparting velocity components to a flow of coolant along said fuel rods, which components are substantially parallel to said band. a matrix of ferrules for surrounding individual fuel rods within a bundle; a substantially horizontally oriented band surrounding said matrix and defining a peripheral wall of the spacer, said band having an upper edge; a plurality of laterally spaced flow tabs extending upwardly from said upper edge, each flow tab having a substantially vertical portion joined to said band and an inclined portion joined to said vertical portion and extending upwardly and inwardly relative to said vertical portion, said inclined portion having a first center line located downwardly and inwardly relative to a pair of lateral free edges of said inclined portion. a matrix of ferrules for surrounding individual fuel rods within a bundle; a substantially horizontally oriented band surrounding said matrix and defining a peripheral wall of the spacer, said band having an upper edge; and a plurality of laterally spaced flow tabs extending upwardly from said upper edge, each flow tab having a lower substantially vertical portion and an upper inclined portion extending away from said vertical portion, said vertical portion having a vertical crease and said inclined portion having an inclined crease, each located centrally of said tab. 2. The spacer of claim 1 and further including second means on said vertical portion for rigidifying said flow tab. 3. The spacer of claim 1 wherein said first means includes a first pair of trapezoidal surfaces angled with respect to one another and said second means includes a second pair of trapezoidal surfaces also angled with respect to each other. 4. The spacer of claim 3 wherein said first and second pair of trapezoidal surfaces have contiguous center lines. 5. The spacer of claim 4 wherein each of said first and second pair of trapezoidal surfaces is substantially planar. 6. A spacer for use with a substantially vertically oriented fuel bundle in a nuclear reactor, the spacer comprising: 7. The spacer of claim 6 wherein said vertical portion has a second center line located outwardly relative to a pair of lateral free edges of said vertical portion. 8. The spacer of claim 6 wherein said inclined portion includes a first pair of substantially planar surfaces angled relative to each other about said first center line. 9. The spacer of claim 7 wherein said vertical portion includes a second pair of substantially planar surfaces angled relative to each other about said second center line. 10. The spacer of claim 9 wherein said first and second center lines are contiguous. 11. The spacer of claim 8 wherein said vertical portion includes a third substantially planar surface below said pair of substantially planar surfaces. 12. The spacer of claim 9 wherein a first included angle between said first pair of substantially planar surfaces is equal to a second included angle between said second pair of substantially planar surfaces. 13. A spacer for use with a substantially vertically oriented fuel bundle in a nuclear reactor comprising: 14. The spacer of claim 13 wherein said vertical crease is located outside said band, and said inclined crease is located inside said band. 15. The spacer of claim 13 wherein each flow tab has a pair of side edges, and wherein said vertical crease lies outside the side edges in said vertical portion and said inclined crease lies inside the side edges on said inclined portion. 16. The spacer of claim 13 wherein said vertical portion includes a first pair of substantially planar surfaces on either side of said vertical crease, and said inclined portion includes a second pair of substantially planar surfaces on either side of said inclined crease. 17. The spacer of claim 16 wherein said portion includes a third substantially planar surface below said vertical crease and connecting said first pair of substantially planar surfaces. |
039830500 | claims | 1. In the method of storing high-level radioactive liquid wastes by calcining the waste to remove the liquid, thereby forming a high-level radioactive calcine, placing the calcine in a metal canister, sealing the metal canister and placing the sealed metal canister in a moisture-containing environment for cooling and long-term storage, the improvement comprising: adding powdered portland cement to the canister containing the calcined wastes so that the cement is in contact with the inner surface of the wall of the canister before the canister is sealed, whereby, should the canister wall fail and develop an opening to the environment, moisture from the environment will enter the canister, mix with the portland cement, forming concrete which will harden, seal the opening and prevent the escape of any radioactivity from the canister to the environment. 2. The method of claim 1 wherein the powdered cement is mixed with the calcine to form an intimate mixture before the mixture is placed into the canister. 3. The method of claim 2 wherein the mixture contains at least 1 part by weight of cement powder per 10 parts by weight of calcine. 4. The method of claim 3 wherein the canister is placed in a water-filled tank for cooling and long-term storage. 5. The method of claim 1 wherein the powdered portland cement is added to the canister to form an outer annulus in contact with the inner wall of the canister and the radioactive calcine is in the center of the canister surrounded by the cement. 6. The method of claim 5 wherein the outer annulus of cement powder is at least about 1 inch in thickness. 7. The method of claim 6 wherein the canister is placed in a water-filled tank for cooling and long-term storage. |
claims | 1. A flexible composition able to stop high fluxes of gamma and neutron radiation and showing resistance to high temperatures, said composition comprising a uniform mixture of: between about 10%-30% by weight an organic polymer selected from the group consisting of silicone rubber, siloxanes, silanols, vinyl elastomers and fluorocarbon polymers for providing a flexible matrix; between about 25%-75% by weight of a powdered gamma radiation shielding material selected from the group consisting of tungsten, lead, tin, antimony, indium and bismuth for increasing gamma radiation shielding of the mixture; between about 5%-10% by weight of a neutron absorbing material selected from the group consisting of boron, cadmium and gadolinium for increasing neutron absorption of the mixture; up to about 5% by weight diamond powder for increasing thermal conductivity of the mixture; up to about 5% by weight powdered silicon dioxide for increasing thermal resistance of the mixture; up to about 5% by weight of barium sulfate powder for increasing neutron absorption and electrical conductivity of the mixture; and between about 2% and 8% by weight of a hydrogen absorbing material selected from the group consisting of palladium, lithium, calcium, titanium, scandium, lithium nickel compounds, lanthanum nickel compounds, yttrium nickel compounds, samarium cobalt compounds and yttrium cobalt compounds for absorbing hydrogen gas. 2. The mixture of claims 1 , wherein the organic polymer comprises a silicone rubber. 3. The mixture of claims 2 , wherein the silicone rubber is formulated to produce a flexible foam upon polymerization. 4. The mixture of claims 1 , wherein the gamma shielding material comprises tungsten. 5. The mixture of claims 4 , wherein the tungsten comprises tungsten carbide. 6. The mixture of claims 1 , wherein the gamma shielding material is metallic. 7. The mixture of claims 1 , wherein the gamma shielding material is a salt. 8. The mixture of claims 7 , wherein the salt comprises an iodide. 9. The mixture of claims 1 , wherein the neutron absorbing material comprises boron. 10. The mixture of claims 9 , wherein the boron comprises one of boron carbide, boron nitride and a mixture of boron carbide and boron nitride. 11. The mixture of claims 1 , wherein the powdered silicon dioxide comprises quartz. 12. The mixture of claims 1 , wherein the hydrogen absorbing material comprises sponge palladium. 13. The mixture of claims 1 , wherein the organic polymer is silicone rubber foam, the gamma radiation shielding material is tungsten carbide, the neutron absorbing material is a mixture of boron carbide and boron nitride, and the hydrogen absorbing material comprises a mixture of titanium, a lanthanum nickel compound and a samarium cobalt compound. 14. A container for highly radioactive material comprising: an inner container; an outer container surrounding the inner container and spaced apart therefrom; and a space between the inner container and the outer container, said space filled with the composition of claim 1 . claim 1 15. A flexible composition able to stop high fluxes of gamma and neutron radiation and showing resistance to high temperatures, said composition comprising a uniform mixture of: between about 10%-30% by weight silicone rubber for providing a flexible matrix; between about 25%-75% by weight of powdered tungsten for increasing gamma radiation shielding of the mixture; between about 5%-10% by weight of powdered boron for increasing neutron absorption of the mixture; up to about 5% by weight diamond powder for increasing thermal conductivity of the mixture; up to about 5% by weight powdered silicon dioxide for increasing thermal resistance of the mixture; up to about 5% by weight of barium sulfate powder for increasing neutron absorption and electrical conductivity of the mixture; and between about 2% and 8% by weight of a hydrogen absorbing material selected from the group consisting of palladium, lithium, calcium, titanium, scandium, lithium nickel compounds, lanthanum nickel compounds, yttrium nickel compounds, samarium cobalt compounds and yttrium cobalt compounds for absorbing hydrogen gas. 16. A container for highly radioactive material comprising: an inner container; an outer container surrounding the inner container and spaced apart therefrom; and a space between the inner container and the outer container, said space filled with the composition of claim 15 . claim 15 |
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description | The methodology of the present invention for determining reduced in-core instrumentation (ICI) patterns is grounded in considerations recommended by the United States Nuclear Regulatory Commission (USNRC) for inclusion in and evaluation of changes to the ICI system. These considerations are made in accordance with federal regulations, specifically, 10 C.F.R. 50.59. The considerations outlined by the USNRC include the following: 1. Detecting inadvertent loading of a fuel assembly into an improper location; 2. Insuring the validity of core power tilt estimates; 3. Maintaining adequate core coverage by instrumentation; 4. Limiting measurement uncertainties to meet plant Technical Specification limits for various measured values, including measured peak linear heat rates, peak pin powers, radial peaking factors, and azimuthal tilts; and 5. Restoring ICI system to full open (or nearly full) service at the beginning of each fuel cycle. Utilizing these considerations, determination of the reduced ICI patterns according to the present inventive methodology includes two main parts. First, candidate ICI patterns having a reduced number of ICIs relative to the existing ICI patterns are selected according to specific selection considerations. Second, the selected candidate ICI patterns are evaluated according to specific criteria. According to the first part of the inventive methodology, candidate ICI patterns having a reduced number of ICIs must be selected. In order to guide selection of the candidate ICI patterns according to the inventive methodology, selection considerations were established to ensure that reduced patterns of ICIs are capable of performing all functions required of a full complement of ICIs. The functions required of the reduced ICI patterns include: i. maintaining the core power tilt measurement capability; ii. providing the capability to perform ex-core detector calibrations; iii. providing full core coverage; iv. detecting fuel misloadings; v. detecting misalignment of the lead control rod bank; and vi. providing sufficient redundancy. First, a reduced ICI pattern maintains core power tilt measurement capability by retaining all ICIs from the existing ICI pattern that belong to tilt groups. In this way, core power tilt measurement capability is not affected by the reduced ICI pattern, because all tilt-related ICIs are retained. Second, the integrity of ex-core detector calibrations is maintained by retaining all ICIs from the existing ICI pattern that provide information about the power distribution in those fuel assemblies that contribute most of the neutron flux impinging on the ex-core detectors. By maintaining measurement capability within those fuel assemblies that contribute most neutron flux to the ex-core detectors, comparisons may be made between the retained ICIs and the ex-core detectors to provide full ex-core detector calibration. Third, full core instrumentation coverage is not impacted through the use of a reduced ICI patterns because those ICIs that provide a uniform distribution of instrumented fuel assemblies from the periphery of the core to the center of the core, in all core quadrants, are retained. In other words, by maintaining a uniform distribution of ICIs, though the number is reduced, full core instrumentation coverage is still provided through the present invention. Fourth, fuel misloadings must be detected by the reduced ICI pattern. According to the inventive method, fuel assembly misleading detection capability of the reduced ICI patterns is verified by simulating fuel misloading during the ICI pattern evaluation process. The evaluation process is discussed in more detail below. Fifth, reduced ICI patterns maintain the capability to detect misalignment of the lead control rod bank. This capability is ensured by retaining those ICIs that are sufficiently close to a control rod location to enable detection of any perturbation in the power distribution associated with movement of the lead control rod bank. By retaining those ICIs that are in a position to detect power distribution changes caused by movement of the lead control rod bank, misalignment of the lead control rod bank is easily detected. Lastly, redundancy of the reduced ICI patterns are proven by evaluating the performance of the reduced ICI patterns in combination with unexpected ICI failures to prove that the reduced ICI patterns still perform their intended functions. The various evaluations are discussed in more detail below. Once a reduced ICI pattern has been selected based upon meeting the various selection considerations set forth above, the reduced ICI pattern configuration is evaluated to ensure that any differences between the predicted core power distributions and those synthesized from the reduced ICI patterns are within licensed limits for the plant. Moreover, the reduced ICI patterns are evaluated to prove that the reduced ICI patterns are able to detect misleading of an improperly located fuel assembly. Further, the reduced ICI pattern is tested with only 75% of the ICIs operable and is forced to meet the full evaluation criteria, even though 25% of the ICIs are inoperable. Thus, under the evaluation portion of the methodology of the present invention, candidate patterns are evaluated according to the following criteria: A. The differences between the predicted core power distributions and those synthesized from the reduced ICI patterns must be in compliance with the limits that have been licensed for the plant; B. The reduced ICI patterns must provide the capability of detecting the misloading of a fuel assembly placed into an improper location; and C. The reduced ICI patterns must be capable of performing all intended functions with only 75% of the ICIs operable, in accordance with the plant Technical Specification. In order to determine that power distribution uncertainties are within licensed limits, comparisons are performed of predicted power distributions obtained for design calculations with measured power distributions obtained from ICI detector signals using accepted licensed methodologies. Two types of uncertainty are evaluated: basic measurement uncertainty and the synthesis uncertainty. Basic measurement uncertainty relates to the local power in instrumented fuel assemblies. Synthesis uncertainty is associated with extrapolating power calculations to non-instrumented fuel assemblies. In order for a reduced ICI pattern to be acceptable, both calculated basic measurement uncertainty and synthesis uncertainty must be within the limits licensed for the plant. In the second part of the evaluation, candidate ICI patterns that satisfy the uncertainty analysis are tested to ensure the ability to detect fuel misloadings. Testing of the fuel misloading detection ability with a reduced ICI pattern is verified by simulating a spectrum of fuel misloadings and examining the differences in the power distributions between a properly loaded core and a simulated misloaded core, as measured by the reduced ICI pattern. Such simulations may be performed using conventionally known simulation techniques. In order to be acceptable, the differences in power distributions obtained from ICI detector signals for the properly loaded core and the simulated misloaded core must be discernable. If the differences are discernable, it is assumed that the reduced ICI pattern is able to detect fuel misloadings. Finally, each reduced ICI pattern is evaluated for its capability to perform its intended functions under a 75% operability Technical Specification requirement. That is, each reduced ICI pattern is further reduced to 75% of the reduced pattern number to demonstrate that the reduced ICI pattern can perform its intended functions with only 75% of the ICIs operable. The ICI operability limit is derived from the plant Technical Specification and the plant license. Existing plant Technical Specifications set forth a 75% ICI operability limit. If the ICI operability limit is ever changed, e.g. increased to 90% or decreased to 70%, then each reduced ICI pattern would be evaluated for its capability to perform its intended functions under that particular Technical Specification ICI operability limit. Applying the methodology described above, several reduced ICI patterns have been determined to be acceptable for various core configurations. Several of these acceptable reduced ICI patterns are set forth below as examples and are shown in FIGS. 1-8. It should be understood that the following Examples are exemplary only, and do not limit the scope of the invention in any way. In FIGS. 1-8, boxes representing fuel assemblies that contain an ICI in existing cores of this configuration are shown with both light and dark cross-hatching, and include a corresponding numeric designation in parentheses. Those boxes including dark cross-hatching correspond to those assemblies from which an ICI has been removed using the methodology of the present invention. Thus, those boxes including light cross-hatching correspond to those assemblies wherein ICIs have not been removed. FIG. 1 shows a plan diagram of a 241 fuel assembly PWR core. Under existing plant designs, a 241 fuel assembly core would ordinarily have an existing ICI pattern of 61 ICIs distributed as shown in FIG. 1. Using the methodology described herein, 12 ICIs (numbers 13, 22, 23, 25, 27, 30, 34, 35, 37, 43, 44 and 47) are removed, leading to a reduced ICI pattern of 49 ICIs. The net ICI reduction in FIG. 1 by the claimed methodology is 20%. A 217 fuel assembly core is shown having an original ICI pattern of 56 ICIs. Using the methodology described herein, 12 ICIs (numbers 2, 4, 9, 15, 18, 20, 37, 39, 42, 48, 53 and 55) are removed from the core. The reduced ICI pattern thereby includes only 44 ICIs, for a net reduction of 21% FIG. 3 shows an alternative 217 fuel assembly core configuration having an original ICI pattern of 45 ICIs. The original ICI pattern in FIG. 3 is different from that shown in FIG. 2. Using the methodology described herein, 10 ICIs (numbers 9, 12, 18, 19, 20, 27, 28, 33, 35 and 37) are removed from the core, yielding a reduced ICI pattern of 35 ICIs, for a net reduction of 22%. In FIG. 4, another configuration is shown for a 217 fuel assembly core with an original ICI pattern of 45 ICIs. Again, the existing ICI pattern shown in FIG. 4 is different from the patterns of either FIGS. 2 or 3. Using the inventive methodology, 10 ICIs (8, 12, 18, 19, 20, 27, 31, 33, 35 and 35) are removed, yielding a reduced ICI pattern of 35 ICIs is shown, yielding a reduction of 10 ICIs, or 22%. In FIG. 5, another 217 fuel assembly core pattern is shown having a 45 ICI original pattern that is different from that shown in either FIGS. 2, 3 or 4. Using the inventive methodology, 10 ICIs (9, 12, 18, 19, 20, 27, 28, 33, 35 and 37) are removed, yielding a reduced ICI pattern of 35 ICIs, yielding a net reduction of 22%. In FIG. 6, a 177 fuel assembly core is shown that includes an original ICI pattern of 45 ICIs. Using the inventive methodology, 6 ICIs (16, 17, 20, 26, 32 and 33) are removed, reducing the number of ICIs to 39, yielding a net reduction of 13%. In FIG. 7, another 177 fuel assembly core configuration is shown having an original ICI pattern of 44 ICIs. Using the inventive methodology, the number of ICIs is reduced by 8 (Numbers 3, 5, 14, 20, 25, 31, 40 and 42) to 36 ICIs, yielding a net reduction of 18%. FIG. 8 shows a 133 fuel assembly core having an original ICI pattern of 28 ICIs. Using the inventive methodology, the number of ICIs is reduced by 3 (Numbers 10, 18 and 20), yielding a net reduction of 11%. In each of examples 1-8, the reduced ICI patterns shown satisfy all requirements of the ICI system while providing substantial reductions in the number of ICIs. Reducing the number of ICIs reduces both plant operating and maintenance costs. If the present methodology is implemented during new plant construction, the amount of additional equipment necessary to support the ICI system would be reduced because fewer ICIs would have to be supported. As a result, a significant reduction in overall capital cost due to construction as well as due to operating and maintenance is realized through implementation of the present invention. Moreover, associated costs from reactor down time during refueling outages may be reduced because a shorter time needed to replace a reduced complement of ICIs. Moreover, because the time along the critical path is correspondingly reduced, potential radiation exposure to plant personnel is also reduced. Various embodiments of this invention have been disclosed. But it should be realized that the various changes and modifications that are possible will be self-evident to those of skill in the art in which the present invention pertains, and may be made without departing from the scope of the invention, which is limited only by the appended claims. |
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047598964 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention This invention relates in general to the field of power producing nuclear reactors and in particular to methods and apparatus for improving neutron flux reduction factors outboard of the core periphery. 2. Description of the Prior Art It is well known that nuclear reactors are both a technical and commercial success. In one type of commercial nuclear power reactors, commonly referred to as a pressurized light water reactor, a reactive region commonly referred to as a nuclear core contains a nuclear fuel such as uranium 235, as well as other fissile materials, which undergo sustained fission reactions and in so doing, generate heat. There are, of course, other materials in the nuclear core, the presence of such other materials, however, is not germane to this invention and, accordingly, will not be discussed. Typically, a group of mechanical components, which are known as reactor internals, structurally support the core within a heremetically sealed pressure vessel. The reactor internals also direct the flow of a cooling medium, such as light water in pressurized, light water nuclear reactors, into the pressure vessel through the nuclear core, and out of the pressure vessel. The cooling medium, which is alternatively called the reactor coolant, removes the heat generated by the fissioning of the nuclear fuel and transfers the heat to another cooling medium within heat exchangers which are typically located external of the pressure vessel. The second cooling medium, which is usually water, is converted into steam in the heat exchangers and is thereafter used to produce electricity by conventional steam turbine-electrical generator combinations. The nuclear core, in the type of nuclear reactor described herein, usually comprises, an array of fuel assemblies stacked together in a side-by-side parallel arrangement to form a shape approximately that of a right solid circular cylinder. Each of such fuel assemblies include a multiplicity of elongated fuel rods and control rod guide tubes held together in a parallel array by grids spaced along the fuel assembly length. Each fuel rod may comprise an elongated slender hollow tube to be filled with nuclear fuel pellets and sealed at each end. Top and bottom nozzles on opposite ends of the fuel assembly and secured to the guide tubes provide for reactor coolant flow into and out of the fuel assemblies. The guide tubes allow for the insertion of elongated control rod assemblies into the nuclear core and dispersed among the nuclear fuel. The control rod assemblies provide for reactor control and serve to accomplish other neutronic purposes. The reactor internals may include a core barrel comprising an elongated cylinder which is interposed between the nuclear core and the cylindrical wall of the pressure vessel. The nuclear core then is positioned within the core barrel. Typically, the reactor coolant enters the pressure vessel through one or more inlet nozzles, flows downward between the pressure vessel and the outside of the core barrel, turns 180.degree., and flows upward through the core and through the space between the outside of the core and the inside of the core barrel. The heated reactor coolant then turns 90.degree. and exits the pressure vessel through one or more exit nozzles and then to the heat exchangers previously mentioned. In the pressurized light water reactors, such as the one described, the fissioning of the nuclear fuel results from the capture of a neutron by the nucleus of the atoms of the nuclear fuel. It is well known that each neutron producing a fission causes heat and the production of more than one other neutron (on the average 2.1 neutrons are released per capture). To sustain the nuclear chain reaction, at least one of the newly produced neutrons must then fission another atom of fuel. Since the neutrons generated are fast neutrons, and fissioning is enhanced by slow neutrons, it is advantageous that the fast neutrons be slowed down or thermalized within the confines of the nuclear core. The light water reactor coolant is an excellent moderator of neutrons; hence, in reactors primarily using U-235 as the nuclear fuel, it is the primary means by which the fast neutrons produced by the fission process are thermalized or slowed down so as to increase the probability that another fission may occur and thereby sustain the chain reaction. The excess neutrons produced by the fissioning of an atom and not used to fission another atom are accounted for in a number of different ways. Some are absorbed by the reactor internals. Others are slowed down and absorbed by a nuclear poison such as boron which is dissolved in the primary coolant. Other neutrons are absorbed by load follow control rods containing nonburnable control poisons which control rods comprise the means for controlling the nuclear reactor. Others are absorbed by special control rods which are interspersed throughout the nuclear core and made of materials specifically selected to absorb neutrons such as burnable poisons which as their name implies are burned during reactor operation and, therefore, become less effective in proportion to the continually reducing reactivity of the nuclear core. Still other neutrons are absorbed by poisons which build up within the nuclear fuel and are caused by the fission process itself. Quite obviously, the accounting for the excess neutrons is a complicated matter which can, however, be summarized by stating that some excess neutrons are purposefully absorbed while the remainder are inadvertently absorbed. And, it is desirable to reduce the number that are inadvertently absorbed. In order to extend the life of the nuclear core as long as is practical so as to minimize time consuming reactor shutdowns for refueling purposes, the fuel assemblies may be provided with enriched nuclear fuel, usually enriched uranium 235. This excessive amount of reactivity is designed into the core at startup so that as the reactivity is depleted over the life of the core, the excess reactivity is then used, thereby extending the life of the core. The amount of enrichment continuously decreases as the reactor operates until such time as the core can no longer sustain the chain reaction. Then the reactor must be shut down and refueled. During the initial stages of reactor operation or during the phase which is known as beginning of life, special neutron absorbing control rods may be inserted within the core and/or additional soluble poisons may be dissolved within the reactor coolant and/or burnable poisons may be included within the fuel assemblies to absorb the excess reactivity. As the excess reactivity decreases due to the nuclear fuel being burned, the amount of insertion of the special control rods and/or the amount of soluble poison and/or the burnable poisons within the special control rods and/or the fuel assemblies are consumed consistent with the reduction in excess reactivity to maintain the chain reaction. In this manner, the excess reactivity is held in abeyance until it is needed. Enriched uranium is extremely expensive. It is preferable, therefore, to reduce the amount of enrichment whenever possible but without reducing the extended operating length of the life of the core. One recognized method to accomplish this result is by making more efficient use of the neutrons produced by the fission process. In general, such techniques are classified as fuel management techniques. Throughout the years of successful reactor operation, a significant amount of experience has been gained in the area of fuel management. As a result, a three-phase core loading plan has been developed whereby the core is divided into three regions, with each region receiving either new, once burned or twice burned fuel. After a period of reactor operation, the most burned fuel is removed and replaced by the now twice burned fuel which in turn is replaced by the now once burned fuel which is replaced by new fuel. Such a fuel management technique has been termed as out-in-in fuel management, which as its name implies moves fuel radially from the outer core regions progressively inward and toward the center of the core. The out-in-in fuel management technique has been gradually replaced with a more economical low leakage loading pattern. While more complicated than the out-in-in pattern, the low leakage pattern significantly improves the neutron economy of fuel reload cycles and has thereby reduced the fuel cycle cost for a given energy output. The low leakage loading fuel management technique is designed to minimize the leakage of fast neutrons from out of the core so that these neutrons, as previously explained, may be used for fissioning purposes. Time and reduced fuel cycle costs have verified the success of the low leakage loading pattern. As an adjunct to the lowering of fuel costs from the use of the low leakage loading pattern, it has been determined that this pattern also results in improved flux reduction factors. That is, the high energy neutron flux which radially emanates from the nuclear core and which may ultimately irradiate the pressure vessel walls is reduced by the low leakage loading pattern. Needless to say, such an effect is obviously advantageous whether achieved by design or otherwise. One complication with the low leakage loading pattern is that it may or may not address a limiting region of the pressure vessel. Indeed, it is even possible that a particular low leakage loading pattern which is designed only to effectuate fuel cycle cost savings may have no effect at all or may even exacerbate neutron irradiation of a particular limiting pressure vessel location such as a weld. Another complication with using the low leakage loading pattern to improve the flux reduction factor is that the improvement may not be sufficient. Indeed, calculations have been made which show this to be the case. In other words, the low leakage loading pattern may not only be a hit or miss situation but does not sufficiently reduce the irradiation of the pressure vessel walls by high energy flux. Where the loading pattern was adjusted in order to specifically achieve adequate flux reductions, it created unacceptable component thermal margins and adversely affected loss of coolant analysis margins. Another ostensible solution to reduce pressure vessel neutron fluence is to replace certain of the peripheral fuel assemblies around the core with rods made, for example, of stainless steel. This plan would remove those fuel assemblies contributing to the irradiation of the pressure vessel and add a neutron absorber material between the core and the pressure vessel. Such a solution is not acceptable because of the corresponding reduction in core power rating and, therefore, lower the power output by the power plant. Another inappropriate solution would be to place additional shielding between the core periphery and the pressure vessel. This solution is unsatisfactory because of space restrictions, requires extensive mechanical redesign and neutronic evaluation of the region outboard of the core, is costly, and requires too much time to effectuate. Reducing the power output by the peripheral fuel assemblies to sufficiently reduce the fast flux level will cause an unacceptable power reduction in the peripheral fuel region in order to achieve a localized fast flux reduction. To maintain the same total core power output, the inboard region power must increase substantially. This creates unacceptable component thermal margins and adversely affects loss of coolant analysis margins. Hence, this is not a viable solution. Accordingly, a primary object of the present invention is to provide a method and apparatus to achieve relatively localized fast flux reductions at the core periphery in both the axial and circumferential directions. Another primary object of the present invention is to provide such localized fast flux reductions on a retrofittable basis; that is, to provide that the method and apparatus may be used with presently built and/or operating nuclear power reactors. Another primary object of the present invention is to achieve the fast flux reduction without materially adversely affecting core rating and/or power output and/or reactor shutdown margins and/or core thermal margins. Still another primary object of the present invention is to achieve the fast flux reduction without materially adversely affecting the fuel cycle costs. The above objects as well as others which are apparent from a reasonable reading and interpretation of this specification, and although they may not be specifically mentioned are all intended to be included within the scope of the invention provided herein. SUMMARY OF THE INVENTION The present invention accomplishes the above objectives and comprises a method and apparatus for forming a curtain at the core periphery to reduce the amount of fast neutrons emanating outward from the core and thereby prevent the pressure vessel from being exposed to such a fast neutron flux. Peripheral fuel assemblies are provided with specially designed displacer rods and/or neutron absorber and/or reflector rods so as to retard the production of fast neutrons and to reflect fast neutrons and to absorb fast neutrons which otherwise would escape from the nuclear core. Thus, a curtain of neutron reflecting and/or absorbing rods is provided at the core periphery. The strength of the absorber and the reflector rods may be varied radially and axially so as to provide a curtain specially adapted to function with particular reactors and/or pressure vessels. |
062326112 | claims | 1. A radiographic intensifying screen, comprising: a) a support; b) a fluorescent layer on the support; and c) a protective layer on the fluorescent layer, comprising: wherein the film-forming resin layer comprises a resin which is different from the resin of the organic macromolecule film. 2. The radiographic intensifying screen according to claim 1, wherein the film-forming resin layer contains a fluorocarbon resin. 3. The radiographic intensifying screen according to claim 1, wherein the organic macromolecule film comprises polyethylene terephthalate, polyethylene naphthalate or aramide. 4. The radiographic intensifying screen according to claim 1, wherein the organic macromolecule film has a thickness of from 1 to 10 .mu.m and the film-forming resin layer has a thickness of from 0.1 to 5 .mu.m. 5. The radiographic intensifying screen according to claim 1, wherein the protective layer has a thickness of from 2 to 10 .mu.m. 6. The radiographic intensifying screen according to claim 1, wherein the film-forming resin layer is formed by coating a solution containing the film-forming resin on the organic macromolecule film. 7. The radiographic screen of claim 1, wherein the organic macromolecule resin film is laminated to the fluorescent layer. 8. The radiographic screen of claim 1, wherein the film-forming resin is formed by applying a coating solution. |
summary | ||
039792552 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT While not limited thereto in its utility, the present invention is particularly well-suited for use in and as a safety system for a nuclear reactor. Accordingly, solely for purposes of explanation, the invention will be described in the environment of a nuclear reactor. Many systems in the nuclear reactor steam supply system require various important parameters to be monitored for safe operation of the system. A select few of these parameters are characterized by the seriousness of the consequences in the event that the parameter exhibit an excursion which violates a certain permissible margin. One system which relies on such a critical parameter is the emergency core coolant injection system. This safety system is designed to respond to an accident such as a main coolant line break which drains the reactor core, called a loss of coolant accident, by injecting emergency coolant into the core so that the core does not melt. A loss of coolant accident is detected by a primary coolant pressure drop which exceeds a predetermined value and which is indicative of a ruptured coolant pipe. U.S. Pat. No. 3,528,884 issued to A. R. Collier et al describes one such system. In that described system, the emergency coolant is injected into the core through a check valve which is activated by a differential pressure. That is, when the pressure of the primary coolant in the reactor pressure vessel is above the pressure of the tank which contains the injection water, the check valve prevents the injection water from entering the pressure vessel. When the pressure of the primary coolant in the reactor pressure vessel drops or undergoes an excursion to a pressure which is below the pressure in the tank containing the emergency coolant, the check valves automatically open and the emergency coolant is forced into the core. One deficiency with this system is that the system necessarily will not work when the reactor is being started or when the reactor is being shut down. During these operations the pressure of the primary coolant in the reactor pressure vessel is below the pressure in the tank containing the emergency coolant. Therefore, a control valve has to be closed which isolates the emergency core coolant and prevents the operation of the emergency core coolant injection system. The present invention enables the generation of a variable setpoint as opposed to a fixed setpoint described in the above U.S. patent so that the emergency core coolant injection system remains operative both during startup and shutdown of the reactor. The present invention also enables pressure maneuvering which allows the pressure to be dropped below the rigidly fixed setpoint characteristic of the prior art system. FIG. 1 illustrates three operating situations: (1) reactor startup from a primary pressure of zero psia to operating pressure of 2,250 psia; (2) reactor shutdown from 2,250 psia; and (3) normal operation of the reactor which occurs at approximately 2,250 psia. In that figure, Curve 1 is the normal operation curve, Curve 2 is the startup curve and Curve 3 is the shutdown curve. The broad dashed horizontal line at about 1600 psia illustrates the fixed setpoint of th prior art system. Therefore, if during operation of the reactor, with the prior art system the primary pressure were to drop from 2,250 psia down to or below 1600 psia, the setpoint would trigger the automatic emergency core injection system. This is not necessarily a desirable result unless the pressure drop was caused by the break of a primary coolant line. It can be seen from the curves in FIG. 1 that Curves 2 and 3 representing reactor startup and reactor shutdown involve pressures well below the fixed setpoint of 1600 psia. Therefore, in order to prevent automatic emergency core coolant injection with the prior art system during startup and shutdown, the injection system must be bypassed. The present invention allows the generation of a variable setpoint as illustrated in the drawing by the dotted lines. The variable setpoint on Curve 2 is a dotted line which is spaced approximately 500 psia below the operating pressure of Curve 2. The pressure setpoint automatically varies upwardly as the Curve 2 increases. The corresponding setpoint for Curve 3 is a horizontal line which starts out at approximately 1750 psia and takes a steplike shape as the Curve 3 decreases. The vertical portion of this setpoint curve cones into existence only when an operator or some other independent device gives permission to the apparatus to decrease the setpoint. When this permission is granted, the setpoint can only be decreased to a value of approximately 500 psia below the existing pressure. Therefore, lacking such permission, the setpoint would remain at 1750 psia and the dropping pressure of Curve 3 would initiate automatic emergency core coolant injection when the Curve 3 dropped below its setpoint. Curve 1 also has a corresponding setpoint curve which is the horizontally dotted line at approximately 1750 psia. From an examination of these curves in FIG. 1, it can be seen that the present invention enables a setpoint generation which automatically is allowed to increase but is not allowed to automatically decrease upon the operating pressure decrease. In this way the operating pressure can increase without triggering emergency core coolant injection but cannot decrease below the predetermined excursion margin without tripping protective action unless independent authorization is given. Such independent authorization will normally be given by the reactor operator after he has checked various gauges and has verified that the prssure drop is intended and has not been caused by a primary coolant leak. While it is contemplated that a human will normally make such a decision, it is conceivable that such authorization can be made by automatic means such as a time clock which automatically allows the periodic downward readjustment of the pressure setpoint: the period depending on the characteristics of the excursion being monitored. Another alternative would be a monitor-logic system which would give authorization only if a given number of criteria were met. The above description describes a situation where the excursion of importance is a downward excursion. The scope of this invention, however, is not so limited and also encompasses the situation where an upward excursion is the event which is to be detected. One example of an upward excursion is a reactor core power excursion which initiates what is called in reactor control terminology an "over power trip". Some reactor protective systems, such as the one described in U.S. Pat. No. 3,791,922 entitled Thermal Margin Protection System by Charles R. Musick, require the assurance that operation of the reactor during excursions greater than a predetermined amount be prevented in order to assure that initial assumptions made by the reactor designers in designing the protection system are met. In this case the reactor power excursion setpoint is allowed to decrease but is not allowed to increase without independent permission, thereby assuring that a reactor trip will occur if the power deviates from its last minimum by a value equal to or greater than the allowable excursion margin. Turning now to FIGS. 2 and 3, the apparatus necessary to accomplish the generation of the variable setpoint which is automatically allowed to change in one direction but not in the other will be described. The pressure of the primary coolant (the invention is equally as applicable to the pressure of the secondary coolant) is determined by some transmitter (not shown). One such pressure transmitter commercially available from Fischer and Porter Company is their model 5OEP1000. The pressure signal P is transmitted through wire 12 and 14 to an algebraic summer 16. Also supplied to summer 16 is a signal indicative of the predetermined excursion margin (P.sub.m) via wire 18. The excursion margin may be predetermined by the system designers by a knowledge of the operating characteristics and tolerances of the particular system. In a pressurized water nuclear reactor a simultaneous reactor-turbine trip results in a primary coolant pressure drop of approximately 450 psia. Therefore, in order not to incur emergency core coolant injection upon such an incident, the predetermined excursion margin (P.sub.m) must be at least greater than 450 psia. It should also be recognized that the excursion margin may be continually changed depending on operating conditions of the system or depending on some other parameter such as time. This excursion margin P.sub.m may be a simple input from a potentiometer and a power source (not shown). The summing element 16, well known in the electronics art, generates a signal commensurate with P-P.sub.m which is then delivered via wire 20, to a digital peak picker 22. The digital peak picker 22 is a well known and commercially available component such as manufactured by Hybrid Systems Corporation under the model No. 5648 or 750 and the particular construction does not form a part of the present invention. However, merely for illustrative purposes, the digital peak picker is depicted in FIG. 3 and will be described briefly hereinafter. Functionally, the peak picker 22 compares in comparator 25 the analogue signal of P-P.sub.m (24) to the peak pickers analogue output 26 and converts their difference into a digital signal 28. As long as a positive difference between P-P.sub.m (24) and the peak picker's output 26 exists, unit bits from oscillator 27 are fed to a counter 30 which counts and stores the total number of bits that it has received. The total number is continually retranslated by digital-analogue converter 31 into an analogue signal 26 which represents the last peak value of the incoming P-P.sub.m (24) signal. This peak value is continually supplied back to the original comparator 25 for comparison with P-P.sub.m (24). When P-P.sub.m (24) falls below the peak pickers output 26, comparator 25 discontinues its signal and bit stream 28 falls to zero so that the counter 30 no longer has an input and can no longer increase its counted and recorded number. Thus, the output signal 26 tracks and holds the highest or peak value of P-P.sub.m (24). Manual reset provision is made in the peak picker 22 by a reset signal 32 which wipes out the counter's (30) memory and returns all memory registers to zero. This in turn immediately drops the output signal 26 to zero and the whole system begins recounting until the output signal 26 has reached the new value of P-P.sub.m (24). The entire reset operation takes an approximate total time ranging from a few seconds to a few micro-seconds. Returning to FIG. 2, the output 26 of the peak picker 22 is symbolized by P.sub.sp and is called the pressure setpoint. This pressure setpoint P.sub.sp is delivered via wire 34 to a summer 36 which also receives an input of the original operating pressure P through wire 38. The summer subtracts the value of the operating pressure P from the pressure setpoint P.sub.sp and when this difference is positive, indicating that the operating pressure has dropped below the permitted maximum deviation, a digital trip signal is sent by bistable 40 to the appropriate reactor tripping and emergency core coolant injection systems. By means of the above described apparatus, a pressure setpoint which represents the maximum allowable deviation of primary coolant pressure from the operating pressure is continually generated. The pressure setpoint is allowed to track or to automatically increase to higher values corresponding with increases in operating pressure but is not allowed to automatically decrease with decreasing operating pressure. For the purposes of this description and appended claims, the term "tracking and holding" means automatically following a change in one direction of the tracked signal while holding that value during changes in the opposite direction. If the operating pressure does not increase, then the pressure setpoint will remain at a constant value until either manually reset or until the primary pressure passes through the setpoint which initiates reactor scram and emergency core coolant injection procedures. A slight variation, which should be obvious to a person ordinarily skilled in the electronics art, enables the above described peak picker to be applied to pick valleys rather than peaks. The slight modification required constitutes some signal preshaping which subtracts the parameter signal from a reference potential before the shaped signal is delivered to the peak picker. The peak picker operates on this inverted signal to pick the inverted valleys. A second step adds back in the reference potential to convert the inverted valleys back into true valleys. This final signal then constitutes the setpoint to which the operating parameter is compared for the generation of a trip signal when the actual parameter exceeds the setpoint. A second embodiment of the apparatus required to practice the present invention appears in FIG. 4. Like the apparatus described above and illustrated in FIG. 2, the embodiment may be applied to generate a continually varying setpoint for either positive or negative excursions. In contrast to the first embodiment, the apparatus illustrated in FIG. 4 tracks not the maximum value of the difference of the monitored parameter and the excursion margin, but it tracks the maximum value of only the monitored parameter and then subtracts the excursion margin to generate the setpoint. An operating parameter signal (pressure P) is first transmitted directly to the peak picker 22. The peak picker operates in a manner similar to that described above and tracks and holds the highest pressure P.sub.h experienced in the system. This highest value (P.sub.h) is transmitted to summer 17 which subtracts the excursion margin P.sub.m to generate a signal commensurate with P.sub. P.sub.m which is the variable setpoint for the system. This setpoint signal, P.sub.h - P.sub.m is then delivered to a second summer 36 which subtracts from P.sub.h - P.sub.m the current value of the incoming pressure signal P which has bypassed the peak picker 22 via wire 38. When this difference is positive, meaning the operating pressure has dropped below the permitted maximum deviation, a digital trip signal is sent by bistable 40 to the appropriate reactor tripping and emergency core coolant injection systems. In a manner similar to that discussed above for the first described embodiment, the circuitry in the second embodiment may also be altered by one skilled in the art so that the second embodiment tracks minimum values of the operating parameter rather than maximum values. In addition, the peak picker of the second disclosed embodiment also has a reset means which allows an independent decision maker to permit the setpoint to be lowered (or raised as the case may be) after it has been determined that independent justification exists. |
abstract | A wafer having a plurality of elements closely arranged thereon is irradiated with an ion beam while being conveyed in one direction by a conveying unit. Each of shutters adjusts an irradiation time during which a target area of the wafer is irradiated with the ion beam. Thus, a frequency in the target area is adjusted. Each of a plurality of mask holes in a pattern mask disposed between the wafer and the shutters corresponds to one area of the wafer. The mask holes are alternately displaced in a wafer conveying direction in which the wafer is conveyed, and are arranged in a plurality of columns perpendicular to the wafer conveying direction. To individually open and close the mask holes, the shutters are arranged to correspond to the respective mask holes. Thus, frequency adjustment, for areas in one column perpendicular to the wafer conveying direction, is performed in multiple steps. |
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047643382 | description | PREFERRED EMBODIMENTS OF THE INVENTION One embodiment of the present invention will be described below, referring to FIG. 3. Corrosion products brought into nuclear reactor pressure vessel 7 through feed water system 11 deposit on the surfaces of fuel rods and are radioactivated into radioactive corrosion products such as cobalt-60, etc. thereon. A portion of the radioactive corrosion products is again dissolved into core water to migrate through the core water, and deposit and accumulate on the machine member and piping in recycle system 13 such as recycle pump 2, etc. in nuclear reactor housing 12, causing an increase in the dose rate. The corrosion products on the surfaces of fuel rods are so-called cruds containing iron oxide as the major component, which mainly takes a chemical form of .alpha.-Fe.sub.2 O.sub.3, i.e. hematite. Minor components such as nickel, cobalt, etc. are adsorbed onto the hematite to form nickel ferrite, cobalt ferrite, etc. FIG. 4 shows dependency on pH of dissolution rate of cobalt from cobalt ferrite having an average particle size of 1 .mu.m (.mu.g/g.hr). When the pH is lowered, that is, shifted to the acidic side, it can be seen therefrom that the cobalt dissolution rate is abruptly increased. FIG. 5 shows cobalt-60 concentration in core water calculated on the basis on pH of dissolution rate of cobalt from cobalt ferrite shown in FIG. 4. The main source for cobalt-60 is the surfaces of fuel rods, as described above, and in spite of remarkable change in the amount of dissolved cobalt, the cobalt dissolution rate itself is too low to change the total amount of cobalt-60 retained on the fuel rods, and thus the total amount of cobalt-60 is kept substantially constant. The cobalt-60 concentration in core water is proportional to the amount of cobalt-60 dissolved from fuel rods, that is, to the amount of cobalt from cobalt ferrite. When the pH in core water is shifted from the neutral to the acidic side, that is, when it is less than pH 7, the cobalt-60 concentration in core water is abruptly increased, whereas, when the pH is shifted to the alkaline side, that is, when it is more than 7, the cobalt-60 concentration is considerably reduced, as shown in FIG. 5. In FIG. 3, the pH in core water is adjusted by adding an involatile alkali such as NaOH to the core water through alkaline chemical injection line 14, where the pH in core water depends on the amount of injected alkali. A portion of the injected alkali is removed in reactor-purifying unit 8, another portion thereof is carried over by steam to turbine 3 and removed from the core water, and further portion thereof is adsorbed onto the machine members and piping of the primary cooling system. Generally, the involatile alkali is removed mainly in the reactor-purifying unit among said three removing means. Let the amount of injected alkali be S (moles/hr), the alkali concentration in core water be C (moles/l), the density of core water be .gamma. (kg/l), the flow rate in the reactor-purifying unit be G.sub.C (kg/hr), the percent alkali removal in the reactor-purifying unit be .epsilon., the flow rate of main steam be Gs (kg/hr), the percent alkali carry-over by steam be .alpha., and the total deposition rate to the machine members and piping be .beta. (l/hr). The alkali concentration in core water can be obtained according to the following formula (1): ##EQU1## In the case of involatile alkali, for example, NaOH, EQU .epsilon./.gamma.G.sub.C >>.alpha./.gamma.G.sub.s, .beta. On the other hand, in the case of volatile alkali, for example, NH.sub.4 OH, EQU .alpha./.gamma.G.sub.s >>.epsilon./.gamma.G.sub.C, .beta. Generally, .epsilon.G.sub.C depends upon the desired degree of removing metallic impurities from core water in the plant. In the case of adding a volatile alkali, the alkali is carried over by the steam, and thus to keep pH in the core water, that is, the alkali concentration, constant, a larger amount of the alkali must be added than in the case of adding the involatile alkali. Thus, it is preferable to add an involatile alkali rather than a volatile alkali. The involatile alkali for this purpose includes alkali metal hydroxides such as NaOH and LiOH, alkaline earth metal hydroxides such as Ca(OH).sub.2, and organic alkali compounds. The organic alkali compounds are liable to disappear through radiolysis in the nuclear reactor, and the alkaline earth metal hydroxides are liable to form insoluble impurities, and are readily depositable mainly on fuel rods. On the other hand, the alkali metal hydroxides are stable at a high temperaure even under irradiation of radioactive rays, and are easiest to handle. Correlation between the amount of NaOH added as an alkali and pH is given below. The NaOH concentration C in core water and pH value H are given according to the following equation (2): EQU H=log {(C+10.sup.-7)10.sup.14 }=14+log (C+10.sup.-7) (2) In the standard type BWR (MWe), the flow rate G.sub.C through the reactor-purifying unit is about 100 tons/hr, and when the percent NaOH removal is presumed to be 100% in the reactor-purifying unit, correlation between the amount of injected NaOH and pH as given in FIG. 6 will be obtained. The carry-over of NaOH by the main steam is negligible in view of the NaOH material balance, but is not always preferable from the viewpoint of turbine side, particularly because Na is radioactivated in the reactor to form .sup.24 Na having a half-life of 15 hours. The carry-over rate is increased proportionally to an increasing NaOH concentration in the core water. Correlation between the .sup.24 Na concentration in condensed water and pH on the basis of the upper limit value of the carry-over rate of involatile component is shown in FIG. 7. When pH is higher than 8.5 in the case of adding NaOH, the radioactivity of condensed water reaches even 10.sup.-4 .mu.Ci/ml, and thus more than the necessary pH is not preferable for the control of condensed water radioactivity. Thus, it is important from the viewpoint of controlling an increase in the dose rate in the primary cooling system and maintenance of the entire system to keep the pH of core water at 7-8.5. It is possible to control dissolution of cobalt-60 from fuel rods without any increase in the radioactivity level of condensed water by keeping the pH of core water in said range, and consequently the cobalt-60 concentration in the core water and furthermore the surface dose rate in the machine members and piping of the primary cooling system can be reduced. Particularly preferable pH range is 7.5-8.0. There is a correlation between the .sup.24 Na concentrations in core water and pH in core water in the case of adding NaOH as given in FIG. 7. The correlation can be given according to the following equation (3) EQU A=2.times.10.sup.16 .delta..phi.(.lambda.Vc/Gc)C (3) where .phi.: average thermal neutron flux in nuclear reactor (n/cm.sup.2.sec.) PA1 .delta.: thermal neutron cross-section of .sup.23 Na (cm.sup.2) PA1 Vc: cooling water holdup in nuclear reactor (kg) PA1 .lambda.: decay constant of .sup.24 Na (hr.sup.-1) PA1 A: .sup.24 Na concentration in core water (.mu.Ci/ml) PA1 C.sub.Fe : concentration of Fe.sup.+2 as typical cation in condensed water (ppb) PA1 .epsilon.: probability of formed NaOH passable through cation exchange resin layer without trapping in the resin layer By monitoring .sup.24 Na concentration from the correlation between .sup.24 Na concentration and pH in core water given by equations (2) and (3), the core water pH can be determined. Another embodiment of adjusting pH of core water by adding an alkali is shown in FIG. 8, where cation exchange resin represented by R--(SO.sub.3 H).sub.2 is filled in condensed water desalter 6 as a filter, and a portion of the cation exchange resin is substituted with Na as a typical alkali metal. When the cation exchange resin undergoes ion exchange with Fe.sup.+2 as a typical cation, an alkali is released according to the following reactions: In the ordinary cation exchange resin, EQU R--(SO.sub.3 H).sub.2 +Fe(OH).sub.2 .fwdarw.R--(SO.sub.3).sub.2 Fe+2H.sub.2 O In the Na-substituted form, cation exchange resin, EQU R--(SO.sub.3 Na).sub.2 +Fe(OH).sub.2 .fwdarw.R--(SO.sub.3).sub.2 Fe+2NaOH If a mixing ratio of Na-substituted form cation resin to the total cation exchange resin is x, the concentration C.sub.Na of NaOH leaking from the outlet of condensed water desalter according to the present embodiment can be obtained according to the following equation (4): EQU C.sub.Na =4.times.10.sup.-8 .epsilon..multidot.xC.sub.Fe (4) where .epsilon. depends on the properties of cation exchange resin and is estimated to have a value of up to 0.1. When the cation concentration is constant, pH of core water can be controlled to any desired value by controlling the mixing ratio x of Na-substituted form cation exchange resin from the following correlation equations (5), (6) and (7): EQU S=Gf C.sub.Na (5) EQU C=S/.epsilon.Gc (6) EQU H=14+log (C+10.sup.-7) (7) where Gc, C, S and H are the same meanings as defined in the equations (1) and (2), and Gf is a feed water rate. Under typical BWR conditions, Gc/Gf is a value of up to 0.02, C.sub.Fe is a value of up to 1.0, and .epsilon. is a value of up to 0.1, where correlation between x and H is shown in FIG. 9. The desired pH of 7.0 to 8.5 can be continuously maintained by setting the mixing ratio x of Na-substituted form cation exchange resin to 0.1 to 0.5 according to equation (8) as will be given later without providing a special means for injecting an alkali. Against any fluctuation in the cation concentration C.sub.Fe in condensed water the pH can be kept in said desired range by changing the mixing ratio of the Na-substituted form cation exchange resin. Furthermore, the pH can be kept in said desired range by keeping the .sup.24 Na level in core water constant as shown above. A preferable embodiment for replacing a portion of cation exchange resin with the Na-substituted form cation exchange resin is shown below. Usually, a condensed water desalter uses a mixture of cation exchange resin and anion exchange resin, and regenerating treatment for recovering the ion exchanging capacity is carried out by separating the cation exchange resin and the anion exchange resin from each other by difference in specific gravity and chemically regenerating the cation exchange resin with H.sub.2 SO.sub.4 and the anion exchange resin with NaOH. FIG. 10 schematically shows the present embodiment. The mixture of ion exchange resins is transferred from condensed water desalter 21 to anion-cation separating column 22, where the anion exchange resin and the cation exchange resin are separated from each other by the difference in specific gravity. The separated anion exchange resin is led to anion exchange resin-regenerating column 23, while the cation exchange resin to cation exchange resin-regenerating column 24, where the former is regenerated with NaOH, and the latter with H.sub.2 SO.sub.4. The regeneration reactions proceed as follows: In the cation exchange resin-regenerating column, EQU R--(SO.sub.3).sub.2 Fe+H.sub.2 SO.sub.4 .revreaction.R--(SO.sub.3 H).sub.2 +FeSO.sub.2 The reaction usually proceeds from the right side to the left side, but the reversed proceding, that is, from the left side to the right side, is possible by using about 1N H.sub.2 SO.sub.4, whereby the cations trapped on the resin can be released as sulfate. Separation of anion exchange resin from cation exchange resin in the anion exchange resin-cation exchange resin separating column is carried out by the difference in specific gravity between the two resins, and the cation exchange resin having a larger specific gravity is withdrawn from the bottom of the separating column, whereas the anion exchange resin having a smaller specific gravity is withdrawn through nozzle 26 on separation level 25 at an intermediate height of the separating column. Separation level 25 differs from one plant to another and can be provided near the bottom or the top upon proper selection in view of the predetermined mixing ratio of anion exchange resin to cation exchange resin. In the present embodiment, the anion exchange resin and the cation exchange resin are mixed at an mixing ratio y' by volume in excess of the predetermined mixing ratio y of the anion exchange resin to the cation exchange resin. As a result, 100% cation exchange resin can be withdrawn from the bottom and a mixture of the anion exchange resin and the cation exchange resin from the separating level. In the anion exchange resin-regenerating column, regeneration is carried out with NaOH, where regeneration of cation exchange resin proceeds as follows: EQU R--(SO.sub.3).sub.2 Fe+NaOH.revreaction.R--(SO.sub.3 Na).sub.2 +Fe(OH).sub.2 EQU R--(SO.sub.3 H).sub.2 +NaOH.revreaction.R--(SO.sub.3 Na).sub.2 +2H.sub.2 O By returning the regenerated anion exchange resin and cation exchange resin to the condensed water desalter, the mixing ratio x of the Na-substituted form cation exchange resin in the condensed water desalter will be as follows: ##EQU2## To meet a fluctuation in the cation concentration in the condensed water or to meet the pH of core water to be predetermined, the mixing ratio x can be adjusted by replacing a portion of the resin mixture with anion exchange resin in the anion exchange resin-regenerating column before the regeneration, thereby reducing the ratio x, or by replacing it with cation exchange resin, thereby increasing the ratio x. By replacing the regenerating solution for the anion exchange resin-regenerating column with a solution of LiOH, or others, leakage of any desired alkali species is made possible. According to further embodiment of the present invention, a portion of cation exchange resin is replaced with Na-substituted form cation exchange resin in the same manner as shown in FIG. 10 at the regeneration of reactor-purifying desalter 8 in place of the condensed water desalter shown in FIG. 8. It is also possible to add NaOH to core water by mixing Na-substituted form cation exchange resin with the cation exchange resin for both condensed water desalter and reactor-purifying desalter. According to still further embodiment of the present invention, a portion of cation exchange resin as powdery resin used in the condensed water desalter or reactor-purifying desalter can be replaced with Na-substituted form cation exchange resin. The non-regenerative use of powdery resin is usual, and thus a portion of Na-substituted form cation exchange resin is made ready before precoating and can be used in mixture with the ordinary H-form cation exchange resin. According to the present invention, it is possible to suppress any increase in the concentration of radioactive corrosion products in core water in a direct cycle type, light water-cooled nuclear reactor without any substantial change in the plant hardware even if there are such disturbances as a resin leakage or lowering of pH in the core water. Particularly, the present invention can be readily applied to the existing plants without any substantial change in the plant hardware. This is a remarkable advantage of the present invention. |
claims | 1. A method for operating a nuclear light water reactor during an operation cycle including an initial control rod cycle and a plurality of subsequent control rod cycles following successively after the initial control rod cycle,said reactor comprising a core, said core comprisinga plurality of cells each including four elongated fuel units and a control rod position, said elongated fuel units being arranged in parallel with each other and each including a plurality of fuel rods containing a nuclear fuel in the form of fissile material,a plurality of control rods, each of said cells including a control rod position, each of said control rods being introduceable in a respective one of said control rod positions, and at least some of said fuel units including a control rod dependent addition of a burnable absorber,said method comprising:introducing substantially all of said control rods in the core before the reactor is started and an operation cycle is initiated,operating the reactor during said operating cycle without replacement or relocation of said fuel units, wherein the operation cycle is at least 18 months,operating the reactor during the initial control rod cycle of said operation cycle with a first control rod configuration with a first group of control rods at least partly introduced and the control rods other than those of the first group extracted,maintaining the control rods in position with substantially no movements thereof during the initial control rod cycle,operating the reactor during a first of the subsequent control rod cycles with a second control rod configuration with the first group of control rods extracted and a second group of control rods at least partly introduced; andoperating the reactor during a second of the subsequent control rod cycles with a third control rod configuration with the second group of control rods extracted and a third group of control rods at least partly introduced;wherein the initial control rod cycle is significantly longer than each of the subsequent control rod cycles. 2. A method according to claim 1, wherein the subsequent control rod cycles also include:operating the reactor during a third of the subsequent control rod cycles with a fourth control rod configuration with the third group of control rods extracted and a fourth group of control rods at least partly introduced. 3. A method according to claim 2, wherein the subsequent control rod cycles also include:operating the reactor during a fourth of the subsequent control rod cycles with a fifth control rod configuration with the fourth group of control rods extracted and a fifth group of control rods at least partly introduced. |
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summary | ||
059636106 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT FIG. 1 is a simplified representation of a conventional nuclear reactor 10 of the type with which the inventive data acquisition system can be used. As shown therein, reactor 10 has a reactor core 12 and a representative two 14, 16 of a multiplicity of control element assemblies (CEAs), each movable by respective control element drive mechanisms 18, 20 through the reactor core. The drive mechanisms such as 18 and 20 are powered by an electronic power supply utilizing silicon controlled rectifiers and have power supply cables leading to a CEDM control cabinet (not shown) located in a cable spreader room (not shown). Means, 22 and 24, such as reed switch position transmitters, are responsive to the movement of the CEA shaft, for generating analog position signals indicative of the CEA position. Each position signal is delivered to a safety control system 30 which, after processing this input signal along with a multiplicity of other signals indicative of plant operating parameters, can generate safety trip signals for delivery to each of the CEA drive mechanisms 18, 20, whereby the shaft of every CEA is released. Turning now to the invention, systems 40 and 40' of FIGS. 2a and 2b, respectively, are general schematic representations of the first and second preferred embodiments of the inventive CEDM data acquisition system shown in combination with CEDM control cabinets such as those of the conventional nuclear reactor 10 described above. The primary difference between the first and second preferred systems 40 and 40' resides in the ability of system 40 to acquire all of the data associated with eight individual CEAs simultaneously, whereas the system 40' is only capable of acquiring data for a single CEA at any given time. Accordingly, system 40' is a more streamlined version of system 40. As shown in FIG. 2b system 40' includes a signal conditioning unit 44' and an associated computer 46'. Signal conditioning unit 44' and computer 46' preferably communicate with one another via a conventional data transmission cable. Signal conditioning unit 44' receives conventional analog coil-current signals from a conventional CEDM control cabinet 42' via a cable. Additionally, signal conditioning unit 44' receives analog position signals from a conventional reed switch position transmitter (RSPT) via a RSPT cable. Thus, within system 40' the flow of information is generally first into signal conditioning unit 44' and subsequently into computer 46' where the data can be manipulated by the user as desired. As shown in FIG. 2b computer 46' is preferably a lap top computer with a PCMCIA card for digitizing conditioned analog signals presented thereto by signal conditioning unit 44'. While the use of a lap top naturally offers the convenience of portability, a desk top PC with an analog to digital (A/D) conversion board installed therein could also be utilized with system 40'. Finally, signal conditioning unit 44' preferably includes a noise suppression network consisting of differential amplifiers and various filters with high common mode rejection to suppress unwanted electrical noise and to prepare the conditioned analog signals for delivery to computer 46'. Since those of ordinary skill in the art will appreciate how to implement the system 40' of FIG. 2b based on the following description of the more elaborate system 40 shown in FIG. 2a, the remainder of this specification will be primarily directed to describing system 40 of FIG. 2a. As shown in FIG. 2a system 40 is a more elaborate embodiment of the inventive data acquisition system which is capable of simultaneously receiving data associated with eight CEAs and eight associated RSPTs. As with the embodiment of FIG. 2b, the flow of information is generally into signal conditioning unit 44 and subsequently into computer 46. As shown, signal conditioning unit 44 and computer 46 transfer information via a conventional data transmission cable. Additionally, conventional CEDM control cabinet 42 is connected to signal conditioning unit 44 with eight cables (one cable per CEA being monitored). Finally, CEA position data is transferred into conditioning unit 44 for up to eight RSPTs simultaneously by using eight cables. Computer 46 can be either a desk top PC or a lap top PC with a cooperating docking station. In either case, computer 46 preferably utilizes Keithly Metrabyte Inc.'s personal computer (PC) analog to digital (A/D) conversion boards in order to digitize the conditioned analog signals entering computer 46 at a rate of about 500 samples per second. Additionally, computer 46 preferably includes a monitor for displaying the digitized signals presenting various display images of the digitized data acquired. Finally, conditioning signal unit 44 includes a noise suppression network consisting of differential amplifiers and various filters with high common mode rejection to suppress unwanted electrical noise originating from the CEDM power supply and to prepare the analog signals for the A/D conversion boards of computer 46. The software utilized to implement the system 40 of FIG. 2a is illustrated on a general level in FIG. 3. The preferred programming language for the software of FIG. 3 is Microsoft Visual Basic Version 4.0. Visual Basic Custom Controls (product VTX-DAS from Keithly Metrabyte, Inc. (VBX)) is preferably utilized to implement all of the data acquisition, data handling and data storage features of the inventive system 40. Additionally, a simple linking program to the Visual Basic Custom Control Program from Scientific Software Tools, Inc. (LABOJX Real Time Chart) is preferably utilized to implement the various graphing functions discussed further below. The software is compatible with the Windows working environment. Naturally, those of ordinary skill will recognize that many other programming languages and software options could also be used to produce the inventive system 40 without departing therefrom. As shown in FIG. 3, the inventive data acquisition system provides the ability to monitor, record and playback newly acquired data. The selection of entering a monitoring mode occurs at block 50 in which case the software then proceeds to block 52 where the data acquisition, data storage in a buffer and graphical display of the data begins. Once data acquisition has begun in block 52 the inventive system, optionally, monitors for a "rod-drop" event at block 58. Once a "rod-drop" has occurred, the display image can automatically change to display a rod-drop measurement screen at block 58 and the data acquisition process terminates. At any time during the data acquisition stage, the user has the option to freeze the display screen to measure the coil-timing and/or coil current, in which case the procedure passes to block 54. Once the data acquisition procedure has begun, the user has the option to permanently record the subsequently acquired data and the procedure passes through the record data block of 53. Also as shown in FIG. 3, the software of the inventive data acquisition system also has the ability to replay previously recorded data for subsequent analysis. The play back procedure begins at block 56 where the user selects to replay previously recorded data. The process then proceeds to block 57 where data is retrieved from the permanent memory, stored in a buffer and passed for graphical display at the monitor. Naturally, each play back terminates at block 58 once a "rod-drop" has occurred and the display image automatically changes to the rod-drop measurement screen. During play back, the user also has the option to freeze the screen at any particular point in time to measure coil timing and/or coil current, in which case the procedure passes to block 54. The stored data can then be replayed any number of times desired by repeating the playback process described above. A more detailed description of the software performance options of the system is illustrated in FIG. 4. As shown therein, the software component of the inventive data acquisition system begins when the software program is launched. A "splash screen" is briefly displayed and then the software displays a main menu which offers the user the options of either replaying a previously recorded trace, acquiring data and creating a new trace, configuring the program or quitting the program. During either the data acquisition mode or the play back mode, the software offers the user a variety of options for displaying acquired and/or recorded data. As can be seen by joint reference to FIGS. 4 through 5c, one display option available to a user includes simultaneous display of coil-current traces for the five coils associated with each CEDM. As shown in FIG. 5a, coil-current traces 61a-61e are recorded/played back at a rate of 300 samples per second with the Y axis representing the coil-current in amps and the X axis representing time in seconds. Mouse-activated buttons 61a'-61e' allow the user to select any one of the coil-current traces thereby freezing the screen and permitting the use of interactive cursors 63a-64b to measure coil-current and/or timing changes. Real time monitoring can be resumed by selecting the mouse-activated continue button 69i. The resulting coil-current and/or timing changes for the selected coil-current trace are displayed in data boxes 65a and 65b, respectively, for simplified and accurate data analysis. The display shown in FIGS. 5a-5c are preferably updated every second as traces 61a-61e progress leftwardly and the user has the ability to view the displayed data in either of the time-expanded (0.5 second, 2 second or 5 second scales) views by selecting mouse-activated buttons 69c. When it is desired that data for a single CEA be recorded/played back, the mouse-activated record rod button 69d can be selected. However, where it is desired that CEDM data for an entire rod-group (up to eight CEAs) be recorded/played back, the user selects mouse-activated, record-group button 69e. In such a case, buttons 68, representing each of up to eight CEDMs appear at the top of the display and become activated. Display 60 then shows coil-current traces 69a-69e for the coils of the selected CEDM and the selected CEDM number appears in box 65c for convenience. Naturally, any one of the other seven CEDMs can be monitored simply by selecting the desired mouse-activated CEDM button 68. As noted in FIG. 4 and shown in FIGS. 5a-5c, a user has the option of directing the inventive data acquisition system to monitor for a rod-drop event and to automatically change the display image upon occurrence of the rod-drop event. This feature is implemented on the display screen 60' of FIG. 5b by selecting the mouse-activated rod-drop button 69a during recordation. With rod-drop button 69a, thus, selected the reed switch position (CEA position) signal 62 will be displayed on screen 60' and the data acquisition system will preferably monitor the upper gripper coil-current trace 61b. Before the upper gripper CEDM coil-current has stabilized to a predetermined holding value (see FIG. 5a) no rod-drop event can occur. However, once coil-current trace 61b stabilizes (see FIG. 5b), trace 61b is monitored to determine whether the coil-current is either above or below the predetermined "holding current" value. If trace 61b falls too far below the "holding current" value (see FIG. 5b) a rod-drop event has occurred and display 60" will automatically change to display a rod-drop measurement screen 60'. The triggering event is depicted in display 60' of FIG. 5b and the resulting rod-drop measurement screen in 60" is depicted in FIG. 5c. As shown in FIG. 5c the rod-drop measurement screen 60" only displays the upper gripper coil-current trace 61b and the position (reed switch position) trace 62 as a function of time, the traces 61b and 62 being indexed on the initiation of the rod-drop event. Also as shown in FIG. 5c an "acceptance" trace 67 will be superimposed on the rod-drop measurement screen that will aid the user in determining whether the upper gripper coil performance is within proper specifications during the rod-drop event. Naturally, the rod-drop event data can be saved (see box 66' of display 60") and the user can return to the normal monitoring display 60 of FIG. 5a. As noted above, while an operational test is being monitored in real time, the user has the option of recording one CEDMs data, all eight CEDMs data and/or the rod-drop event data. Further, the user has the option of printing display data by selecting the mouse-activated print button 69f. Moreover, operational test data can be replayed any number of times by selecting the mouse-activated restart button 69h. Finally, a user may choose to quit the program at any time by selecting the mouse-activated quit button 69j. Turning now to the signal conditioning aspect of the system, the block diagram of one signal conditioning unit 70 of the inventive system is shown in FIG. 6. Signaling conditioning unit 70 receives unprocessed analog coil-current and/or reed switch position signals at the input 71 thereof. The signals then pass to the differential amplifier 72 and low pass filter 74 where elimination of extraneous noise introduced into the signals by the electronic circuit powering the CEDMs is removed. Preferably, the differential amplifier 72 is a single integrated circuit which is preferably an isolation amplifier with supporting circuitry designed to provide high common mode rejection at the lowest of the frequencies of the extraneous noise (about 300 Hz and above) to be removed. Low pass filter 74 preferably has a cut-off frequency of about 300 Hz so as to further reduce extraneous noise from signals passing therethrough. Use of the isolation amplifier as shown offers the highly desirable feature that voltage spikes, or other erroneous electrical signals, which may occur downstream in the system, are not fed back to the control element drive mechanisms. After being, thus, conditioned the analog signals pass through an output 75 to an appropriate computer such as computers 46 or 46' depicted in FIGS. 2a and 2b, respectively. A schematic representation of the preferred signal conditioning unit 70 is shown in greater detail in FIG. 7. As shown, integrated circuit 73 is preferably an isolation amplifier having supporting circuitry 77 selected to provide high common mode rejection at or about the frequency of the alternating current supplied to the CEDMs. Additionally, signal conditioning unit 70 provides a low pass filter which has a cut off frequency of about 300 Hz. This is lower than the frequencies of noise contained in the DC current supplied to the CEDMs, but higher than an AC ripple component of the DC power. The preferred integrated circuit 73 can be purchased from Burr-Brown as Model No. ISO 165 and is preferred for its low power and high electrical isolation characteristics. Additionally, the Burr-Brown ISO 165 provides for signal gain. The Burr-Brown ISO 103 (which is depicted in FIG. 7) is also an acceptable alternative isolation amplifier. However, it is less desirable than the ISO 165 because it is more expensive, consumes more power, generates more heat and does not offer signal gain. Also as shown in FIG. 6, the preferred supporting circuitry 77 includes a 20 ohm resistor 76 connected across the input terminals of integrated circuit 73. Resistor 76 provides the very advantageous feature of allowing the data acquisition system to be calibrated (to remove an offset component of the signals exiting the signal conditioning unit 44) without the physical manipulation of any components of the system. Instead, the system can be calibrated, without signal conditioning unit 44 (FIG. 1) being connected to the CEDM control cabinet 42, simply be selecting the appropriate option in the program main menu of the accompanying software. Resistor 76 does not otherwise effect conditioning of the signals entering input 71. A more extensive schematic of a complete circuit board containing, inter alia, a plurality of signal conditioning units 70 is shown in FIG. 8. As indicated by the use of comparable reference numerals, each signal conditioning unit 70 depicted in FIG. 8 should be understood to include the supporting circuitry depicted in FIG. 7. The circuit board 80 of FIG. 8 includes signal conditioning units for conditioning coil-current signals and reed position switch (i.e., position) signal for two CEDMs. Accordingly, the complete multi-rod data acquisition system 40 of FIG. 2a should be understood to include four nearly identical circuit boards 80. As shown in FIG. 8, five signal conditioning units 70 are allocated for the five coils of each CEDM and these signal conditioning units receive input signals via input connector 71'. An additional signal conditioning unit 70' is allocated to condition the reed switch position signals which enter unit 70' via input connector 82. It will be appreciated that removal of an offset component of the position signal exiting the signal conditioning unit 44 is not necessary. Thus, no calibration resistor 76 is included in signal conditioning unit 70'. However, an amplitude reducing resistor 83 is connected to the output of signal conditioning unit 70' in order to ensure that the signals exiting unit 70' are compatible with the digital to analog converters of the computer into which signals will be entering. It will be appreciated that the lower lefthand portion of FIG. 8 is a repetition of the upper lefthand portion of FIG. 8 and is dedicated to signal conditioning the signals acquired from another CEDM. Master circuit board 80 of FIG. 8 further includes means for receiving electric power 86 to operate signal conditioning units 70 and 70'. Also, an output connector 85' is utilized to connect a master oscillator 84 and a 5 volt regulator 87 of master circuit board 80 with the three other signal conditioning circuit boards. Similarly, an output signal connector 75' of each of the four circuit boards is in electrical communication with a downstream computer. The single master oscillator 84 generates identical synchronization signals 79 to operate all of the isolation amplifiers of system 40 in unison. While the present invention has been described in connection with what is presently considered to be the most practical and preferred embodiments, it is to be understood that the invention is not limited to the disclosed embodiment, but is intended to cover the various modifications and equivalent arrangements included within the spirit and scope of the appended claims. |
abstract | The strip is of the type comprising a wall portion for delimiting a cell for receiving a fuel rod and allowing flow of a coolant upwardly through the spacer grid, a spring provided on the wall portion for biasing a fuel rod extending through the cell away from the wall portion, the spring being cut out in the strip and delimited by a slot and a motion limiter formed in the strip on the wall portion to limit motion of a fuel rod received in the cell towards the wall portion against action of the spring. According to one aspect of the invention, the motion limiter is located on an edge of the slot opposite the spring and defines a risen portion on the edge. |
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claims | 1. A method for fabricating a coated zirconium alloy article, the method comprising steps of:pre-treating a surface of a zirconium alloy parent material (step 1);providing a pure metallic material of silicon (Si) or chromium (Cr) in powder form (step 2);maintaining a distance between a plasma gun and the parent material at 10 cm and plasma spraying only the pure metallic material of Si or Cr on the surface of the pre-treated parent material of step 1 with the plasma gun (step 3) to form a coating layer on the pre-treated parent material; andthermally-treating the coated parent material coated in step 3 in an inert atmosphere (step 4) to create a diffusion layer comprising zirconium and the pure metallic material of Si or Cr between the parent material and the coating layer,wherein the pure metallic material of Si or Cr is coated on the parent material to a thickness ranging between 20 μm and 500 μm to form the coating layer, and the thermal treatment of step 4 is performed at 350° C. for 4 hours. 2. The method as set forth in claim 1, wherein the pre-treating at step 1 comprises grinding the surface of the parent material using particles of an oxide, a intermetallic compound or a silicon compound. 3. The method as set forth in claim 1, wherein the article is one of a cladding tube, a guide tube, an instrumentation tube and a spacer grid of a nuclear reactor. 4. The method as set forth in claim 1, wherein the zirconium alloy parent material has a resistance against oxidation by the pure metallic material of silicon (Si) or chromium (Cr) being oxidized to form silicon dioxide (SiO2) or chromium oxide (Cr2O3). 5. The method as set forth in claim 1, wherein the pure metallic material of Si or Cr is deformable plastically, thereby restricting cracking or scraping off of the coating layer and improving bonding with the parent material. 6. A method for fabricating a coated zirconium alloy article, the method comprising steps of:pre-treating a surface of a zirconium alloy parent material (step 1);providing a pure metallic material of silicon (Si) or chromium (Cr) in powder form (step 2);plasma spraying only the pure metallic material of Si or Cr on the surface of the pre-treated parent material of step 1 (step 3) to form a coating layer on the pre-treated parent material; andthermally-treating the coated parent material coated in step 3 in an inert atmosphere (step 4) to create a diffusion layer comprising zirconium and the pure metallic material of Si or Cr between the parent material and the coating layer,wherein the pure metallic material is coated on the parent material to a thickness ranging between 20 and 500 μm to form the coating layer, andthe thermal treatment of step 4 is performed at 350° C. for 4 hours. |
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059784300 | abstract | The length gauge includes an elongated standards rod, an indicator location block carrying a gauge and an indicator set block. The set block is applied to the location block and the gauge is zeroed out at a known distance between the movable element on the gauge and a reference surface. Water rods are assembled to a tie plate and the gauge is used to measure the length of the adjustable length water rods. The location block receives in a recess a portion of the tie bar of the water rod. The standards rod engages the tie plate and the gauge pin engages the opposite end of the standards rod. Any deviation in the length of the water rod from a designed length is indicated on the gauge. |
053368940 | summary | BACKGROUND OF THE INVENTION The present invention relates generally to a universal infrared heat source controller, and more particularly to a controller for a single infrared source capable of being programmed to act as a target for a missile target seeker. A previous test system in use requires two controllers and two heat sources. Each controller must use its own heat source. It is necessary to switch from the operation of one controller to the other. The switching procedure involves: 1. Removing the heat source that is in the collimating tube and placing it on the provided holding shelf. 2. Unplugging the interface cable from the one controller and plugging it into the one to be used. 3. Performing the recommended calibration after switching controllers. The disadvantage to the system is that it is time consuming and also causes wear on the cabling. The disadvantages to the controllers are: 1. The heat sources are not provided with adequate protection. Heat sources burn up on a yearly basis. 2. Lack of support for repair and calibration of the controllers from the manufacturer. 3. Some of the circuitry in the controllers is unreliable. Repeatability of the controllers is not as good as it needs to be and they need frequent repairs. United States patents of interest include U.S. Pat. No. 4,480,372, to Wirick et al, which describes a target for calibrating and testing infrared devices. Barnett et al, in U.S. Pat. No. 3,960,000, supply energy to a heat source 13 used to test a missile 11. The heat source is operated by a controller 14 connected to a programmer 15. Heat energy impinging on IR cell 17 actuates a detector 18. In U.S. Pat. No. 4,482,252, Lorenz discusses a calibration method and apparatus for optical scanners used to scan cloudscapes or landscapes from above the earth in an aircraft or spacecraft. Operation in the infrared area of the optical spectrum is described. U.S. Pat. No. 4,621,265 to Buse et al describes a simulator array and method for evaluating the tracking capability of a passive target seeker. In U.S. Pat. No. 4,737,792 to Grone, a counter-based simulated target generator is used to generate signals for testing a radar system. SUMMARY OF THE INVENTION An objective of the invention is to correct a heat source problem and to eliminate controller problems. The invention relates to a controller for a single infrared source capable of being programmed to act as a target for an AIM-9target seeker GCS (Guidance and Control Section). It is constructed for operation as part of an automatic test equipment system for testing a guidance and control section of missiles of different types, and can be operated manually or remotely. In either mode the controller is capable of selecting any temperature, aperture and shutter-filter combination. Four different kinds of data are read by the controller, namely (a) temperature data, (b) aperture data, (c) shutter-filter data, and (d) missile ID data. After data is read and decoded by a microprocessor, action is taken on each kind of data by separate circuitry. The unit contains internal protection circuitry for the most critical components. Advantages of the invention are: 1. A single heat source is used to do what the previous system does with two heat sources. 2. The controller is programmable and can accommodate future needs. A new feature relates to protection circuitry for the heat source. The heat source is protected from being over heated by component failure or miscalibration. The controller includes a CPU using a microprocessor, with a program stored in a programmable read only memory. The heat source includes an aperture wheel having a plurality of apertures rotated by a motor. A potentiometer is mechanically connected to the motor and aperature wheel to indicate its position. The controller reads aperture data from the stand, compares the data to digital signals derived from the potentiometer by analog to digital conversion, operates the motor to select an aperture as designated by the data, and then sends an aperature ready signal to the stand. The controller reads missile identification data from the stand. Operation is controlled depending on missile type (AIM-9P, AIM-9L, or AIM-9M). For temperature control, the black body is part of a voltage divider which is one side of a resistance bridge circuit. A voltage divider forming the other side of the bridge provides a reference voltage. A plurality of MOSFETs are used to select a value of resistance to determine the value of the reference voltage. A transistor circuit between a 24-volt power supply and the top of the bridge controls the power to the black body. An instrumentation amplifier is connected across the diagonal of the bridge to compare the reference voltage to a voltage determined by the resistance of the black body, which is a function of its temperature. The CPU reads temperature data from the stand, converts it to a temperature code, and uses it to select the MOSFETs. The instrumentation amplifier output is coupled to the transistor circuit which controls power to the bridge, which thereby controls the temperature and resistance of the black body. When the black body is at the designated temperature, the differential voltage across the diagonal of the bridge and the inputs of the instrumentation amplifier is approximately zero. The CPU reads shutter-filter data from the stand, and uses the data to generate signals to control the solenoids for a shutter and two filters which are part of the IR heat source. The control of the solenoids depends on the missile type. The controller includes a temperature ready circuit coupled to a thermocoupe integrated into the black body for indicating to the CPU whether the black body is heating, cooling, or stabilized. A temperature-ready signal is sent to the stand when the temperature has stabilized. A black body protection circuit opens a solid state relay to disable the 24-volt power supply to the bridge when the thermocouple indicates a temperature of approximately 95 degrees C. The results of testing this circuit shows that the fuse for the 24-volt supply will blow or the black body temperature will be maintained at approximately 95 degrees. A light on the front panel of the controller will indicate the overheat condition. |
054024552 | claims | 1. An improved shielding composite comprising: (a) a fibrous mat layer having a first and second face, the mat layer comprising a mat having, a thickness of from 1.2 cm (0.5 in.) to 10 cm (4 in.) and comprising an interwoven matrix of metal fibers effective to provide a radioactive shielding effect and a concrete-based material encasing the fibers and permeating the matrix, and filling at least 50 percent by volume of the void spaces within the matrix; (b) a first concrete-based layer located proximate to the first face of the mat layer, and (c) a second concrete-based layer located proximate to the second face of the mat layer. (a) providing a mat having a thickness of from 1.2 cm (0.5 in.) to 10 cm (4 in.) comprising an interwoven fiber matrix of metal fibers, the mat having a first and a second face; and (b) pouring a fluid concrete-based mixture into and adjacent to the mat to encase the fibers in the concrete-based mixture and permeate the matrix, and fill at least 50 percent by volume of the void spaces within the matrix, and to provide a first concrete-based layer proximate to the first face of the mat, where the metal fiber matrix is effective to provide a radioactive shielding effect. (a) a waste container comprising a side wall defining an enclosed space for storing waste materials, said side wall comprising a composite material comprising (i) a fibrous mat layer having a first and second face, the mat layer comprising a mat having a thickness of from 1.2 cm (0.5 in.) to 10 cm (4 in.) and comprising an interwoven metal fiber matrix effective to provide a radioactive shielding effect and a concrete-based material encasing the fibers and permeating the matrix, and filling at least 50 percent by volume of the void spaces within the matrix; and (ii) a first concrete-based layer located proximate to the first face of the mat layer; and (b) a top wall and a bottom wall located proximate to the side wall, the top wall and bottom walls enclosing the enclosed space for storing the waste materials. 2. The composite of claim 1 wherein the first concrete-based layer further comprises at least one additive selected from the group consisting of barite, magnetite, taconite, depleted uranium, vitrified glass-like materials, and mixtures of these additives. 3. The composite of claim 2 further comprising an impermeable coating layer located adjacent to the first concrete layer. 4. The composite of claim 1 wherein the interwoven fibers have a thickness of from about 10 to about 100 .mu.m. 5. The composite of claim 2 wherein the second concrete-based layer further comprises at least one additive selected from the group consisting of barite, magnetite, taconite, depleted uranium, vitrified glass-like materials, and mixtures thereof. 6. The composite of claim 5 further comprising an impermeable coating layer located adjacent to the second concrete layer. 7. The composite of claim 1 wherein the mat layer has a thickness of from about 2.5 cm (1 in.) to about 5 cm (2 in.), the mat has a fiber volume of from about 1 volume percent to about 10 volume percent, the concrete-based material encasing the fibers and permeating the matrix fills at least 80 percent by volume of the void spaces within the matrix and contains only up to 10 weight percent of particulate material below about 500 .mu.m, and where the composite comprises shielding for radioactive waste materials. 8. A method of constructing a containment storage structure comprising: 9. The method of claim 8 wherein the concrete-based mixture comprises from about 15 to about 40 weight percent cement; from about 5 to about 15 weight percent water; and from about 0.5 to about 0.1 weight percent plasticizer; and from about 25 to about 75 weight percent shielding additives. 10. The method of claim 9 wherein the concrete-based mixture further comprises at least one shielding additive selected from the group consisting of barite, magnetite, taconite, depleted uranium, vitrified glass-like materials, and mixtures thereof. 11. The method of claim 9 wherein the concrete-based mixture comprises from about 0.5 to about 0.1 weight percent plasticizer. 12. The method of claim 9 further comprising placing an impermeable layer adjacent to the first concrete-based layer. 13. The method of claim 9, wherein the concrete-based mixture further comprises fibers having a thickness of from about 10 to about 100 .mu.m and contains only up to 10 weight percent of particulate material below about 500 .mu.m, the pouring step further comprises use of a vibration process effective to provide high permeation of the matrix, wherein the concrete-based mixture is also poured to provide a second concrete-based layer proximate to the second face of the mat, and an impermeable coating layer is placed on the exposed face of either concrete-based layer to prevent liquids from contacting the concrete-based layer, and wherein a liner is placed inside the structure adjacent to the inner layer of the structure. 14. The method of claim 9 wherein the mat is freestanding, and has a thickness of from about 2.5 cm (1 in.) to about 5 cm (2 in.). 15. The method of claim 9 wherein the interwoven fibers comprise from about 1 to about 10 volume percent of the mat. 16. The method of claim 9 wherein the concrete-based mixture permeates at least 90 volume percent of the mat. 17. A waste container for storage of hazardous, radioactive, or mixed waste materials, comprising: 18. The container of claim 17 containing waste and stored at a storage site. 19. The container of claim 18 wherein the first concrete-based layer further comprises at least one additive selected from the group consisting of barite, magnetite, taconite, depleted uranium, vitrified glass-like materials, and mixtures of these additives, where a second concrete-based layer is located proximate to the second face of the mat layer and where an impermeable coating layer is disposed on the exposed face of either concrete-based layer, to prevent liquids from contacting the concrete-based layer. 20. The container of claim 18 wherein the concrete-based mixture further comprises from about 0.5 to about 0.1 weight percent plasticizer. 21. The container of claim 18 wherein the concrete-based mixture further comprises from about 15 to about 40 weight percent cement; from about 5 to about 15 weight percent water; and from about 0.5 to about 0.1 weight percent plasticizer; and from about 25 to about 75 weight percent shielding additives. 22. The container of claim 18 wherein the concrete-based mixture further comprises individual metallic fibers. 23. The container of claim 22 wherein the fibers are made from recycled metal. 24. The container of claim 18 wherein the container side wall is cylindrical, square, or hexagonal. 25. The container of claim 18 wherein the mat layer has thickness of from about 2.5 cm (1 in.) to about 5 cm (2 in.), the mat has a fiber volume of from about 1 volume percent to about 10 volume percent, the concrete-based material encasing the fibers and permeating the matrix, fills at least 80 percent by volume of the void spaces within the matrix and contains only up to 10 weight percent of particulate material below about 500 .mu.m, and where the container provides shielding for radioactive waste materials. 26. The container of claim 18, having a liner placed inside the container adjacent to the inner layer of the container. |
description | Over the past several years, there has been a proliferation of sensor network deployments into our environments. Sensors in such networks capture data pertaining to weather, traffic, parking, security, real-time views, and many other data items and conditions. The sensor data portal model has been designed to expose the data for search and querying by a general audience. Sensor data web portals enable browsing of disparate collections of sensor networks. Such portals function as a rendezvous point for sensor network administrators to publish their data, and through which a large base of users may pose individual queries to the sensor network. Thus, the sensor data web portal provides a platform with which users can interact and visualize content being generated by autonomous sensor networks. However, sensor data web portals commonly experience time delays due to the continuously changing nature of sensor readings and unavailability of data. Therefore, there remains a need to improve the way sensor data is collected and presented to users via a web portal. This summary is provided to introduce simplified concepts relating to spatial searches, and these concepts are further described below in the detailed description. This summary is not intended to identify essential features of the claimed subject matter, nor is it intended for use in determining the scope of the claimed subject matter. Techniques for collecting and displaying sensor data captured by a spatially representative sample of sensors requested in a search query are described. The sensors are represented in an index structure (e.g., a data tree). In response to the search query, the index structure is leveraged to identify a subset of sensors that exhibits a similar spatial distribution to the original full set of sensors. Sensor data is then collected from the subset of sensors by probing the sensors, or retrieving recently cached data located by the index, and finally returned to satisfy the query. In this manner, the number of sensors to be probed is reduced, thereby reducing latency involved with polling a large number of sensors and making the search process more efficient. This disclosure is directed to techniques for conducting a spatial search for sensor data using a sensor data web portal. The portal provides users with continuous access to sensor readings obtained from various sensors spread throughout disparate sensor networks. Users specify certain spatial regions from which they wish to receive readings, and a summarization of the sensor readings satisfying these conditions is produced and returned, rather than values from individual sensors. The summarization may be performed in many ways, including by simple aggregates such as an average value over groups of sensors in the query region, a maximum value, a minimum value, or sum of the values. Generally, query processing in the sensor portal involves two high-level components: (1) a portal web service to facilitate query processing and (2) a backend database to maintain sensor metadata and provide core search functionality. In this architecture, data is not maintained in any persistent archive, but is instead collected in an on-demand fashion during query processing. First, a portal service translates any queries issued by clients into a sequence of declarative queries. The first query asks the backend database to determine a set of relevant sensors. The database processes the queries by inspecting the sensor locations it maintains. Upon receipt of a list of sensors, the web service first checks a cache to ascertain whether there is any recent and current sensor data that can be returned to satisfy the query. If fresh data in the cache is not sufficient to answer the query, the web service polls the sensors for updated readings and issues a second query to the database to insert the new readings into the cache maintained in the backend. The web service then queries the database for all relevant readings from its cache, presenting the results to the user. Moreover, the backend database may routinely poll sensors as data in the caches become stale. This polling may occur responsive to a search query or as part of a separate process. The polling is thus conducted in a manner transparent to the user, as the user merely submits a query and receives results from the web portal. One challenge for the data portal is to provide timely processing of user queries using a data collection mechanism that efficiently retrieves readings from the sensor networks, with large numbers of sensors and large volumes of queries present in the system. The techniques described herein address this challenge by providing a layered sampling algorithm that facilitates probing of random subsets from the complete list of sensors relevant to a query. Layered sampling leverages the index structure to produce a random subset of sensors that exhibits a similar spatial distribution to the original full set of sensors. By reducing the number of sensors to probe, the latency involved with polling a large number of sensors is reduced, thereby making the process more efficient. The techniques described herein may be used in many different operating environments and systems. Multiple and varied implementations are described below. An exemplary environment that is suitable for practicing various implementations is discussed in the following section. Exemplary systems and methodologies for obtaining the sensor data from various sensors through the sensor data web portal are described in the general context of computer-executable instructions (program modules) being executed by a computing device such as a personal computer. Program modules generally include routines, programs, objects, components, data structures, etc., that perform particular tasks or implement particular abstract data types. While the systems and methods are described in the foregoing contexts, acts and operations described hereinafter is implemented in hardware or other forms of computing platforms. Exemplary Environment FIG. 1 illustrates an exemplary architecture 100 in which a communication efficient spatial search sensor data web portal may be implemented. Architecture 100 is shown in a client-server environment where a server 102 receives queries from any number of client devices 104-1, 104-2, and 104-3 (collectively referred to as devices 104) over a network 106. The queries contain requests for sensor data from sensors in distributed sensor networks. Server 102 processes the queries by employing a sensor index maintained locally or on a remote storage 108 that is accessible over network 106. After processing, server 102 returns results to the client devices 104. Server 102 may be implemented in many ways including, for example, as a standalone general purpose computing device or mainframe, or as a cluster of servers (e.g., arranged in a server farm). Client devices 104 may be implemented in any number of ways including, for example, as general purpose computing devices, laptops, mobile computing devices, PDAs, communication devices, GPS-equipped devices, and/or so on. Network 106 may include, but is not limited to, a Local Area Network (LAN), a Wide Area Network (WAN), and a Metropolitan Area Network (MAN). Further, network 106 is representative of a wireless network, a wired network, or a combination thereof. Storage 108 may be implemented in any number of ways, including as a remote database server or as an accessible networked storage device, such as a RAID system or the like. Server 102 includes one or more processor(s) 110 coupled to a system memory 112. Processor(s) 110 may include, for example, microprocessors, microcomputers, microcontrollers, multi-core processors, and so forth. The processor(s) 110 are configured to fetch and execute computer-program instructions stored in system memory 112. System memory 112 includes computer-readable media in the form of volatile memory, such as Random Access Memory (RAM) and/or non-volatile memory, such as Read Only Memory (ROM) or flash RAM. Server 102 hosts a web portal 114 that facilitates user access to sensor data obtained from the various sensors. The web portal 114 provides a user interface (UI) that may be rendered on client devices 104 to support submission of user queries for sensor data and presentation of results. One example UI 116 is illustrated as a browser-rendered graphical UI titled “SensorMap” on client device 104-1. UI 116 allows users to specify spatial regions to identify a collection of sensors from which they wish to receive readings. The sensors may be of any type including, for example, temperature sensors, video cameras, humidity sensors, wind sensors, traffic sensors, parking sensors, security sensors, and so on. In FIG. 1, representative sensors include a temperature sensor 118-1, a camera 118-2, a video camera 118-3, and any other sensors 118-4 (collectively referred to as sensors 118). Upon receiving a query submitted by a client device 104, server 102 analyzes and translates the query into a sequence of declarative queries that may be used to identify a set of sensors that can contribute to the sensor data requested by the user. Server 102 implements a query processing module 120, which is stored in system memory 112 and executed by processor(s) 110. To aid in identifying the appropriate set of sensors to respond to the user's query, query processing module 120 discovers attributes of the set of sensors. The attributes may include, for example, sensor location in terms of latitude and longitude, type of sensor, schemas for sensor readings, and so on. In one implementation, query processing module 120 ascertains sensor attributes from a database 122, which is illustrated as being stored in remote storage device 108 (although it may be stored in other locations, such as within system memory 112). Database 122 stores an index of the sensors, and the index maintains metadata describing the attributes of the sensors. Server 102 has an indexing module 124 stored in system memory 112 and executable on processor(s) 110 to create such index structures. In one implementation, the index is structured as a data tree having plural nodes arranged in layers. More particularly, the tree may be configured as a Collection R-Tree (or “COLR-Tree”), which is a spatial index built according to the techniques described herein, but loosely based on a classic R-Tree structure. One example technique of creating a COLR-Tree structure is described below in more detail with reference to FIGS. 3 and 4. Once the set of sensors are identified, the appropriate sensors are polled and updated readings are inserted into the database. Caches may be maintained within the indexing tree to hold the data. The collected updated readings may then be retrieved from the database and presented to client devices 104 using the UI 116. To illustrate this architecture 100, consider the following example. Suppose a user at client device 104-1 wants to know the temperature at a specific geographic location. The user submits a query for the temperature at the target location using UI 116 rendered on client device 104-1. Server 102 receives the query and query processing module 120 examines the query to determine the latitude and longitude of the target location. Subsequent to examination, query processing module 120 reviews the data tree stored in database 122 to identify a set of sensors having latitudes and longitudes within the target location. Upon identifying the set of sensors from various sensor networks, query processing module 120 probes the set of sensors to obtain updated sensor readings, including temperature data. The updated sensor data is then formatted and presented to the user through UI 116. Although the sensors 118 are shown accessible by server 102, the sensors 118 may be accessible directly by remote storage device 108. This allows the sensor metadata to be stored directly in database 122 in response to probing the sensors by indexing module 124. The sensor metadata stored in remote storage device 108 is then indexed to form the data tree. FIG. 2 illustrates the sensor search user interface (UI) 116 in more detail. In this example, search UI 116 is configured in the environment of depicting geographic locations of sensors within a certain region. Although not illustrated in this rendering, the UI 116 allows the user to draw arbitrary polygons to specify a search region. Thus, the user can form a box or rectangle, for example, around a portion of the map displayed in UI 116 to designate an area of interest. Alternatively, the user may specify a search region by entering search keywords (e.g., “Seattle”, “King County”) or by entering phrases or sentences (e.g., “weather in Seattle”, or “What is the traffic on I-90 and I5 interchange?”). Once the user submits the query, server 102 collects the appropriate data and returns it to UI 116 for depiction in a results pane 202. In this illustration, a regional map of the cities of Seattle and Bellevue in Washington State is shown in results pane 202, as well as various freeways and highways within this region. It should be noted that the sensor search UI 116 may be implemented in other environments, and display results of searches in other formats, including non-graphical and non-geographical presentations. Next to the results pane 202 in search UI 116 is a control panel 204 that enables a user to refine the search or navigate throughout the results. In the illustrated implementation, control panel 204 has a sensor selection area 210, a view name entry area 212, and location viewing controls 214. Sensor selection area 210 provides various options that allow a user to define which sensor types to display in results pane 202, since there may be many types of sensors located throughout a region. Representative sensors shown in sensor selection area 210 include options for temperature sensors 220, video or web cameras 222, weather sensors 224, traffic sensors 226, parking sensors 228, and other generic sensors 230. Here, the user may check a corresponding box to have the sensor type displayed. The sensor types selected by the user are then represented by graphical icons on the map in results pane 202. View name entry area 212 enables the user to name the search results view. Upon entry of the name, the user may actuate a control (e.g., “Save View” button or return key) to save that name. Once a particular search is named, the user may quickly rerun future searches for the same sensor data by simply selecting the view name. This is useful, for example, for the user who might like to know traffic conditions on a daily basis for the morning or evening commutes. Location viewing controls 214 facilitates user navigation through various locations within the region in results pane 202. Viewing controls 214 also provide additional control features (i.e. panning and zooming) to view sensors in each of the locations. Thus, the user may be provided with a facility to identify sensor densities (i.e. number of sensors) and types of sensors at different locations, thereby enabling the user to make a decision for selecting sensors prior to posting a request for sensor data. Control panel 204 further allows users to refine the search. Notice, for example, that there may be a high density of sensors in certain key areas of the region, such as at freeway interchanges. A user may want to know the traffic status at a particular stretch of freeway within the region depicted in results pane 202. Control panel 204 can be used to navigate through the location to find the stretch of road, limit display of only traffic sensors 226, and then ascertain the traffic sensor density (i.e. number of traffic sensors present in the particular location). A facility may be provided to identify locations of the traffic sensors 226 (i.e. latitude and longitude of locations of traffic sensors) within the target freeway or a list may be provided to allow the user to choose specific sensors along this stretch of road. Upon identifying target sensors, a new query may be submitted and processed to collect data only from the specified traffic sensors 226. In one implementation, UI 116 enables the user to specify freshness constraints of the sensor data in terms of how stale (or how recent) the sensor data is to be acceptable. For example, the user may specify that that sensor data older than a particular time period (e.g., 10 minutes, 1 hour, 1 day, etc.) may not be acceptable. Thus, the user can obtain the latest sensor readings from all the locations. Exemplary Server Implementation FIG. 3 shows certain functional aspects of server system 102 in more detail. As noted earlier, server system 102 includes processor(s) 110 and system memory 112. Server system 102 further includes network interfaces 302 to provide connectivity to a wide variety of networks, such as network 106, and protocol types such as wire networks (e.g., LAN, cable, etc.) and wireless networks (e.g., WLAN, cellular, satellite, etc.). Input/output interfaces 304 provide data input and output capabilities for server system 102. Input/output interfaces 304 may include, for example, a mouse port, a keyboard port, etc. System memory 112 stores program modules 306 and program data 308. Program modules 306 include, for example, a web portal 114 and other application program modules 310 (e.g., an Operating System (OS) to provide a runtime environment, networked communications between multiple users, and so forth). Web portal 114 has several responsibilities including acting as a front end or proxy to the sensor networks, performing the data collection by communicating with the sensor networks, and initiating query processing by the backend database by performing query translation. Web portal 114 also presents query results back to the client interface, such as through UI 116. The presentation of results typically includes aggregating sensor readings, since the user is usually not interested in visualizing the results of every single sensor node, especially in regions with a high sensor density. In one implementation described below in more detail, the choice of sensors over which to aggregate is decided dynamically using a pixel-based clustering algorithm that groups sensors lying within a distance corresponding to an n×n block of pixels (where n is a system parameter, but is typically small such as 5×5). As noted earlier, system 102 may be employed for identifying, capturing, and indexing the metadata associated with external sensors 118 of various sensor networks as well as for subsequently providing and presenting desired sensor data specified by a user via a UI 116. As such, web portal 114 includes a query processing module 120 to handle user queries, an indexing module 124 to index the metadata into a structure, and a data acquisition module 312 to acquire data from the sensors. In the following discussion, components of the server system 102 used to identify and index the sensor metadata are described first, followed by an explanation of components involved in presenting desired sensor data specified by a user. Server system 102 collects and stores metadata received from external sensors 118 of various sensor networks. The metadata may be provided and registered by sensor network administrators. The metadata may include, for example, locations of external sensors 118 and schemas that define how sensor readings are formatted and how to retrieve that information. From the metadata, indexing module 124 creates an index structure, such as data tree(s) 314, and stores that data trees) 314 as program data 308 in system memory 112. In one implementation, indexing module 124 is configured to cluster nodes in data tree(s) 314 into groups of nodes such that the tree has a hierarchical structure with one or more layers. More specifically, indexing module 124 includes a clustering module 316 configured to cluster the metadata of sensors 118 into various groups of metadata based on various factors, such as sensor location and sensor type. Subsequent to creating the groups, clustering module 316 constructs layers of the groups of nodes to form data tree(s) 314. Each layer includes multiple nodes, and individual nodes store metadata of corresponding external sensors 118. This process may be performed on a layer by layer basis, where each layer is created successively beginning with the lowest layer. Within the hierarchical arrangement, the nodes may be said to include parent nodes associated with external sensors 118 and child nodes associated with locations of external sensors 118. Further, each node of data tree(s) 314 is provided with associated caches to store sensor readings collected from the respective external sensors 118. As an alternative, indexing module 124 may allocate a single cache for a layer of nodes in data tree(s) 314. FIG. 4 shows one example structure of data tree(s) 314 in more detail. Data tree(s) 314 has multiple layers of nodes 402-418. Each node is associated with at least one external sensor 118. As shown, data tree(s) 314 has a top layer with a parent node 402 corresponding to a main sensor in a sensor network. Parent node 402 is illustrated with two child nodes 404 and 406 that form a middle layer. These middle-layer child nodes 404 and 406 may each have one or more children, as represented by two sets of three child nodes in which nodes 408, 410, and 412 are children of node 404 and nodes 414, 416 and 418 are children of node 406. These last nodes 408-418 form leaf nodes of the data tree(s) 314 in that there are no further child nodes dependent from them. The leaf nodes 408-418 form a bottom layer of the tree. Each node in each layer of data tree(s) structure 314 has an associated cache to store sensor readings obtained from external sensors 118. During a process of collecting sensor readings, sensor readings from a plurality of external sensors 118 are stored in the caches associated with leaf nodes 408-418. Further, nodes 404 and 406 of the middle layer may store a processed version of the sensor readings of respective child nodes. For instance, node 404 stores a processed version of the sensor readings stored in child nodes 408, 410, and 412. The processed version may be an average of the sensor readings. Further, root node 402 at the upper layer may store data obtained by processing the sensor readings stored in child nodes 404 and 406 in the middle layer, such as averaging the sensor readings in nodes 404 and 406. In one implementation, the data tree(s) structure 314 is configured as a Collection R-Tree (or “COLR-Tree”), which provides a spatial index of sensors in the sensor networks. Each layer 0, 1, 2 in the data tree(s) 314 has an associated table, as represented by table 420 for layer 2 and table 422 for layer 1. The layer tables 420 and 422 have columns representing an identifier for a parent node, an identifier for a child node, and any metadata corresponding to that child node. Here, the metadata is spatial metadata indicating a bounding box in terms of (x, y) coordinates of the child node and the number of sensors below the child node. Notice that the bounding box of the root node 402 is roughly the combined size of the boxes for the leaf nodes 408-418 descendant from the root node 402. The bounding boxes represent geographical regions within which sensors associated with the nodes are positioned. Thus, the root node 402 contains sensor data that is an aggregate (e.g., average) of data obtained from multiple sensors in lower layers. There is one row in the layer tables for each child node. Layer tables 420 and 422 are associated with each other in that child identifiers in an upper level layer table are present as node identifiers in a lower level layer table. The data tree(s) 314 is traversed by obtaining following successive child identifiers through the layer tables 420 and 422. Each layer table has a corresponding cache table, as represented by cache tables 424 and 426. Each cache table stores cached sensor readings of nodes within the layer. Cache tables 424 and 426 contain a node ID, a value representing the (possibly aggregated) reading, a slot ID, and the number of descendant leaves in the subtree. The data tree(s) 314 (COLR-tree structure) may be constructed by indexing module 124 in a bottom-up fashion from the lower layer of nodes to the upper layer. In one approach, indexing module 124 clusters sensor metadata using a k-means clustering technique (i.e., an algorithm to cluster objects based on attributes into k partitions or clusters) based on the geographic proximity of the sensors. This process is repeated to construct the data tree(s) 314 one layer at a time, terminating at the construction of a single cluster which becomes the root of our index. In another implementation, clustering module 316 constructs multiple data tree(s) 314 simultaneously. Further, one or two sets of data tree(s) 314 may be linked to each by clustering module 316, assuming a relation between the sets of data tree(s) 314 may be identified. After data tree(s) 314 has been constructed and stored, system 102 is ready to accept queries and process those queries to return sensor data requested by the users. User input queries are received from client devices 104 over the network via network interfaces 302, or obtained from input/output interfaces 304, and are passed on to query processing module 120 for processing. Input queries 318 are shown stored as program data 308 in FIG. 3 and accessible during processing. Query processing module 120 examines input queries 318 to identify a target data tree that has a set of external sensors 118 that are likely to satisfy information specified in input queries 318 by the users. The information may include, for example, location of sensors 118, spatial properties of external sensors 118 (i.e. spatial skewed distribution of sensors in each sensor networks or angle of location of the sensors), number of sensor readings, and so forth. Once a target data tree is identified, query processing module 120 identifies a set of external sensors 118 represented by nodes within the data tree(s) 314. In one implementation, query processing module 120 employs a sampling module 320 to conduct a layered sampling process that leverages the index structure of the target data tree to produce a random subset of sensors with a similar spatial distribution to the original set of sensors. As a result, the sampling process reduces the number of external sensors 118 to be probed for sensor readings, and yet still provide an accurate response to user queries (i.e. input queries 318). Thus, the sampling process reduces the communication cost and end-to-end latency in providing the sensor data from the time of receipt of input queries 318 by system 102. Generally, layered sampling allows siblings in the data tree(s) 314 to independently choose samples from their descendants. A user specifies a sample target size (i.e., the number of sensors to be read). Starting at the root node, with the user-specified sample target size, the sampling module 320 first examines the cache tables at each layer to determine whether current data exists. Depending upon the data in the cache, the sampling module 320 decides how many additional samples are needed. Thereafter, the sampling module 320 traverses through the target data tree, descending along nodes relevant to the query and splitting the target size amongst the child nodes. Thus, each child node is asked to return a sample smaller than the original target size, so that consequently when the samples from each child node are combined, the original size goal is met. The sampling traversal terminates when a node is assigned a target size smaller than a value of one. More particularly, in the present implementation, sampling module 320 may assign a target size at a node as ‘r’ and define a target size at one of its child nodes as ‘ri’. The target size of the child nodes may be computed as: r i = r × w i ∑ i w i ( 1 ) where ‘wi’ is a number of sensors that are descendents of child node ‘i’ of the current node (parent node or root node of the data tree). Thus, the above equation (1) denotes that the sample target size is divided and allocated to each sensor present at the branches of the current node. Such allocation process results in the identification of a smaller sample of nodes associated with lesser number of sensors than a desired number of external sensors 118. This smaller sample of nodes is sent to data acquisition module 312 for probing the sensors to collect the sensor readings. The sensor readings collected are stored into caches of respective nodes and displayed to the user on UI 116 through display devices 324. For example, sampling module 320 may need to identify a set of prominent temperature sensors from a location having high density temperature sensors (i.e. higher number of temperature sensors). Sampling module 320 may review the target data tree and distribute the target size specified by the user amongst the nodes on a layer by layer fashion. Implementation of such a process results in the elimination of nodes having target sizes less than one. Thus, the sensor data may be collected from a set of sensors having fewer sensors, thereby reducing a time delay in collection and presentation to the user. In such a scenario, a user may prefer to obtain the sensor data at a shorter time interval. In one exemplary implementation, sampling module 320 may be configured to eliminate some child nodes that may not possess any sensor readings that can contribute to the sensor data. In such a case, sampling module 320 may allocate a larger fraction of the sample target size to each child node. Thus, the larger fraction of the target size assigned to the child nodes can be computed as: r i = r × w i ∑ i w i × [ overlap ( BB ( i ) , q ) ] ( 2 ) where ‘wi’ is a number of sensors that are descendents of child node ‘i’ of the current node, ‘overlap’ denotes an indicator function that represents a maximum number of nodes in the target data tree that may satisfy the query rather than the total number of sensors descending the current node. For example, sampling module 320 may split the target size of temperature sensors (i.e. for a specified temperature data in a location) unequally amongst the child nodes by assigning a larger target size and a smaller target size to higher weighted nodes and lower weighted nodes, respectively. Such a process may be implemented to further reduce the number of sample sensors by eliminating the sensors that may not overlap with the user's query (i.e. may not contribute to the user's query) thereby reducing the end-to-end latency. Thus, a smaller sample of nodes determined by the above reduction process is send to data acquisition module 312 for probing the sensors to collect the sensor readings. The sensor readings collected are stored into caches of respective nodes and displayed to the user through UI 116 on client devices 104 or on other display devices 324. In yet another implementation, child nodes possessing sensor readings in the cache, stored during any of the previous traversals through the target data tree by sampling module 320, may be eliminated to reduce the number of sample sensors. In such a scenario, sampling module 320 may deduct the child nodes having the sensor readings in the cache. Thus the target size of the child nodes ‘i’ can be computed as: r i = r × w i - c i ∑ i w i × [ overlap ( BB ( i ) , q ) ] ( 3 ) where ‘wi’ is a number of sensors that are descendents of child node ‘i’ of the current node, ‘overlap’ denotes an indicator function that represents a maximum number of nodes in the target data tree that may satisfy the query rather than the total number of sensors descending the current node and ‘ci’ denoting the sampling weight. The sampling weight corresponds to an aggregate of the values of number of sensors having the sensor readings pre-stored in the cache. For example, upon traversing through the target data tree to identify a set of temperature sensors, sampling module 320 may determine the nodes having the latest temperature readings stored in the cache. In such a case, nodes associated with the set of temperature sensors may be deducted from the number of nodes that are descendants of a child node of the any of parent nodes associated with the set of temperature sensors. Thus, a smaller sample of nodes determined by the above reduction process is send to data acquisition module 312 for probing the sensors to collect the sensor readings. The sensor readings collected are stored into caches of respective nodes and displayed to the user on UI 116 through display devices 324. In one exemplary implementation, sampling module 320 may determine the smaller sample of nodes by performing all the techniques simultaneously, as described above according to equation (1), (2) and (3) thereby reducing the end-to-end latency. Further, it may be noted that sampling module 320 can determine the smaller sample of nodes employing the techniques as described in equation (1), (2) and (3) in various combinations. Operation Exemplary processes for operating a sensor data web portal to conduct communication efficient spatial searches for sensor data are described in this section with additional reference to FIGS. 1-4. FIG. 5 provides an overview of performing communication efficient spatial searches, while FIGS. 6 and 7 offer more detailed implementations of creating an index of sensors and processing queries using the index. The exemplary processes may be described in the general context of computer executable instructions. Generally, computer executable instructions can include routines, programs, objects, components, data structures, procedures, modules, functions, and the like that perform particular functions or implement particular abstract data types. The processes may also be practiced in a distributed computing environment where functions are performed by remote processing devices that are linked through a communications network. In a distributed computing environment, computer executable instructions may be located in both local and remote computer storage media, including memory storage devices. FIG. 5 illustrates a general overall process 500 for performing communication efficient spatial searches for sensor data based on a user's query. The process 500 is illustrated as a collection of blocks in a logical flow graph, which represents a sequence of operations that can be implemented in hardware, software, or a combination thereof. In the context of software, the blocks represent computer instructions that, when executed by one or more processors, perform the recited operations. The order in which the process is described is not intended to be construed as a limitation, and any number of the described blocks can be combined in any order to implement the process, or an alternate process. Additionally, individual blocks may be deleted from the process without departing from the spirit and scope of the subject matter described herein. For discussion purposes, the process 500 is described with reference to environment shown in FIG. 1 and system 102 shown in FIG. 2. At 502, an index of geographically-distributed sensors is constructed and stored. In one context, the sensors capture data pertaining to weather, traffic, parking, security, real-time views, and so forth. In building this index, a sensor registration process may be conducted, where sensor network administrators provide metadata including sensor locations and schemas of sensor readings. The metadata may include, for example, sensor location, type of sensor, schema for sensor readings, and so forth. In one approach, index construction is initiated with the hierarchical clustering algorithm described above, which constructs one layer of the index at a time from the metadata. According to one implementation, the index is formed as a data tree with multiple layers of nodes, where individual nodes are associated with caches to store metadata of external sensors 118. One example data tree(s) 314 is shown in FIG. 4. The administrator may modify the index by eliminating certain sensors which may not be deemed useful, or alternatively may specify a selected number of sensors for any particular location. At 504, a query for sensor data is received. The query may be submitted by users via UI 116 on client devices 104 and routed to server 102 via network 106. Query processing module 120 may review the query to identify certain information, such as location of sensors, number of sensors (i.e. external sensors), and so forth. The query may further include the identity of certain external sensors 118 or indicate a physical boundary of a particular area in which the sensors exist. For example, if the user is interested in a location with a high density of traffic sensors, the user may input a geographical range within which traffic sensors may be sampled. At 506, a subset of the sensors from which to obtain data that would satisfy the query is identified using the index. In one implementation, the smallest possible number of nodes is determined. The number of sensors in the subset may be specified in the user query, or chosen by the system administrator as a controllable parameter. The sensors forming the subset are then selected randomly from the index to exhibit a spatial distribution similar to the complete set of sensors. At 508, sensor data is obtained from the subset of sensors. The sensor data may be retrieved from caches associated with the sensors, or alternatively data acquisition module 312 may poll the subset of sensors and collect updated sensor readings on demand. At 510, the sensor data is formatted and presented to the user in a manner responsive to the user query. The sample sensor data may be displayed through UI 116, as shown in FIG. 2. FIG. 6 illustrates an exemplary process 600 for creating a data tree in more detail. At 602, sensors to be included in an index are identified. This may be accomplished through a registration process, for example, where sensor network administrators register sensors with the web portal. The administrators may submit metadata on the sensors including sensor type, location, schemas, and so on. At 604, sensors to be represented in the index are clustered together to form groups. In one implementation, clustering module 316 runs a clustering algorithm to form the groups of sensors. The clustering may be based on one or more criteria, such as sensor type, location, and so forth. As shown in FIG. 4, for example, clustering module 316 creates a bottom layer (i.e., layer 0) of all possible sensors to be represented by the index. The clustering process may then be repeated to form successively higher layers up to top layer with a single root node. At 606, various groups of sensors that may be interlinked in some manner are identified. As one example, external sensors 118 located in smaller areas within a particular location may be aggregated to form a group. At 608, the various groups of sensors can be combined to form new larger groups of sensors. In one technique, a clustering algorithm employs a pixel-based clustering algorithm that enables grouping of sensors lying within a distance corresponding to an n×n block of pixels, where n denotes a system parameter (e.g., 5×5). Various groups of external sensors 118 are reviewed by clustering module 316 using the pixel-based clustering algorithm to determine whether distances between groups of external sensors 118 tallies with a distance of an n×n block of pixels. If found that the distances tallies, the groups of external sensors 118 can be associated with one another to form new groups of external sensors 118. At 610, one or more data trees are formed from the groups of sensors including newly combined groups. The data trees are created by clustering module 316 and stored in database. FIGS. 7A and 7B illustrate an exemplary process 700 for processing a query for sample sensor data. At 702, a query for sample sensor data is received from a user. The query may be received by query processing module 120, which identifies a set of constraints specified by the user. The constraints may include a number of external sensors 118 that to be probed to obtain sensor readings corresponding to the sample sensor data. At 704, a target size of how many external sensors should be probed to retrieve the desired sensor data is determined. The target size may be specified in the input query, or it may be a parameter specified by the system administrator. It may also be inferred or calculated based on information provided in the query. At 706, a determination is made whether the target size needs to be distributed among a group of multiple nodes of the data tree. If so (i.e., the “yes” path from block 706), the target size if divided into parts and nodes from a top to bottom layer are assigned a smaller target size at 708. If the query does not include any information regarding dividing target size (i.e., the “no” path from block 706), a determination is made whether nodes deficient of sensor data corresponding to the requested sensor data should be eliminated. The query processing module may review the query to identify whether the user has specified to eliminate such nodes. If specified (i.e., the “yes” path from block 710), such nodes are eliminated from the group of nodes of the data tree at 712. If the query does not specify the elimination such nodes (i.e., the “no” path from block 710), a determination is made whether any cached sensor data would satisfy the query at 714. If such cached data exists (i.e., the “yes” path from block 714), such nodes can be excluded from the group of nodes in the data tree which will be probed at 716. Conversely, if the query does not specify elimination of such nodes (i.e., the “no” path from block 714), the subset of nodes from which to provide the sensor data are identified and the corresponding sensors are probed at 718. The data may then be returned to the user, such as via UI 116. Conclusion Although embodiments of techniques of identifying and displaying a sample sensor data based on a user's query have been described in language specific to structural features and/or methods, it is to be understood that the subject of the appended claims is not necessarily limited to the specific features or methods described. Rather, the specific features and methods are disclosed as exemplary implementations techniques of identifying and displaying a sample sensor data based on a user's query. |
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053393406 | abstract | A baffle is provided between a relatively hot containment vessel and a relatively cold silo for enhancing air cooling performance. The baffle includes a perforate inner wall positionable outside the containment vessel to define an inner flow riser therebetween, and an imperforate outer wall positionable outside the inner wall to define an outer flow riser therebetween. Apertures in the inner wall allow thermal radiation to pass laterally therethrough to the outer wall, with cooling air flowing upwardly through the inner and outer risers for removing heat. |
062630385 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a nuclear reactor core arrangement. More particularly, this invention relates to a nuclear reactor core arrangement which is adapted to combust plutonium along with uranium fuels and which utilizes a plurality of fuel assemblies that include mixed-oxide (MOX) fuel rods. 2. Discussion of the Related Art The Department of Energy (DOE) has a large excess of plutonium resulting from the retirement of nuclear weapons and is considering options for its disposal. One option recommended by the National Academy of Sciences (NAS) for the disposal of the excess weapons-grade plutonium is conversion to spent fuel. In this approach, the excess weapons plutonium is converted to plutonium oxide (PuO.sub.2) and used in a mixed oxide (PuO.sub.2 --UO.sub.2) form without reprocessing as fuel for existing nuclear reactors. This results in a spent fuel form which is "proliferation resistant" and that meets the "spent fuel standard" which is recommended by the NAS and which is being used by the DOE. However, this mixed oxide (MOX) approach requires: 1) conservative, realistic core performance characteristics which are similar to those for current uranium core designs; 2) that the technique minimize licensing risks by avoiding any erosion of safety margins compared to those for currently licensed conventional uranium core designs; 3) that impacts on plant operation be minimized or totally avoided; and 4) that the energy extracted from the MOX fuel be maximized, thus providing the best economics. Accordingly, ground rules were established by the DOE in light of the above objective. Namely, it is required that: There is no mixing of MOX and burnable absorber in the same fuel rod. This allows manufacture of lead test assemblies in existing European MOX fuel fabrication facilities. PA1 The fuel and core designs are developed using existing fuel and core design methodologies. PA1 The equilibrium cycle core design characteristics using MOX matches current uranium oxide (UO.sub.2) reload core design characteristics as much as possible. PA1 The cycle length of the MOX core design is essentially the same as that of the UO.sub.2 core design. PA1 There is no significant (if any) plant modifications necessary. PA1 There is no significant impact on plant systems or operation. PA1 Plant parameters should remain within existing plant technical specifications to the greatest extent possible. Accordingly, there exists a need for a nuclear core arrangement which can used in existing facilities and which enables an acceptably high throughput of plutonium in the form of MOX, while remaining within the above constraints. SUMMARY OF THE INVENTION Accordingly, an object of the present invention is to provide a novel core design that allows the use of mixed oxide (MOX) fuel containing weapons-grade plutonium in an existing nuclear reactor. Another object of the present invention is to provide a novel reload core design that enables the disposal of a large quantity of weapons-grade material in existing nuclear reactors with no significant plant modifications or impact on plant systems or plant operation. Another object of the present invention is to provide novel core designs for pressurized water reactor which uses MOX fuel, which maximizes the loading and throughput of weapons-grade plutonium and which is capable of disposing of a predetermined amount of weapons-grade plutonium in a given number of years of plant operation. Another object of the present invention is to provide a novel reload MOX core design that has essentially the same cycle length and combustion characteristics as existing UO.sub.2 core designs. Another object of the present invention is to provide a novel core design that would ensure that fuel assemblies can be manufactured in existing MOX fuel fabrication facilities and meet the requirement that the MOX and a burnable absorber are not present in the same fuel rod. Another object of the present invention is to provide a novel reload MOX core design that will allow plant parameters to remain within the existing plant technical specifications to the greatest extent possible. In brief, in order to achieve the above objects and to use up the above mentioned stockpile of weapons-grade plutonium, the plutonium is converted into a mixed oxide (MOX) fuel form wherein it can be disposed in a plurality of different fuel assembly designs. Depending on the equilibrium cycle that is required, a predetermined number of one or more of the fuel assembly types is selected and arranged in the core of the reactor in accordance with a selected loading schedule. Each of the fuel assemblies is designed to produce different combustion characteristics whereby the appropriate selection and disposition in the core enables the resulting equilibrium cycle to closely resemble that which is produced using conventional urania fuel. The arrangement of the MOX fuel and burnable absorber rods within each of the fuel assemblies, in combination with a selective control of the amount of plutonium which is contained in each of the MOX rods, is used to tailor the combustion characteristics of the assembly. More specifically, a first aspect of the invention resides in a fuel assembly for use in a nuclear reactor comprising: a plurality of MOX fuel rods; and a plurality of burnable absorber rods; each of the MOX fuel rods and each of the burnable absorber rods being disposed at a predetermined location within the fuel rod assembly. A second aspect of the invention resides in an equilibrium cycle core arrangement of a nuclear reactor comprising: a plurality of fuel assembly types, each type comprising: a plurality of MOX fuel rods, and a plurality of burnable absorber rods, each of the MOX fuel rods and each of the absorber rods being disposed at a predetermined location within a rod matrix for that type; wherein each fuel assembly type has a different number of MOX fuel rods and burnable absorber rods, respectively. A further aspect of the invention resides in a nuclear reactor core comprising a first predetermined number of mixed oxide (MOX) fuel assemblies which are arranged in a predetermined pattern in the core; each of the fuel assemblies being selected from a plurality of different fuel assembly designs wherein the MOX fuel is arranged differently and which, when arranged in the predetermined pattern, combust to produce an equilibrium cycle which is essentially the same as an equilibrium cycle produced using fuel assemblies containing only urania fuel. Another aspect of the invention comes in a method of fueling a nuclear reactor comprising the steps of: loading a first group of fresh unburnt MOX fuel rod assemblies into a first set of predetermined positions in a core of the reactor, in accordance with a predetermined location schedule; loading a second group of MOX fuel rod assemblies which have been burned once, into a second set of predetermined positions which are selectively arranged in the core with respect to the first set of predetermined positions, in accordance with the predetermined location schedule; and loading a third group of MOX fuel rod assemblies which have been burned twice, into a third set of predetermined positions which are selectively arranged in the core with respect to the first and second set of predetermined positions, in accordance with the predetermined location schedule. An important feature of the above method comes in the step of selecting the first group of fuel rod assemblies so as to comprise one or more of a plurality of predetermined octantly symmetrical assembly designs which each contain different amounts of plutonium and/or wherein the plutonium is distributed between the fuel rods of the assembly in a manner wherein an equilibrium cycle for the core exhibits a predetermined relationship with a predetermined equilibrium cycle produced using urania fuel. Another important feature of the above method comes in the step of distributing the amount of plutonium which is contained in the fuel rods of each of the plurality of octantly symmetrical assembly designs in accordance with a plurality of predetermined distribution schedules. Yet another aspect of the invention resides in a nuclear reactor core comprising: a first group of fresh MOX fuel rod assemblies which are unburnt and which are loaded into a first set of predetermined positions in the core, in accordance with a predetermined location schedule; a second group of MOX fuel rod assemblies which have been burned once, and which are loaded into a second set of predetermined positions which are selectively arranged in the core with respect to the first set of predetermined positions, in accordance with the predetermined location schedule; and a third group of MOX fuel rods assemblies which have been burned twice, and which are loaded into a third set of predetermined positions which are selectively arranged in the core with respect to the first and second set of predetermined positions, in accordance with the predetermined location schedule. An important feature of the above structure comes in that the first group of fuel rod assemblies comprise one or more of a plurality of predetermined octantly symmetrical assembly designs which each contain different amounts of plutonium, and so that an equilibrium cycle for the core exhibits a predetermined relationship with a predetermined equilibrium cycle produced using urania fuel. A further important feature of the above structure comes in that the amount of plutonium, which is contained in the fuel rods of each of the plurality of octantly symmetrical assembly designs, is distributed in accordance with a respective plurality of predetermined distribution schedules. |
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. |
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claims | 1. A contamination barrier that passes through radiation from a radiation source and captures debris coming from the radiation source, said contamination barrier comprising:an inner ring;an outer ring; anda plurality of lamellas extending in a radial direction from a main axis, each of said lamellas being positioned in a respective plane that comprises said main axis,wherein at least one outer end of each of said lamellas is slidably connected to at least one of said inner and outer ring. 2. A contamination barrier according to claim 1, wherein said lamellas are thermally connected to at least one of said inner and outer ring. 3. A contamination barrier according to claim 1, further comprising a first shield that protects said inner ring from being hit by radiation from said radiation source. 4. A contamination barrier according to claim 3, further comprising a second shield that blocks thermal radiation from said first shield. 5. A contamination barrier according to claim 4, further comprising a third shield that reduces heating of the first shield caused by direct radiation from the radiation source, wherein said third shield is disposed upstream of said first shield with respect to the direction of propagation of the radiation emitted by the radiation source along the main axis. 6. A contamination barrier according to claim 5, wherein said third shield is substantially thermally isolated with respect to said first shield. 7. A contamination barrier according to claim 6, wherein said third shield is connected to said first shield. 8. A contamination barrier according to claim 3, further comprising at least one cooling spoke to support said first shield, said at least one cooling spoke being thermally connected to said outer ring. 9. A contamination barrier according to claim 8, wherein said first shield comprises a plurality of shield members, each shield member being connected to said outer ring via a separate cooling spoke. 10. A contamination barrier according to claim 4, further comprising a first cooling device arranged to cool at least one of said first and second shields. 11. A contamination barrier according to claim 10, further comprising a second cooling device arranged to cool said inner ring. 12. A contamination barrier according to claim 11, further comprising a third cooling device arranged to cool said outer ring. 13. A contamination barrier according to claim 1, wherein said lamellas are curved in said respective planes, and said inner and outer ring are shaped as slices of a conical pipe. 14. A contamination barrier according to claim 1, wherein a first side of said lamellas facing the radiation source is thicker than the rest of said lamellas. 15. A contamination barrier that passes through radiation from a radiation source and captures debris coming from the radiation source, said contamination barrier comprising:a plurality of lamellas; anda support structure that slidably engages said lamellas,wherein said lamellas and said support structure are configured and arranged to allow said lamellas to expand and contract in response to changes in temperature. 16. A contamination barrier according to claim 15, wherein said support structure comprises an inner ring and an outer ring and said plurality of lamellas are slidably connected to at least one of said inner and outer ring. 17. A contamination barrier that permits radiation to pass therethrough and captures debris from a radiation source generated by the radiation source, said contamination barrier including a support structure and a plurality of thin plate members mounted on said support structure, said radiation propagating along an optical axis and said thin plate members being disposed along a plane that includes said axis, said plate members being slidably movable relative to said support structure. 18. A radiation system comprising:a contamination barrier that passes through radiation from a radiation source and captures debris coming from the radiation source; anda collector that collects radiation passing said contamination barrier,wherein said contamination barrier comprises an inner ring, an outer ring, and a plurality of lamellas extending in a radial direction from a main axis, each of said lamellas being positioned in a respective plane that comprises said main axis, and at least one outer end of each of said lamellas is slidably connected to at least one of said inner and outer ring. 19. A lithographic projection apparatus comprising:a radiation system to provide a beam of radiation;a support structure to support a patterning structure to be irradiated by a beam of radiation to pattern said beam of radiation;a substrate support to support a substrate; anda projection system to image an irradiated portion of the patterning structure onto a target portion of the substrate,wherein said radiation system comprises a contamination barrier that passes through radiation from a radiation source and captures debris coming from the radiation source, said contamination barrier comprising an inner ring, an outer ring, and a plurality of lamellas extending in a radial direction from a main axis, each of said lamellas being positioned in a respective plane that comprises said main axis, and at least one outer end of each of said lamellas is slidably connected to at least one of said inner and outer ring; and a collector for collecting radiation passing said contamination barrier. 20. A method of manufacturing an integrated structure by a lithographic process, said method comprising:radiating a beam of radiation through a radiation system;providing a support structure to support a patterning structure to be irradiated by the beam of radiation to pattern said beam of radiation;providing a substrate support to support a substrate; andproviding a projection system to image an irradiated portion of the patterning structure onto a target portion of the substrate,wherein said radiating the beam of radiation through the radiation system comprises passing radiation from a radiation source through a contamination barrier comprising an inner ring, an outer ring, and a plurality of lamellas extending in a radial direction from a main axis, wherein each of said lamellas being positioned in a respective plane that comprises said main axis, and at least one outer end of each of said lamellas is slidably connected to at least one of said inner and outer ring; and collecting radiation passing said contamination barrier. 21. A method of manufacturing an integrated structure by a lithographic process, said method comprising:generating a beam of radiation with a radiation source;capturing debris from the radiation source;collecting radiation passing said contamination barrier;patterning said beam of radiation with a patterning structure; andimaging an irradiated portion of the patterning structure onto a target portion of a substrate,wherein said capturing debris comprises providing a support structure and a plurality of lamellas that are slidably engaged with the support structure so as to allow the plurality of lamellas to expand and contract in response to changes in temperature. |
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044951396 | description | SPECIFIC DESCRIPTION As seen in the drawing a vessel 1 of spherulitic cast iron has a cover 2 of the same material. The vessel has a mouth formed with a shoulder 23 lying in a plane perpendicular to the vessel center axis, a cylindrical intermediate surface extending up from its outer periphery, and another shoulder 24 parallel to the shoulder 23. Other than these formations, some bolt holes, and a groove 26, the vessel 1 is not machined much, but can be a raw casting. The cover 2 is basically formed of a flange part 4 and a plug part 6. The plug part 6 forms an inner shoulder 5 closely juxtaposed with and axially confronting the shoulder 23, and an intermediate cylindrical surface 7 complementary to the surface 27. The flange forms another shoulder surface 25 confronting and complemtnary to the surface 24. Bolts 21 extending through the outer regions of the flange 4 secure the cover 2 to the vessel 1. A safety cover 21 is secured by further bolts 22 to the rim of the vessel 1 in the groove 26 and serves principally to protect the cover 2 from physical harm. The surfaces 5 and 7 are formed with respective axially downwardly and radially outwardly open grooves receiving respective O-ring seals 10 and 9 that tightly engage the surfaces 23 and 27, and that form an annular compartment 15. The surface 24 of the flange 4 is formed with two concentric and radially spaced grooves that receive respective C-section seals 8a and 8b of an outer seal 8. The rings 8a and 8b together form an annular outer compartment 17, and the ring 8b forms with the ring 9 an intermediate compartment 16. In addition the cover 2 is formed with respective passages 11, 12, and 13 opening into the respective chambers 15, 16, and 17 and provided at their other ends with valves 14 of a monitoring means 3. The cover is formed with an axially upwardly open recess 19 in which the valve 14 of the intermediate chamber 16 opens, although normally it is covered by a cap 28. Another cover 29 closes this recess 19 for maximum protection, and the outer passage 13 of the outer chamber 17 opens directly into this recess 19, so the valve 14 in its cover 29 can be tapped to test for leaks. Normally the interior 18 of the vessel 1 is filled with a pressurized, easily detectable tracer gas above the radioactive material in it. If this gas is detected though the monitoring means 3 in any of the chambers 15-17, the container can be refitted. In any case, the outermost chamber 17 can be sampled easily by removing the safety cover 20, then pulling the cover 30 off the valve 14 in the cover 29 and connecting up to this valve 14. If no leak is detected one can be sure that the cover 29 can be removed to sample the chambers 15 and 16. This is an extremely safe procedure. Thus the container according to the instant invention can be made quite a bit more cheaply than the prior-art one, as all of the tricky machining is done on the relatively portable cover 2. In addition three chambers are provided in a row to test for leakage in the statutorily required failsafe manner, and all three of these chambers are formed by structure on the cover 2. These chambers can be individually sampled and/or charged at superatmospheric pressure. |
abstract | Provided are a silicotitanate molded body having high strength and reduced generation of fine powder, a production method thereof, an adsorbent comprising the silicotitanate molded body, and a decontamination method of radioactive cesium and/or radioactive strontium by using the adsorbent. The silicotitanate molded body comprises: crystalline silicotitanate particles that have a particle size distribution in which 90% or more, on volume basis, of the particles have a particle size within a range of 1 μm or more and 10 μm or less and that are represented by a general formula of A2Ti2O3(SiO4).nH2O wherein A represents one or two alkali metal elements selected from Na and K, and n represents a number of 0 to 2; and an oxide of one or more elements selected from the group consisting of aluminum, zirconium, iron, and cerium. |
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039740285 | abstract | A nuclear reactor having a flattened reactor activity curve across the reactor includes fuel extending over a lesser portion of the fuel channels in the central portion of the reactor than in the remainder of the reactor. |
claims | 1. A material for radiation detection, comprising a scintillator material comprising a halide ofa rare-earth metal; anda group-13 element,wherein the group-13 element forms covalent bonds with the halogen; the halide beingA′(1−x)B′xCa(1−y)EuyC′3,A′(1−x)B′xM′2Br7(1−y)C′7y,A′(1−x)B′xM″(1−y)EuyI3,A′3(1−x)B′3xM″(1−y)EuyI5,A′(1−x)B′xM″2(1−y)Eu2yI5,A′(1−x)B′xM′2Cl7,M′(1−x)B′xC′3, orany combination thereof,wherein:A′=Li, Na, K, Rb, Cs or any combination thereof,B′=B, Al, Ga, In, Tl or any combination thereof,C′=Cl, Br, I or any combination thereof,M′ consist of Ce, Sc, Y, La, Lu, Gd, Pr, Tb, Yb, Nd or any combination thereof,M″ consists of Sr, Ca, Ba or any combination of thereof,where 0<x<1, and where 0<y<1. 2. The material of claim 1, wherein the group-13 element comprises thallium (Tl). 3. The material of claim 2, made from a rare-earth metal halide comprising LaBr3, LaCl3, CeBr3, CeCl3 or LuI3 or a combination thereof, and a halide of a group-13 element in stoichiometric amounts. 4. The material of claim 3, made from a rare-earth metal halide comprises LaBr3 and a halide of a group-13 element in stoichiometric amounts, and cerium (Ce). 5. The material of claim 2, wherein the rare-earth metal comprises at least two rare-earth metal elements. 6. The material of claim 1, made from a rare-earth metal halide comprising LaBr3, LaCl3, CeBr3, CeCl3, LuI3 or a combination thereof, and a halide of a group-13 element in stoichiometric amounts. 7. The material of claim 1, wherein the rare-earth metal comprises at least two rare-earth metal elements. 8. The material of claim 1, wherein the halide defines a crystal lattice having a symmetry that is different from a symmetry of a crystal lattice defined by a halide of the rare-earth halide without the group-13 element. 9. The material of claim 1, wherein the halide is a stoichiometric halide of the formulaA′(1−x)B′xM′2Br7(1−y)C′7y,orA′(1−x)B′xM′2Cl7. 10. The material of claim 1, the scintillator material being a single crystal or polycrystal. 11. A radiation detector, comprising:a material of claim 1 adapted to generate photons in response to an impinging radiation; anda photon detector optically coupled to the scintillator material, arranged to receive the photons generated by the scintillator material and adapted to generate an electrical signal indicative of the photon generation. 12. An imaging method, comprising:using at least one radiation detector of claim 11 to receive radiation from a plurality of radiation sources distributed in an object to be imaged and generate a plurality of signals indicative of the received radiation; andbased on the plurality of signals, deriving a spatial distribution of an attribute of the object. 13. The material of claim 1, wherein the halide is a stoichiometric halide. 14. The material of claim 13, wherein the halide is single crystalline or polycrystalline. 15. A method of making a scintillation material, comprising:making a melt by heating a stoichiometric mixture of:a rare-earth metal halide, anda salt of a group-13 element; andgrowing a single crystal from the melt, wherein the rare-earth metal halide and salt of a group-13 element are present in the stoichiometric mixture in a ratio to produce a single crystal of:A′(1−x)B′xCa(1−y)EuyC′3,A′(1−x)B′xM′2Br7(1−y)C′7y,A′(1−x)B′xM″(1−y)EuyI3,A′3(1−x)B′3xM″(1−y)EuyI5,A′(1−x)B′xM″2(1−y)Eu2yI5,A′(1−x)B′xM′2Cl7, orany combination thereof,wherein:A′=Li, Na, K, Rb, Cs or any combination thereof,B′=B, Al, Ga, In, Tl or any combination thereof,C′=Cl, Br, I or any combination thereof,M′ consist of Ce, Sc, Y, La, Lu, Gd, Pr, Tb, Yb, Nd or any combination thereof,M″ consists of Sr, Ca, Ba or any combination of thereof,where 0<x<1, and where 0<y<1. 16. The material of claim 15, wherein the rare-earth metal halide and a salt of a group-13 element are present in the stoichiometric mixture in a ratio to produce a single crystal of:A′(1−x)B′xM′2Br7(1−y)C′7y,A′(1−x)B′xM′2Cl7, ora combination thereof. 17. A material for radiation detection, comprising a rare-earth metal halide scintillator compound co-doped with a group-13 element where the group-13 element forms covalent bonds with the halogen of the halide; and where the halide is:A′(1−x)B′xCa(1−y)EuyC′3,A′(1−x)B′xM′2Br7(1−y)C′7y,A′(1−x)B′xM″(1−y)EuyI3,A′3(1−x)B′3xM″(1−y)EuyI5,A′(1−x)B′xM″2(1−y)Eu2yI5,A′(1−x)B′xM′2Cl7,M′(1−x)B′xC′3, orany combination thereof,wherein:A′=Li, Na, K, Rb, Cs or any combination thereof,B′=B, Al, Ga, In, Tl or any combination thereof,C′=Cl, Br, I or any combination thereof,M′ consist of Ce, Sc, Y, La, Lu, Gd, Pr, Tb, Yb, Nd or any combination thereof,M″ consists of Sr, Ca, Ba or any combination of thereof,where 0<x<1, and where 0<y≤1. 18. The material of claim 17, wherein the group-13 element comprises Tl. 19. The material of claim 18, made from a rare-earth metal halide comprising LaBr3, LaCl3, CeBr3, CeCl3, LuI3 or a combination thereof, and a halide of a group-13 element in stoichiometric amounts. 20. The material of claim 17, wherein the rare-earth metal halide scintillator material comprises at least two rare-earth metal elements. |
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060977876 | abstract | The present invention relates to a fast and accurate method for calculating fluence of a calculation plane over a patient. According to an embodiment of the present invention, only a subset of the collimator leaves are analyzed for the fluence calculation, thus reducing the number of calculations required. Additionally, pre-integrated values of scatter strips, associated with each point of the calculation plane, may be referenced in a lookup table. The use of these pre-integrated values allows the avoidance of adding the fluence contribution of each square on the scattering plane. Rather, pre-calculated values of a subset of the scattering plane (scatter strip) may be referenced and combined, thus reducing the number of calculations required for a final scatter contribution to a point on the calculation plane. Further, the thickness of the collimator leaves is considered in the fluence calculation, thus providing a more accurate model for the scatter contributions of points on the scattering plane. |
052316548 | claims | 1. A collimator for collimating radiation beams emitted from a radiation point source comprising: a collimator body adapted to be situated adjacent to a radiation imager having an array of detector elements, said collimator body comprising a photosensitive material and a layer of radiation absorbent material overlying at least portions of said photosensitive material and having a first surface disposed closest to said radiation point source and a second surface disposed closest to the detector element array, said first and second surfaces being substantially coplanar; said collimator body having a plurality of channels therein, each of said channels extending from an opening in said first surface to an opening in said second surface and positioned so that the opening of each of said channels in said second surface is in substantial alignment with a respective one of said detector elements, the longitudinal axis of each of said channels having a selected orientation angle substantially aligned with a direct beam path between said point source and the respective detector element underlying said channel; each of said channels having substantially smooth sidewalls comprising said radiation absorbent material along their length. a radiation point source; a radiation detector comprising an array of detector elements, said array being disposed to detect radiation emitted from said point source; and a collimator disposed between said detector element array and said radiation point source and having a substantially planar surface adjoining said array of detector elements, said collimator comprising a photosensitive material and a layer of radiation absorbent material overlying at least portions of said photosensitive material, and further having a plurality of channels therein to pass radiation emitted by said point source to respective ones of said detector elements, said channels having respective longitudinal axes aligned along respective selected orientation angles, said orientation angles corresponding to respective direct paths from said point source to respective ones of said detector elements, said channels having substantially smooth sidewalls comprising said radiation absorbent material along their length. 2. The collimator of claim 1 wherein the cross-sectional shape of each of said channels corresponds with the cross-sectional shape of each of said respective detector elements. 3. The collimator of claim 1 wherein said collimator body further comprises a radiation absorbent material. 4. The collimator of 1 wherein said radiation absorbent material is selected to substantially absorb radiation of the wavelength distribution emitted by said radiation point source. 5. The collimator of claim 3 wherein said collimator body comprises a photosensitive glass substrate on which a layer of said radiation absorbent material is applied. 6. The collimator of claim 1 wherein said collimator body comprises a plurality of layers, each of said layers having passages formed therein, said layers being joined together so as to align the respective longitudinal axes of said channels. 7. The collimator of claim 1 wherein said detector elements are arranged in a two-dimensional array. 8. A radiation imaging device comprising: 9. The device of claim 8 wherein said radiation point source comprises an x-ray source. 10. The device of claim 9 wherein said collimator further comprises an x-ray absorbent material. 11. The device of claim 10 wherein said radiation absorbent material comprises a material chosen from the group consisting of tungsten, lead, and gold. 12. The device of claim 10 wherein said collimator comprises a plurality of photosensitive glass substrates joined together in layers. 13. The device of claim 12 wherein said x-ray absorbent material is applied at least on all surfaces of said glass substrates exposed to radiation in the assembled imaging device. 14. The device of claim 8 wherein said sloped sidewalls of each respective one of said channels are substantially aligned with the respective selected orientation angle of each of said channels. 15. The device of claim 8 wherein said selected orientation angles of said channels range between about 0.degree. and 10.degree.. 16. The device of claim 14 wherein said sloped sidewalls of each of said channels have a substantially uniform slope along their length. 17. The device of claim 8 wherein said detector elements are arranged in a two-dimensional array. |
description | This is a divisional application of application Ser. No. 09/665,452, filed Sep. 19, 2000 now U.S. Pat. No. 6,885,718; which was a continuing application, under 35 U.S.C. §120, of International application PCT/DE99/00617, filed Mar. 8, 1999; the application also claims the priority, under 35 U.S.C. §119, of German patent application No. 198 12 071.0, filed Mar. 19, 1998; the prior applications are herewith incorporated by reference in their entirety. The invention relates to an apparatus for transferring an article, in particular a nuclear fuel element, from a fluid-filled first vessel into a fluid-filled second vessel or in the opposite direction, with a connecting element connecting the interiors of the vessels, and with a transport device for moving the article through the connecting element. In nuclear power stations, the nuclear fuel is located in the fuel elements that are arranged in the reactor core. The nuclear fuel generates heat in the reactor core, which is delivered for further utilization with a reactor coolant and a reactor cooling circuit. The reactor coolant is mixed with a neutron absorber, in particular with boron or boric acid, in order, together with other devices, to ensure the subcriticality of the reactor core. As soon as a defined fraction of nuclear fuel in the fuel elements has been consumed, fuel elements in the reactor core must be exchanged for new fuel elements. The spent fuel elements are usually stored intermediately in a fuel element storage pond that is filled with water in order to recool the spent fuel elements. The new fuel elements are customarily kept in reserve in the same fuel element storage pond. In the fuel element storage pond, the subcriticality of a fuel element is ensured, as a rule, solely by the geometry and material composition of the storage racks. Therefore, the water need not be treated with boron. An exchange of the spent fuel elements, then, necessitates a transfer of a spent fuel element out of the reactor core into the fuel element storage pond and a transfer of a new fuel element out of the fuel element storage pond into the reactor core. For reasons of fuel element cooling, the transfer must in each case be carried out under water. Townsend, et al., describes a fuel transfer system suitable for the purpose of the transfer in U.S. Pat. No. 5,291,532. This system has a conveyor truck arranged above the reactor core. With the aid of this conveyor truck, a spent fuel element first can be transported vertically upward out of the reactor core and subsequently, along with the forward movement of the conveyor truck, be transported into a transfer pond arranged laterally next to the reactor core. Finally, the fuel element is transported from there further on into a fuel element storage pond. Such a fuel element transfer system, in which the fuel elements are transported above the reactor core, and, in the case of the pressurized water nuclear reactor described, also above the reactor pressure vessel, into the laterally distant fuel element storage pond, is complicated and requires a large quantity of boron-containing water. Another concept therefore provides for the reactor pit and the fuel element storage pond to be arranged at approximately the same height and to be connected to one another by means of a pipe arranged near the bottom. The reactor pit is defined as the part of the containment that contains the reactor pressure vessel and the reactor core. A fuel element transfer system operating according to this concept and suitable for Katz discloses use in water-cooled reactors, et al. in U.S. Pat. No. 4,053,067. This fuel element transfer system provides for the containment region containing the reactor core to be connected to the fuel element store via a tube arranged essentially horizontally below the water surface. For transferring a fuel element, the containment, in particular the reactor pit, is filled with boron-treated water up to a defined level which is sufficient to draw the fuel elements arranged vertically in the reactor core out of the core and position them next to the core in front of the tube. With the aid of a pivoting mechanism, the fuel element is moved into a horizontal position and at the same time laid onto a truck. The rail-bound truck subsequently transports the fuel element out of the containment through the tube to the fuel element store. Transport in the opposite direction functions in a similar way. A fuel element transfer system which operates in a similar way and likewise has a transfer tube between the containment and the fuel element store and a conveyor truck capable of being moved through the transfer tube was described on 01.08.1998 at 15:26 hours under the Internet address “http://www.nrc.gov/nrc/educate/reactor/12-refuel/indexfr.html”. Fuel element transfer systems having a transfer tube between the reactor pressure vessel and a storage vessel have also been developed for use in sodium-cooled nuclear reactors. The transfer tube illustrated by Wade in U.S. Pat. No. 4,096,031 is arranged between the storage vessel and the reactor pressure vessel at an inclination relative to the horizontal. Wade, U.S. Pat. No. 4,069,099, shows a nonrectilinear transfer tube. This transfer tube is V-shaped, so that the fuel elements do not have to be moved into an essentially horizontal position before they can be transported through the transfer tube. Instead, fuel elements within the V-shaped transfer tube only need to be tilted slightly sideways out of the reactor pressure vessel so that the fuel element can be further transported. For the European pressurized water reactor (EPR), the fuel element exchanges with the aid of a transfer tube between the containment interior and the fuel element storage pond. For example, the journal “Nuclear Engineering International,” October 1997, page 14 ff., and the accompanying poster respectively describe and depict a transfer system with a horizontally arranged transfer tube. All tube-based fuel transfer systems share the disadvantage that a transfer tube for exchanging liquids connects the liquid-filled vessels. As already mentioned, the reactor coolant in the reactor pressure vessel and in the reactor cooling circuit contains boron-treated water (boron water). In a situation where there is a fuel element exchange, the reactor pit in the containment is also flooded with boron water of the same or a similar concentration. The boron water is supplied from a separate boron water supply vessel, so that the subcriticality of the fuel elements remaining in the reactor core continues to be ensured after the reactor pressure vessel has been opened. In addition, an exchange of liquid between the fuel element storage pond and the reactor core may also take place via the transfer tube and the flooded interior of the containment during the fuel element transfer. In such an exchange, the boron concentration in the liquid in the fuel element storage pond must be equal to the boron concentration in the reactor pit and in the reactor core prior to the opening of the transfer tube in order to ensure the subcriticality of the reactor core. Therefore, a large quantity of costly boron-containing water is required to operate the known fuel element transfer systems. Particularly when the nuclear power station uses high reactor core burnup, particularly high neutron absorption in the reactor coolant must continue to be ensured during the fuel element exchange. This requires a particularly high concentration of the boron isotope active as neutron absorber, with the mass number 10 (B10), in the reactor cooling circuit. That is, boric acid with a fraction of this isotope, which is artificially increased by means of an isotope separation process, has to be used. Boric acid of this kind is exceedingly costly. Therefore, the quantity required should be minimized. It is accordingly an object of the invention to provide an apparatus for transferring an article between fluid-filled vessels, which overcomes the above-mentioned disadvantages of the heretofore-known devices and methods of this general type. With the foregoing and other objects in view there is provided, in accordance with the invention, a method for transferring an article, in particular a nuclear fuel element. The method having the following steps: providing a fluid-filled first vessel and a fluid-filled second vessel, the interiors of the vessels connected by a connecting element, the connecting element having a first part facing the first vessel and a second part facing the second vessel, and a transport device for moving the article through the connecting element; maintaining a first fluid flow out the first vessel in a first part of the connecting element; and transporting the article through the connecting element with the first fluid flow being maintained. In accordance with another feature of the invention, the method can include maintaining a second fluid flow flowing out of the second vessel in a second part while the article is being transported through the connecting element. In accordance with another feature of the invention, the method can include providing a first issue of the connecting element in the first vessel and a second issue of the connecting element in the second vessel; and setting an essentially identical static pressure before the first fluid flow and the second fluid flow are generated. In accordance with another feature of the invention, the method can include discharging fluid from the connecting element. In accordance with another feature of the invention, the method can include supplying fluid is supplied to one of the vessels with a flow intensity while discharging a fluid with the same flow intensity from the connecting element. In accordance with another feature of the invention, the method can include supplying a fluid to the first vessel at a first flow intensity and to the second vessel with a second flow intensity; and discharging the fluid from the connecting element with an extraction flow intensity which corresponds to the sum of the flow intensities of fluid supplied to the vessels. In accordance with another feature of the invention, the article can be a nuclear fuel element, the first vessel can be a reactor pit of a nuclear power station, and the second vessel can be a fuel element storage pond of the nuclear power station. The invention also provides for an apparatus for transferring an article. The apparatus features a fluid-filled first vessel and a fluid-filled second vessel, each having an interior; a connecting element connecting the interiors of the vessels; a transport device for moving the article through the connecting element; and an extraction device for the discharge of the fluid located on the connecting element. In accordance with another feature of the invention, the extraction device includes a measuring and regulating device for measuring and setting an extraction flow intensity. In accordance with another feature of the invention, the apparatus can feature a collecting vessel for receiving the discharged fluid. In accordance with another feature of the invention, the apparatus can feature an extraction device including an extraction line leading upward as far as an apex point. The apex point can be below a fluid level in one of the vessels. In accordance with another feature of the invention, the apparatus can feature a first issue of the connecting element in the first vessel; and a first pressure measuring device for measuring a first pressure in the first vessel level with the first issue. In addition, the apparatus can include a second issue of the connecting element in the second vessel; and a second pressure-measuring device for measuring a second pressure in the second vessel level with the issue of the connecting element. In accordance with another feature of the invention, the apparatus can include an evaluation unit connected to the first pressure measuring device and the second pressure-measuring device determining the pressure difference between the first pressure and the second pressure. In accordance with another feature of the invention, the apparatus can feature a line conducting the discharged fluid to a preparation plant, in which water contained in the fluid is separated from a boron-containing substance contained in the fluid. In accordance with another feature of the invention, the apparatus can further feature a first metering valve mounted at a first inflow into aid first vessel, through which a predeterminable first fluid flow is set. In addition, the apparatus can include a second metering valve mounted at a second inflow into the second vessel, through which a predeterminable second fluid flow can be set. The object of the invention is to provide a method and an apparatus, with the aid of which, an article can be transported through a tube arranged between two vessels, without fluid in one of the vessels being intermixed with fluid from the other vessel. In terms of the nuclear technology sector, a fuel element transfer that requires only a small amount of neutron-absorbing coolant, in particular, a small amount of boric acid or of B10-enriched boric acid. At least, a smaller quantity of the neutron-absorbing coolant is required than in the tube-based fuel element transfer systems known hitherto. The object relating to a method is achieved, according to the invention, in that a first fluid flow flowing out of the first vessel is maintained in a first part of the connecting element, the first part facing the first vessel, while the article is transported through the connecting element, with the first fluid flow being maintained. The first part of the connecting element opens, for example, directly into the first vessel. The article is, in particular, a fuel element. The invention proceeds from the consideration that the first vessel and the second vessel can be connected by a connecting element, without fluid passing from the second vessel into the first vessel, if fluid flows constantly from the first vessel into the connecting element. This affords the advantage that the article can be transported through the connecting element, without the fluid in the first vessel being intermixed with fluid from the second vessel. If the two vessels were connected by the connecting element, without the first fluid flow prevailing, then, for example if only due to the transport of the article from the second vessel into the first vessel, fluid would also be introduced from the second vessel into the first vessel. Furthermore, even without transport taking place, small pressure differences between the vessels would lead to fluid exchange between the vessels according to the principle of communicating tubes. The first fluid flow, acting virtually as a blocking flow, prevents fluid from flowing out of the second vessel into the first vessel. According to a preferred refinement of the method, while the article is being transported through the connecting element, a second fluid flow flowing out of the second vessel is maintained in a second part of the connecting element, the second part facing the second vessel. This advantageously also prevents the situation where fluid is introduced from the first vessel into the second vessel during the transport of the article in any direction between the vessels. According to a development of the method, before the first or the second or the first and the second fluid flows are generated, an essentially identical static pressure is set at a first issue of the connecting element in the first vessel and at a second issue of the connecting element in the second vessel. This ensures that there is no driving force that could drive a fluid stream from one vessel into the other vessel. The effects of possibly different air pressures across the (upwardly open) vessels or of different temperatures or densities of the fluids in the vessels can be taken into account at the same time. After the pressure compensation described has been accomplished, for example, the connecting element that was still closed up to then can be opened, without a (strong) fluid flow commencing immediately between the vessels. Subsequently, for example, the first and, if appropriate, additionally the second fluid flow can then be built up and then, at most, have to compensate for a very slight static pressure difference between the vessels. The pressure compensation therefore reinforces the effect of the method. The pressure compensation described entails the advantage that virtually no fluid exchange can take place between the vessels, and that the connecting element can therefore be kept constantly open during the exchange of all the fuel elements. That is, no sluice slides or the like are necessary. Through this method, fuel element exchange times are shortened and costs are reduced. The first or the second or the first and the second fluid flows are preferably generated and/or maintained by a fluid being discharged from the connecting element. As a result, both the first and the second fluid flow can be generated in a simple way. The first part of the connecting element then extends between the first vessel and a point at which the fluid is discharged, and the second part of the connecting element extends between the second vessel and this point. For example, fluid is supplied to one of the vessels with one flow intensity and fluid is discharged from the connecting element with the same flow intensity. What is achieved by the supply of fluid is that the fluid level in the vessels does not fall, even though fluid is constantly discharged from the connecting element. Setting the same flow intensities during supply and discharge ensures that the fluid level remains constant. In the event that the first fluid flow is maintained solely in the first part of the connecting element, the first part facing the first vessel, fluid is preferably supplied to the first vessel. “Flow intensity” means a volumetric flow per unit time. According to a particularly preferred refinement of the method, fluid is supplied to the first vessel with a first flow intensity and to the second vessel with a second flow intensity and fluid is discharged from the connecting element with an extraction flow intensity which corresponds to the sum of the flow intensities of fluid supplied to the vessels. This refinement is suitable particularly for the situation where both the first and the second fluid flows are maintained. In a particularly preferred refinement of the method, a nuclear fuel element transfers from the first vessel, which is a reactor pit of a nuclear power station, into the second vessel, which is a fuel element storage pond of a nuclear power station, or in the opposite direction. In this case, the fluid in the first vessel and optionally the fluid in the second vessel do not intermix with the other fluid due to the transport of the fuel element through the connecting element. Thus, for example, the fuel element storage pond can be filled with (pure) water and remain filled with this even during a fuel element exchange, and the reactor pit and reactor pressure vessel can be filled with boron-treated water. Intermixing of the boron-treated water in the reactor pit with the water in the fuel element storage pond is prevented. Therefore, the water in the fuel element storage pond does not have to be treated with boron. The non-boron-treated water of the fuel element storage pond cannot pass into the reactor pit and therefore cannot lead to a criticality of the fuel elements located therein. In an appropriate refinement of the method, the boron-treated water of the reactor pit cannot be intermixed with the large quantity of pure water in the fuel element storage pond, from which the boron-treated water could be separated again only at enormous outlay. By contrast, the preparation of a small quantity of fluid which is extracted from the connecting element and in which boron-treated water and pure water is intermixed does not present any problems. The object relating to an apparatus is achieved, according to the invention, by means of an apparatus according to the preamble of patent claim 8, which is characterized by an extraction device for the discharge of fluid located on the connecting element. This apparatus is suitable preferably for carrying out the method according to the invention. The extraction device is preferably arranged in such a way that fluid is extracted from the connecting element at a point outside the interiors of the vessels. The extraction device is linked, for example, to the connecting element at a point outside the interiors of the vessels. The point is located on the connecting element preferably approximately centrally between the vessels. The extraction device comprises, for example, a measuring and/or regulating device for measuring and for setting an extraction flow intensity. Consequently, fluid can be discharged from the connecting element in a metered manner, in order to set fluid flows in the connecting element, for example the first and/or the second fluid flow, accurately. For receiving the discharged fluid, in particular a collecting vessel may be provided. According to a preferred refinement of the apparatus, the extraction device comprises an extraction line that leads upward as far as an apex point. This affords the advantage that the fluid level in each of the vessels cannot fall below a height at which the apex point is located. To improve this effect, the apex point may be capable of being connected to the surrounding atmosphere via a shutoff fitting. The apex point is located, for example, slightly below a fluid level in one of the vessels. Such a fluid level is, for example, a fluid level that is to be set and/or maintained in one of the vessels. Other developments of the apparatus have a first pressure-measuring device for measuring a first pressure in the first vessel level with the issue of the connecting element. The apparatus also could include a second pressure-measuring device for measuring a second pressure in the second vessel level with the issue of the connecting element. When the first and the second pressures are known, the driving force, which could lead to a fluid exchange between the vessels, can be deduced. For this purpose, for example, an evaluation unit connects to the two pressure measuring devices for determining the pressure difference between the first pressure and the second pressure. If the pressure difference prior to the generation of a fluid flow in the connecting element is set virtually at zero, as is provided according to a refinement of the method according to the invention, then the driving force for a fluid exchange between the vessels can be minimized. However, even during a fluid flow in the connecting element, the driving force and a flow driven by it between the vessels can thereby be kept low. Another development of the apparatus likewise allows a compensation of a pressure difference that may possibly be present between the vessels. For this purpose, an outflow orifice of an outflow pipe is present in each case level with the issues of the connecting element. The outflow pipes issue into a common pipe. In turn, the common pipe leads to the extraction device and, in particular, issues in the extraction line. The pipe may have, in particular, a valve. Preferably, the flow resistances of the outflow pipes are essentially the same. For example, the pipe cross sections are identical. The diameters of the outflow pipes are small, compared to a diameter of the connecting element. For example, they can amount to less than 10%, preferably less than 5%, of a diameter of the connecting element. With the valve open, pressure compensation takes place between the vessels, without fluid being capable of passing from one vessel into the other. In particular, the first vessel is a fuel element storage pond or a reactor pit of a nuclear power station and the second vessel is a reactor pit or a fuel element storage pond of a nuclear power station respectively. This affords the advantage that the fuel element storage pond and the reactor pit may contain water with different boron contents, without the possibility of the change in the boron content in one of the vessels and therefore, in particular, of a reduction in the boron content in the boron-treated water in one of the vessels, in particular in the reactor pit. For example, there is a line for conducting the discharged fluid further on to a preparation plant, in which water contained in the fluid can be separated from a boron-containing substance contained in the fluid. The boron content in the fluid discharged by means of the extraction device does not, as a rule, correspond to either the boron content of the fluid from the first vessel or to the boron content of the fluid from the second vessel. Therefore, an immediate reuse of the discharged fluid is usually not possible. It is therefore advantageous for a boron-containing costly substance contained in the fluid to be supplied to a preparation plant, so that a reuse of the separated boron-containing substance becomes possible. Further refinements of the apparatus have a metering valve mounted at a first inflow into the first vessel. The metering valve sets a predeterminable first fluid stream. A second metering valve can be mounted at a second inflow into the second vessel. The second metering valve sets a predeterminable second fluid stream. Therewith, the fluid level in the two vessels is kept approximately equal and to set an overall supply flow intensity which corresponds to the extraction flow intensity. This also prevents the fluid level in one of the vessels or in both vessels from falling during the discharge of fluid from the connecting element. Other features that 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 an apparatus for transferring an article between fluid-filled vessels, 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. A first vessel 10 in FIG. 1 symbolizes the reactor pit of a nuclear power station and is also designated as such below. The reactor pit 10 is located in the interior 11 of a reactor building, for example a containment, which has a wall 14. A second vessel 15 diagrammatically represents a fuel element storage pond and is also designated as such below. When the fuel elements located in the reactor core (not illustrated) are being changed, the reactor pit 10 is filled to a fluid level 20 with a fluid B which, for example, is boron-containing water and is designated as such below. The reactor pit 10 is illustrated diagrammatically in the filled state mentioned. For receiving exchanged spent fuel elements and for keeping in reserve fresh fuel elements, the fuel element storage pond 15 is filled to a fluid level 25 with a fluid D which is demineralized water and is referred to as such below. For feeding boron-containing water B into the reactor pit 10, there is a first inflow 30 with a first metering valve 31. There is likewise a second inflow 35 with a second metering valve 36 for the fuel element storage pond 15, with the aid of which inflow demineralized water D can be fed into the fuel element storage pond 15. The interior 43 of the reactor pit 10 and the interior 44 of the fuel element storage pond 15 are connected, in each case via an orifice in a side wall 40 of the reactor pit 10 and in a side wall 41 of the fuel element storage pond 15, with the aid of a connecting element 48 mounted between the orifices. The connecting element 48 is in the form of a tube. A first closing-off device 52 is located at a first issue 50 of the connecting element 48 into the reactor pit 10. A second closing-off device 57 is located at a second issue 55 of the connecting element 48 into the fuel element storage pond 15. A transport device 60 is movable in the interior 58 of the connecting element 48, by means of which transport device an article 62 can be transported through the connecting element 48. In the exemplary embodiment illustrated, the article 62 is a fuel element and is designated as such below. The transport device 60 may, for example, be a rail-bound truck, on which, as illustrated, the fuel element 62 is transported, lying horizontally, through the connecting element 48. During the times when the connecting element 48 is not used for the transport of fuel elements, the connecting element 48 can be closed off with the aid of the first and the second closing-off devices 52 and 57, and the interiors 43, 44 of the reactor pit 10 and of the fuel element storage pond 15 can thus be separated from one another in a fluid-type manner. For the rest of the description of the design and functioning of the apparatus illustrated, then, it is assumed that the closing-off devices 52, 57 are first closed. Before the closing-off devices 52, 57 are opened in order to transport a fuel element 62 through the connecting element 48, pressure compensation is first carried out between the reactor pit 10 and the fuel element storage pond 15. For this purpose, a first pressure measuring device 70 and a second pressure measuring device 75 are provided respectively in the reactor pit 10 and in the fuel element storage pond 15. The first pressure measuring device 70 comprises a first effective-pressure line 70A which, by means of a first orifice 70B, receives a first pressure p1 in the interior 43 of the reactor pit 10 (in the boron-containing water B) in the vicinity of the first issue 50. The first effective-pressure line 70A is connected to an evaluation unit 77 via a first valve 70C. Likewise connected to the evaluation unit 77 is a second effective-pressure line 75A. The second effective-pressure line 75A belongs to the second pressure-measuring device 75 and terminates with a second orifice 75B in the vicinity of the second issue 55 of the connecting element 48 in the fuel element storage pond 15. The second pressure measuring device 75 records a second pressure P2 in the demineralized water D. The second pressure-measuring device 75 has a second valve 75C. The orifices 70B, 75B are arranged exactly at the (geodetic) height of the issues 50 and 55, in order to avoid the influence of different densities at the measuring locations. The evaluation unit 77 forms the difference between the first pressure p1 and the second pressure p2 (Δp=p2−p1). The fluid levels 20, 25 in the reactor pit 10 and in the fuel element storage pond 15 are changed via the first inflow 30 and/or via the second inflow 35 in such a way that the differential pressure Δp determined by the evaluation unit 77 becomes approximately zero. After this pressure compensation, the closing-off devices 52, 57 are opened. As a result of the pressure compensation, no driving force exists, at least no strong driving force, for a flow between the reactor pit 10 and the fuel element storage pond 15. However, slight flows possibly remain. For example, slight flows driven by temperature and/or concentration gradients which despite the pressure compensation could lead to fluid exchange between the reactor pit 10 and the fuel element storage pond 15. After the closing-off devices 52, 57 have been opened, the transport of a fuel element 62 through the connecting element 48 by a transport device 60 is begun. However, the movement of the transport device 60 together with the fuel element 62 through the connecting element 48 would move fluid from one of the vessels into the other. Because even the fluid in the connecting element 48 that is filled with fluid after the opening of the closing-off devices 52, 57, is set into motion. An extraction device 99 for discharging fluid A from the connecting element 48 is present at an extraction point 80. The extraction point 80 divides the connecting element 48 into a first part 90 and a second part 95. The first part 90 of the connecting element 48 faces the reactor pit 10 and the second part 95 faces the fuel element storage pond 15. The extraction device 99 includes an extraction line 99A. The extraction line 99A starts from an extraction connection piece 99B mounted on the connecting element 48 at the extraction point 80. The extraction line 99A leads first vertically upward as far as an apex point 99C. At the apex point 99C, a line 99D branches off, which can be closed with the aid of a shutoff fitting 99E. After the apex point 99C, the extraction line 99A leads downward and connects to a collecting vessel 101 via a measuring and/or regulating device 99F. A return line 103 for discharged fluid A leads from the collecting vessel 101 to a preparation plant 105. The pressure that drives the flow of discharged fluid A is obtained from the height difference between the fluid levels 20, 25 and the position of the collecting vessel 101. As a result of the siphon effect, the discharged fluid A automatically flows through the U-shaped extraction line 99A into the collecting vessel 101. By opening the shutoff fitting 99E, the siphon effect and therefore the flow of discharged fluid A into the collecting vessel 101 is prevented. The discharged fluid A consists of a mixture of demineralized water D and boron-containing water B, that is say water with a lower boron content than the water B in the reactor pit 10. The discharged fluid A therefore cannot be supplied directly to the reactor pit 10 because this would lead to a change in the boron content of the boron-containing water B in the reactor pit 10. As a result, the reactor core no longer would be supplied with sufficient boron. In turn, lack of boron could create criticality in the reactor core due to lack of neutron absorption. The preparation plant 105 splits the discharged fluid A, for example, with the aid of suitable ion exchanges, into a fraction with highly boron-containing water B and a fraction with demineralized water D. These fractions are supplied to the reactor pit 10 and to the fuel element storage pond 15 respectively via a return line 107 for boron-containing water B and via a return line 109 for demineralized water D. In the example depicted, the return lines 107, 109 issue in the first inflow 30 and in the second inflow 35 respectively. As a result of the discharge of fluid A from the connecting element 48 with the aid of the extraction device 99, a first fluid flow 110 and a second fluid flow 115 are generated in the connecting element 48. The first fluid flow 110 flows out of the reactor pit 10 to the extraction point 80, and the second fluid flow 115 flows out of the fuel element storage pond 15 to this extraction point 80. This configuration guarantees that even while a fuel element 62 is being transported through the connecting element 48 with the aid of the transport device 60, a flow of demineralized water D out of the fuel element storage pond 15 into the reactor pit 10 and a flow of boron-containing water B out of the reactor pit 10 into the fuel element storage pond 15 are prevented. This applies both to a transfer of a spent fuel element out of the reactor pit 10 into the fuel element storage pond 15 and to a transfer of a fresh fuel element out of the fuel element storage pond 15 into the reactor pit 10. The first fluid flow indicated by means of the arrow 110 acts as a blocking flow for the reactor pit 10 and the second fluid flow indicated by the arrow 115 acts as a blocking flow for the fuel element storage pond 15. The first fluid flow 110 and the second fluid flow 115 are combined at the extraction point 80 and are discharged jointly from the connecting element 48 by the extraction device 99. Only a small volume of fluid A is discharged, in comparison with the volume of the reactor pit 10 and of the fuel element storage pond 15. In order to prevent the fluid levels 20, 25 in the reactor pit 10 and in the fuel element storage pond 15 to fall undesirably during the transport of a fuel element 62 through the connecting element 48, that is say during the discharge of fluid A from the connecting element 48 with the aid of the extraction device 99, boron-containing water B is supplied to the reactor pit 10 through the first inflow 30 with a first fluid flow intensity I1 and demineralized water D is supplied to the fuel element storage pond 15 through the second inflow 35 with a second fluid flow intensity I2. Fluid A with an extraction flow intensity IE is discharged from the connecting element 48 with the aid of the extraction device 99. The extraction flow intensity IE can be measured and/or set at the measuring and/or regulating device 99F. The first fluid flow intensity I1, the second fluid flow intensity I2, and the extraction flow intensity IE are set in such a way that the sum of the first fluid flow intensity I1 and of the second fluid flow intensity I2 gives the extraction flow intensity IE: (IE=I1+I2). Preferably, there is in this case an extraction flow intensity IE giving rise to a flow velocity of approximately 0.01 m/s in the connecting element 48. The time sequence in the setting of the fluid flows is, for example, such that, after the opening of the closing-off devices 52, 57, first the first fluid flow intensity I1 and the second fluid flow intensity I2 are set and then the extraction flow intensity IE is set. The resetting of the first fluid flow intensity I1 and/or the second fluid flow intensity I2 compensates for the Long-term changes in the fluid levels 20, 25. Should an unforeseeable malfunction occur unexpectedly during transfer of a fuel element through the connecting element 48 open on both sides, one of the closing-off devices 52, 57 or both are then closed. The closing-off devices 52, 57 are designed to be redundant for safety reasons. For the same reasons, the transport device 60 cannot be moved through the closing-off devices 52, 57, so that these can be closed at any time. The second exemplary embodiment of an apparatus according to the invention, as illustrated in FIG. 2, demonstrates another possibility for bringing about pressure compensation between the vessels 10, 15, for example, as already described, before the closing-off devices 52, 57 are opened. In contrast to the exemplary embodiment illustrated in FIG. 1, no pressure measuring devices 70, 75 are necessary. Instead, first outflow pipe 120 and a second outflow pipe 130 are inserted. The outflow orifices 120A, 130A of which are arranged respectively in the vicinity of the first issue 50 and of the second issue 55 of the connecting element 48. The outflow orifices 120A, 130A are located exactly at the geodetic height of the issues 50 and 55. The diameters of the outflow pipes 120, 130 are considerably smaller than a diameter of the connecting element 48. The outflow pipes 120, 130 are led, via a first nonreturn valve 122 and a second nonreturn valve 132 and also via a first flowmeter 124 and a second flowmeter 134, so as to issue into a common pipe 140. The pipe 140, in turn, issues in the extraction line 99A in the vicinity of the extraction point 80. The pipe 140 is led via a valve 142. The hydraulic resistance between the outflow orifice 120A of the first outflow pipe 120 and the issue of the first outflow pipe 120 into the pipe 140 and the hydraulic resistance between the outflow orifice 130A of the second outflow 130 and the issue of the second outflow pipe 130 into the pipe 140 are equal. Therefore, the line cross sections of the outflow pipes 120, 130 and the flow resistances of the nonreturn valves 122, 132 and of the flowmeters 124, 134 are approximately equal. After the opening of the valve 142, the pressures in the vessels 10, 15 are equalized. For example, fluid flows out of the vessel with the higher pressure through the pipe 140 into the extraction line 99A and therefore to the collecting vessel 101. The nonreturn valves 122, 132 prevent a flow of this liquid into the other vessel in each case. The flowmeters 124, 134 serve for additionally monitoring the pressure compensation. The apparatus according to the second exemplary embodiment compensates the pressure compensation the vessels in a simple, cost-effective, and reliable way. |
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claims | 1. Device for switching and controlling an electron dose emitted by a micro-emitter, characterized in that it comprises:a sensor module that receives the output current from the micro-emitter and a voltage to adjust the polarization point of the said device, a comparator module that receives the output signal from the said sensor module, and a threshold voltage to adjust the quantity of electrons to be emitted, a logical module that receives the output signal from the comparator module, and a start signal to initialize the electron emission, and a logical signal to define whether or not the micro-emitter should emit, a control module that receives the output signal from the said logical module that generates the voltages necessary for initialization and extinction of the micro-emitter current pulse, means of varying the threshold voltage such that the sum S=Nstart+Nmeasure+Noff remains substantially constant during the electron emission, where Nstart is the number of electrons at the current pulse start time, Nmeasure is the number of electrons at the measurement time of this current pulse, and Noff is the number of electrons at the extinguishing time of this current pulse. 2. Device according to claim 1, comprising means of modulating the threshold voltage (V2) in time starting from the initialization signal (start) so as to program an electron dose control that is variable in time such that excess electrons emitted during the start (tstart) and extinguishing (toff) times are strictly compensated by a reduction of the programmed dose in time. 3. Device according to either claim 1 or 2, comprising:a module for detecting the micro-emitter current, capable of reproducing the tip current Itip exactly, or adding a gain on the current, a variable voltage generation module (68) that outputs a set voltage V2=f(Itip). 4. Device according to claim 3 in which the micro-emitter is a microtip. 5. Device according to claim 2 in which the micro-emitter is a microtip. 6. Device according to claim 1 in which the micro-emitter is a microtip. 7. Linear or matrix switching and controlling device for electron doses emitted by a set of micro-emitters, characterized in that it comprises the following for each micro-emitter:a sensor module that receives the output current from the micro-emitter and a voltage to adjust the polarization point, a comparator module that receives the output signal from the said sensor module and a threshold voltage to adjust the quantity of electrons to be emitted, a logical module that receives the output signal from the comparator module, and a start signal to initialize the electron emission, and a logical signal to define whether or not the micro-emitter should emit, a control module that receives the output signal from the said logical module that generates the voltages necessary for initialization and extinction of the micro-emitter current pulse, means of varying the threshold voltage such that during the electron emission, the sum S=Nstart+Nmeasure+Noff remains substantially constant, where Nstart is the number of electrons at the current pulse start time, Nmeasure is the number of electrons at the measurement time of this current pulse, Noff is the number of electrons at the extinguishing time of this current pulse. 8. Device according to claim 7 in which the micro-emitter is a microtip. 9. Process for switching and controlling an electron dose emitted by a micro-emitter comprising:a step to convert the current output by the micro-emitter and to adjust the operating polarization point, a step to compare the signal obtained at the output from the previous step with a threshold voltage for adjustment of the quantity of electrons to be emitted, a logical step to initialize the electron emission, and to define whether or not the micro-emitter should emit, a control step that generates the voltages necessary for initialization and for extinction of the micro-emitter current pulse, characterized in that it comprises:a step to vary the threshold voltage (V2) such that during the emission of electrons, the sum S=Nstart+Nmeasure+Noff remains approximately constant, Nstart being the number of electrons at the current pulse start time, Nmeasure being the number of electrons at the measurement time of this current pulse, Noff being the number of electrons at the extinguishing time of this current pulse. 10. Process according to claim 9, comprising a step in which the threshold voltage (V2) is modulated with time starting from the initialization signal (start) so as to program an electron dose control that is variable in time such that excess electrons emitted during the start (tstart) and extinguishing (toff) times are all or partly compensated by a reduction of the programmed dose in time. 11. Process according to claim 9, comprising:a step for detecting the tip current, capable of reproducing the tip current Itip exactly, or adding a gain on the current, a step to generate a variable voltage (68) that outputs a set voltage V2=f(Itip). |
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claims | 1. A beam conversion system comprising: a photoconverter on which is incident a light image, thereby to convert the light image into a charged particle emission pattern; an electron optical system that demagnifies and focuses the emission pattern onto a resist coated substrate; and a displacement device comprising a motor, the displacement device being coupled to rotate the photoconverter, wherein the light image illuminates a rotating surface of the photoconverter. 2. The system of claim 1 wherein the photoconverter comprises a photocathode. claim 1 3. The system of claim 2 , further comprising: claim 2 a stage that moves the axis of the rotating photoconverter within the plane of the photoconverter. 4. The system of claim 2 wherein the motor is magnetically coupled to the photoconverter. claim 2 5. The system of claim 4 wherein the photoconverter is encased in a vacuum casing and wherein the motor is positioned outside of the vacuum casing and wherein the motor is magnetically coupled to the photoconverter. claim 4 6. The system of claim 1 , wherein the displacement device comprises: claim 1 a stage that moves the photoconverter in an X-Y direction and within the plane of the photoconverter. 7. The system of claim 1 further comprising: claim 1 a regeneration source positioned to continuously regenerate a portion of the photoconverter while another portion is being used to convert an image. 8. The system of claim 7 wherein the regeneration source sputters material onto the photoconverter. claim 7 9. The system of claim 7 wherein the regeneration source deposits material onto the photoconverter by molecular beam epitaxial deposition. claim 7 10. The system of claim 7 wherein the regeneration source exposes the photoconverter to a molecular beam. claim 7 11. The system of claim 7 wherein the regeneration source deposits material onto the photoconverter by ion beam deposition. claim 7 12. The system of claim 7 wherein the regeneration source deposits material onto the photoconverter by condensation from a gas. claim 7 13. The system of claim 7 wherein the regeneration source comprises: claim 7 a regeneration chamber a regeneration source positioned within the regeneration chamber, wherein the regeneration source provides regenerative material; a pump that controls the pressure in the regeneration chamber; and a nozzle opening, spaced apart from the photoconverter, wherein the nozzle controls the direction and shape of material emitted to the photoconverter. 14. The system of claim 13 wherein the pressure within the regeneration chamber is higher than the pressure in a region proximate the photoconverter in which the light image is converted into the emission pattern. claim 13 15. The system of claim 13 wherein the nozzle opening is shaped to provide uniform regeneration of the photoconverter. claim 13 16. The system of claim 13 wherein the nozzle opening is trapezoid shaped. claim 13 17. The system of claim 7 wherein the regeneration source continuously regenerates a portion of the photoconverter while another portion of the photoconverter converts the light image into an emission pattern. claim 7 18. The system of claim 1 , wherein the movement of the photoconverter by the displacement device is rotation about an axis of the photoconverter. claim 1 19. The system of claim 1 , wherein the photoconverter is ring-shaped. claim 1 20. The system of claim 1 , wherein the motor rotates the photoconverter about an axis, the motor being coaxial to the axis. claim 1 21. A method of generating a charged particle beam image from a light optical image, the method comprising the acts of: generating a light image; providing a motor to rotate a photoconverter; directing the image onto the photoconverter whereby the image is incident on a rotating surface of the photoconverter; converting, at the photoconverter, the image into a charged particle emission pattern; demagnifying the emission pattern; and focusing the demagnified emission pattern onto a substrate. 22. The method of claim 21 wherein the photoconverter comprises a photocathode. claim 21 23. The method of claim 20 , further comprising the act of regenerating a portion of the photoconverter. claim 20 24. The method of claim 23 wherein the act of regenerating comprises the act of sputtering. claim 23 25. The method of claim 23 wherein the act of regenerating comprises the act of molecular beam deposition. claim 23 26. The method of claim 23 wherein the act of regenerating comprises the act of exposing the photoconverter to a molecular beam epitaxy. claim 23 27. The method of claim 23 wherein the act of regenerating comprises the act of ion beam deposition. claim 23 28. The method of claim 23 wherein the act of regenerating comprises the act of providing condensation from gas. claim 23 29. The method of claim 23 wherein the act of regenerating comprises removing undesired material. claim 23 30. The method of claim 23 wherein the act of regenerating comprises the act of using a nozzle opening shaped to provide uniform exposure to the photoconverter. claim 23 31. The method of claim 21 , wherein the photoconverter device continuously moves. claim 21 32. The method of claim 21 , wherein the movement of the photoconverter is rotation. claim 21 33. The method of claim 21 , wherein the photoconverter is ring-shaped. claim 21 34. The method of claim 21 , wherein the rotation of the photoconverter is about an axis, and wherein the motor is coaxial to the axis to rotate the photoconverter about the axis. claim 21 35. A method of generating a charged particle pattern, comprising the steps of: generating a light image; providing a photoconverter; providing a motor to rotate the photoconverter; rotating the photoconverter relative to an axis of the light image; directing the light image onto the rotating photoconverter, whereby the image is incident on a rotating surface of the photoconverter; converting, at the photoconverter, the light image into a charged particle pattern; demagnifying the charged particle pattern; and focusing the demagnified charged particle pattern onto a target. |
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summary | ||
summary | ||
abstract | Methods and apparatus are disclosed for reducing thermal deformation of xe2x80x9cupstreamxe2x80x9d marks (as used for alignment and/or calibration) situated on a reticle or on a reticle plane (e.g., on the reticle stage), thereby facilitating more accurate transfer of the reticle pattern to a sensitized substrate (e.g., semiconductor wafer) using a charged particle beam (e.g., electron beam). The charged particle beam illuminates an upstream mark situated on the reticle or on a reticle plane and projects an image of the illuminated upstream mark onto a corresponding xe2x80x9cdownstreamxe2x80x9d mark situated on a substrate plane. A shield is situated upstream of the upstream mark and serves to block downstream passage of the charged particle beam except to illuminate the upstream mark or a portion of the upstream mark. The upstream mark can be situated on the reticle or on a mark member situated in the reticle plane. |
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description | This Application is a continuation of U.S. patent application Ser. No. 13/497,737 filed on Nov. 5, 2012, now U.S. Pat. No. 9,824,781, which is a national phase application of International Application Number PCT/US2010/50397 filed on Sep. 27, 2010, which claims the priority benefit of U.S. provisional Patent Application No. 61/245,881 filed on Sep. 25, 2009, the disclosures of which is expressly incorporated herein in its entirety by reference. The present invention generally relates to systems and methods for handling massive containers and, more particularly, handling storage casks for nuclear waste material. Nuclear power plants are required to have systems and methods for removing spent nuclear fuel from the plants so that it can be stored and/or processed. The spent nuclear fuel is typically stored in casks. While the current systems and methods may handle the casks, they have a number of problems. Existing systems have little documentation, require significant man hours, and use out-dated technology. These current methods also require a relatively large number of single use components that makes these systems expensive and difficult to maintain. Accordingly, there is a need in the art for improved systems and methods for handling casks containing nuclear waste material. The present invention provides a system and method that overcomes at least some of the issues of the related art. Disclosed is a method for removing spent nuclear fuel comprising the steps of moving a cask below a penetration using a transporter, raising the cask from the transporter using a handling mechanism engaging only upper trunnions of the cask so that the cask self-aligns with the penetration using gravity, securing the cask to the penetration, inserting the spent fuel into the cask, unsecuring the cask from the penetration, and lowering the cask onto the transporter using the handling mechanism. Also disclosed is an upper handling mechanism for handling a sent nuclear fuel cask having pairs of upper and lower trunnions. The mechanism comprises, in combination, a fixed position frame, a tool movable in the vertical direction relative to the frame, a plurality of hydraulic cylinders for vertically moving the tool relative to the frame, and a pair of paddles pivotably attached to the tool for selectively engaging the upper trunnions of the cask. Also disclosed is a method for removing spent nuclear fuel comprising the steps of moving a cask below an opening at a first station using a self-powered transporter, rotating the cask from a horizontal orientation to a vertical orientation at the first station, moving the cask below hoist at a second station using the self-powered transporter, moving the cask below a penetration at a second station using the self-powered transporter, raising the cask from the self-powered transporter to the penetration, securing the cask to the penetration, inserting the spent fuel into the cask, unsecuring the cask from the penetration, and lowering the cask onto the self-powered transporter. Further disclosed is a self-powered vehicle for transporting a spent nuclear fuel cask having pairs of upper and lower trunnions. The vehicle comprises, in combination, a body, an upender secured to the body for holding the cask and moving the cask between vertical and horizontal orientations, and a plurality of independently driven and independently steered wheels on each lateral side of the body. From the foregoing disclosure and the following more detailed description of various preferred embodiments it will be apparent to those skilled in the art that the present invention provides a significant advance in the technology of systems and methods for spent nuclear fuel removal. Particularly significant in this regard is the potential the invention affords for providing an, reliable and effective system and method for handling spent nuclear fuel casks. Additional features and advantages of various preferred embodiments will be better understood in view of the detailed description provided below. It should be understood that the appended drawings are not necessarily to scale, presenting a somewhat simplified representation of various preferred features illustrative of the basic principles of the invention. The specific design features of the cask handling system as disclosed herein, including, for example, specific dimensions, orientations, locations, and shapes will be determined in part by the particular intended application and use environment. Certain features of the illustrated embodiments have been enlarged or distorted relative to others to facilitate visualization and clear understanding. In particular, thin features may be thickened, for example, for clarity or illustration. All references to direction and position, unless otherwise indicated, refer to the orientation of the cask handling system illustrated in the drawings. In general, up or upward refers to an upward direction within the plane of the paper in FIG. 3 and down or downward refers to a downward direction within the plane of the paper in FIG. 3. It will be apparent to those skilled in the art, that is, to those who have knowledge or experience in this area of technology, that many uses and design variations are possible for the improved systems and methods disclosed herein. The following detailed discussion of various alternative and preferred embodiments will illustrate the general principles of the invention with reference to preferred embodiments. Other embodiments suitable for other applications will be apparent to those skilled in the art given the benefit of this disclosure. Referring now to the drawings, FIGS. 1 to 5 illustrate a fuel building 10 having a fuel transfer or cask handling system according to the present invention 12. The illustrated cask handling system 12 handles a spent fuel storage cask 14 through the process of removing spent nuclear fuel from the fuel building 10 including providing an unloaded cask 14, preparing and opening the cask 14, loading spent fuel into the cask 14, sealing the cask 14, and removing the loaded cask 14 from the fuel building 10. The cask handling system 12 includes a self-powered mobile cask handling vehicle or cask transporter 16, an upper handling mechanism 18, a penetration cover 20, a seismic restraint 22, and a valve system 24. A preferred method according to the present invention for removing spent fuel assemblies from a fuel building 10 and transporting them to on-site facilities for the next stage of disposal is as follows. First, a complete empty cask 14 is placed onto the cask transporter 16 in the horizontal or vertical orientation by an overhead gantry crane. The cask 14 is securely attached to an upender structure 26 of the cask transporter 16 which can pivot the cask 14 about a horizontal and laterally extending pivot axis 28 so that the cask 14 can be moved between horizontal and vertical positions. Precise positioning of the cask 14 onto the cask transporter 16 is not necessary because locating the cask 14 with respect to a fuel pool 30 and penetration 32 in the building 10 is accomplished by the other equipment as described hereinafter. With the cask 14 positioned in its horizontal position, the cask transporter 16 drives to the fuel building 10. The cask transporter 16 has the ability to drive anywhere on site and can be operated by an on-board driver or by radio remote control. The cask transporter 16 has a hydraulic power system that is powered by a self-contained motor and generator 36 (no external tractor or tugger is required). When inside the cask transfer facility 10, the cask transporter 16 has the ability to run on remote power via an umbilical cord. The cask transporter 16 enters a cask loading hall or fuel hall 34 of the fuel building 10 and aligns itself with a pair of parallel, embedded floor rails 40. When the cask transport 16 is aligned with the imbedded floor rails 40 and completely with the cask loading hall 34, isolation doors are shut and temporary power is connected to the cask transporter 16 via the umbilical cord. Because the cask transporter 16 is aligned with the rails 40, side-to-side or lateral positioning of the cask transporter 16 is automatically accomplished and precise positioning from front to back in a linear direction within the cask loading hall 34 can be obtained. The upender 26 on the cask transporter 16 repositions the cask from its traveling horizontal position to its vertical position, engages upper seismic constraints, and positions the cask 14 under a first processing station which is the cask prep station 42. At the first station 42, a shock absorbing cover, protection lid, and fixing flange of the cask 14 are each manually removed using an auxiliary crane located in the fuel building 10. Personnel are located above the fuel hall 34 and access the cask 14 through a hole 44 in the floor. This provides a controlled and safe work area for removing the covers and lids from the cask 14. The cask components are stored on sliding shelves located adjacent the hole 44. Once bolts for the biological lid of the cask 14 have been removed, the cask transporter 16 is moved by radio control to the second station which is the biological lid station 46. At the second station 46, a hoist 48 with a grapple device is manually operated, aided with cameras, to maneuver the grapple to engage and remove the biological lid of the cask 14. With the biological lid moved out of the way, a thorough visual inspection of all seals and sealing surfaces of the cask 14 is conducted by an operator using cameras. Redundant piping and hosing is connected into ports of the cask 14 at this time. The cask transporter 16 is then moved by radio control to the third station which is the cask loading station 50. At the third station 50, the cask transporter 16 locates the cask 14 under the penetration 32 of the fuel pool 30 and personnel disconnect the cask 14 from the cask transporter 16. In this position, the cask is located under the upper handling mechanism 18. Hydraulically powered paddles 52 of the upper handling mechanism 18 have key slots 54 which are extended and slide over upper trunnions 56 of the cask 14 to lock the cask 14 to the upper handling mechanism 18. With the cask 14 securely held by the paddles 52, the cask transporter 16 is backed away and a vertical guide system or seismic restraint 22 rises from the floor and engages lower trunnions 58 of the cask 14. As the cask 14 is raised by the upper handling mechanism 18 from the cask transporter 16, the lower trunnions 58 engage a keyed structure 60 in the vertical guide system 22, preventing a swinging pendulum motion in a seismic event. The cask 14 is lifted up by the upper handling mechanism 18 and proper alignment of mating surfaces is visually verified using cameras. A multi-stage redundant bladder system engages an inner face of the cask opening. Mechanical locking means engage and the paddles 52 locate the cask 14 in alignment (similar to a plumb bob) using gravity. The redundant bladder system is then inflated to secure the seal. After successful docking of the cask 14, the penetration 32 is filled with borated or de-mineralized water. Using vent and drain valves, the cask 14 is filled with water and pressure is equalized on the two sides of the penetration upper cover 20. At this time, all personnel are exited from the loading hall 34. The penetration upper cover 20 is opened and remains opened and monitored by cameras as spent fuel is loaded into the cask 14. As the cask has been loaded with spent fuel and the cameras verify that the spent fuel bundles are located properly, the penetration upper cover 20 is closed. The area below the penetration upper cover 20 is drained, rinsed with de-mineralized water and allowed to dry. The water in the cask is lowered to the necessary level for the biological lid. The cask transporter 16 is then moved back to the cask loading station, the bladder seals are depressurized and the cask 16 is lowered from the seal and onto the cask transporter 16. The paddles 52 retract from the cask 14 and mechanical means secure the cask to the cask transporter 16. The cask transporter 16 then moves the cask 14 back to the biological lid station 46 where the biological lid is placed back onto the cask 14 and the remaining cask restraints are secured. Personnel are then allowed back into the loading hall 34. Redundant piping and hosing is disconnected from the cask ports and all ports are properly sealed. The cask transporter 16 then moves the cask 14 back to the cask prep station 42. Remaining cask components are reassembled and properly engaged on the cask 14. Remaining cask constraints are secured and the cask 14 is down ended to its horizontal orientation. Radiological tests are performed and decontamination is performed as necessary. The doors in the loading hall 34 are opened and temporary power to the cask transporter is removed, that is, the umbilical cord is removed. The cask transporter 16 then drives out of the fuel building 10 under its own power. The cask transporter 16 takes the cask 14 to a handling area for final disposal. As best shown in FIGS. 6 to 8, the illustrated cask transporter 16 is a diesel/electric, self-propelled, wheeled vehicle that transports the storage cask 14 which weighs 125 tons. The illustrated cask transporter 16 includes sixteen wheels 62 which are driven by industrial hydraulic motors 64 with integral brakes for total control and greater flexibility. The illustrated cask transport 16 has four pairs of wheels 62 on each lateral side of the cask transporter 16. A diesel powered electric generator 36 provides power to operate the cask transporter 16. The cask transporter 16 preferably is designed to safely hold a TN32 cask 14 during a seismic event. A dynamic multiplier of 1.15 is preferably considered for impact loading during normal operations. Hydraulic fluids are preferably suitable for outdoor operation at 0 degree Fahrenheit and are preferably non-flammable with a flashpoint> or =100 degrees Fahrenheit. High pressure hydraulic lines are preferably secured and protected to prevent whipping in the unlikely event of failure. Hydraulic systems preferably carry the rated load, including a 15% hoist factor. Calculated safety margins for cylinder buckling and hoop stress are preferably a minimum of 2:1 versus the buckling load limit and the material yield strength respectively. The cask transporter 16 is sized and shaped so that it is stable to ensure that an upset will not occur during normal or off normal events. The illustrated cask transporter 16 can shuttle loaded and unloaded storage casks 14 between the fuel handling hall 34 and any other accessible location at the site. The illustrated cask transporter 16 has a unique turning mechanism and wheel design allows significantly more maneuverability over prior systems. The cask transporter 16 preferably includes the following features: twenty year design life; all weather design; OSHA compliant design; auto-rotating, fully loaded on concrete or other hard surface; key start switch; switch type speed control; diesel fuel tank of about forty to fifty gallons; heaters (sump pump, fuel tank, and hydraulic reservoir); dead man controls (brakes applied upon release of control, loss of fluid pressure, or loss of power); traverse speed of 0.4 mph+/−0.05 mph on level ground; manual lowering capability without power; warning lights and audible alarm (30 foot range); provisions to prevent uncontrolled lowering; portable fire extinguisher; float battery charger; access ladders and fall protection; control panel capacity nameplate (rated load, empty weight, temperature limitations); ability to traverse two inch lip of obstructions at the site; durable outdoor paint system; and non-slip walkway surfaces. The illustrated cask transporter 16 includes a body 66 which is the main weldment vehicle frame. The body 66 is the center structure that ties the entire machine together. It is constructed from welded plates and structural shapes. The body 66 serves as the mounting point for all other systems of the cask transporter 16 and also serves to support the cask 14. The body 66 is preferably a weldment constructed primarily from mild steel and structural shapes (ASTM A572 and A500C with yield strengths of 50,000 psi and welded per AWS D1.1). Welding complies with AWS D1.1. The structure is evaluated for both static and seismic load requirements. As best shown in FIGS. 9 to 13, rubber tire propulsion/support systems of the illustrated cask transporter 16 include the wheels 62, rotation mechanisms 68, and hydraulic drive units 70. The illustrated eight pairs of dual-rubber wheels 62 (four pairs on each side and sixteen total wheels) are mounted on the underside of the body 66. The wheels 62 are preferably foam-filled aircraft tires such as those available from Michelin or equivalents that are designed for high capacities and high speeds. Because the cask transporter 16 is traveling at very low speeds, these wheels 62 are conservatively designed for this function. The foam-filled tires ensure that there is never a flat tire that could challenge the safety of the fuel assembly with a transported cask 14. Each dual tire set is driven by the hydraulic motor 64. Based on a dirt surface, a rotational speed of 3.056 rpm and 5% grade, each hydraulic motor 64 is approximately 5HP and is independently controlled by the PLC. Each dual wheel set is independently steered using commercially available rotary actuators 72. The rotary actuators 72 are used to pivot a joint where a conventional mounting proves impractical due to space, weight, or motion restrictions. These rack and pinion actuators 72 provide high torque output, zero leakage drift-free positioning, and excellent shock load resistance. These types of wheel sets are highly reliable. A control system provides the signals to drive, turn and rotate the wheels 62. Using a PLC that independently controls each of the dual wheel assemblies, the cask transporter 16 can turn as needed and drive around the entire site. The steering system provides the operator with the capacity to rotate the cask transporter 16 on itself, that is pivoting about its center (best shown in 13D). As best shown in FIG. 14, the illustrated cask transporter 16 includes the diesel powered generator 36 located at the rear of the body 66 to provide electrical power. The generator 16 includes a diesel engine, generator, diesel fuel tank, and all of the equipment to support the operation of the engine and generator and are all contained within a frame of a module 74. The engine and generator are sized to manage the most demanding function as limited by the control system. The diesel engine drives the generator, which is selected to provide 460V/3-phase/60 Hertz power to the cask transporter. This electricity powers and electric motor/hydraulic pump module for the lift function of the upender 26 and either another electric motor/hydraulic pump module for the propel function of the wheels 62 or two electric propel motors. Noise suppression systems are included with the system to reduce the dba levels workers are exposed to below OSHA limits. Operation of the cask transporter 16 requires that each function (propulsion, upending, etc.) be operated separately to maximize safety. As best shown in FIG. 15, the illustrated cask transporter 16 employs automatic drop protection to prevent uncontrolled lowering of the cask 14 during any system failure, such as loss of pressure to the cylinders or other catastrophic failure of the lifting system 26. The cask transporter 16 preferably is equipped with a separate safety system. This safety system holds the cask 14 in a safe condition in the unlikely event that a hydraulic cylinder fails or other structural parts of the lifting system 26 fail to function. Separately mounted from the hydraulic cylinder, the safety system employs two commercially available SITEMA safety catchers 76. Conventional locking devices fitted to the hydraulic presses (such as locking bolts or latches) often operate at the top, or a few more positions. Form fitting systems have a gap in safety between where power is disrupted and the hole slide hits a locking point. These obvious disadvantages are avoided by using SITEMA safety catchers 76. These safety catchers 76 prevent the cask 14 from crashing down at any stage of ascent or descent, are mechanically safe and reliable, and do not have a ratchet. A high safety standard, along with improvements in productivity, is achieved through: the load is supported on a holding shaft separate from the cylinder; the SITEMA safety catcher clamps without a ratchet, so that a safe clamping condition is attained throughout the entire stroke and a productively increase is offered as the actual stroke can be limited to the length that is absolutely necessary; the clamping system is held open by hydraulic or pneumatic means so that when pressure drops, the cask 14 is immediately secured; the energy of a falling or sinking load is used to generate the clamping force which only happens if the load starts to move downward from the secured position (when the safety catcher is without pressure). In this case, the cask 14 is securely stopped almost instantly with help of the self-intensifying clamp movement; and SITEMA braking operations are fully operational at all cylinder speeds and usually a deceleration of 1 to 3 g (acceleration due to gravity) is achieved and the resulting braking distance is not more than a few centimeters. The illustrated cask transport 16 includes operator control system 78 including control panels and a generator module console. The operator control system is ergonomically mounted on the top deck 18 of the cask transporter 16 to provide user friendly operation from a swiveling operator's chair 82, in a location providing an unobstructed view of cask handling operations. Next to the operator's chair 82 is a stationary control console that has auxiliary indications. The operator's chair 82 can rotate approximately 270 degrees and automatically reverses the joy stick controls based on the orientation of the chair 82. The operator's control is provided with a protective cover to prevent weather damage. Hydraulics are operated by manipulation of solenoid valves that port fluid to extend and retract from commercially available hydraulic cylinders, such as those available from Parker. Counter-balance and pressure compensated flow valves ensure that the hydraulic system only operates when commanded, and is fail safe on the loss of pressure from leaks or pump failure. Operating pressure will be displayed on the stationary console plus additional warning lights for low hydraulic level and other fault conditions. The speed of the cask transporter 16 is controlled by a joy stick that is located on the operator's chair 82. Based on the position of the joystick, a 0-10 VDC signal is sent to a proportional valve that drives the eight hydraulic motors 64 either in forward or reverse. The joystick is spring-returned to neutral (0 position) to act as a dead-man switch. Steering is controlled by a multi-axis joystick that feeds a proportional signal though a PLC, such as those available from Allen Bradley, or equivalent that separately steers the eight pairs of wheels 62. The PLC program individually controls the wheels 62 so that they are rotated correctly based on their position on the cask transporter 16. Hydraulic fluid drives the eight rotary actuators 72, such as Parker HTR series hydraulic rotary actuators, with electronic feedback to properly position the wheels 62. A separate 75 HP motor drives a 28 gallon piston pump that is connected to a 80 gallon HPU reservoir for steering and propulsion. The tank comes with heat exchanger and heaters to accommodate any environmental extreme. Strainers and filters are preferably provided. Controls for the cask transporter 16 are designed to be fail-safe, so that loss of power will shut down the system and prevent an uncontrolled movement of the cask 14. All safety interlocks and controls of the cask transporter 16 are hard wired between the specific relays, drives, circuit breakers, and other electrical equipment. The control system is designed per NEC standards and mounted within a minimum of NEMA 4 enclosures. Wiring is mounted in rigid conduit except for necessary flexible connections and at the interface between the conduit and the equipment. The cask transporter 16 is also grounded for personnel and equipment protection. The upender 26 is powered by dual hydraulic brake-motors coupled top a planetary gear set to drive a pinion/bull gear ensemble. Encoders are integrated into each drive and set up as a master/slave configuration to ensure the upending is done in unison. Rotation is about a point approximately within three inches or about 80 centimeters of the center of gravity, therefore necessary power is kept to a minimum. In case of failure of one drive system, the other brake motor can hold the cask 14 by itself and can be driven to lower the cask 14 back down to a safe position. In addition, the fuel building crane can also be used to lower the cask 14 in case of a catastrophic failure. To prevent shock to the fuel assembly and cask 14, shock absorbers have been incorporated into the bed for safety. The upender 26 can be extended approximately forty inches or about one meter so that the cask 14 can be raised to the upper elevation at the cask preparation station 42. Dual eight inch double acting cylinders lift the cask 14 using non-flammable hydraulic fluid at a pressure of greater than 80% of the maximum operating pressure. Safety catchers 76 are incorporated into the cylinders so that a failure of a cylinder rod will nit be catastrophic. On loss of power, the cylinders can be manually lowered to put the cask 14 in a safe condition. When the cask 14 is on the upender 26, it is captivated in several locations. On the bottom of the cask 14, an “L” shaped platform 84 is hydraulically operated to latch the lower portion of the cask 14. This prevents the cask 14 from sliding and keeps the trunnions 56, 58 in their respective pockets in the bed. A second hydraulic assembly latches the rear upper trunnion 56 and prevents the cask 14 from tipping forward under even the worst anticipated seismic event. The locks fail safely in case of loss of power or loss of hydraulic fluid. On the bottom of the upender carriage is an alignment or guide tool or assembly. This hydraulically activated alignment assembly lowers onto the rails 40 that are embedded in the floor of the fuel hall 34 to guide the cask transporter 16 in precise alignment. This assembly is only a guide and does not have driven wheels. A single hydraulic cylinder is used to raise (store) and lower (engage) the assembly. At the first or cask prep station 42, the cask 14 is moved to the vertical position. The cask transporter 16 aligns the cask 14 with the hole 44 in the ceiling of the fuel hall 34 and the fuel building crane is used to perform cask component removal/replacement work. The cask 14 is positioned so that the crane can take each lid out of the cask 14, bring it up through the hole 44 and place it on a rolling shelf. Operators can easily access the top of the cask 14 to remove bolts and prepare the cask for insertion of the fuel assemblies. After the biological lid's bolts are removed, the cask 14 proceeds to the second or biological lid station 46 to have the biological lid removed and the seals inspected. Using the Hevi-Lift Hoist 48 or the equivalent mounted onto a bridge and trolley assembly, a grapple can be maneuvered to attach to the biological lid and remove it from the cask 14. The Hevi-Lift Hoist 48 is a 7.5 to 10 ton unit that has multiple single failure proof components in order that the lid cannot be dropped onto spent fuel. The hoist 48 has multiple brakes (CD brake, load brake and regenerative braking) coupled with a duel rope system to ensure that the breakage of rope will not drop the load. The hoist 48 is operated with a variable frequency drive, such as a Smartorque drive, or equivalent for precise positioning. The bridge and trolley are very short spans providing approximately one foot (or about 0.3 meters) of travel in the X and Y plane. The bridge and trolley are over sized to allow for a 10:1 design factor based on ultimate strength and are operated using a standard starter and relay rather than VFD. The grapple is designed to meet the requirements of ASME N14.6-1993, “Special Lifting Devices for Shipping Containers Weighing 10,000 pounds or More” and ASME BTH-1, “Design of Below the Hook Lifting Devices”. The grapple is designed to interface with the round lug on the top of the geological shield. The device has jaws that meet the standard configuration profile. The jaws of the grapple pass through the opening (ID) in the canister lifting lug and come to rest on the top of the lid. As the weight of the grapple shifts from being held by the hoist due to being carried by the lid, the linkage of the system of the grapple moves downward and disengages the mechanical latch. The mechanical latch works by using a T-shaped rod and cam profile that has the ability to move up and down, and to rotate. Similar to operating a ball point pen, the cam mechanism in the latch alternates from extending and retracting the T-shaped rod. When the grapple travels downward, it activates the latch to move the wedge configuration to drive the jaws outward until full stroke is obtained (approximately 2 inches). Once the grapple is attached to the lug, it is mechanically locked and cannot open as a result of operator error. This is efficient because the mechanical principle of wedges (incline planes) gives a mechanical advantage based on the weight of the load lifted. The jaws cannot disengage while lifting the load. When disengaging the cask, the reverse sequence occurs. On the downward motion of the grapple, the weight of the unit applies a vertical force on a linage series which in turn applies a horizontal force to retract the jaws. This all occurs simultaneously leaving the jaws retracted and the grapple in the unlatched position. The grapple can then be lifted free of the lid. Once the biological lid has been removed, the cask transporter 16 moves the cask 14 to the third or upper cask handling station 50 where the cask 14 is positioned against a penetration seal. As best shown in FIG. 16, the upper cask handling station 18 includes a weldment 86 that has four hydraulic cylinders 88 that raise and lower the cask engagement tool 90. The cask engagement tool 90 includes the two pivoting paddles 52 with the key slots 54 that fit over the upper trunnions 56 on the side of the cask 14. With the cask 14 aligned under the upper cask handling station 18, the ten inch diameter cylinders 88 lower the paddles 52, and an electro-mechanical actuator pivots the paddles 52 down about their horizontal pivot axes and over the trunnions 56. The cylinders 88 then rise slightly to ensure proper fit and take some initial cask load. The cask 14 is disengaged from the cask transporter 16, which backs away from the cask 14. With the entire cask 14 suspended from the upper cask handling station 18, the cylinders 88 raise the cask 14 into the penetration seal. The cylinders 88 rise together based on a linear encoder in each rod that feeds back to the control system to ensure proper alignment. In addition to the linear encoders, the upper cask handling system 18 ensures proper alignment with guide tubes that are positioned at each of the four corners. Gravity ensures the cask 14 hangs straight down since the round trunnions 56 seated in the round key slots 54. As the cask 14 is raised, it interfaces with a stainless steel penetration lower flange that has a multi-level seal system. Once the cask 14 is seated, the seals are filled with air to seal the interface between the penetration lower flange and the cask 14 so that there is no leakage even with the pressure resulting from a significant water column. In between the seals are leak detection sensors that provide assurance that the main and backup seals are tight. With the cask 14 properly seated, tapered shear pins are inserted between the stationary structure and the lift frame 18 to lock the cask 14 in place. This provides assurance that even during a seismic event, the cask 14 will not become disengaged from the penetration seal. Once the cask 14 has been raised and seated on the penetration seal, the lower seismic restraint 22 engages the lower trunnions 58 of the cask 14 to securely hold the assembly. This carbon steel weldment 92 is mounted permanently to the floor in the fuel hall 34 below the penetration 32. As best shown in FIG. 17, the seismic restraint 22 includes two horizontally moving arms 94 that extend out at the height of the trunnions 58. The cask transporter 16 straddles the lower seismic restraint 22 when it delivers the cask 14. After the cask transporter 16 has released the cask 14 and backed out of the way, the restraint 22 actuates to engage the trunnions 58 using an ACME screw to bring the arms 94 over the trunnions 58. A separate locking plate operated by a mechanical-electrical actuator locks both arms to the cask 14 so the unit can handle seismic forces in all three planes. Once the cask 14 is filled with fuel assemblies, the lower seismic restraint 22 releases the cask 14 by shifting locks and retracting the arms 94 from the cask 14. The cask 14 can then be lowered onto the cask transporter 16 and the loaded cask 14 can be removed from the fuel hall 34. As best shown in FIGS. 18 to 20, the penetration upper hatch or cover 20 includes a rim 96, a cover 98 with latches 100, o-rings 102, latch cylinders 104, a hatch cylinder 106, a hydraulic power unit 108, piping 110, and leak sensors 112. The rim 96 is a stainless steel weldment sized to fit the hole at the upper penetration 32. It houses the o-ring seals 102, provides a base for installation of the hatch cylinder 106 and the latch cylinders 104, and offers a pivot for the cover 98. The cover 98 is a stainless steel weldment. It mates with the rim 96 at the pivot points, through the hatch cylinder 106, through the latch cylinders 104, and at the o-ring seals 102 where it provides sealing. The o-rings 102 are fabricated o-rings of about a 1.0 inch cross section. The o-rings 102 are fabricated to three different diameters to provide three concentric sealing surfaces. Material is compatible with the water of the spent fuel building and a high radiation application. The latch cylinders 104 are stainless steel water hydraulic cylinders of 3.25 inch bore and 3.5 inch active stroke. They are front flange mounted and rear flange retrained to decrease deflection when operating. The rod is 2.0 inches in diameter with a ¾×15 degree end taper. This taper forces the cover tight against the o-ring seals providing a positive seal. The hatch cylinder 106 is a stainless steel water hydraulic cylinder of 4.0 inch bore, 16.0 inch active stroke, and 1.5 inch diameter rod. It is mounted to the rim 96 at its base end and to the cover 98 at its rod end providing the force to open and close the cover 98. The hydraulic power unit (HPU) 108 is a motor driven water hydraulic pump which provides flow and pressure to operate the cylinders 104, 106. It incorporates water hydraulic valves to operate the latch cylinders 104 or the hatch cylinder 106. Piping 110 to the cylinders 104, 106 is stainless steel tubing fabricated to the dimensions of the SFB transferring flow and pressure from the HPU 108 to the cylinders 104, 106. The leak sensors 112 are switches which provide a signal to the system when sensing a leak through the o-rings 102. Operation of the penetration upper hatch cover 20 begins with the cover 98 closed and locked. When it is desired to open the cover 98, the operator activates the valve operating the latch cylinders 104. These cylinders 104 retract, pulling their rods (pins) from the cover latches 100. Sensors confirm when the cover is unlatched. The operator then activates the hatch cylinder 106. This cylinder 106 pulls on the cover lever and opens the cover 98. The cover rotates from zero degrees through about 105 degrees at full open. Sensors confirm that the cover is fully open and the penetration 20 is ready for passage of the fuel assemblies. Fuel is passed through the penetration 20 until the spent fuel cask 14 is full, and must be removed. To close the penetration 20, the operator activates the hatch cylinder 106 to close the cover 98. The cylinder 106 moves the lid 98 until the CG is past center and then restrains the lid 98 as it lowers down onto the o-ring seals 102 of the rim 96. A hatch cylinder pin may be manually pulled to allow the cover 98 to close in an emergency. The illustrated embodiment has three o-rings 102 arranged circumferentially about the hatch opening. These o-rings 102 seal on their tops and bottoms against the cover 98 and the rim 96. The operator activates the latch cylinders 104 which drive their tapered rods (pins) into the latches 100 of the cover 98. This taper further forces the lid 98 tight against the o-rings 102 ensuring their complete seal. Sensors indicate when the latches 100 are fully engaged. Since there is no residual force attempting to release the latch cylinders 104, the lid 98 will remain closed and sealed during any unforeseen conditions. Hand pumps can release the latches 100 during emergency situations. FIG. 21 illustrates piping of the cask handling room 50. The piping system includes valves for filling the cask 14 including venting, valves to spray down the annulus and cask 14 with de-mineralized water, and a pressure gauge and level indicator for the cask 14. Double valves are provided so that a failed unit can be isolated. Most of the valves are manual and located outside the fuel hall 34. Those valves and gauges inside the fuel hall 34 that are not accessible when personnel are not allowed in the vicinity are electrically operated. FIG. 22 is an electrical schematic of the control system of the cask handling system 12. The cask handling system control is housed in a floor mounted NEMA 12 enclosure/main control console. The enclosure/main control console contains two PLCs, an operator interface and all video camera controls with LED flat screen monitors. The control system employs two independent PLCs. The first PLC is an Allan Bradley ControLogix PLC or the equivalent and is dedicated to the control and operation of the cask handling system 12. The second PLC is an Alan Bradley dual processor GuardLogic PLC or the equivalent and is used for monitoring all safety related devices and functions. This PLC, when used with safety I/O blocks is safety certified SIL-3 per IEC 61508. Both PLC processors will communicate over an Ethernet/IP network. The operator interface consists of an Allan Bradley Panelview Plus LED touch screen monitor or the equivalent that is in direct communication with the operational PLC over an Ethernet/IP network. This interface is programmed with various operator control screens as well as screens for operational interlocks, fault messages, and troubleshooting aids. All motion is interlocked in the PLC program to assure all operations are performed in the proper sequence. A hard wired safety emergency stop pushbutton is located at each of the three working stations as well as at the remote main control console. When any of the emergency stop buttons are pressed, all motion relating to the cask handling system will stop. The camera system will consist of several strategically placed video cameras for monitoring various cask loading operations and overall cask handling status. Where necessary, cameras will be radiation hardened and incorporate a pan/tilt/zoom feature. Camera joystick controls along with the associated flat panel color viewing monitors are located at the remote man control console. Once inside the fuel hall, the cask transporter is powered via a plugged in power cable and control and control of the cask transporter will be accomplished by means of a control chief radio remote control box. In addition, the main control console PLC will monitor various functions of the on board cask transporter PLC over a connected network communication cable. It is apparent from the above disclosure that the improved cask handling system 12 utilizes a number of innovations to reduce the time to perform the task and significantly reduces the number of components. The sealing process where the cask 12 is interfaced to the spent fuel pool is simplified to allow gravity to help align the system to prevent any leakage. The self-powered mobile cask handling vehicle 16 handles the cask 14 at a number of stations and transports the casks 14 throughout the site. From the foregoing disclosure and detailed description of certain preferred embodiments, it will be apparent that various modifications, additions and other alternative embodiments are possible without departing from the true scope and spirit of the present invention. The embodiments discussed were chosen and described to provide the best illustration of the principles of the present invention and its practical application 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 present invention as determined by the appended claims when interpreted in accordance with the benefit to which they are fairly, legally, and equitably entitled. |
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abstract | Compositions and processes for forming radiopaque polymeric articles are disclosed. In one embodiment, radiation inspection apparatuses and methods are then used to determine the presence and attributes of such radiopaque polymeric articles. A radiopaque polymeric article of the present invention can be created by mixing a radiopaque material, such as barium, bismuth, tungsten or their compounds, with a powdered polymer, pelletized polymer or liquid solution, emulsion or suspension of a polymer in solvent or water. In addition to creating radiation detectable objects, the radiopaque polymeric materials of the present invention can be used to create radiation protective articles, such as radiation protective garments and bomb containment vessels. Enhanced radiation protection can also be achieved through the use of nano-materials. The principals of the present invention can be used to provide protection against other types of hazards, including fire, chemical, biological and projectile hazards. |
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061817627 | abstract | A nuclear fuel bundle has differential peak power limits for the edge or peripheral fuel rods and the interior fuel rods. Also, the magnitude of nuclear fuel in the edge or peripheral rods is decreased in comparison with the magnitude of nuclear fuel in the interior rods. The nuclear reactor can thus operate at higher power output by decreasing the margins between the power outputs and peak power limit of the interior rods, as well as by decreasing the margins between the power outputs of the edge or peripheral rods and the increased peak power limit of those rods. The outer or peripheral edge rods can also be enriched for enhanced power output. |
description | This application claims the benefit of U.S. Provisional Patent Application No. 60/737,018, filed on Nov. 14, 2005, the disclosure of which is hereby incorporated by reference in its entirety and for all purposes. The present invention relates generally to structural health monitoring. More specifically, the present invention relates to a method and apparatus for switching among sensing elements in a structural health monitoring system. The diagnostics and monitoring of structures, such as that carried out in the structural health monitoring field, are often accomplished by employing arrays of monitoring elements such as sensors, actuators, and/or transducers. While many advances have been made, the field continues to face challenges. For example, such arrays often require large numbers of monitoring elements in order to be effective, as structures often must have a variety of sensing elements placed at various locations for accurate monitoring. Because individual sensing elements must often be placed separately, affixing a large array of such sensing elements can be tedious and time consuming. In addition, as each individual sensing element can require one or, commonly, multiple wires, large arrays of sensing elements can require a large number of individual wires, which may be difficult to handle and keep track of. The securing of such large numbers of wires can often be painstaking and time consuming, as well. It is therefore desirable to reduce the number of wires used in arrays of structural health monitoring elements. The invention can be implemented in numerous ways, including as a method, apparatus, or computer readable medium. Several embodiments of the invention are discussed below. In one embodiment of the invention, a structural health monitoring system comprises a plurality of monitoring elements configured for coupling to a structure, and a plurality of switches each in electrical communication with an associated monitoring element of the plurality of monitoring elements, and each configured to switch the associated monitoring element to an on state and an off state. Also included are a control line configured for coupling to the structure, and in electrical communication with each switch of the plurality of switches, a signal line configured for coupling to the structure, and in electrical communication with each monitoring element of the plurality of monitoring elements, and a controller. The controller is configured to transmit a control signal along the control line so as to switch selected ones of the monitoring elements to the on state or the off state, and configured to perform at least one of transmitting a monitoring signal along the signal line to those monitoring elements switched to the on state so as to initiate a monitoring of the structure, or receiving a sensing signal along the signal line from those monitoring elements switched to the on state so as to facilitate the sensing of the structure. In another embodiment of the invention, and in a structural health monitoring system including a plurality of monitoring elements configured for coupling to a structure, the monitoring elements each having an on state and an off state, a method of monitoring the health of a structure comprises selecting ones of the monitoring elements, and switching the selected ones of the monitoring elements to the on state. Also included is at least one of transmitting a monitoring signal to those monitoring elements switched to the on state so as to initiate a monitoring of the structure, or receiving a sensing signal from those monitoring elements switched to the on state so as to facilitate the sensing of the structure. In another embodiment of the invention, and in a computer readable medium having computer executable instructions thereon for a method of monitoring the health of a structure in a structural health monitoring system including a plurality of monitoring elements configured for coupling to the structure, the monitoring elements each having an on state and an off state, the method comprises selecting ones of the monitoring elements and switching the selected ones of the monitoring elements to the on state. Also included is at least one of transmitting a monitoring signal to those monitoring elements switched to the on state so as to initiate a monitoring of the structure, or receiving a sensing signal from those monitoring elements switched to the on state so as to facilitate the sensing of the structure. Other aspects and advantages of the invention will become apparent from the following detailed description taken in conjunction with the accompanying drawings which illustrate, by way of example, the principles of the invention. Like reference numerals refer to corresponding parts throughout the drawings. Also, it is understood that the depictions in the figures are diagrammatic and not necessarily to scale. In one sense, the invention relates to the use of a single line for switching multiple monitoring elements on/off, and a single line for sending signals to, or receiving signals from, those elements that are switched on. Monitoring elements each have an associated switching element, and each switching element is connected to a common switching line, or control line. A signal from the control line turns each switch on or off. Each monitoring element is also connected to a single signal line, and only those monitoring elements that are turned on can transmit/receive data signals along this signal line. In this manner, even large arrays of monitoring elements need use only two lines: a single control line, and a single signal line. This yields a structural health monitoring system that uses very few wires, and is thus simple and easy to install. Initially, it should be noted that it is often preferable to employ the methods and apparatuses of the invention in conjunction with a flexible sensing layer. More particularly, while the invention is typically carried out as an array of monitoring elements, it is often preferable to affix these monitoring elements, and at least some of their wiring and/or control elements, to a flexible sensing layer that can be attached to a structure. The layer holds the monitoring elements, letting them carry out their monitoring functions upon the structure they are attached to. In this manner, only a single sensing layer need be attached to the structure, rather than a number of individual monitoring elements and their associated circuitry. It should also be noted, however, that the methods and apparatuses of the invention need not necessarily utilize such a flexible layer, but that the invention instead encompasses embodiments in which no layer is used. FIG. 1A illustrates a flexible sensing layer for use in accordance with embodiments of the present invention. A diagnostic layer 100 is shown, which contains an array of sensors 102. The sensors 102 can be sensors capable of receiving signals used in structural health monitoring such as stress waves, and are connected to conductive traces 104. The traces 104 connect (or interconnect, if necessary) sensors 102 to one or more output leads 106 configured for connection to a processor or other device capable of analyzing the data derived from the sensors 102. The diagnostic layer 100 and its operation are further described in U.S. Pat. No. 6,370,964 to Chang et al., which is hereby incorporated by reference in its entirety and for all purposes. Construction of the diagnostic layer 100 is also explained in U.S. patent application Ser. No. 10/873,548, filed on Jun. 21, 2004, which is also incorporated by reference in its entirety and for all purposes. It should be noted that the present invention is not limited to the embodiments disclosed in the aforementioned U.S. patent application Ser. No. 10/873,548, but instead encompasses the use of flexible sensor layers having any configuration. For illustration, FIG. 1B further describes aspects of the operation of the diagnostic layer 100. In operation, the output leads 106 are electrically connected to an analysis unit such as a microprocessor 108, suitable for analyzing signals from the sensors 102. In certain embodiments, the flexible layer 100 is first attached to a structure in a manner that allows the sensing elements 102 to detect quantities related to the health of the structure. For instance, the sensors 102 can be sensors configured to detect stress waves propagated within the structure, and emit electrical signals accordingly. The microprocessor 108 then analyzes these electrical signals to assess various aspects of the health of the structure. For instance, detected stress waves can be analyzed to detect crack propagation within the structure, delamination within composite structures, or the likelihood of fatigue-related failure. Quantities such as these can then be displayed to the user via display 110. In one embodiment, the sensors 102 can be piezoelectric transducers capable of reacting to a propagating stress wave by generating a voltage signal. Analysis of these signals highlights properties of the stress wave, such as its magnitude, propagation speed, frequency components, and the like. Such properties are known to be useful in structural health monitoring. FIG. 1C illustrates a circuit diagram representation of such an embodiment. This embodiment can often be represented as a circuit 112, where each sensor 102 is represented as a voltage source 114 in series with a capacitor 116 (impedance circuitry) used to adjust signal strength. This pair is in electrical contact with a data acquisition unit 118, such as a known data acquisition card employed by microprocessors 108 (the data acquisition unit 118 can be thought of as a component interface to the microprocessor 108). Propagating stress waves induce the sensor 102 to emit a voltage signal that is recorded by the data acquisition unit 118, where it can be analyzed to determine the health of the structure in question. As discussed below, these piezoelectric transducers can also act as actuators, converting an applied voltage to a stress wave signal. In another embodiment, the sensors 102 can be known fiber optic sensors that convert stress waves to optical signals. FIG. 2A illustrates further details of a sensing layer 100. It should be noted that the invention includes sensing layers 100 configured in any number of ways. For instance, the sensors 102 can be distributed in any manner throughout the layer 100. Here, six such sensors 102 are shown regularly distributed in a two-dimensional array, each with a single trace 104 extending to the contacts 200. However, one of skill will observe that the sensors 102, traces 104, and contacts 200 can be distributed in any manner, and in any number, without departing from the scope of the invention. For example, the sensors 102 can also be configured in a one-dimensional array such as that shown in FIG. 2B. Here, instead of two rows of sensors 102, a single row is employed. Such a one-dimensional array finds uses in, for example, the monitoring of areas too narrow to fit a two-dimensional array. In the following description, one of ordinary skill in the art will observe that the systems and methods described can be implemented in conjunction with a flexible layer described above, and can also be implemented without one. As one example, FIG. 3 illustrates a system employing an array of actuators, a single control line, and a single signal line. Here, the sensor system 300 includes a controller 302 connected to a control line 304 and a signal line 306. A number of switches 308 are connected to the control line 304 and the signal line 306. Also, each switch 308 has an actuator 310 connected to it. In operation, the actuators 310 are affixed to a structure, and controller 302 transmits a control signal along the control line 304, identifying the actuators 310 it desires to turn on. The control signal switches the corresponding switches 308 to the on state, whereupon the controller 302 can transmit a signal along the signal line 306. As the actuators 310 are each connected to the signal line 306 through their respective switches 308, this signal only reaches those actuators 310 whose switches 308 are turned on. In this manner, a single control line 304 can be used to select actuators 310 for activation, after which a single signal line 306 can be used to transmit monitoring signals to only those actuators 310 that have been selected. It can be seen that different actuators 310 can be turned on and off in this same manner as necessary, so as to perform structural health monitoring operations on different parts of the structure at different times while utilizing only a single pair of wires. The actuators 310 and signals can be any elements, and their corresponding signals, employed in monitoring the health of a structure. In one embodiment, the actuators 310 are piezoelectric transducers that convert electrical signals to stress waves that propagate within the structure they are attached to, and likewise also convert received stress waves to electrical signals. In this embodiment, the signals transmitted along the signal line 306 correspond to a stress wave having a desired profile. The transducers whose switches 308 are turned on would thus receive this signal, and convert it to stress waves within the structure. These stress waves could then be picked up, perhaps by other transducers whose switches 308 are also in the on state. The detected stress waves are then converted by these transducers back into an electrical signal that is sent along the signal line 308 to the controller 302 for processing. One of ordinary skill in the art will observe that the invention can be employed in sensing, as well as actuation. FIG. 4 illustrates a system employing an array of sensors, a single control line, and a single signal line. Here, the system 400 includes a controller 402, connected to a control line 404 and a signal line 406. A number of switches 408 are connected to the control line 404 and the signal line 406. Also, each switch 408 has a sensor 410 connected to it. The system 400 is configured similar to the system 300 of FIG. 3, except that the system 400 employs sensors 410 instead of actuators 310. In operation, the controller 402 sends a control signal along its control line 404, switching the switches 408 of desired sensors 410 to their on state. The sensors 410 whose switches 408 are switched on can thus monitor the structure they are attached to, and send corresponding electrical signals along the signal line 406 to the controller 402. In this manner, a network of sensors 410 can be controlled using only a signal line 406 and control line 404, instead of one or more wires for each individual sensor 410. The actuators and sensors can also be connected to separate networks, as shown in FIG. 5. Here, the actuator system 300 of FIG. 3 is employed along with a sensing system 500, which includes a controller 502 connected to a signal line 504 that has sensing elements 506. In the embodiment shown, the sensing elements 506 shown are known fiber optic sensors, and the signal line 504 is a fiber optic line, however the invention includes the use of any sensors. In operation, the system 300 and the system 500 are attached to the same structure. The system 300 can then be employed to generate stress waves in the structure as above, where the stress waves are detected by the sensing elements 506 of the sensing system 500. Similarly, as shown in FIG. 6, the invention can employ an actuator system 300 and a sensing system 400 such as that desexibed in FIG. 4. Here, the actuator system 300 can transmit stress waves through a structure as described above, and these stress waves can be detected by the sensors 410 of the sensing system 400. While FIGS. 3 and 4 illustrate networks of actuators and sensors respectively, it should be recognized that the invention can also employ networks containing both actuators and sensors simultaneously. For example, in system 400 of FIG. 4, the sensors 410 can be either actuators (such as actuators 308) or sensors 410. Those actuators whose switches 408 are in the on state can thus receive signals from the signal line 406 and transmit corresponding stress waves through the structure, whereupon the transmitted stress waves can be detected by sensors 410 whose switches 408 are in the on state. The detected stress waves are then converted to electrical signals and sent back along the signal line 406. It is sometimes also desirable to employ separate signal and control lines for actuators and sensors, i.e., one signal line/control line pair for the actuators, and a separate signal line/control line pair for the sensors. FIG. 7 illustrates such a configuration. Here, a sensor/actuator system 700 has a controller 702 connected to a first control line 704 and a first signal line 706. These lines 704, 706 are connected to switches 708, which are in turn connected to actuators 710. Similarly, the controller 702 is also connected to a second control line 712 and second signal line 714. The lines 712, 714 are connected to switches 716, which are connected to sensors 718. Typically, the actuators 710 and sensors 718 are attached to a structure at differing locations. As above, the controller 702 can then switch on certain actuators 710 via the first control line 704 and switches 708, and can then transmit signals along the first signal line 706 to these actuators 710, thus generating a stress wave in the structure. Simultaneously, the controller 702 can also switch on certain sensors 718 via the second control line 712 and switches 716. These sensors 718 can detect the stress wave after it has propagated through the structure, converting the sensed wave to electrical signals and transmitting these signals along the second signal line 714 for processing. In embodiments employing transducers or other combined sensor/actuators, the same monitoring elements can act as both actuators and sensors, while still being controlled by a single switch. FIG. 8 illustrates a system 800 employing an array of transducers, a single control line, and a single signal line. Here, a controller 802 has a first control line 804 and a first signal line 806, as well as a second control line 808 and second signal line 810. Each of these lines 804-810 are connected to switches 812, which are each connected to a transducer 814. In operation, the controller 802 can transmit control signals along the first control line 804 to switch selected transducers 814 on/off for actuation, and can then transmit signals along the first signal line 806 to cause those transducers to generate stress waves in the structure being monitored. The controller 802 can also transmit control signals along the second control line 808 to switch other transducers 814 on/off for sensing. These transducers 814 then sense stress waves, convert them to electrical signals, and transmit them along the second signal line 810 back to the controller 802. Configurations like this one allow a network of transducers to be controlled by only four lines 804-810, rather than one or more lines for each transducer 814. While the switches described above have been shown as switching a single monitoring element on/off, it should be recognized that the invention can also employ switches capable of switching multiple monitoring elements. FIG. 9 illustrates an embodiment employing such switches. System 900 includes a controller 902 with a control line 904 and signal line 906. These lines 904-906 are connected to switches 908, which each switch a number of monitoring elements 910 to their on/off states. Control signals sent from the controller 902 along the control line 904 identify certain monitoring elements 910 for switching on/off. The switches 908 then switch these monitoring elements 910 (which can each be actuators, sensors, transducers, or the like) on or off accordingly. If actuation is desired, the controller 902 then transmits signals to those elements 910 that are switched on, so as to cause them to generate stress waves or other diagnostic signals. If sensing is desired, those elements 910 that are switched on can monitor the structure they are attached to, and transmit signals along the signal line 906 for detection by the controller 902. One of ordinary skill in the art will realize that the invention is not limited to the various embodiments described above. For example, a flexible layer can be employed for convenience, but need not necessarily be used. Various components can also be attached to the flexible layer, or left off, i.e., only the various monitoring elements and lines can be attached to the layer, or their associated controllers or other circuitry can be attached as well. Also, the monitoring elements described above can be piezoelectric transducers, fiber optic sensors, or any other sensors, actuators, or sensor/actuators employed in structural health monitoring. The various lines can also be bundled into single cables for further convenience. For instance, in FIG. 8, all four lines 804-810 can be bundled into a single cable. The foregoing description, for purposes of explanation, used specific nomenclature to provide a thorough understanding of the invention. However, it will be apparent to one skilled in the art that the specific details are not required in order to practice the invention. Thus, the foregoing descriptions of specific embodiments of the present invention are presented for purposes of illustration and description. They are not intended to be exhaustive or to limit the invention to the precise forms disclosed. Many modifications and variations are possible in view of the above teachings. For example, as described above, sensor networks and actuator networks can be utilized separately or in conjunction with each other, and a single network can employ both sensors and actuators. Also, networks can utilize any suitable components as monitoring elements, such as piezoelectric transducers, fiber optic sensors, and the like. The embodiments were chosen and described in order to best explain the principles of the invention and its practical applications, to thereby enable others skilled in the art to best utilize the invention and various embodiments with various modifications as are suited to the particular use contemplated. |
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claims | 1. A shield structure configured to protect a head and/or neck of a patient during a radiologic procedure, comprising:a bottom shield that includes radiation attenuating material and that is configured to be positioned between the head and/or neck of the patient and a radiation source so as to shield the patient from radiation directed toward the bottom of the patient, wherein the bottom shield is of a general size to shield the head and/or neck of the patient;a side shield that includes radiation attenuating material and that is configured to extend upward relative to the bottom shield so as to shield the patient from radiation directed toward a side of the patient, the side shield comprising a front portion configured to be disposed proximal to the top of the head of the patient, a first side portion configured to be disposed proximal to a first side of the head of the patient, and a second side portion configured to be disposed proximal to a second side of the head of the patient;a first opening configured to receive the head and/or neck of the patient; anda second opening disposed in opposing relationship to the bottom shield and configured to allow access to the patient by extending from the front portion to the first opening. 2. The shield structure of claim 1, wherein the bottom shield and the side shield are a continuous, unitary wall structure. 3. The shield structure of claim 1, wherein the bottom shield and the side shield are separate structures. 4. The shield structure of claim 1, wherein the front portion, the first side portion, and the second side portion are a continuous, unitary structure. 5. The shield structure of claim 1, wherein the front portion, the first side portion, and the second side portion are separate structures. 6. The shield structure of claim 1, wherein the side shield includes a movable portion that is movable substantially vertically relative to the bottom shield between a retracted position and an extended position, wherein the side shield provides greater shielding to the patient when the movable portion is in the extended position. 7. The shield structure of claim 1, wherein at least one of the bottom shield and the side shield include a shielding layer that includes the radiation attenuating material, and a first structural layer that supports and at least partially covers the shielding layer. 8. The shield structure of claim 7, wherein at least one of the bottom shield and the side shield further includes a second structural layer that supports and at least partially covers the shielding layer. 9. The shield structure of claim 7, wherein the radiation attenuating material includes lead. 10. The shield structure of claim 7, wherein the first structural layer is formed of a material that provides a resilient barrier to, and that will not be denatured by, EPA-registered hospital disinfectants. 11. The shield structure of claim 10, wherein the first structural layer includes carbon fiber. 12. The shield structure of claim 1, further comprising a first skirt disposed on a first side of the bottom shield and that hangs downward relative to the bottom shield, and a second skirt disposed on a second side of the bottom shield and that hangs downward relative to the bottom shield, wherein each of the first and second skirts includes radiation attenuating material. 13. A method of protecting a head and/or neck of a patient during a radiologic procedure, comprising:positioning the head and/or neck of the patient in a shield structure, wherein the shield structure has a bottom shield that includes radiation attenuating material and is positioned between the head and/or neck of the patient and a radiation source, and a side shield that includes radiation attenuating material and extends upward relative to the bottom shield, wherein the side shield comprises a front portion that is disposed proximal to the top of the head of the patient, a first side portion that is disposed proximal to a first side of the head of the patient, and a second side portion that is disposed proximal to a second side of the head of the patient, wherein the head and/or neck of the patient is received in a first opening in the shield structure, and access to the patient is allowed by a second opening disposed in opposing relationship to the bottom shield and configured to extend from the front portion to the first opening; andexposing the patient to radiation to conduct the radiologic procedure. 14. The method of claim 13, further comprising extending a movable portion of the side shield from a retracted position to an extended position to provide greater shielding to the patient during the radiologic procedure. 15. The method of claim 13, further comprising disposing a body part of medical personnel between a first skirt disposed on a first side of the bottom shield and that hangs downward relative to the bottom shield, and a second skirt disposed on a second side of the bottom shield and that hangs downward relative to the bottom shield, wherein each of the first and second skirts includes radiation attenuating material. 16. A shield structure configured to protect a head and/or neck of a patient during a radiologic procedure, comprising:a bottom shield that includes radiation attenuating material and that is configured to be positioned between the head and/or neck of the patient and a radiation source so as to shield the patient from radiation directed toward the bottom of the patient, wherein the bottom shield is of a general size to shield the head and/or neck of the patient;a side shield that includes radiation attenuating material and that is configured to extend upward relative to the bottom shield so as to shield the patient from radiation directed toward a side of the patient; andan opening configured to receive the head and neck of the patient,wherein the side shield has a length such that an area immediately below the neck of the patient remains exposed from the shield structure. 17. The shield structure of claim 16, wherein the bottom shield and the side shield are a continuous, unitary wall structure. 18. The shield structure of claim 16, wherein the bottom shield and the side shield are separate structures. 19. The shield structure of claim 16, further comprising a first skirt disposed on a first side of the bottom shield and that hangs downward relative to the bottom shield, and a second skirt disposed on a second side of the bottom shield and that hangs downward relative to the bottom shield, wherein each of the first and second skirts includes radiation attenuating material. |
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claims | 1. A dry storage canister that stores spent nuclear fuel rods, comprising:an elongated outer housing, the elongated outer housing extending from a first end to a second end; andan elongated internal basket provided within the housing, the elongated internal basket extending in a region between the first end and the second end of the elongated outer housing, the internal basket defining multiple elongated tubes forming multiple discrete storage cells, each of the cells and comprising a plurality of the spent nuclear fuel rods that have been separated from their fuel rod assemblies; andwherein the canister has no neutron absorber material. 2. The canister of claim 1, wherein the outer housing is an elongated cylindrical housing. 3. The canister of claim 1, wherein the storage cells are configured such that a rod packing density of approximately 4 to 6 of the spent fuel rods per square inch can be achieved so that the canister is in less danger of reaching nuclear criticality if a neutron moderator were to enter the canister and so that the nuclear absorber material is not needed in the canister. 4. The canister of claim 1, wherein the internal basket is made of a metal material having a high thermal conductivity. 5. The canister of claim 1, wherein the internal basket is made of one or more of carbon steel, aluminum, or copper. 6. The canister of claim 1, wherein the tubes are cylindrical tubes. 7. The canister of claim 1, wherein the internal basket further comprises at least one elongated spacer disk that extends between the tubes. 8. The canister of claim 1, further comprising a cask in which the canister is placed. 9. The canister of claim 1, wherein the spent nuclear fuel rods are contiguous within each of the cells. 10. The canister of claim 1, further comprising a corrugated divider within at least one of the cells that separates a plurality of the spent nuclear fuel rods. 11. The canister of claim 1, wherein the internal basket is made of a metal material having a high thermal conductivity. |
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claims | 1. A fuel assembly comprising: at least, a first fuel rod containing plutonium and not containing burnable poison; a plurality of second fuel rods containing uranium and burnable poison and not containing plutonium; a water rod; and a channel box having a rectangular shape in horizontal section and accommodating the first fuel rod, the plurality of second fuel rods, and the water rod in a bundle, whereinthe plurality of second fuel rods is disposed on an outermost periphery and/or adjacent to the water rod, of a fuel rod array in the horizontal section,N2<N1 (N2 is a positive integer or zero) is satisfied where a number of the second fuel rods arranged on the outermost periphery is N1 and a number of the second fuel rods arranged adjacent to the water rod is N2, andof the second fuel rods arranged on the outermost periphery, W2<N2+WO<W1 (W2 is a positive integer) is satisfied where a number of the second fuel rods arranged without being vertically and/or horizontally adjacent to any of other of second fuel rods in the horizontal section is WO, a number of the second fuel rods arranged vertically and/or horizontally adjacent to only one second fuel rod in the horizontal section is W1, and a number of the second fuel rods arranged vertically and/or horizontally adjacent to two second fuel rods in the horizontal section is W2, andthe second fuel rods arranged vertically and/or horizontally adjacent to the two second fuel rods in the horizontal section are arranged at a corner portion of the channel box. 2. The fuel assembly according to claim 1, whereinaverage fissile plutonium enrichment in the horizontal section is 4.0 wt % to 7.8 wt %. 3. The fuel assembly according to claim 2, whereinthe burnable poison contained in the second fuel rod is gadolinium. 4. The fuel assembly according to claim 1, whereinthe burnable poison contained in the second fuel rod is gadolinium. 5. The fuel assembly according to claim 3, whereinthe fuel assembly includes a nine-row by nine-column fuel grid array, andthe first fuel rod includes thirty four fuel rods with fissile plutonium enrichment of 6.1 wt %, twenty fuel rods with fissile plutonium enrichment of 4.2 wt %, five fuel rods with fissile plutonium enrichment of 2.5 wt %, and eight partial length fuel rods with fissile plutonium enrichment of 6.1 wt %. 6. The fuel assembly according to claim 3, whereinthe fuel assembly includes a nine-row by nine-column fuel grid array, andthe first fuel rod includes thirty two fuel rods with fissile plutonium enrichment of 9.3 wt %, twenty one fuel rods with fissile plutonium enrichment of 6.5 wt %, one fuel rod with fissile plutonium enrichment of 3.0 wt %, two fuel rods with fissile plutonium enrichment of 5.5 wt %, four partial length rods with fissile plutonium enrichment of 8.0 wt %, and four partial length fuel rods with fissile plutonium enrichment of 9.3 wt %. 7. The fuel assembly according to claim 3, whereinthe fuel assembly includes a nine-row by nine-column fuel grid array, andthe first fuel rod includes thirty two fuel rods with fissile plutonium enrichment of 10.9 wt %, twenty fuel rods with fissile plutonium enrichment of 7.5 wt %, one fuel rod with fissile plutonium enrichment of 2.5 wt %, three partial length fuel rods with fissile plutonium enrichment of 10.0 wt %, and four partial length fuel rods with fissile plutonium enrichment of 10.9 wt %. 8. The fuel assembly according to claim 3, whereinthe fuel assembly includes a ten-row by ten-column fuel grid array, andthe first fuel rod includes forty fuel rods with fissile plutonium enrichment of 9.3 wt %, twenty four fuel rods with fissile plutonium enrichment of 6.5 wt %, one fuel rod with fissile plutonium enrichment of 3.0 wt %, two fuel rods with fissile plutonium enrichment of 5.5 wt %, eight partial length rods with fissile plutonium enrichment of 8.0 wt %, and six partial length fuel rods with fissile plutonium enrichment of 9.3 wt %. 9. A reactor of a nuclear reactor loaded with a plurality of fuel assemblies, each of the fuel assemblies comprising:at least, a first fuel rod containing plutonium and not containing burnable poison;a plurality of second fuel rods containing uranium and burnable poison and not containing plutonium;a water rod; anda channel box having a rectangular shape in horizontal section and accommodating the first fuel rod, the plurality of second fuel rods, and the water rod in a bundle, whereinthe plurality of second fuel rods is disposed on an outermost periphery and/or adjacent to the water rod, of a fuel rod array in the horizontal section,N2<N1 (N2 is a positive integer or zero) is satisfied wherea number of the second fuel rods arranged on the outermost periphery is N1 anda number of the second fuel rods arranged adjacent to the water rod is N2, andof the second fuel rods arranged on the outermost periphery,W2<N2+WO<W1 (W2 is a positive integer) is satisfied wherea number of the second fuel rods arranged without being vertically and/or horizontally adjacent to any of other second fuel rods in the horizontal section is WO,a number of the second fuel rods arranged vertically and/or horizontally adjacent to only one second fuel rod in the horizontal section is Wi, anda number of the second fuel rods arranged vertically and/or horizontally adjacent to two second fuel rods in the horizontal section is W2, and the second fuel rods arranged vertically and/or horizontally adjacent to the two second fuel rods in the horizontal section are arranged at a corner portion of the channel box. 10. The reactor according to claim 9, whereinaverage fissile plutonium enrichment in the horizontal section is 4.0 wt % to 7.8 wt %. 11. The reactor according to claim 10, whereinthe burnable poison contained in the second fuel rod is gadolinium. 12. The reactor according to claim 9, whereinthe burnable poison contained in the second fuel rod is gadolinium. |
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claims | 1. A reactor vessel handling method comprising the steps of removing a polar crane mounted on an annular rail in a reactor containment vessel of a pressurized water reactor and having a girder and a trolley, and then carrying out a reactor vessel through an opening provided in a top portion of said reactor containment vessel. 2. A reactor vessel handling method according to claim 1 , wherein said reactor vessel is carried out together with core internals. claim 1 3. A reactor vessel handling method according to claim 1 , wherein said reactor vessel is carried out in a state in which a radiation shield case is combined with said reactor vessel. claim 1 4. A reactor vessel handling method according to claim 1 , wherein carrying-out of said reactor vessel or carrying-in of said new reactor vessel is performed using a self-propelled heavy-duty crane. claim 1 5. A reactor vessel handling method comprising the steps of removing a polar crane mounted on an annular rail in a reactor containment vessel of a pressurized water reactor and having a girder and a trolley, and carrying out a reactor vessel through an opening provided in a top portion of said reactor containment vessel; and then carrying in a new reactor vessel to a predetermined position within said reactor containment vessel through said opening. 6. A reactor vessel handling method comprising the steps of, in a state in which a polar crane mounted on an annular rail in a reactor containment vessel of a pressurized water reactor and having a girder and a trolley, is removed, carrying in a new reactor vessel to a predetermined position within said reactor containment vessel through an opening provided in a top portion of said reactor containment vessel. 7. A reactor vessel handling method according to claim 6 , wherein after carrying in said new reactor vessel, said polar crane is restored to an original state. claim 6 |
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043839690 | abstract | Removal of small amounts of .sup.14 CO.sub.2, .sup.14 CO and corresponding alkanes produced in nuclear power plants from the exhaust gases of the purification plants by conversion of the radioactive compounds into .sup.14 CO.sub.2, and removing this .sup.14 CO.sub.2 from the main gas stream of exhaust gases. |
summary | ||
053848136 | summary | BACKGROUND OF THE INVENTION The invention relates to storage racks for storing nuclear fuel assemblies both during transport and during stationary storage. Preferably, the racks are highly overdamped, enabling them to best withstand vibrations caused by seismic events or rough handling. Fuel for nuclear reactors is typically configured in the form of elongated fuel rods, which may be separate, stand-alone elements, or may be positioned within canisters. Hereinafter, the fuel rods and rod/canister combinations are referred to as fuel assemblies. Both before and following use, the fuel assemblies must be stored and/or transported with great care. To assure that such care is achieved, storage racks are often used to support a plurality of fuel assemblies in a generally parallel, spaced-apart configuration, while maintaining the fuel assemblies in a subcritical array environment. During storage, the racks and the fuel assemblies contained therein, may be completely submerged in a pool of water. The water provides cooling and additional shielding from nuclear radiation. The fuel storage racks of the prior art typically consist of an assembly of hollow cells, each defined by an array of elongated rectangular cross-section boxes or compartments. The boxes are typically made by forming sheets of stainless steel into elongated rectangular cross-section tubes and welding the corners of the elongated tubes together to form a matrix of elongated hollow cells, each adapted the receive a single fuel assembly. Exemplary storage racks are disclosed in U.S. Pat. Nos. 4,695,424, 4,857,263, 4,948,553, and 4,366,115. A neutron absorbing (or "poison") material, such as borated stainless steel, is typically welded or otherwise rigidly affixed to each of the walls of boxes to absorb neutron flux from the fuel assemblies which may be positioned within the boxes, thereby avoiding an undesirable concentration of neutrons. These prior art storage racks suffer from several disadvantages. For example, neutron absorbing elements, and particularly those made borated stainless steel, are expensive and difficult to form and weld to the walls of the boxes. Further, the individual cells are known to be weak along the top edge and have little torsional or crush strength. In addition, storage racks constructed in this way have little resistance to vibration, such as may be caused by seismic events. Due to the reactive nature of the nuclear fuel assemblies, such damage to the storage racks can be disastrous. Accordingly, it is an object of the present invention to provide an improved storage and/or transport rack for nuclear fuel assemblies. Another object of the present invention is to provide an improved storage rack for nuclear fuel assemblies which is highly overdamped to enable the rack to withstand the vibration of seismic events or rough handling such as may be encountered during transportation of the rack. It is another object to provide a storage rack for nuclear fuel assemblies which has improved torsional and crush strength. A further object is to provide an improved storage rack for nuclear fuel assemblies which may De easily and inexpensively manufactured. Other objects of the invention will in part be obvious and will in part appear hereinafter. SUMMARY OF THE INVENTION According to the present invention, a rack structure is provided for long term storage and/or transport of nuclear fuel assemblies. The storage rack includes an array of individual storage cells. The cells of the array are defined by a plurality of substantially polygonal cross-section, elongated cell housings, each extending along an elongated central axis, wherein the central axes are substantially parallel to each other. In accord with an important aspect of the invention, a slab of neutron absorbing (or "poison") material is biased against the outer surfaces of the cell housings. Preferably, the cell housings are positioned in alternate points of a rectangular grid configuration, so that each cell housing defines one cell in its interior and so that the outer walls of three or more adjacent cell housings define one cell. A stiffener wall may be welded to the adjacent cell housings along the perimeter of the rack to enclose the open cells along the perimeter. The cell housings and the stiffener walls are held in parallel alignment by support bars affixed thereto, for example by welding, at both the top and bottom ends of the array of cell housings. Preferably, the support bars are positioned at the top and bottom ends between each row of cell housings and along the outer perimeter of the rack. The support bars may be recessed on one side or on alternating sides to provide positioning of the cell housings prior to affixation. A base plate is affixed to bottom of the array to define the lower boundary of the respective cells and to support the fuel assemblies therein. To facilitate water flow for cooling of the nuclear fuel assemblies, the base plate may include holes at positions within each cell. Pedestals extending from the base plate may be used to raise the rack above a floor. With this configuration, a new or spent fuel assembly may be placed in each of the cells. When in place, neutron flux from the fuel assemblies is absorbed by the poison material on the cell housing walls. Retaining devices, or clamp assemblies, hold the poison slabs in position, while preloading (i.e. forcing) the slabs against the walls, and permit easy assembly of the rack without requiring welding of the poison material to the housings. The retaining devices press the slabs firmly against the walls of the cell housings. The resulting friction between the slab and cell walls results in a coulomb damping function that has proven extremely effective in deadening vibration. The slabs also serve to strengthen the cell walls making them more resistant to deformation or "oil canning". Thus, the slabs, as held in place by the preload forces, preferably establish an overdamped characteristic for the cell housings, resulting in a substantially stronger, vibration resistant configuration as compared with the prior art. In the preferred form of the invention the slabs, which are preloaded against the cell walls, are made of a neutron absorbing material. Alternatively, the invention may be configured with other materials that are preloaded against the cell walls, which merely provide the coulomb damping function. In the latter configurations, other forms of neutron absorption may or may not be used. Alternatively, the retaining device may include a single cover plate which extends over and protects the entire poison slab. The cover plate is provided with flanges along the perimeter which are fixed to the housing to preload or force the slab against the outer surface. Additionally, the cover plate may be provided with raised bumps or ridges which bear on the poison slab and further preload it against the housing in a substantially uniform manner. In view of the potential danger inherent in handling and storing nuclear fuel assemblies, it is critical that the storage racks effectively isolate and support the nuclear fuel assemblies under adverse conditions. In use, the racks are subject to stresses resulting from normal insertion and removal of the nuclear fuel assemblies, rough handling during transportation of the rack and possibly natural phenomena such as an earthquake. By providing the rack with a highly overdamped characteristic, it is better able to withstand these stresses and insure long term stability, as compared to prior art racks. Typically, in prior art storage racks, individual cells are welded together at the corners of their respective cell housings. This method has proven to be undesirable not only because is it difficult and therefore expensive to manufacture but also because it has proven to be structurally inferior. In contrast, by welding the cell housings to support bars at the top and bottom of the rack, and with little or no other cell-to-cell fixation, the present invention avoids both the difficulty and expense of the prior art process and provides improved torsion and crush strength. The support bars also reinforce the upper edges of the cell housing to protect them from damage during normal insertion and removal of the nuclear fuel assemblies. |
summary | ||
abstract | A device (1) for treating a packaging material (2) by means of UV radiation, the device having a source (6) of such radiation, and a protective screen (8) interposed between the source (6) and the material (2) for treatment; the screen (8) has a film (9) of a polymer resistant and permeable to UV radiation, and which is flexible, does not tear, and does not form fragments. |
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051805448 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS Referring to FIG. 1 which is a perspective view of a first embodiment of the control blade of the invention for use in nuclear reactors, the control blade generally denoted by a numeral 10 has a central tie rod 14 interconnecting an upper structure 12 provided with a handle 11 and a lower structure 13. The central tie rod 14 has radial projections which provide a substantially cross-shaped cross-section of the central tie rod 14. A substantially U-shaped sheath plate having a considerable depth and made of stainless steel is secured to the end of each projection of the central tie rod 14. The space in each sheath plate 15 receives plate-shaped long-life neutron absorber 18 made of hafnium (Hf). Each sheath 15 and the long-life neutron absorber 18 in combination constitute a wing 16 of the control blade 10. Thus, the control blade 10 has four wings 16. The control blade 10 is designed to have substantially the same size, shape and weight as those of conventional control blades charged with boron carbide (B.sub.4 C), so that it can be back-fitted in existing nuclear reactors. For instance, the control blade 10 has an effective length of about 3.83 m, a blade width of about 250 mm, a blade thickness of about 8 mm, a sheath plate thickness of about 1 mm and a total weight of about 100 kg. The neutron absorber 18 is divided along the axis of the tie rod 14 into a plurality of elements or sections, e.g., four neutron absorber elements or sections 18a, 18b, 18c and 18d as shown in FIG. 2. In FIG. 2, the left half part of the control blade 10 is loaded with the neutron absorber elements, while the right half part is shown in a neutron absorber elements 18a, 18d and 18c other than the element 18d adjacent to the lower structure 13 are supported by absorber element supports 20 which are formed on each projection of the central tie rod 20 at a suitable interval in the direction of axis of the central tie rod 14 so as to prevent the neutron absorber elements 18a to 18c from moving up and down. The neutron absorber elements 18a to 18d are so designed that the neutron absorber 18 composed of these elements exhibits neutron absorption characteristics which are progressively decreased from the end adjacent to the upper structure 12 towards the end adjacent to the lower structure 13. More particularly, in the illustrated embodiment in which the neutron absorber 18 is divided into four elements 18a to 18d, each element has a constant thickness but the thicknesses are changed in a stepped manner such that the uppermost neutron absorber element 18a adjacent to the upper structure has the greatest thickness and the lowermost neutron absorber element adjacent to the lower structure has the smallest thickness. This stepped change in the thickness of the neutron absorber 18 causes a correspondingly stepped change in the reactivity worth, i.e., the neutron absorption characteristics, as shown in FIG. 3. The design may be such that, depending on the design or the manner of operation of the control blade, the extreme end portion of the uppermost neutron absorber element 18a adjacent to the upper structure, e.g., the region within 35 cm as measured from the end extremity, has specifically increased neutron absorption characteristics so as to improve the scramming performance of the reactor or specifically decreased neutron absorption characteristic so as to suppress any drastic variation of the reactor output which may be caused when the control blade is extracted. In addition, the neutron absorption characteristics are so varied in at least the uppermost neutron absorber element 18a such that the portion of the element adjacent to the central tie rod 14 has greater neutron absorption capacity. In general, a long-life control blade 10 used in a nuclear reactor tends to suffer from embrittlement of the upper structure 12 because of an extremely heavy neutron exposure of the upper structure. Therefore, the upper structure is usually made of a stainless steel having a specifically high purity, so as to suppress any tendency for the upper structure to become fragile. In order to minimize the weight of the control blade, the upper structure 12, the lower structure 13 and a speed limiter 22 attached to the lower structure have thicknesses reduced as possible. As will be seen from FIGS. 1 and 2, the control blade 10 has a vacant portion 23 below the upper structure 12. This vacant portion 23 may be utilized as an auxiliary handle. The vacant portion 23 is formed at such a position where no neutron absorption is necessary due to the design of the control blade. The provision of the vacant portion 23 contributes to a further reduction in the weight of the control blade. It has been confirmed through experiments that the amount of fast neutron exposure at the upper part of the auxiliary handle is as small as 1/5 to 1/3 of that at the upper part of the handle portion. This suggests that the degree of embrittlement at the auxiliary handle portion 23 is as small as 1/5 to 1/3 of that at the upper part of the handle portion, so that the provision of the auxiliary handle portion 23 provides an effective back-up for the handle 11 during handling of the control blade. Each of the neutron absorber elements, e.g., element 18a, disposed in the sheath plate 15 is composed of a pair of neutron absorber plates or sheets 18a.sub.1 and 18a.sub.2 made of hafnium films or sheets and arranged so as to oppose each other as shown in FIG. 4. These neutron absorber plates 18a.sub.1 and 18a.sub.2 are spaced from each other by spot-like spacers 24. These spacers 24 improve the mechanical strength of the neutron absorber element 18a and preserve a flat water gap or space 25 between the opposing neutron absorber plates 18a.sub. 1 and 18a.sub.2 for allowing a moderator to flow therethrough. A plurality of water passage holes 26 communicating with the water gap 25 are formed in the walls of the sheath plate 15 and the corresponding portions of the neutron absorber element 18a. The water passage holes 26 as a rule are not formed in such a way as to penetrate the wing 16 linearly. In other words, these holes 26 are formed in a staggered manner. Each of the neutron absorber plate in each of the elements 18a to 18d has the form of a thin plate or sheet of 0.5 to 2.0 mm thick and is curved at its edge extending along the end of the wing 16. A small gap is formed between the curved end extremities of the pair of neutron absorber plates 18a.sub.1 and 18a.sub.2 at the end of the wing 16, in order to ensure sufficient flexibility of these neutron absorber plates 18a.sub.1 and 18a.sub.2. The neutron absorber 18 incorporated in the nuclear reactor control blade 10 of the invention may be sectioned in the axial direction of the tie rod 14 into eight stages or elements 18a to 18h, as shown in FIG. 5. The neutron absorber supporting element of each stage is supported by a plurality of supporting spacers 30 which are fixed to the sheath plate 15 at a suitable interval as shown in FIG. 5, thereby preventing the neutron absorber elements 18a to 18h from moving up and down. The neutron absorber elements 18a to 18h are so designed that the neutron absorber 18 composed of these elements exhibits neutron absorption characteristics which are progressively decreased from the end adjacent to the upper structure 12 towards the end adjacent to the lower structure 13. More particularly, in this embodiment in which the neutron absorber 18 is divided into eight elements 18a to 18h, each element has a constant thickness but the thickness is changed in a stepped manner such that the uppermost neutron absorber element 18a adjacent to the upper structure has the greatest thickness and the lowermost neutron absorber element 18h adjacent to the lower structure has the smallest thickness. This stepped change in the thickness of the neutron absorber 18 causes a correspondingly stepped change in the reactivity worth, i.e., the neutron absorption characteristics, as shown in FIG. 6A. In the arrangement shown in FIG. 6A, all the neutron absorber elements 18a to 18h have different thicknesses such that the neutron absorber 18 as a whole exhibits a thickness distribution which progressively decreases towards the end adjacent to the lower structure. This, however, is not exclusive and the thickness distribution may be such that a plurality of adjacent neutron absorber elements, e.g., two elements as shown in FIG. 6B, have an identical thickness, or such that each of the neutron absorption elements has the greatest thickness at its end adjacent to the upper structure 12 and the smallest thickness at its end adjacent to the lower structure 13, so that the neutron absorber 18 as a whole exhibits a substantially linear or rectilinear change in the thickness as shown in FIG. 6C. It is also possible to use the arrangements shown in FIGS., 6A to 6C in combination. As shown in FIG. 8, the neutron absorber elements 18a to 18h of the respective stages have pairs of neutron absorber sheets or plates 18a.sub.1, 18a.sub.2 ; 18b.sub.1, 18b.sub.2 ; . . . ; 18h.sub.1, 18h.sub.2 constituted by sheets of hafnium, the neutron absorber plates of each pair being arranged to oppose each other in the direction of thickness of the wing 16. These neutron absorber plates of each pair are spaced from each other by supporting spacers 30 which has, as shown in FIG. 7, a disk-like spacing portion 30a and supporting legs 30b projecting axially from the center of the spacing portion 30a at both sides thereof. As will be seen in FIG. 8, the supporting legs 30b loosely penetrate corresponding holes in the opposing neutron absorber plates 18a.sub.1 and 18a.sub.2 for example and are fixed to the inner wall surfaces of the sheath 15 by, for example, welding. The holes 31 formed in the opposing neutron absorber plates of each neutron absorber element have a diameter which is slightly greater than the supporting legs 30b so as to allow thermal expansion or contraction of the neutron absorber plates caused by a change in the temperature. These supporting spacers 30 securely hold the neutron absorber plates such as the plates 18a.sub.1, 18a.sub.2 within the sheath while preserving a flat water gap 25 between these opposing neutron absorber plates 18a.sub.1, 18a.sub.2 so as to guide the flow of a moderator. Thus, the water gap 25 provides a flow passage for the moderator. A plurality of water passages 26 communicating with the water gap 25 are formed in the walls of the sheath 15 and the corresponding portions of the neutron absorber element of each stage of the neutron absorber 18. As a rule, the water passage holes 26 are arranged such that the wing 16 is not penetrated linearly, i.e., in a staggered or zig-zag manner as shown in FIG. 9. In this embodiment, the thickness of each wing 16 of the nuclear reactor control blade 10 is about 8 mm, and each neutron absorber plate constituting each of the neutron absorber elements 18a to 18h is constituted by a metallic neutron absorber plate having a very small thickness of, for example, 0.5 to 2.0 mm. With this arrangement, as shown in FIG. 6A, the neutron absorber plate in the neutron absorber element adjacent to the upper structure 12 has a thickness of 1.5 to 2.0 mm, while the neutron absorber plate in the neutron absorber element adjacent to the lower structure has a thickness of 0.5 to 1.0 mm. The neutron absorber plates in the intermediate neutron absorber elements have intermediate thicknesses. A critical experiment was conducted by inserting an experimental flux-trap-type control blade into a BWR core simulator. The control blade used in this experiment was composed of pairs of hafnium plates as the neutron absorbers arranged in each sheath made of stainless steel with a water gap formed between opposing hafnium plates, as shown in FIGS. 8 and 9. A relationship between the water gap width and the reactivity worth was confirmed through the experiment, as shown in FIG. 11. This experimental result teaches that a large flux trapping effect is obtained even with a small water gap width of 2 to 5 mm. It will be noted also that a reduction in the thickness of the hafnium plate increases the water gap width correspondingly, thus enabling any reduction in the reactivity worth due to reduction in the plate thickness to be compensated to a certain extent. From these facts, it is understood that a reduction in the weight of the control blade is attainable, while maintaining high reactivity worth, by enlarging the water gap width through reducing the thickness of the hafnium plate at portions of the control blade other than the upper portion where a specifically large hafnium plate thickness is required in view of the neutron exposure distribution and in consideration of the reactor shut down margin. FIGS. 10A and 10B show the portion marked at C in FIG. 5 with the supporting spacers removed. FIG. 10B shows a section taken along the line D--D in FIG. 10A. It will be seen that gaps 33a, 33b are formed between the adjacent neutron absorber plates 18a.sub.1, 18a.sub.2 ; 18b.sub.1 ; 18b.sub.2 ; . . . 18h.sub.1, 18h.sub.2 of the successive neutron absorber elements 18a, 18b, . . . 18h arranged in the axial direction of the control blade 10. It will also be seen that the gaps 33a and the gaps 33b on opposite side of the water gap are staggered such that they are masked by the neutron absorber plates. Namely, the gaps 33a and the gaps 33b formed between the adjacent neutron absorber elements at both sides of the water gap are arranged in a staggered manner such that these gaps are masked by the neutron absorber plates on the opposite sides of the water gap and such that the adjacent gaps on both sides of the water gap do not occupy the same plane of a level. FIGS. 12 to 14 show modifications of the arrangement of neutron absorber plates in adjacent neutron absorber elements. In these modifications, the gaps 34a, 35a, 36a formed in the front side of the wing are disposed so as to intersect, when viewed in the direction normal to the plane of the wing, the gaps 34b, 35b, 36b formed in the rear side of the wing, in such a manner that the areas over which the gaps on the front and rear side cross each other are minimized. By minimizing these areas, it is possible to avoid any local reduction in the reactivity worth along the length of the control blade. Other modifications of the shape and position of the gaps between the adjacent neutron absorber elements will be obvious to those skilled in the art. The operation of the described embodiment of the nuclear reactor control blade is as follows. A curve as shown in FIG. 15 exemplarily shows the fissile nuclide concentration distribution along the axis of a boiling water reactor core in which the fuel has been burnt up to a certain degree Since the control of burn-up in the reactor core is divided into four sections in the direction of axis of the reactor core, it is convenient that the control blade 10 for controlling the burn-up also is divided into four sections or sections of a number which is a multiple of four. The burn-up of fuel is comparatively slow in the lower end portion of the core of the nuclear reactor, so that the concentration of fissile nuclides is large in this portion of the nuclear reactor. Representing the axial length of the reactor core by L, the upper portion above the mid portion 2/4.multidot.L experiences a phenomenon known as hardening of neutron spectrum due to voids generated in this portion. As a result, the plutonium production reaction is promoted in this portion. At the same time, the voids reduce the thermal neutron flux so as to retard the burning of the fuel. For these reasons, the reactor core usually exhibits the fissile nuclide distribution pattern as shown in FIG. 15. Where the fissile nuclide distribution pattern as shown in FIG. 15 is exhibited by the reactor core, the reactor core in the shut-down state shows a neutron multiplication factor distribution along the axis of the reactor core as shown by a curve B in FIG. 16. In general, the greater the neutron multiplication factor, the smaller the reactor shutdown margin, i.e., the smaller the subcriticality factor. The reduction in the multiplication factor at the lower and upper ends of the reactor core as shown by curve B is attributable to leakage of neutrons at these portions of the reactor core. In FIG. 17, a curve C shows the distribution of amounts of the neutron exposure of the nuclear reactor control blade along the axis of the reactor control blade, as observed when the nuclear reactor control blade 10 is used. From this curve, it will be seen that the amount of neutron exposure is drastically increased in a limited region of a certain height (usually about 30 cm) from the upper end extremity of the control blade 10. In other portions of the control blade 10, the amount of neutron exposure is progressively decreased towards the lower end of the control blade 10. The control blade 10 in accordance with the present invention is so designed as to provide a satisfactory control effect under the neutron multiplication factor characteristics and the amount of neutron exposure characteristics shown in FIGS. 15 and 16. Namely, the control blade 10 of the embodiment is so designed that the upper end portion thereof, corresponding to 1/4.multidot.L (about 90 to 95 cm) is designed to cope with the local reduction in shut-down margin attributable to the rise in the neutron multiplication factor, as well as decreasing tendency of the shut-down margin due to drastic increase in the amount of neutron exposure, which is observed in the upper portion of the reactor core as shown in FIGS. 16 and 17. As shown in FIG. 3, the neutron absorber elements are designed such that the neutron absorber as a whole is progressively thinned from the end adjacent to the upper structure 12 towards the end adjacent to the lower structure 13, thus decreasing the neutron absorption effect in a corresponding manner. It is to be noted, however, the neutron absorption power in the region of 1/4.multidot.L from the lower end of the control blade 10, i.e., from the upper end of the lower structure 13, is determined to be slightly smaller than that in the region between 1/4.multidot.L and 2/4.multidot.L, because in the region of 1/4.multidot.L, the neutron multiplication factor is greater than the region between 1/4.multidot.L and 2/4.multidot.L as shown in FIG. 16, though the amount of neutron exposure in the region of 1/4.multidot.L is smaller than that in the region between 1/4.multidot.L and 2/4.multidot.L. FIG. 18 shows a curve D which represents a typical example of the amount of neutron exposure in the breadthwise direction of each wing 16. As will be seen from the curve D, the amount of neutron exposure is drastically increased in the region near the outer end of the wing and is slightly increased in the inner region adjacent to the tie rod 14. It is, therefore, possible to obtain a reactivity worth distribution as shown in FIG. 19, by varying the neutron absorption characteristics of the neutron absorber 18 in the breadthwise direction of the wing 16. In the nuclear reactor control blade 10 of the described embodiment, the above-mentioned variation in the neutron absorption characteristics can be attained by employing thin neutron absorber plates in each of the elements 18a to 18d (FIG. 2) or 18a to 18h (FIG. 5) of the long-life neutron absorber 18 and arranging these neutron absorber plates such that a flat water gap serving as a passage for the moderator is defined between the opposing neutron absorber plates. It is thus possible to reduce the weight of the long-life neutron absorber 18 in the wing 16 as compared with the case where no water gap is formed within the heavy long-life neutron absorber. This in turn contributes to a reduction in the overall weight of the nuclear reactor control blade 10 as a whole, thus enabling existing control rod drive mechanism to serve without any change or modification in design. Other embodiments of the nuclear reactor control blade of the present invention will be described hereinunder. FIGS. 20 to 22 show a second embodiment of the nuclear reactor control blade in accordance with the invention, with means for reinforcement of the wings of the control blade. The control blade denoted by 10A has a plurality of wings each having a plate-like long-life neutron absorber 18A composed of pairs of opposing neutron absorber plates or sheets 38a and 38b which are spaced from and held on each other by means of spot-like spacers 39. These neutron absorber plates are fixed at their end portions corresponding to the outer end of the wing to a common tie rod 40, thus assuring high mechanical strength or stability. In general, the outer end portions of the neutron absorber plates 38a, 38b receive a greater amount of neutron exposure as compared with other portions. The tie rod 40 provided between these end portions of the neutron absorber plates effectively increases the reactivity. The other ends of the neutron absorber plates 38a and 38b, i.e., the facing ends adjacent to the central tie rod 14, are curved to approach each other but are spaced from each other so as to absorb any thermal expansion of the neutron absorber plates 38a and 38b. Other portions of this embodiment are materially the same as those of the embodiment shown in FIGS. 2 and 4 and, therefore, are denoted by the same reference numerals and a detailed description thereof is omitted. FIG. 23 illustrates a third embodiment of the nuclear reactor control blade in accordance with the present invention. The control blade, generally designated by a numeral 10B, has a plurality of wings each of which is composed of a sheath 15 and a neutron absorber 18B in the sheath. The neutron absorber 18B has pairs of neutron absorber plates or sheets 41a and 41b, each of which is formed by bending a hafnium plate into a deep U-shaped form. These neutron absorber plates 41a and 41b received in the sheath 15 such that their open ends oppose each other across a stiffener 42 which also serves as a spacer. The space between both walls of each U-bent neutron absorber plate constitutes a water gap 25 which serves as a passage for allowing a moderator to flow therethrough. The water gap 25 is preserved by steps formed on the stiffener 42 on which the adjacent ends of both walls of the neutron absorber plate 41 rest, or by a corrugated sheet 43 which is preferably made of a long-life neutron absorbing material such as hafnium. Preferably, the corrugated sheet 43 made of the long-life neutron absorber is positioned in the radially outer portion of the wing 16. FIG. 24 shows a fourth embodiment of the nuclear reactor control blade in accordance with the present invention. In this embodiment, the control blade denoted by a numeral 10C has a long-life neutron absorber 18C received in the sheath 15 of each wing 16 and composed of a plurality of neutron absorber plates 46 each of which is composed of a long-life neutron absorber plate which is bent into a deep U-shaped form. The opposing walls of this neutron absorber plate 46 define therebetween a water gap 25 for guiding the flow of a moderator therethrough. More specifically, the neutron absorber plate 46 is made from a hafnium plate and both walls thereof is convexed inward at a substantially mid portion along the breadth of the wing 16. The inward convexities 46a, 46b serve to preserve the water gap 25 and affords any exposure growth of the neutron absorber when exposed to neutrons. The ends of of both walls of the neutron absorber plate 46 on the open end of the latter are tapered such that the distance between both walls is gradually increased towards the central tie rod 14 so as to engage tapered surfaces on the end of corresponding projection of the central tie rod 14. FIG. 25 shows a fifth embodiment of the nuclear reactor control blade in accordance with the present invention. In this embodiment of the control blade denoted by a numeral 10D, the sheath 15 of each blade 16 receives a stiffener 47, and long-life neutron absorber 18D on each side of the stiffener 47. Each neutron absorber 18D is composed of a pair of neutron absorber plates or sheets 48a and 48b arranged to oppose each other and bent inwardly at both ends so as to form therebetween a water gap 25 for a moderator. The bends of the neutron absorber plates are done at the portions where the amount of neutron exposure is high, so that the effective thickness of the neutron absorber and, hence, the reactivity worth (neutron absorption characteristics) is effectively enhanced at these portions. The length over which the neutron absorber plate 48b in the radially outer portion is bent is preferably 1 cm to 3 cm. FIG. 26 shows a sixth embodiment of the nuclear reactor control blade in accordance with the present invention. This embodiment of the control blade, denoted by a numeral 10E, has in the sheath 15 of each wing 16 thereof a long-life neutron absorber composed of a pair of plate-shaped neutron absorber plates or sheets 49, 49 each being bent into a deep U-like form. These neutron absorber plates 49 and 49 are disposed in the sheath 15 such that their open ends oppose each other. One of the walls of each plate 49 is extended beyond and bent over the other wall at the open end of the plate 49, thus preserving a water gap 25 between two walls of the plate 49, while affording a margin for the exposure growth when exposed to neutrons. FIG. 27 shows a seventh embodiment of the nuclear reactor control blade in accordance with the present invention. The control blade, denoted by a numeral 10F, has a stiffener 50 disposed in the sheath 15 of each wing 16 thereof, and long-life neutron absorber 18 arranged on each side of the stiffener 50. Each of the neutron absorber 18F is composed of a pair of opposing neutron absorber plates 51a and 51b such that a water gap 25 is defined therebetween. Each of the neutron absorber plates 51a, 51b is lightly bent at one of its ends and bent largely at the other of its ends, and is arranged so that the largely bent end of each plate embraces the lightly-bent end of the other plate. The bends of both neutron absorber plates 51a, 51b serve to preserve the water gap 25 between these plates while affording a margin for the growth of the neutron absorber plates when exposed to neutrons. In the second to the seventh embodiments described hereinbefore, the neutron absorber is divided in the direction of axis of the central tie rod into a plurality of neutron absorber elements each of which is arranged such that a water gap 25 for guiding the flow of a moderator is defined between opposing walls or opposing plates of the neutron absorber material. In consequence, the weight of the neutron absorber is reduced by an amount corresponding to the volume of the water gap, so that the weight of the control blade as a whole is effectively and securely reduced to enable the control blade to be handled by the existing control rod drive mechanism without requiring any change or modification of the control rod drive mechanism. Since the moderator is allowed to flow through the water gap defined between the neutron absorbers, the reactivity is enhanced to allow a reduction in the amount of the neutron absorber. Furthermore, the neutron absorber elements can be positioned effectively at portions where the provision of the neutron absorber is significant from the view point of the reactor shut-down margin. It is thus possible to effectively increase the reactivity while improving also the shut-down margin of the reactor. FIGS. 28 to 31 show eighth to eleventh embodiments of the nuclear reactor control blade in accordance with the present invention. The embodiment shown in FIG. 28, denoted by a numeral 10G, has a neutron absorber 18 in each wing thereof. The neutron absorber 18 constituted by, for example, hafnium metal plates, is divided into a plurality of elements in the axial direction, one of which is shown and indicated at 18A. In this embodiment, the neutron absorber element represented by 18A is divided into two sections: namely, an inner section composed of opposing neutron absorber plates 18Aa, 18Aa and an outer section composed of opposing neutron absorber plates 18Ab, 18Ab, in the breadthwise direction of the wing 16, i.e., in the radial direction of the control blade. The neutron absorber plates 18Aa and 18Aa, as well as the neutron absorber plates 18Ab and 16Ab, are disposed to oppose each other in the thicknesswise direction of the wing 16, thereby defining therebetween a water gap 25 for guiding the flow of a moderator. It will be seen that the width of the water gap 25 is changed in a stepped matter in the breadthwise direction of the wing 16, because the neutron absorber plates 18Aa of the inner section has a smaller thickness than the neutron absorber plates 18Ab in the outer section. In the embodiment shown in FIG. 29, the control blade denoted by 10H has in each wing thereof a plurality of neutron absorber elements 18B composed of a pair of opposing neutron absorber plates or sheets 18Ba the thickness of which is progressively decreased from the radial end of the wing 16 towards the central tie rod 14. In the embodiment shown in FIG. 30, the control blade denoted by 10I has in each wing thereof a neutron absorber 18 divided in the axial direction into a plurality of elements 18C which is further divided in the breadthwise direction into an inner section composed of neutron absorber elements 18Ca, 18Ca and an outer section composed of neutron absorber elements 18Cb, 18Cb. In order to avoid any gap formed between the inner and outer sections, engaging steps 55 are formed on the ends of the neutron absorber plates or sheets 18Cb, 18Cb adjacent to the inner section, so that the ends of the neutron absorber plates or sheets 18Ca, 18Ca of the inner section fit the engaging steps 55. Such engaging steps may be formed both on the neutron absorber plates of both the inner and outer sections. With this arrangement, since the neutron absorber plates 18Ca, 18Ca of the inner section and the neutron absorber plates 18Cb, 18Cb of the outer section are partially overlapped, it is possible to prevent any leak of neutrons from the boundary between both sections. In the embodiment shown in FIG. 31, the control blade denoted by 10J has a neutron absorber element 18D in each wing 16 thereof. The neutron absorber element 18D is composed of three neutron absorber plates or sheets 18Da, 18Db and 18Dc which are arranged in the thicknesswise direction of the wing 16. These neutron absorber plates 18Da, 18Db and 18Dc are held together by means of spacers 56 so as to form water gaps 25 between the adjacent neutron absorber plates. FIG. 32 shows a twelfth embodiment of the nuclear reactor control blade in accordance with the present invention. The control blade of this embodiment, denoted by a numeral 10K, has a neutron absorber 18E in each wing 16 thereof. The neutron absorber 18E is divided in the axial direction into a plurality of elements 18Ea, 18Eb, 18Ec and so forth, each of which is composed of opposing neutron absorber plates or sheets. In this embodiment, the neutron absorber plates of the adjacent elements are partially overlapped at their adjacent ends by steps formed on both or either one of these elements. For instance, the neutron absorber plates of the uppermost neutron absorber element 18Ea are provided at their ends adjacent to the next element 18Eb with steps 57, which mate steps 58 formed in the adjacent ends of the neutron absorber plates of the element 18Eb. The neutron absorber plates of the neutron absorber element 18Eb are provided on their other ends with steps 59 so as to overlay the adjacent ends of the neutron absorber plates or sheets of the next element 18Ec. In the element 18Ec, no step is formed in the ends of the neutron absorber plates adjacent to the element 18Eb because the thickness of the neutron absorber plates or sheets in the element 18Ec is smaller than that in the element 18Eb. The shapes and the arrangements of the steps illustrated in FIG. 32 are only illustrative and may be modified in various forms. A description will be made hereinunder as to an embodiment of the nuclear reactor control blade of the invention which employs an anti-crevice measure for the purpose of preventing any electrochemical corrosion. FIGS. 33 is a perspective view of the control blade employing anti-crevice measure, while FIG. 34 is a side elevational view of the control blade in which the right-side wing is partly-sectioned. The control blade, which is generally denoted by a numeral 100, has an upper structure 102 provided with a handle 101, a lower structure 103 and a central tie rod 104 having a cross-shaped cross-section and integrally interconnecting the upper structure 102 and the lower structure 103. A sheath 105 having a U-shaped cross-section is secured to each projection on the central tie rod 102. Each sheath accommodates a plate-like long-life neutron absorber 106 which is typically made of hafnium plate. The sheath 105 and the long-life neutron absorber housed therein constitutes a wing 107. Thus, the control blade 100 has four wings 107. As will be seen from FIG. 35, the neutron absorber 106 includes neutron absorber plates 106a in the form of rectangular sheets and arranged to oppose each other in the thicknesswise direction of the wing 107. These opposing neutron absorber plates 106a, 106a are spaced from each other by supporting spacers 108 such that a water gap 110 for guiding the flow of moderator is defined therebetween. Washer-like spacers 109 are mounted on both sides of each supporting spacer 108 so that a water passage 111 of a predetermined width is formed between the outer surface of each neutron absorber plate 106a and the adjacent inner wall of the sheath 105. At the same time, a water passage spaces 112a are formed between the side surfaces of the neutron absorber plates 106a adjacent to the central tie rod 104 and the opposing surface of the central tie rod 104. These water passage spaces 102a are formed by, for example, chamferring edges of the projection of the tie rod 104, as at 104a. The cross-sectional areas of the water passage spaces 102a and, hence, the cooling effect will be increased by chamfering also the edges of the neutron absorber plates 106a, as at 113. Side surfaces of the upper structure 102 and the end structure 103 also are chamferred as at 104a, so that water passage spaces 112a are formed between these chamferred surfaces and the opposing side surfaces of the neutron absorber plates 106a. Preferably, water passage spaces 112b are formed also between the inner surface of the sheath 105 and the neutron absorber plates 106a at the outer end portion of the wing 107. Such water passage spaces 112b can be formed by chamferring the portions of the neutron absorber plates 106a as at 113b, in the end portion of the wing 107. The arrangement may also be such that the chamfer 113b is provided over the entire circumference of each neutron absorber plate 106a so that the water passage spaces 112a are formed between the chamferred circumferential edge of the neutron absorber plate 106a and the opposing surfaces of the central tie rod 104, upper structure 102 and the end structure 103. The construction for securing the neutron absorber plate 106a will be explained in detail with reference to FIGS. 36 and 37. As explained before, two neutron absorber plates 106a, 106a are disposed in each sheath 105 so as to oppose each other and are spaced by a predetermined distance from each other by means of the supporting spacers 108. Each supporting spacer 108 has a spacing portion 117 engaging with the opposing neutron absorber plates 106a, 106a so as to preserve the predetermined gap therebetween and supporting legs 118 projecting from centers of both ends of the spacing portion 117. Both end surfaces of the spacing portion 117 are provided with water passage groove 119, as will be best seen from FIG. 37A. The supporting legs 118 are secured in mounting holes formed in the walls of the U-shaped sheath 105 by welding. However, since the spacer 108 and the sheath 105 are made of the same stainless steel, and since a surface finishing treatment is conducted after the welding, no weld line appears on the outer side of the sheath 105. Washer-like spacers 109 are provided on the supporting legs 118 of the supporting spacer engaging with both neutron absorber plates 106a, 106a. Namely, the supporting legs 118 of the supporting spacer 108 loosely penetrate holes 120 formed in the neutron absorber plates 120 and fit in the above-mentioned mounting holes 121 formed in the walls of the sheath 105. The supporting legs 118 are then welded to the sheath from the external surface of the sheath 105. The spacing portion 117 of the supporting spacer 108 serve to preserve a water gap 110 between opposing neutron absorber plates 106a, 106a, while the washer-like spacers 109 serve to preserve water passages 111 of a predetermined width between the outer surfaces of both neutron absorber plates 106a and the adjacent inner surfaces of the sheath 105. Preferably, the surfaces of the washer-like spacers 109 are provided with water passage grooves 119a, as in the case of the supporting spacers 108. The operation and advantage of this embodiment will be described hereinunder. The reactor water serving as the moderator is introduced into each sheath past passage holes 122a, 122b formed in the walls of the sheath as shown in FIG. 34. The water then flows through the water gap 110 and the water passages 111, while carrying away the heat generated in the neutron absorber plates 106a, 106a. The reactor water is sufficiently distributed also to the water passage spaces 112a formed between the neutron absorber plates 106a and the adjacent constituents of the control blade, as well as to the water passage spaces 112b formed between the neutron absorber plates 106a, 106a and the inner surface of the sheath 105 at the outer end portion of each wing 107. In consequence, there is no stagnation of water or dead water space. Thus, a smooth flow of the water is ensured over the entire region in the control blade so that any local heating of the neutron absorber plates 106a, 106a and the sheath 105 is avoided. The supporting spacers 108 are provided with water passage grooves 119 formed therein so that the reactor water is allowed to flow along the surfaces of the supporting spacers 108 contacting the neutron absorber plates 106a, 106a, so that overheating of the spacers 108 is prevented. In this embodiment, therefore, cooling effect of the reactor water is ensured by virtue of the water passages 111 formed between the outer surfaces of the neutron absorber plates 106a, 106a and the adjacent inner walls of the sheath 105 and water passage spaces 112a and so forth formed along the side surfaces of the neutron absorber plates 106a, 106a, in addition to the provision of the water gap 110 formed between both neutron absorber plates 106a and 106a. This effectively eliminates any risk that the sheath 105 and the neutron absorber plates 106a, 106a may be damaged by local overheating. It is also to be noted that the constituents such as the sheath 105, neutron absorber plates 106a, 106a and the central tie rod 104 are spaced from each other by the water gap 110, water passages 111 and the water passage spaces 112a. Therefore, the risk that these constituents may be electrochemically corroded due to direct contact of different metallic materials can be avoid almost completely, so that the nuclear reactor control blade can operate for a long period in sound state. A nuclear reactor control blade employing another example of anti-crevice measure will be described hereinunder with reference to FIG. 38. In this embodiment, a multiplicity of dimples 123 are formed in the wall of the sheath 105 such that the reverse side of these dimples project inwardly of the sheath 105. The ends of the projections provided by the dimples contact the adjacent surface of the neutron absorber plate 106a so as to keep this surface of the neutron absorber plate 106a from the inner surface of the sheath 105, thereby forming water passage 111 of a predetermined width therebetween. The same effect is produced by forming the dimples in the neutron absorber plate 106a such that the reverse side of the dimples project towards the outer side into contact with the surface of the sheath. In this embodiment, it is not necessary to employ any specific member for the purpose of regulating the width of the water passages 111, such as the washer-like spacer 109 (see FIG. 35) employed in the preceding embodiment, so that the work for fabricating the spacers, as well as the assembly of the wing 107, can be facilitated advantageously. The nuclear reactor control blades shown in FIGS. 34 to 38 offer the following advantages. Namely, these control blades enable the reactor water to flow smoothly without any stagnation and without forming any dead water space, by virtue of the provision of the water passages 111 between the outer surfaces of the neutron absorber plates 106a, 106a and the inner surfaces of the sheath 105, as well as the water passage spaces 112a, 112b formed between the side surfaces of the neutron absorber plates 106a, 106a and the adjacent surfaces of the central tie rod 104, upper structure 102 and the lower structure 102. In consequence, heat exchange is conducted without any impediment in the regions where heat is generated as a result of the neutron absorption, so that any damage of the sheath due to local heating is avoided. In addition, the water passages 111 serve to keep the outer surfaces of the neutron absorber plates 106a away from the inner surfaces of the sheath 105, while water passage spaces serve to prevent direct contact between the neutron absorber plates and members therearound. Therefore, any risk of electrochemical corrosion occurring due to direct contact of different metallic materials is prevented, thus ensuring the nuclear reactor control blade to operate for a long period of time in the sound state. A description will be made hereinunder as to an embodiment of the nuclear reactor control blade which employs both an anti-crevice measure and an anti-earthquake measure, with reference to FIG. 39 in which the same reference numerals are used to denote the same parts or members as those of the control blade 100 shown in FIGS. 33 to 37. General arrangement of the control blade 100A shown in FIG. 39 is similar to that of the control blade 100 explained before. In this control blade 100A, a long-life neutron absorber 130 such as of hafnium sheet accommodated in the sheath 105 is divided into a plurality of stages or neutron absorber elements 130a along the axis of the central tie rod 104. The neutron absorber element 130a of each stage is composed of a plurality of, e.g., two, neutron absorber plates 130b arranged so as to oppose to each other and integrally connected through a plurality of spacers 108. A water gap 110 for allowing a moderator to flow therethrough is defined between the opposing neutron absorber plates 130b. Recesses 131 are formed in the walls of the sheath 105 of each wing 107 so as to extend in a direction perpendicular to the axis of the sheath 105. As will be seen from FIGS. 39 and 40, each of the recesses 131 is formed by depressing inwardly the wall of the sheath 105 along a line extending in the breadthwise direction of the wing 107, so that the inner surface of the sheath wall project into a gap d formed between the neutron absorber plates 130b of the adjacent neutron absorber elements 130a. It is assumed here that each wing 107 is divided in the axial direction into three regions: namely, an uppermost first region, an intermediate second region and a lowermost third region. The recesses 131 are provided such that at least one recess 131 is formed at the upper end of the second region so as to extend in a direction perpendicular to the axis, i.e., in the breadthwise direction of the wing. Each recess 131 is so formed that it provides a communication between a notch 132a formed in the outer end of each wing 107 and a notch 132b formed in the portion of the sheath where the sheath is connected to the central tie rod. FIG. 41 illustrates a process for fabricating the sheath 105 having the recesses 131. A blank sheet 105a of the sheath steel in the developed state is notched at its both side edges as at 132a and an aperture 132b is formed as illustrated. Then, a suitable mechanical processing is conducted to cause the blank sheet 105a to be bent along bend lines Bl--Bl which interconnect the notches 132a and the aperture 132b, whereby the recess 131 is formed. The bending operation is facilitated by virtue of the presence of the notches 132b, 132b on both sides edges of the blank sheet 105a and the central aperture 132a. Then, the blank sheet 105a is bent along a vertical line Cl which passes the aperture 132a into a form like U, whereby a deeply-bent U-shaped sheath 105 is formed. The thus formed U-shaped sheath 105 with the recesses 131 is fixed to the corresponding projection of the central tie rod 104 by, for example, spot welding, as shown in FIG. 40. In this state, the recess 131 is so positioned that the reverse side of the neutron absorber plate corresponding to the recess projects into a gap d formed between the adjacent neutron absorber plates 130b, 130b of the adjacent neutron absorber elements 130a. In order to accomplish a smooth flow of the moderator into and out of the sheath 105, a plurality of water passage holes 133 are formed in predetermined portions of the sheath 105 as will be seen from FIG. 39. In this embodiment, since recesses extending in the direction perpendicular to the axis of each wing are formed in the walls of the sheath such that the inner surfaces of the sheath wall project inwardly, undesirable outward expansion of the sheath is prevented even when an excessively large bending stress on the central blades and/or any excessive stress due to large acceleration in the axial direction is caused in the event of a heavy earthquake. In consequence, smooth movement of the control blades is ensured even in the case of such a heavy earthquake. Namely, when a large external force is applied to the control blades due to, for example, an earthquake, the force transmitted to the sheath 5, tending to deform the sheath, is effectively absorbed by the expansion or contraction of the sheath wall along the linear recesses, so that any outward deformation of the sheath wall is avoided. It will be understood that the outward deformation of the sheath will cause a serious problem in that the movement of the control blade is hindered due to mechanical interference between the sheath of the control blade and the fuel assemblies around the control blade. The prevention of the outward deformation of the sheath, therefore, offers a great advantage from the view point of safety. As stated before, the advantage of the recess 131 is remarkable particularly when it is provided in the second of three regions defined along the axis of the wing 107. This is because the deformation of the sheath, attributable to the stress in the control blade 100A due to, for example, an earthquake is greatest in the second region, i.e., the axial mid region of the control blade 100A. In this embodiment, therefore, the gap d for receiving the inward projection formed as a result of the recessing of the linear recess 131 is provided by an axial discontinuity of the neutron absorber only in the second region of the wing 107, while the neutron absorber is arranged without discontinuity in the uppermost first region of the wing which receives the heaviest neutron exposure. This in turn ensures a sufficiently large value of the reactor shut down margin. The provision of the linear recess 131 in the lowermost third region of the wing is not so significant because this lowermost region receives only a small external force as compared with the second region. The number and the positions of the linear recesses 131 in the second region can be determined suitably taking into account the factors such as the load condition. It is also possible to control the rigidity of the blade by designing such that the positions of the linear recesses are varied according to the wings 107. It is also possible to form the linear grooves 131 on both sides of each wing 107 at different heightwise or axial positions, as shown in FIG. 43. The sheath used in such a wing may be formed by the same process as that explained before in connection with FIG. 41, though the aperture 132a which is to provide the notch in the outer end of the wing has a vertically elongated form. FIG. 44 is a sectional view of a wing which is obtained by bending the blank sheet 105a shown in FIG. 43 into a deep U-like form so as to form a sheath, and securing the thus formed sheath to the central tie rod with the neutron absorber 130b received therein. In this case, the gaps d, d of about 10 mm, formed between the adjacent neutron absorber plates 130b, 130b at both thicknesswise ends of the wing, are staggered in the heightwise direction from each other such that each gap d is masked by the opposing neutron absorber plate 130a. The construction shown in FIG. 44 offers, unlike the arrangement shown in FIG. 42 in which the linear recesses on both sides of the wing are positioned at the same level, an advantage that neutron fluxes passing through each gap d is effectively masked by the opposing neutron absorber plate 130b so that local reduction in the reactivity worth at the position of the gap d is avoided, thus preventing any reduction in the reactor shut-down margin. Preferably, the corners of the neutron absorber plates 130b facing the projections on the back side of the recesses 131 are chamferred as shown in FIGS. 42 and 44, so that application of local stress to bent portions of the sheath is avoided. FIGS. 45 to 48 show modifications of the nuclear reactor control blades explained before in connection with FIG. 39. In these Figures, the control blade is generally designated at a numeral 100B and the same reference numerals are used to denote the same parts or members as those used in the control blade 100A shown in FIG. 39, with the description of such parts or members being omitted. The nuclear reactor control blade 100B has a sheath 105 accommodating a long-life neutron absorber 130 which is divided into a plurality of elements in the axial direction of the central tie rod. The sheath 105 also is divided at portions corresponding to the division of the neutron absorber 130a so that the sheath 105 is composed of a plurality of sheath elements arranged in the direction of axis of the central tie rod 105. The arrangement is such that at least one discontinuity of the sheath, extending in the direction perpendicular to the axis, i.e., the breadthwise direction of the wing 107, is located within the second one of the axial three regions of the wing 107, i.e., the intermediate one of three regions which are assumed along the height of the wing 107. This is because the stress generated in the sheath wall caused by, for example, an earthquake and, hence, an amount of deformation of the sheath wall are greatest in the second region of the sheath. The arrangement may be such that, as shown in FIG. 45, the different wings have different positions of division of the sheath. Such an arrangement offers an advantage that the rigidity or strength of the sheath 105 as a whole is increased because the discontinuities are not concentrated to the same height along the axis of the sheath 105. As will be seen from FIG. 46, a gap d of a predetermined size is formed between the adjacent sheath elements 105a and 105a. The size of the gap d is so determined that any deformation of the control blade due to an external force is sufficiently absorbed by this gap. As will be seen from FIGS. 47 and 48, a holding member 136 having a fitting groove 135 extending in the breadthwise direction of the sheath element 105a is fitted in the end of each sheath element 105a. Thus, the fitting grooves 135 in the opposing ends of the adjacent sheath elements oppose each other and slidably receive a tabular neutron absorber constituted by, for example, a hafnium plate, such that the neutron absorber 137 is movable within the grooves 135 in the direction of axis of the control blade. That is, the adjacent sheath elements 105a and 105a are connected through the intermediary of the neutron absorber 137 in a manner capable of expanding and contracting in the axial direction. The overall width of the fitting grooves 135, 135 is greater than the overall width of the neutron absorber 137, and a space for absorbing the axial displacement of the neutron absorber 137 is left behind the neutron absorber, i.e., at the innermost end of each groove 135. The neutron absorber 137 fitting in the grooves 135 may have a substantially T-shaped cross-section as shown in FIG. 47. In such an arrangement, the step 139 of the tabular neutron absorber 137 abuts the end surface of the holding member 136 thereby limiting the movement of the neutron absorber 137 in the breadthwise direction. In consequence, the tabular neutron absorber 137 is prevented from coming off and from contacting the outer end of the sheath. It is also preferred that the end surfaces of the sheath elements 105a and the corners of the holding member 136 are chamferred so as to prevent any damage which may occur due to mechanical interference between the adjacent sheath elements. In the event that a large bending force is applied to the nuclear reactor control blade due to, for example, a heavy earthquake, the greater stress occurs in the axially mid portion of the control blade having an elongated form, so that the greatest strain or deformation appears in this portion of the control blade. This deformation, however, is effectively absorbed by the gap d formed between the adjacent sheath elements 105a, as well as by the fitting grooves 135. Namely, the tabular neutron absorber 137 is allowed to slide in the axial direction of the control blade by an amount corresponding to the deformation, thus preventing any stress to occur in the sheath wall. This conveniently avoids any outward deformation of the sheath, which may otherwise occur to cause a mechanical interference between the control blade and the adjacent fuel assemblies, seriously impeding the vertical movement of the control blade. In this embodiment, no discontinuity of the sheath is provided in the heightwise first region corresponding to 1/3 of the entire height of the control blade as measured from the upper end, so that no substantial reduction in the neutron absorbability takes place in this first region. As explained before, the upper end of the control blade and the outer ends of wings of the control blade are generally subjected to the heaviest neutron exposure. It is, therefore, not preferred to provide any discontinuity of the sheath, i.e., discontinuity of neutron absorber, in such portions of the control blade. In this embodiment, since the discontinuity due to the division of the sheath is located in the second region, i.e., in the intermediate one of three regions assumed in the axial or heightwise direction of the control blade, it is possible to avoid any reduction in the neutron absorbability in the first region and a large margin for shutting down the reactor is ensured. A description will be made hereinunder as to a modification of the embodiment shown in FIG. 45, with specific reference to FIG. 49. This modification of the control blade, denoted by a numeral 100C, has two elongated hafnium neutron absorber bars 140 of hafnium extending axially of the control blade along the outer edge of each wing 107. A space for accommodating any thermal expansion of the neutron absorber bars 140 are provided at the upper side of these neutron absorber bars 140. Other portions are materially the same as those of the embodiment shown in FIG. 45. According to this arrangement, the neutron absorbability is increased by the presence of the neutron absorber bars 140 along the outer edge of each wing 107 where the consumption of the neutron absorber is specifically large. In consequence, the nuclear life of the control blade can be prolonged. The neutron absorber bars 140 extending axially along the outer edges of the series of sheath elements 105a effectively serve also as structural members so as to enhance the mechanical strength of each wing. The neutron absorbability is locally reduced in the regions where there is any discontinuity of the neutron absorber plates 130b. It will be understood that such a local reduction in the absorbability is effectively compensated by the presence of the neutron absorber bars 140. The neutron absorber bars 140 also provide an advantage in that they effectively restrains relative movement between the sheath elements 105a in the same wing 107. It will be seen that, in this embodiment, the sheath of the control blade is divided into a plurality of sheath elements in the axial direction of the control blade, and the adjacent sheath elements are connected for free expansion and contraction by means of the tabular neutron absorber fitting in grooves formed in the opposing ends of the adjacent sheath elements. In consequence, the deformation of the sheath wall, which may be caused by a large external force applied to the control blade due to, for example, an earthquake is effectively suppressed. Namely, any deformation of the sheath caused by a stress in the control blade can effectively be absorbed by the gaps between the successive sheath elements and the fitting grooves which loosely and slidably receive the interconnecting tabular neutron absorber. This in turn eliminates any risk of the control blade expanding outward and a consequent interference between the control blade and fuel assemblies around the control blade, thus eliminating any impediment to the smooth movement of the control blade. This embodiment, therefore, ensures a smooth movement of the control blade even in the event of application of a large force to the control blade. A description will be made hereinunder as to different embodiments of the present invention which are hybrid-type control blades incorporating both anti-crevice and anti-earthquake measures. FIG. 50 is a fragmentary front elevational view of a hybrid-type control blade for use in a nuclear reactor, constructed in accordance with the present invention. This embodiment of the nuclear reactor control blade, generally denoted by a numeral 100D, has an upper structure 102 provided with a handle 101, a lower structure 103 and a central tie rod 104 having a cross-shaped cross-section and integrally interconnecting both structures 102 and 103. A sheath 105 having a U-shaped cross-section is secured to the end of each projection of the central tie rod 104. A plate-like long-life neutron absorber 130, typically made of hafnium, is disposed within each sheath 105 at a portion adjacent to the central tie rod 104. The sheath 105 and the long-life neutron absorber 130 in combination constitute a wing 107. Thus, the control blade is provided with four such wings 7 on the central tie rod 104. An elongated hafnium bar 140 is disposed adjacent to the neutron absorber 106 so as to extend along the outer edge of the wing 107. The elongated hafnium bar 140 may include 2 to 5 rods of hafnium having circular cross-section and disposed in parallel so as to extend over the entire axial length of the wing along the outer edge of the latter, thus enhancing the strength at the outer end of the wing. In addition, a space for absorbing the thermal expansion of the hafnium bar 140 is formed within the sheath on the upper side of the hafnium bar 140 as shown in from FIG. 50. On the other hand, the neutron absorber 130 is divided in the axial direction of the tie rod 104 into a plurality of stages or neutron absorber elements 130a. As will be seen from FIG. 51, the neutron absorber element 130a of each stage is constituted by a plurality of neutron absorber plates on sheets 130b arranged such as to oppose in the thicknesswise direction of the wing 107. A water gap 110 for guiding the flow of a moderator is defined between these neutron absorber plates 130b. In addition, water passages 111 are formed between the outer surfaces of the respective neutron absorber plates 130b and the adjacent inner surfaces of the sheath 105. The neutron absorber plates 130b and 130b are held on the sheath 105 at a predetermined distance from each other preserved by end spacers 145 and a central spacer 108. The end spacers 145 and the central spacer 108 are made of hafnium and a stainless steel, respectively. A detailed description will be made hereinunder as to the construction for securing the neutron absorber plate 130b with specific reference to FIG. 52. The end spacers 145 having a comparatively small axial length are disposed between both neutron absorber plates 130b and 130b at both longitudinal ends of the neutron absorber plates. The neutron absorber plates 130b, 130b are partly fixed to the end spacers 145 by, for example, welding. Each end spacer 145 has a spacing portion 145a the width of which determines the size of the water gap 110 between both neutron absorber plates 130b, 130b. At the same time, water passages 111 of a predetermined width are formed between the outer surfaces of the respective neutron absorber plates 130b, 130b and the adjacent inner surfaces of the sheath 105. It is preferred that the corners of the projection of the central tie rod 104, as well as the adjacent corners of the end spacer 45 are chamferred as shown in FIG. 52 so as to provide a passage for water so that the reactor water as the moderator encounters a reduced resistance to accomplish a large resistance, whereby the local overheating of the neutron absorber plates 130b is avoided. The central portions of the neutron absorber plates 130b, 130b are fixed at predetermined positions by means of a central spacer 108 and ring-shaped spacers 109. As will be seen from FIGS. 37A, 37B and 37C, the central spacer 108 has a collar-like seat portion 117 having a thickness corresponding to the width of the channel constituted by the water gap 110. Water passage grooves 118 are formed in both surfaces of the seat portion 117 so that the reactor water as the moderator is allowed to flow along these grooves. The opposing neutron absorber plates 130b, 130b rest on the respective surfaces of the seat portion 117 of the central spacer 108 and are fixed to the sheath 105 through the intermediary of the respective ring-shaped spacers 109. The central spacer 108 has both axial ends received in mounting holes 146 formed in the walls of the sheath 105 and are secured to the latter by, for example, welding. The ring-shaped spacers 109 may be provided in both surfaces thereof with water passage grooves 119a as in the case of the central spacer 108. In operation, as shown in FIG. 50, the reactor water is introduced into the sheath 105 through passage holes 122a, 122b formed in the wall of the sheath and then flows along the water gap 110 and the water passages 111 so as to carry away the heat generated in the neutron absorber plates 130b, while serving also as the moderator. The water then comes out of the control blade through passage holes 122a, 122b formed in an upper portion of the wing 107. The control blade 100D of this embodiment, therefore, is a hybrid-type control blade which employs both a plate-like neutron absorber 130 made of a sheet of hafnium and an elongated hafnium bar 140 provided along the outer edge of each wing 107. In consequence, the mechanical strength of the control blade as a whole is improved and the durability of the control blade against any external force is improved. Thus, the control blade of this embodiment is suitable for use in nuclear reactors which are designed for a long-term operation. In this embodiment, the reactor water as the coolant can be distributed to every portions in the control blade by virtue of the water passages 111 formed between the outer surfaces of the neutron absorber plates 130b and the adjacent inner surfaces of the sheath 105. In consequence, any damage of the sheath 105 due to local heating of the neutron absorber plates 130b can be avoided advantageously. In addition, the outer surfaces of the neutron absorber plates 130b and the inner surfaces of the sheath 105 are spaced apart from each other across the water passage 111 preserved by the end spacers 145, central spacer 108 and the ring-shaped spacers 109. It is therefore possible to suppress electrochemical corrosion which otherwise may be caused due to direct contact between two different metallic materials, so that the control blade 100D can operate for a along period of time in the sound state. A different nuclear reactor control blade, which also is of hybrid type, will be described hereinunder with reference to FIG. 53. This control blade incorporates an elongated hollow hafnium bar or tube 140a disposed along the outer edge of each wing 107 so as to extend in the axial direction of the wing 107. In this case, the central bore 151 in the elongated hafnium bar 140a provides a passage for the water as the moderator. The water flowing through this passage exhibits a certain degree of neutron absorbing effect thus substituting for the hafnium. It is therefore possible to appreciably reduce the total weight of hafnium employed in the control blade. It is possible to form a multiplicity of apertures in the wall of the hafnium bar 140a so as to diversify the flow paths of the moderator and so as to reduce the flow resistance, thus attaining higher cooling efficiency. In another form of this embodiment, a plurality of dimples 150 are formed in the walls of the sheath 105 such that the reverse side of the dimples project inward into contact with the adjacent neutron absorber plates 130b so as to serve as spacers which keep the outer surfaces of the neutron absorber plates 130b away from the inner surfaces of the sheath thereby preserving a predetermined distance of passages 111 therebetween. In this embodiment, the end spacers 145 is required only to determine the width of the water gap 110 formed between opposing neutron absorber plates 130b at the center of the sheath. The end spacer, therefore, can have a simple form which is easy to machine. This embodiment offers an advantage in that the ring-like spacers 109 for determining the width of the water passages 111 can be dispensed with so that the assembly of the wing 107 can be simplified. It will be noted also that this embodiment employs a plurality of end spacers 145 provided between the opposing neutron absorber plates at their both ends, at a predetermined interval in the axial direction of the wing 107 as will be seen from FIG. 54. Each end spacer 145 is fixed by, for example, welding only to one of the opposing neutron absorber plates 130b, 130b. It is, therefore, possible to avoid any deformation of both neutron absorber plates in the event that one of these neutron absorber plates is accidentally deformed by an unexpected reason. Thus, the described embodiments of the nuclear reactor control blade, designed as hybrid type control blades, can exhibit improved structural strength by virtue of the provision of elongated hafnium members along the outer edges of the wings, so that the control blade as a whole exhibits a greater resistance to external force. This offers a higher reliability of operation of the nuclear reactor despite earthquakes and scramming operations which are expected to occur or be executed for many times during longterm operation of the nuclear reactor. In addition, the coolant can effectively distributed to every portions in the control blade by virtue of the provision of the water passages between the outer surfaces of the neutron absorbers and the adjacent inner surfaces of the sheath, so that the risk of the sheath or the neutron absorber plates being damaged by local overheating is avoided advantageously. Furthermore, the water passages serve to keep the neutron absorber plates away from the inner surfaces of the sheath so that any electrochemical corrosion, which may otherwise be caused due to direct contact between different metallic materials can be eliminated advantageously, whereby the soundness of the nuclear reactor control blade can be maintained for a long period of time. A description will be made hereinunder as to an embodiment of a flux-trap type control blade for boiling water reactors in accordance with the present invention, which is specifically designed to reduce the weight of the sheath and to prevent any tendency for buckling, with specific reference to FIG. 55. This control blade, generally designated by a numeral 200, has an upper structure 202 provided with a handle 201, a lower structure 203, and a central tie rod 204 having a cross-shaped cross-section and integrally interconnecting the upper and lower structures 202 and 203. A sheath 205 having a U-shaped cross-section is fixed to each of four projections of the central tie rod 204. The sheath accommodates a long-life neutron absorber 206 made of hafnium plate. The sheath 205 and the neutron absorber 206 in combination constitute a wing 207. Thus, the control blade 200 as a whole has four such wings 207. Guide rollers 208 for guiding the movement of the control blade into and out of the reactor core are provided on both sides of the portion of the upper structure 202 corresponding to each wing 207, while the lower structure 203 is provided with a speed limiter 209. The neutron absorber 206 is divided into a plurality of stages or neutron absorber element 206a in the direction of the axis of the central tie rod 204. As will be seen from FIG. 56, each neutron absorber element 206a is constituted by a pair of opposing neutron absorber plates or sheets 206b accommodated by the sheath 205. A plurality of spacers 210 are disposed between these neutron absorber plates at suitably dispersed locations, so as to form a water gap of a predetermined width between these neutron absorber plates 206a. The spacers 210 also serve to reinforce the wing. As will be seen from FIGS. 55 and 57, water passage holes 211 are formed in the walls of the sheath 205 and in the neutron absorber plates 206b so as to introduce the flow of a moderator (coolant) into the water gap between two neutron absorber plates. The sheath 205 also is provided with water passage holes 212 adapted for guiding the flow of the moderator into channels defined between the neutron absorber plates 206b and the adjacent surfaces of the sheath. Furthermore, water passage holes 213 adjacent the upper structure are formed in a portion of the sheath 205 near the upper end thereof, while water passage holes 213 adjacent to the lower structure are formed in a portion of the sheath 205 near the lower end thereof. The sheath 205 is further provided at a portion near the inner end thereof with water passage holes 214 adjacent to the tie rod. The neutron absorber plates 206b are made of hafnium, while the sheath 205 and the central tie rod 204 are made of a stainless steel. The neutron absorber plates 206b, therefore, exhibit a different value of thermal expansion coefficient from that of other structural members. In order to absorb any difference in the amount of thermal expansion due to the difference in the thermal expansion coefficient between the neutron absorber plates 206b and other structural members, the neutron absorber element 206a is sectioned into a plurality of sections in the longitudinal directions, i.e., in the vertical direction, as shown in FIG. 59, and a gap G for absorbing the difference in the amount of thermal expansion is formed between the neutron absorber plates 206b, 206b of the adjacent sections. In order to compensate for any reduction in the mechanical strength due to the presence of the gaps G formed as a result of division of the neutron absorber plate, the spacers 210 are disposed in a staggered manner in the direction of axis of the central tie rod 204 in such a manner as to meet the following condition: EQU L.sub.1 =L.sub.2 =L.sub.3 =L.sub.4 >L.sub.5 where, L.sub.1, L.sub.2, L.sub.3 and L.sub.4 represent the pitch of the spacers within each section 206a, 206a, 206a . . . while L.sub.5 represents the distance between adjacent spacers of the adjacent two sections. Considering that each wing exhibits a smaller mechanical strength at the free end portion than at the central or inner portion which is stiffened by the central tie rod 204, it is advisable that the number of the spacers is selected to be larger in the end portion of each wing than in the central or inner portion. For instance, in the embodiment shown in FIG. 59, each section of the neutron absorber plate has two spacers at its inner region and three spacers at its outer region. The clearance of channel S between the each wall of the sheath and the adjacent neutron absorber plate 206b is intended for preventing any stagnation of water. This channel S can be formed by dimpling the wall of the sheath 205 inwardly in a depth of 0.2 to 0.3 mm as at 205a, as shown in FIG. 60. The pitch of the dimpling 205a is, for example, about 10 cm in the axial direction of the central tie rod. The channel S preserved by such dimpling has such a size that the water in this channel is replaced in one to several days. The channel S may be formed by other means than the described dimpling, e.g., by employing a washer-like spacers of the type shown in FIG. 36. The water passage holes 213 and 214 effectively prevent the water from stagnating in the regions near the upper structure 202, end structure 203 and the central tie rod 204. The side edges of the central tie rod and the corners of the neutron absorber plate 206b adjacent to the central tie rod are chamferred as at 204a and 206c, as shown in FIG. 60, thereby reducing the resistance encountered by the flow of water. The spacer 210 used in this embodiment may be of the type which is shown in FIGS. 37A, 37B and 37C. Namely, the spacer 210 has a disk-shaped seat portion 210a and mounting leg portions 210a projecting from both sides of the seat portion 210a. Grooves 210c of a suitable number and depth are formed in both surfaces of the disk-shaped seat portion 210a in a crossing form as shown in FIG. 37B, in such a manner as not to cause any significant reduction in the strength of the spacer 210. As will be readily understood from FIG. 60, the grooves 210c serve to provide communication between the inner and outer surfaces of each neutron absorber plate. The mounting leg portions 210b of the spacer 210 fit in mounting holes 206a provided in the neutron absorber plates 206a such as to leave a gap G', so as to enable any difference in the thermal expansion to be absorbed. It is possible to substitute some of the spacers 210 by hafnium spacers 216 as shown in FIG. 61. Each hafnium spacer is fixed to either one of the opposing plate by, for example, shrink fit or welding, and has a height or thickness that the other end thereof touches the inner surface of the opposing neutron absorber plate, as will be seen from FIG. 62. It is also possible to use, as the spacer, a wire-type spacer 217 made of stainless steel and having a diameter of 3 to 5 mm, as shown in FIG. 63. These wire-type spacers are fixed to the spacers 210 so as to stiffen the wing of the control blade against lateral bending force, while preserving the gap G' between the opposing neutron absorber plates. Although the illustrated embodiment employs only two wire-type spacers, the number of the wire-type spacers may be increased or decreased as occasion demands The control blade of this embodiment exhibits a long life by virtue of the use of hafnium which is a typical long-life neutron absorbing material. The neutron absorber is arranged in the form of flat sheets or plates such as to form, between the opposing neutron absorber plates, a water gap into which water serving as a coolant and a moderator is introduced. Both the neutron absorber plates and the water serve to enhance the reactivity worth, so that the control blade as a whole can exhibit a large value of the reactivity worth. Alternatively, for attaining a required level of reactivity worth, the amount of expensive hafnium having a large density (13.3 g/cm.sup.3) may be reduced. In addition, the spacers in each wing are dispersed in the region near the central tie rod and in the region remote from the central tie rod, so that a linear flow passage extending in the axial direction of the central tie rod is formed between both neutron absorber plates in the central region thereof. These spacers are arranged in a staggered manner substantially at a constant interval such that the distance between the two adjacent spacers on different neutron absorber plates is slightly smaller than the axial pitch of the spacers in each neutron absorber plate. In consequence, the wing exhibits a substantially uniform distribution of strength against lateral bending force over its entire length. FIG. 64 shows another form of the flux-trap-type control blade embodying the present invention. This control blade, generally denoted by a numeral 200A, is basically the same as the control blade 200 shown in FIG. 55 so that the same parts or members as those in the control blade 200 are denoted by the same reference numerals and detailed description thereof is omitted. The construction of the control blade 200A is shown in detail in FIGS. 64 to 67. The control blade 200A has four wings 207 each having a plurality of spacers 210 which may be of the same type as that used in the control blade 200 explained before. These spacers 210 are arranged at a higher density in the outer region of the wing than at the inner or central region which is stiffened by the central tie rod 204. For instance, in the embodiment shown in FIG. 66, three spacers 210 are disposed in the outer region, while the inner region employs two spacers. Preferably, a wire-type spacer 220 is disposed in the region near the outer end of the wing. The wire-type spacer is fixed to the spacers 210 by, for example, welding. The clearance of channel S between the each wall of the sheath 205 and the adjacent neutron absorber plate 206b is intended for preventing any stagnation of water. This channel S can be formed by dimpling the wall of the sheath 205 inwardly in a depth of 0.2 to 0.3 mm as at 205a, as shown in FIGS. 65 and 67. The channel S preserved by such dimpling has such a size that the water in this channel is replaced in one to several days. The channel S may be formed by other means than the described dimpling, e.g., by employing a washer-like spacers (not shown). The water passage holes 213 to 214 effectively prevent the water from stagnating in the regions near the upper structure 202, end structure 203 and the central tie rod 204. The side edges of the central tie rod and the corners of the neutron absorber plate 206b adjacent to the central tie rod are chamferred as at 204a and 206c, as shown in FIG. 60, thereby reducing the resistance encountered by the flow of water. The spacer 210 used in this embodiment may be of the type which is shown in FIGS. 37A, 37B and 37C. The mounting leg portions 210b of the spacer 210 fit in mounting holes 206a provided in the neutron absorber plates 206a such as to leave a gap G', so as to enable any difference in the thermal expansion to be absorbed. It is possible to substitute some of the spacers 210 by hafnium spacers 221 as shown in FIG. 68. Each hafnium spacer is fixed to either one of the opposing plate by, for example, shrink fit or welding, and has a height or thickness that the other end thereof touches the inner surface of the opposing neutron absorber plate. The embodiment shown in FIG. 68 employs two wire-type spacers 220 for each neutron absorber plate 206b. These wire-type spacers are fixed at their upper and lower ends to the hafnium spacers 221. The wire-type spacer is constituted by a hafnium wire of a diameter ranging from 3 mm to 5 mm, and provides a reinforcement to the wing of the control blade against lateral bending force, while preserving the gap G' between the opposing neutron absorber plates 206a, 206a. It is possible to use spacers made of a stainless steel, in place of the hafnium spacers 221. In such a case, the wire-type spacers 220 also are preferably made of a stainless steel, from the view point of easiness of welding. In the control blade of the described embodiment, the neutron absorber is arranged in the form of flat plates such that a gap for guiding the flow of water, serving as a moderator and a coolant, is defined between the opposing neutron absorber plates. Both the neutron absorber plates and the water in the water gap serve to enhance the reactivity worth, so that the control blade as a whole can exhibit a large value of reactivity worth. Alternatively, the amount of expensive hafnium having a high density (13.3 g/cm.sup.3) is reduced for attaining a given reactivity worth. In addition, the wire-type spacers arranged to extend in the axial direction of the central tie rod, together with the spot-like spacers dispersed over the entire region of the neutron absorber plates, serve to increase the strength of each wing of the control blade against bending external force. Although preferred embodiments of the nuclear reactor control blade of the invention have been described, it is to be noted that these embodiments are only illustrative and various changes and modifications may be imparted thereto without departing from the scope of the invention which is limited solely by the appended claims. |
claims | 1. A system for producing heat comprising:a proton beam source for producing a proton beam, wherein the proton source is adapted to vary the energy level of the produced proton beam between a first energy level of at least approximately 4.5 MeV and a second energy level of at least approximately 2.4 MeV, wherein the proton source includes a power input for receiving power to drive the proton beam source;a Thorium molten salt assembly comprising:a main assembly body;a tubular member positioned within the main assembly body,a lid coupled to the main assembly body in the form of a circular disk defining a plurality of openings passing therethrough, the plurality of openings including:a first opening passing through a top lid having a diameter larger than the other openings, the first opening defining a top window opening through which protons from the proton source may pass;a window element positioned within the first opening; the window element being formed of a material permitting the passage of protons therethrough;a molten salt solution contained within the main assembly body; the molten salt solution containing Thorium and Lithium;a plurality of Thorium fuel rods positioned within the tubular member body and arranged such that the top portions of each Thorium fuel rod is below the first opening in the lid, each solid Thorium fuel rod comprising:an inner member comprising Beryllium andan outer member formed from a solid that comprises at least some solid Thorium,wherein the outer member defines an opening passing through the Thorium fuel rod and the inner member is located within the opening;a primary heat exchange assembly positioned within the main assembly body, the primary heat exchange assembly comprising:an input manifold the input manifold comprising a manifold element defining an inlet port on the top surface of the manifold element of a first diameter and a plurality of openings on the bottom surface of the manifold element, each of the openings on the bottom surface of the manifold element having a diameter less than the diameter of the inlet port;an outlet manifold, the outlet manifold comprising a manifold element defining an outlet port on the top surface of the manifold element of a first diameter and a plurality of openings on the bottom surface of the manifold element, each of the openings on the bottom surface of the manifold element having a diameter less than the diameter of the outlet port; anda plurality of coiled tubular elements, each of the plurality of coiled tubular elements having a first end coupled to one of the openings on the bottom surface of the input manifold and a second end coupled to one of the openings on the bottom surface of the outlet manifold;an input pipe having a first end coupled to the inlet port of the input manifold, the input pipe passing through an opening in the top lid; andan outlet pipe having a first end coupled to the outlet port of the outlet manifold, the outlet pipe passing through an opening in the top lid;wherein the proton beam source is adapted to direct a proton beam of the first energy level such that the beam is directed to a Beryllium core of at least one Thorium fuel rod to promote the generation of fast neutrons and the fission of Thorium within the Thorium fuel rod to generate heat, and wherein at least a portion of the Thorium in the molten salt is converted to Uranium;wherein the proton beam is further adapted to direct a proton beam of the second energy level into the molten salt within the main assembly body to promote the generation of thermal neutrons and the fission of Uranium within the molten salt to generate heat; andwherein the manifolds and pipes comprising the primary heat exchange assembly are adapted to circulate molten salt to remove heat from the interior of the main assembly body. 2. The system for producing heat of claim 1 wherein the window element comprises titanium. 3. The system of producing heat of claim 1 wherein the outer surface of each of the Thorium rods defines a spiral shape. 4. The system of producing heat of claim 1 wherein each Thorium fuel rod extends at least ⅔ of the way down into the main assembly body. 5. The system of producing heat of claim 1 wherein the main assembly body is rounded at the bottom. |
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description | With reference to FIG. 1, an accelerator 10 is controlled by a beam voltage and current controller 12 to generate a beam of electrons with a preselected energy (MeV) and beam current. In the preferred embodiment, the electrons are generated by a Rhodotron brand name accelerator in the range of 1-10 MeV. A sweep control circuit 14 controls electromagnets or electrostatic plates of a beam deflection circuit 16 to sweep the electron beam, preferably back and forth in a selected plane. A titanium or aluminum window 18 of a vacuum horn 20 defines the exit from the vacuum system from which the electron beam 22 emerges for the treatment process. An electron absorbing plate 24 collects electrons and channels them to ground. A conveying system conveys items 30 through the e-beam 22. In the illustrated embodiment, the conveyor system includes a horizontal belt conveyor 32 which is driven by a motor 34. A motor speed controller 36 controls the speed of the motor. Of course, other types of conveyor systems are contemplated, including overhead conveyors, pneumatic or hydraulic conveyors, spaced palettes, and the like. In the illustrated belt conveyor system, the items 30 are positioned one after another on the conveyor belt closely packed with a minimal gap in between. Preferably, the items are packages or palettes of fixed size which hold individual items to be irradiated. A plurality of radiation detector arrays 40a, 40b, are positioned in the path of the e-beam 22. The first detector array 40a is in array that measures the strength (energy) of the electron beam after it has exited the item. The optional second detector array 40b detects the energy of the e-beam before it enters the product, if the energy is not otherwise known. The outputs of both the detector arrays 40a, 40b are conveyed to an amplifier section 44 for amplification. In the preferred embodiment, the outputs are digitized 46, serialized 48, converted into optical signals 50, and conveyed to a remote location. The amplifier section 44 is shielded to protect the electronics from stray electrons and static fields that might interfere with the electronic processing. The optical signal is conveyed to a location remote from such stray charges where it is converted to selected electronic format 52 and analyzed by a processor 54, such as a computer. Preferably, the beam control 12 provides the energy of the electrons entering the product. The computer subtracts or otherwise compares the strength of the electron beam before and after it enters each item. The processor 54 further compares the strength of the beam at various distances from the conveyor (heights in the illustrated embodiment) to identify regions in which high density materials may be interfering with. complete irradiation of the downstream material. The processor determines the dose received by each region of each item and forwards that dose information to an archival system 56 such as a computer memory, a tape, or a paper printout. In a first alternate embodiment, the processor 54 compares the measured dose information with preselected dose requirements. Based on differences between the selected and actual dosage, a parameter adjustment processor 58 adjusts one or more of the beam energy, the beam sweep, the conveyor speed, and the like. For example, when the detectors detect that near portions of the items are absorbing too much radiation leaving far portions of the items under irradiated, the parameter adjustment processor 58 increases or adjusts the accelerator to increase the MeV or the electron beam current, up to maximum values set for the items being irradiated. Once the maximum dose is reached, the adjustment processor 58 controls the motor speed controller 36 to reduce the speed of the conveyor. When the items have small regions of higher density, the sensing of an increase in the absorbed radiation causes the parameter adjustment processor 58 to increase the energy of the electron beam or decrease the speed of the conveyor until the region of higher density has passed through the beam. Thereafter, the beam power can be reduced or the conveying speed can be increased. Analogously, when the region of higher density is localized vertically, in the illustrated horizontal conveyor embodiment, the parameter adjustment processor 56 causing the sweep control circuit 14 to adjust the sweep such that the electron beam is directed to the higher density region for a longer duration. Preferably, the beam strength and the conveying speed are also adjusted to maintain the appropriate dosing in other regions of the package. Analogously, in response to regions of little absorption of the electron beam, the sweep circuit can be controlled to dwell for a shorter percentage of the time on these regions. In the preferred embodiment, the detectors are inductive detectors that detect the increases and decreases in electron beam energy. That is, although the electron beam may be viewed as a beam that is the full width of the horn 20, more typically the beam of electrons is focused into about a pulsed two centimeter diameter ray. This ray is swept up and down rapidly compared to the speed of the conveyor such that the electron beam is effectively a wall. More specifically to the preferred embodiment, and with reference to FIG. 2 each detector array includes a first coil or current transformer 60 and a second coil or current transformer 62. Between them, a metal foil 64, aluminum in the preferred embodiment with a selected energy absorption profile, is disposed. Both current transformers 60, 62 and the metal foil 64 are located within a vacuum chamber 66. The pulsed electron beam passes through a collimator 68 equipped with a cooling system and passes through the first current transformer 60. The sweeping electron beam 22 sends electron beam current pulses through the first transformer which induces currents circumferentially therearound in the first transformer which induced current is measured and the measurement held or stored. The beam passes through the metal foil, which is 3xc3x9710xe2x88x924 to 6xc3x9710xe2x88x924 m thick aluminum in the preferred embodiment. The beam passes through the second current transformer 62, again inducing currents. The second induced current is less than the first induced current by the amount of absorption in the foil which is based on the thickness of the metal foil 64. The currents are compared, and from that information, the energy of the electron beam is determined. The energy of the electron beam can be determined empirically by measuring the current drop between the two coils with electron beams of different known energies. Alternately, the energy can be calculated from the physics of the detector including foil thickness, atomic number of the metal in the foil, number of turns in the transformer coil, and the like. More specifically, the scanning mode of the electron accelerator leads to a pulsed character of the electron beam in cross-section. The primary electron beam has a current I0 and kinetic energy E0. After propagation of the electron beam across the irradiated product, the electron beam has a kinetic energy E1. The number of electrons is the same on both sides of the product, because electrons only lose kinetic energy. In the detector, the measurement of the electron beam current in front and behind the absorption foil 64 by the transforms 60, 62 enables the determination of an absorption factor K of the electron beam within the foil: K=I2/I1=f(E)xe2x80x83xe2x80x83(1) where, I1, is the beam current in front of the foil and I2 is the current behind the foil. The charge Q of the beam after the foil is: Q=Q0*exe2x88x92(m/p)*dxe2x80x83xe2x80x83(2) where Q is charged after the foil and Q0 is the charge before the foil. M/p is the mass absorption coefficient for the foil and is a function of the energy, f(E), and d is the thickness of the foil. Recognizing that current is charge per unit time, Q=Q0*exe2x88x92(m/p)*d yields: I2=I1*exe2x88x92(m/p)*dxe2x80x83xe2x80x83(3) From measurements with a plurality of different foil thicknesses, the dependence of K on the kinetic energy of the electrons can be calibrated. Hence, the kinetic energy of the measured electrons can be determined. Looking to FIG. 3, a standard dependency for the coefficient of partial transmission of energy for aluminum foils of 300 and 500 xcexcm is illustrated. After the determination of E1from these measurements, the energy absorbed in the product Ep is calculated by: Ep=E0xe2x88x92E1xe2x80x83xe2x80x83(4) From the beam current which the accelerator is controlled to put out, the scanning rate and other parameters of the electron beam in the scan horn, and a diameter of the hole in the collimator 68, one can determine the number of electrons Ne passing through the detector. The absorbed Joule""s energy Ej, in the product: Ej=Ep*Ne*1.6*10xe2x88x9219 [J]xe2x80x83xe2x80x83(5) Because the total mass of the product or package is known, the mass of the product along the ray in front of the detector with the diameter of the collimator hole is: M=0.8D2c*L*pxe2x80x83xe2x80x83(6) where p is the density of the product, L is the thickness of the product, and Dc is beam diameter after collimation. Hence, the absorb dose D is: D=Ej/Mxe2x80x83xe2x80x83(7) The processor 54 calculates this factor. The processor is preferably preprogrammed with lookup tables to which this factor is compared. Based on this comparison, the parameter adjustment processor 58 makes appropriate adjustments to process controls, a human readable display indicative of dosing is produced, data is stored in the archival system 56, or the like. Although illustrated relatively large in comparison to the items, it is to be appreciated that the individual detectors can be very small compared to the items. The array 40a may, for example, include hundreds of individual detectors. The array 40b may, for example, be only a single detector. It is also to be appreciated that the electron beam can be swept in other dimensions. For example, the beam can also be swept parallel to the direction of motion of the conveyor. When the beam is swept in two dimensions, it cuts a large rectangular swath. The electron density entering a unit area of the item per unit time is lower, but the product remains within the beam longer. The side to side movement of the beam allows for the placement of a two dimensional array above or below the items to measure absorbed dose in two dimensions. It is further to be appreciated that this detection system can be used to detect charged beams in numerous other applications. For example, this detector can be used in conjunction with electron beams that are used to create coatings by the synthesis of powdered material, such as diamond like coatings (dlc) on tools, nanophase silicon nitrite coatings, high purity metal coatings, and the like. It can be used with charged particle beams for surface modification such as cleaning of metals, surface hardening of metals, corrosion resistance, and other high temperature applications. The detector can also be used for electron beams which are used in the destruction of toxic gases such as the cleaning of flue gases for oxides of sulfur and nitrogen, removal of exhaust gases from diesel engines, destruction of fluorine gases, destruction of aromatic hydrocarbons, and the like. The detector may also be used with charged particle beams for treating liquid materials such as for the destruction of organic wastes, the breaking down of potentially toxic hydrocarbons such as tricloroethylenes, propanes, benzenes, phenols, halogenated chemicals, and the like, and for drying liquids, such as ink in printing machines, lacquers, and paints. The detector may also be used to monitor charged particles beams in the food industry such as the disinfection of food stuffs such as sugar, grains, coffee beans, fruits, vegetables, and spices, the pasteurization of milk or other liquid foods, sanitizing meats such as poultry, pork, sausage, and the like, inhibiting sprouting, and extending storage life. It will also find application in conjunction with monitoring electron and other charged particle beams used to form other particles or other types of radiation, such as the generation of ultraviolet irradiation, conversion of the electron beam to x-rays or gamma rays, the production of neutrons, eximer lasers, the production of ozone, and the like. The present system may also be used to monitor charged particles beams in conjunction with polymers and rubbers. The e-beam irradiation can be used for the controlled cross linking of polymers, degrading of polymers, drafting of polymers, modification of plastics, polymerization of epoxy compounds, sterilization of polymer units, vulcanization of rubber, and the like. It is to be appreciated that the determination of dose absorption can also be used to determine the local mass of the product. The invention has been described with reference to the preferred embodiment. 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. |
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059050144 | abstract | A radiation image storage panel is provided having a support, an intermediate layer and a phosphor layer comprising a binder and a stimulable phosphor dispersed therein, said panel being colored with a colorant so that the mean reflectance of said panel in the wavelength region of the stimulating rays for said stimulating phosphor is lower than the mean reflectance of said panel in the wavelength region of the light emitted by said stimulable phosphor upon stimulation thereof, characterized in that said colorant is a triarylmethane dye having at least one aqueous alkaline soluble group and is present in at least one of said support, said phosphor layer or an intermediate layer between said support and said phosphor layer. |
claims | 1. An apparatus comprising:a nuclear reactor including a pressure vessel and a nuclear reactor core disposed in the pressure vessel;a subterranean containment structure containing the nuclear reactor; andan ultimate heat sink pool disposed at grade level wherein an upper portion of the subterranean containment structure defines at least a portion of the bottom of the ultimate heat sink pool;wherein the upper portion of the subterranean containment structure comprises an upper dome that defines at least a portion of the bottom of the ultimate heat sink pool; andwherein an uppermost extremity of the upper dome of the subterranean containment structure extends above the surface of the ultimate heat sink pool to define an island surrounded by the ultimate heat sink pool. 2. The apparatus of claim 1, wherein the upper dome of the subterranean containment structure includes grooves or undulations. 3. The apparatus of claim 1, wherein the subterranean containment structure comprises steel. 4. The apparatus of claim 1, wherein the nuclear reactor comprises a pressurized water reactor (PWR) and the subterranean containment structure is large enough to simultaneously accommodate both the PWR and at least one steam generator designed to operate in or with the PWR. 5. The apparatus of claim 4, wherein the PWR is an integral PWR and the subterranean containment structure is large enough to simultaneously accommodate both the PWR and an internal steam generator disposed outside of the PWR but designed to operate in the integral PWR. 6. An apparatus comprising:a nuclear reactor including a pressure vessel and a nuclear reactor core disposed in the pressure vessel;a subterranean containment structure containing the nuclear reactor; andan ultimate heat sink pool disposed at grade level wherein an upper portion of the subterranean containment structure defines at least a portion of the bottom of the ultimate heat sink pool;a secondary containment structure containing the subterranean containment structure and the ultimate heat sink pool, the secondary containment structure having vents arranged to allow water evaporated or boiled off of the ultimate heat sink pool to escape from the secondary containment structure; andgutters disposed in the secondary containment structure to admit surface water from outside the secondary containment structure into the ultimate heat sink pool. 7. An apparatus comprising:a nuclear reactor including a pressure vessel and a nuclear reactor core disposed in the pressure vessel;a subterranean containment structure containing the nuclear reactor; andan ultimate heat sink pool disposed at grade level wherein an upper portion of the subterranean containment structure defines at least a portion of the bottom of the ultimate heat sink pool; andwherein the portion of the bottom of the ultimate heat sink pool defined by the upper portion of the subterranean containment structure has an area of at least Q decay heat U · ( T max - T UHS ) where Tmax denotes the maximum allowable temperature inside the subterranean containment structure, TUHS denotes the maximum allowable temperature of the ultimate heat sink pool, Qdecay heat denotes the highest postulated value for heat generated by fission product decay following reactor shutdown, and U denotes the overall heat transfer coefficient for heat transfer from the a subterranean containment structure to the ultimate heat sink pool. 8. The apparatus of claim 7, wherein the ultimate heat sink pool has a capacity of at least 300,000 gallons. 9. An apparatus comprising:a nuclear reactor including a pressure vessel and a nuclear reactor core disposed in the pressure vessel;a subterranean containment structure containing the nuclear reactor; andan ultimate heat sink pool disposed at grade level wherein an upper portion of the subterranean containment structure defines at least a portion of the bottom of the ultimate heat sink pool;a condenser comprising a heat exchanger including hot and cold flow paths disposed inside the subterranean containment structure; andcooling water lines operatively connecting the condenser with the ultimate heat sink pool. 10. The apparatus of claim 9, wherein ends of the cooling lines disposed in the ultimate heat sink pool terminate in one of (i) open ends and (ii) connections with a heat exchanger disposed in the ultimate heat sink pool. 11. An apparatus comprising:a pressurized water reactor (PWR) including a pressure vessel and a nuclear reactor core disposed in the pressure vessel;a subterranean containment structure containing the nuclear reactor; andan ultimate heat sink pool having a bottom defined at least in part by an upper portion of the subterranean containment structure, wherein the upper portion of the subterranean containment structure comprises an upper dome and the upper dome of the subterranean containment structure protrudes above the surface of the ultimate heat sink pool to define an island surrounded by the ultimate heat sink pool. 12. An apparatus comprising:a pressurized water reactor (PWR) including a pressure vessel and a nuclear reactor core disposed in the pressure vessel;a subterranean containment structure containing the nuclear reactor; andan ultimate heat sink pool having a bottom defined at least in part by an upper portion of the subterranean containment structure, wherein a contact area Awet between the ultimate heat sink pool and the upper portion of the subterranean containment structure satisfies a criterion U·Awet·ΔTmin≧Qdecay heat where Qdecay heat denotes the highest postulated value for heat generated by fission product decay following reactor shutdown, and U denotes the overall heat transfer coefficient for heat transfer from the a subterranean containment structure to the ultimate heat sink pool, and ΔTmin denotes the minimum temperature difference between the subterranean containment structure and the ultimate heat sink pool postulated to occur during any accident scenario under consideration. 13. An apparatus comprising:a pressurized water reactor (PWR) including a pressure vessel and a nuclear reactor core disposed in the pressure vessel;a subterranean containment structure containing the nuclear reactor;an ultimate heat sink pool having a bottom defined at least in part by an upper portion of the subterranean containment structure;a condenser comprising a heat exchanger including hot and cold flow paths disposed inside the subterranean containment structure; andcooling water lines operatively connecting the condenser with the ultimate heat sink pool. 14. An apparatus comprising:a nuclear reactor including a pressure vessel and a nuclear reactor core disposed in the pressure vessel;a containment structure containing the nuclear reactor;an ultimate heat sink pool disposed on top of the containment structure wherein the containment structure defines a bottom of the ultimate heat sink pool;a condenser comprising a heat exchanger including hot and cold flow paths disposed inside the containment structure; andcooling water lines operatively connecting the condenser with the ultimate heat sink pool. |
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041697600 | claims | 1. A nuclear reactor comprising in combination: a. a reactor core having an active region containing fissionable material; b. a group of moveable control rods insertible into said core for controlling the power of said nuclear reactor core; c. moveable control means including at least one rod normally maintained partially inserted in the core for controlling asymmetrical axial power distributions, said control rod having a length substantially equal to the length of said active region of the core and having a first portion at said rod's first end containing a first neutron poison, a second portion at said rod's second end containing a second neutron poison, and a third portion whose opposite ends are adjacent to and between said first and second portions, said third portion having a small reactivity control worth relative to said first and second portions; and d. a plurality of scrammable control rod drive mechanisms for moveably positioning each of said control rods of said group and of said control means, said control rod drive mechanisms all being of the type which rapidly insert their associated control rods into said reactor core to their full-in positions when an emergency shut down of the reactor is desired. 2. The nuclear reactor as recited in claim 1 wherein said second neutron absorbing material has a smaller macroscopic neutron cross-section than said first neutron absorbing material. 3. The nuclear reactor as recited in claim 2 wherein said first neutron absorbing material is boron carbide (B.sub.4 C). |
050200836 | description | DETAILED DESCRIPTION OF PREFERRED EMBODIMENTS FIG. 1 shows an intermediate structure in the production of one embodiment of the membrane according to the invention. Here, a wafer 10 includes a monocrystalline silicon substrate 12 and an upper region 14 which is 1 to 2 microns deep. The upper portion of the region 14 contains a silicon dioxide stratum 16 deposited or grown to a depth of 100 Angstroms to several thousand (e.g. 3000) Angstroms, and a lower stratum 18 of heavily boron-doped silicon. An etched resist 20 and the oxide stratum 16 contain patterned recesses 22. The structure of FIG. 1 is prepared by (1) forming the monocrystalline silicon wafer 10 with a 1,0,0 orientation; (2) forming the heavily boron-doped stratum 18, about 10.sup.20 /cc, in the upper surface of the wafer 10 by ion implantation or diffussion to a depth of 1 micron or 2 microns (depending on the desired thickness of the membrane to be produced); (3) creating the silicon dioxide stratum 16 to a depth of 100 Angstroms to several thousand Angstroms on the upper surface of the doped region 14 by heating in an oxygen or steam atmosphere so as to grow the oxide stratum 16 from the stratum 18, or by vapor depositing the oxide stratum 16 on the stratum 18; (4) applying the resist 20 over the silicon dioxide stratum 16; (5) generating a latent chemical image in the form of a pattern in the shape of recesses 22 in the resist 20; and (6) etching the pattern through the oxide stratum 16. The fabricating process then converts the wafer 10 to the condition shown in FIG. 2. This is done by (7) stripping the resist 20 and cleaning the wafer 10 to expose the oxide stratum 16 with its recesses 22; and (8) "trenching" the patterned recesses 22 into the doped silicon stratum 18 in a reactive ion etching (RIE) machine to a depth less than the depth of the stratum 18. This produces the "trenches" 28 in the shape of the patterned recesses 22 in the stratum 18. The resulting etched wafer 10 includes the substrate 12 which supports pattern etched strata 16 and 18. The process continues by (9) simultaneously sputter depositing platinum into, and back sputtering platinum from, the recesses 22 and trenches 28 at about 350.degree. C. to 600.degree. C. in an Argon atmosphere as shown in FIG. 3. Here, a platinum "cathode" 32 faces the stratum 16 and a voltage source 34 applies an RF potential between the cathode and the wafer 10 through a capacitor 36. A heat source 38 heats the wafer 10, the cathode 32 to about 350.degree. C. to 600.degree. C. Depending on the temperature and voltage, platinum sputters from the cathode 32 onto the upper surfaces of the wafer 30 and then back sputters onto the platinum cathode 32. The term "cathode" is used to describe the platinum source 32 even though the voltage actually applied is RF. The circuit partially rectifies the applied RF voltage. It generates a DC voltage of 25 v to 250 v. The applied voltage varies the sputtering rate from the platinum cathode and the back sputtering rate from the wafer 10 surface. In FIG. 3, the voltage is selected to establish a back sputtering rate equal to the sputtering rate. This keeps the silicon dioxide surface of the stratum 16 clear of platinum. However, sputtering platinum onto the walls of the silicon trenches 28 in the stratum 18 causes formation of platinum silicides PtSi and PtSi.sub.2 40, which line the trench walls. The formation of platinum silicides 40 continues during sputtering until the silicides fill and overflow the trenches 28. The back sputtering prevents deposition of platinum on the surface of the oxide stratum 16. The fabricating process then resumes with (10) etching away the substrate 12 and the SiO.sub.2 layer 16 to obtain the membrane 44 in FIG. 4. The membrane 44 includes platinum free surfaces on X-ray transparent, boron-doped, stratum 18, and X-ray opaque protrusion 42 of silicides 40 following the shape of the pattern on the resist 20. This makes the membrane 44 eminently suitable for creating densely packed patterns on semiconductor chips. FIG. 5 illustrates the use of the membrane 44 in a chip manufacturing arrangement. Here, a holder 46 positions the membrane 44 between an X-ray source 48 and a raw chip 50 having an X-ray sensitive resist 52 to produce an image pattern on the chip. An aligner 54 aligns the chip relative to the mask 44. This chip is then completed by further known steps, including diffusion and etching steps, and possibly including further exposure to X-rays through another membrane. FIGS. 6 to 8 illustrate another process and product embodying the invention. In FIG. 6, a silicon wafer 60 includes a silicon substrate 62 supporting a region 64, 1 to 2 microns deep. The upper portion of the region 64 contains an upper silicon dioxide stratum 66 having a depth of 100 .ANG. to 3000 .ANG., and a lower stratum 68 of boron-doped silicon. A resist 70 includes an upper silicon rich layer 72 in the shape of a pattern having elevations 74 and recesses 76. The structure of FIG. 6 is prepared by: (1) forming the silicon wafer 60; (2) forming the heavily doped stratum 68, about 10.sup.20 /cc, in the upper surface of the wafer 60 by ion implantation or diffussion to a depth of 1 micron to 2 microns (depending on the desired thickness of the membrane to be produced); (3) creating the silicon dioxide stratum 66 to a depth of 100 .ANG. to 3000 .ANG. in the upper surface of the doped region 64 by heating air in an oxygen or steam atmosphere; (4) applying the resist 70 over the silicon dioxide stratum 66; (5) forming a latent chemical "image" in the shape of a pattern with a low level electron beam; (6) exposing the latent image to a hexamethyl disilozane (HMDS) reagent to produce the silicon rich layer 72; (7) exposing the resist 70 to oxygen reactive ion etching to etch away the regions 76 not affected by HMDS. Where the resist 70 contains the silicon rich layer 72, i.e. where it is doped, the etch rate is low; the resist is not silicon rich, i.e. where it is not doped, the etch rate is high. This produces resist elevations 74 and depressions 76. The process continues by: (8) etching through the SiO.sub.2 layer 66 in an atmosphere that etches oxides, such as fluoride compounds. This results in the structure of FIG. 6. The wafer 60 of FIG. 7 results from: (9) stripping the resist 70; (10) subjecting the wafer surface to an atmosphere of WF.sub.6 +SiH.sub.4 +H.sub.2 at a temperature of 200.degree. C. to 500.degree. C. This deposits tungsten by displacement of Si, partially from the SiH.sub.4 (silane) and partially from the silicon stratum 68, and forms tungsten penetrations 78. The WF.sub.6 +SiH.sub.4 +H.sub.2 atmosphere will not replace the oxide in the stratum 66 and therefore the deposition follows the imaged pattern of the depressions 76. Once the deposition has penetrated in the stratum 68, the tungsten deposition continues upwardly and produces tungsten elevations or plugs 80. The next steps involves: (11) stripping off the oxide stratum 66 and the main central portion 82 of the substrate 62 to form the membrane 84 in FIG. 8. The membrane 84 is used in the same way as the membrane 44 in FIG. 5. That is, the membrane 84 is used in place of the membrane 44 of FIG. 5. An advantage of the process in steps 6 to 8 resides in that successive steps can be carried out in a vacuum atmosphere and continue in a vacuum atmosphere. FIG. 9 is a flow chart illustrating the process shown with respect to FIGS. 1 to 4. FIG. 10 is a flow chart illustrating the process shown with respect to FIGS. 6 to 8. While the drawings show only simple opaque protrusions 42 and plugs 80, these represent portions of complex protruding opaque patterns of the type used to form complex chips. The latent chemical image of step 5 in both instances is produced by known means such as those using light or E-beams. The resists 20 and 70 are chosen to accommodate the known means. While embodiments of the invention have been described in detail, these are only examples, and it will be evident to those skilled in the art that the invention may be embodied otherwise. |
claims | 1. A method of manufacturing a mask, the method comprising:designing a second mask data pattern for forming a first mask data pattern;creating a first emulation pattern, which is determined from the second mask data pattern, using a first emulation;creating a second emulation pattern, which is determined from the first emulation pattern, using a second emulation;comparing a pattern, in which the first and second emulation patterns overlap, with the first mask data pattern; andmanufacturing a mask layer, which corresponds to the second mask data pattern, according to results of the comparison. 2. The method of claim 1, wherein the first emulation and the second emulation comprise a general emulation and a self-aligning double patterning (SADP) emulation, respectively. 3. The method of claim 2, wherein creating the second emulation pattern comprises:defining one or more parameters related to respective step processes of an SADP process;modeling each of the step processes using the one or more parameters; andcreating the second emulation pattern by applying step process models to the first emulation pattern. 4. The method of claim 3, wherein the step processes of the SADP process comprises:forming an etching target layer and a first hard mask layer pattern on a substrate;conformally forming a sacrificial layer on the first hard mask layer pattern and an exposed etching target layer;forming a second hard mask layer on the sacrificial layer;planarizing the second hard mask layer and the sacrificial layer so that an upper surface of the first hard mask layer pattern is exposed;removing the sacrificial layer exposed between the first hard mask layer pattern and the planarized second hard mask layer; andremoving the etching target layer exposed between the first hard mask layer pattern and the planarized second hard mask layer. 5. The method of claim 3, wherein the one or more parameters comprise a deposition thickness of at least one layer formed through each of the step processes, and a degree of skew that occurs after performing each of the step processes. 6. The method of claim 1, wherein comparing the pattern, in which the first and second emulation patterns overlap, with the first mask data pattern comprises:finding conflict points at which the pattern, in which the first and second emulation patterns overlap, and the first mask data pattern do not match with each other; andclassifying the conflict points. 7. The method of claim 1, further comprising:comparing the pattern, in which the first and second emulation patterns overlap, with the first mask data pattern; andmodifying the second mask data pattern according to the results of the comparison. 8. The method of claim 1, further comprising:comparing the pattern, in which the first and second emulation patterns overlap, with the first mask data pattern; anddesigning third mask data patterns, which are used to create partial patterns that are not formed using the second mask data pattern, according to the results of the comparison. 9. The method of claim 8, wherein each of the third mask data patterns comprises at least one of trimming patterns, connection patterns, or dummy patterns. 10. The method of claim 1, further comprising creating a third emulation pattern, which is determined from the second mask data pattern, using a layout-based self-aligning double patterning (SADP) emulation. 11. The method of claim 10, further comprising:comparing the third emulation pattern with the first mask data pattern; andmodifying the second mask data pattern according to the results of the comparison. 12. The method of claim 10, further comprising:comparing the third emulation pattern with the first mask data pattern; anddesigning the third mask data patterns, which are used to create the partial patterns that are not formed using the second mask data pattern, according to the results of the comparison. 13. The method of claim 12, wherein each of the third mask data patterns comprises at least one of trimming patterns, connection patterns, or dummy patterns. 14. The method of claim 1, further comprising, before performing the first emulation, performing Optical Proximity Correction (OPC) on the second mask data pattern. 15. The method of claim 1, wherein the first emulation pattern and/or the second emulation pattern have a hierarchical structure. 16. A method of manufacturing a mask, the method comprising:designing a first mask data pattern;designing a second mask data pattern for forming the first mask data pattern;creating a first emulation pattern, which is determined from the second mask data pattern, using a layout-based self-aligning double patterning (SADP) emulation;comparing the first emulation pattern with the first mask data pattern;modifying the second mask data pattern according to results of the comparison;performing an Optical Proximity Correction (OPC) on the modified second mask data pattern;creating a second emulation pattern, which is determined from the second mask data pattern, on which the OPC has been performed, using a general emulation;creating a third emulation pattern, which is determined from the second emulation pattern, using an SADP emulation;comparing a pattern, in which the second and third emulation patterns overlap, with the first mask data pattern; andmanufacturing a first mask layer, which corresponds to the second mask data pattern, according to results of the comparison. 17. The method of claim 16, wherein creating the third emulation pattern comprises:defining one or more parameters related to each of step processes of an SADP process;modeling each of the step processes using the parameters; andcreating the third emulation pattern by applying step process models to the second emulation pattern. 18. The method of claim 16, further comprising:comparing the pattern, which is obtained through overlapping, with the first mask data pattern;and modifying the second mask data pattern according to the results of the comparison. 19. The method of claim 16, further comprising:comparing the pattern, which is obtained through overlapping, with the first mask data pattern;and designing third mask data patterns, which are used to create partial patterns that are not formed using the second mask data pattern, according to results of the comparison. 20. The method of claim 16, wherein the comparing the first emulation pattern with the first mask data pattern and modifying the second mask data pattern according to the results of the comparison comprises:modifying the second mask data pattern; anddesigning third mask data patterns for creating partial patterns that are not forming using the second mask data pattern. 21. A method of manufacturing a mask, the method comprising:designing a first mask data pattern;designing a second mask data pattern for forming the first mask data pattern;creating a first emulation pattern, which is determined from the second mask data pattern, using a layout-based self-aligning double patterning (SADP) emulation;comparing the first emulation pattern with the first mask data pattern;designing third mask data patterns, which arc used to create partial patterns that are not formed using the second mask data pattern, according to results of the comparison;performing an Optical Proximity Correction (OPC) on the second and third mask data patterns;creating a second emulation pattern, which is determined from the second mask data pattern on which the OPC has been performed, using a general emulation;creating a third emulation pattern, which is determined from the second emulation pattern, using an SADP emulation;creating a fourth emulation pattern, which is determined from the third mask data patterns on which the OPC has been performed, using the general emulation;comparing a pattern, in which the second to fourth emulation patterns overlap, with the first mask data pattern; andmanufacturing first and second mask layers, which respectively correspond to the second and third mask data patterns, according to results of the comparison. 22. The method of claim 21, wherein creating the third emulation pattern comprises:defining one or more parameters related to each of step processes of an SADP process;modeling each of the step processes using the parameters; andcreating the third emulation pattern by applying step process models to the second emulation pattern. 23. The method of claim 22, wherein the step processes of the SADP process comprise:forming an etching target layer and a first hard mask layer pattern on a substrate;conformally forming a sacrificial layer on the first hard mask layer pattern and an exposed etching target layer;forming a second hard mask layer on the sacrificial layer;planarizing the second hard mask layer and the sacrificial layer so that an upper surface of the first hard mask layer pattern is exposed;removing the sacrificial layer exposed between the first hard mask layer pattern and the planarized second hard mask layer; andremoving etching target layer exposed between the first hard mask layer pattern and the planarized second hard mask layer. 24. The method of claim 22, wherein the parameters include a deposition thickness of at least one layer formed through each of the step processes, and a degree of skew that occurs after performing each of the step processes. 25. The method of claim 21, wherein comparing the pattern, which is obtained through overlapping, with the first mask data pattern comprises:finding conflict points at which the pattern, which is obtained through overlapping, and the first mask data pattern do not match each other; andclassifying the conflict points. 26. The method of claim 21, further comprising:comparing the pattern, which is obtained through overlapping, with the first mask data pattern; andmodifying the second and third mask data patterns according to the results of the comparison. 27. The method of claim 21, wherein comparing the first emulation pattern with the first mask data pattern and designing the third mask data patterns comprises:modifying the second mask data pattern; anddesigning the third mask data patterns for creating partial patterns that are not formed using the second mask data pattern. 28. The method of claim 21, wherein at least one of the first to third emulation patterns has a hierarchical structure. 29. A method of manufacturing a mask, the method comprising:defining one or more parameters related to respective step processes of a self-aligning double patterning (SADP) process;modeling each of the step processes using the parameters; andcreating a second emulation pattern by applying step process models to a first emulation pattern. 30. The method of claim 29, wherein the step processes of the SADP process comprise:forming an etching target layer and a first hard mask layer pattern on a substrate;conformally forming a sacrificial layer on the first hard mask layer pattern and an exposed etching target layer;forming a second hard mask layer on the sacrificial layer;planarizing the second hard mask layer and the sacrificial layer so that an upper surface of the first hard mask layer pattern is exposed;removing the sacrificial layer exposed between the first hard mask layer pattern and the planarized second hard mask layer; andremoving the etching target layer exposed between the first hard mask layer pattern and the planarized second hard mask layer. 31. The method of claim 29, wherein the parameters include a deposition thickness of at least one layer formed through each of the step processes, or a degree of skew that occurs after performing each of the step processes. 32. The method of claim 29, wherein the first emulation pattern and/or the second emulation pattern have a hierarchical structure. |
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claims | 1. A transmitter device comprising:a neutron detector structured to generate electrical current from neutron flux;an oscillator circuit comprising an electrostatic switch electrically connected to said neutron detector, wherein said electrostatic switch is moveable based on the neutron detector; andan antenna electrically connected with said electrostatic switch,wherein said oscillator circuit is structured to pulse said antenna based on the neutron detector,wherein a period between pulses is related to the neutron flux, andwherein said antenna is structured to emit a signal corresponding to a number of characteristic values of said oscillator circuit. 2. The transmitter device of claim 1 wherein said oscillator circuit further comprises a capacitor and an inductor configured to be electrically connected with said capacitor; and wherein said capacitor is electrically connected with said neutron detector. 3. The transmitter device of claim 2 wherein said electrostatic switch comprises a first terminal, a second terminal, a first vane electrically connected with said first terminal, and at least one other vane electrically connected with said second terminal; wherein said capacitor is electrically connected with one of said first terminal or said second terminal; and wherein said antenna is electrically connected with the other of said first terminal or said second terminal. 4. The transmitter device of claim 3 wherein said electrostatic switch further comprises a first conductor and a second conductor; wherein said first conductor is electrically connected with said first terminal and said first vane; wherein said second conductor is electrically connected with said second terminal and said at least one other vane; wherein said electrostatic switch is structured to move from an OPEN position to a CLOSED position; wherein said electrostatic switch further comprises a bracket extending from a portion of said second conductor; wherein, when said electrostatic switch is in the OPEN position, said first conductor is spaced from said bracket; and wherein, when said electrostatic switch moves from the OPEN position to the CLOSED position, said first conductor moves into engagement with said bracket. 5. The transmitter device of claim 4 wherein said electrostatic switch further comprises a housing and a support fiber; wherein said first vane is disposed internal with respect to said housing; wherein said support fiber is coupled to said first vane and said housing and is configured to provide a torsional preload on said first vane; wherein, when said electrostatic switch moves from the OPEN position to the CLOSED position, electrostatic attractive forces between said first vane and said at least one other vane overcome the preload of said support fiber in order to move said electrostatic switch to the CLOSED position. 6. The transmitter device of claim 4 wherein said portion of said second conductor is disposed substantially parallel to said first conductor; wherein, when said electrostatic switch is in the CLOSED position, current flows through said first conductor in a first direction; and wherein, when said electrostatic switch is in the CLOSED position, current flows through said portion of said second conductor in a second direction generally opposite the first direction, thereby creating a repulsive electromagnetic force between said first conductor and said portion of said second conductor in order to maintain said electrostatic switch in the CLOSED position. 7. The transmitter device of claim 4 wherein said electrostatic switch further comprises a support post mechanically coupled to said first vane and electrically connected to said first terminal; and wherein said first conductor extends from said support post. 8. The transmitter device of claim 7 wherein said electrostatic switch further comprises a container and an electrically conductive substance disposed internal with respect to said container; and wherein said container electrically connects said support post to said first terminal. 9. The transmitter device of claim 4 wherein said at least one other vane comprises a second vane and a third vane each electrically connected with said second terminal; wherein said first vane is disposed between said second vane and said third vane; and wherein, when said electrostatic switch moves from the OPEN position to the CLOSED position, said first vane rotates toward said second vane and said third vane. 10. The transmitter device of claim 9 wherein said second vane and said third vane form a unitary component made from a single piece of material. 11. The transmitter device of claim 9 wherein, when said electrostatic switch is in the OPEN position, said first vane is disposed parallel to said second vane and said third vane. 12. The transmitter device of claim 2 wherein said oscillator circuit further comprises a resistance temperature detector electrically connected in series with said inductor; and wherein said resistance temperature detector is structured to alter the signal emitted by said antenna. 13. The transmitter device of claim 2 wherein said oscillator circuit further comprises a second inductor electrically connected in series with said inductor; and wherein said second inductor is structured to alter the signal emitted by said antenna. 14. The transmitter device of claim 13 wherein said second inductor is a variable inductor. 15. The transmitter device of claim 1 wherein said transmitter device is devoid of a semiconductor. 16. The transmitter device of claim 1 wherein said transmitter device comprises only one single powering mechanism; and wherein said one single powering mechanism is said neutron detector. 17. A nuclear reactor system comprising:a fuel assembly having a fuel rod; anda transmitter device comprising:a neutron detector disposed within said fuel rod, said neutron detector being structured to generate electrical current from neutron flux,an oscillator circuit comprising an electrostatic switch electrically connected to said neutron detector, wherein said electrostatic switch is moveable based on the neutron detector, andan antenna electrically connected with said electrostatic switch,wherein said oscillator circuit is structured to pulse said antenna based on the neutron detector,wherein a period between pulses is related to the neutron flux, andwherein said antenna is structured to emit a signal corresponding to a number of characteristic values of said oscillator circuit. |
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summary | ||
description | This application is a continuation of PCT Application PCT/US2003/023412, filed Jul. 25, 2003, and published under the PCT Articles in English as WO 2004/013867 A2 on Feb. 12, 2004. PCT/US2003/023412 claimed priority to U.S. Provisional Application No. 60/400,809, filed Aug. 2, 2002. The entire disclosures of PCT/US2003/023412 and U.S. Ser. No. 60/400,809 are incorporated herein by reference in their entirety. The U.S. Government has a paid-up license in this invention and the right in limited circumstances to require the patent owner to license others on reasonable terms as provided for by the terms of Contract #1 R43 RR14935-01 awarded by the National Institutes of Health. This invention relates generally to devices and methods for diffracting or focusing high-energy electromagnetic radiation. Specifically, the present invention provides improved methods and apparatus for directing or focusing x-rays using devices having a plurality of crystal optics having varying atomic diffraction planes. Implementation of x-ray analysis methods has been one of the most significant developments in twentieth-century science and technology. The use of x-ray diffraction, x-ray spectroscopy, x-ray imaging, and other x-ray analysis techniques has led to a profound increase in knowledge in virtually all scientific fields. In areas of x-ray spectroscopy, high x-ray beam intensity is an essential requirement to reduce sample exposure times and, consequently, to improve the signal-to-noise ratio of x-ray analysis measurements. In the past, expensive and powerful x-ray sources, such as rotating anode x-ray tubes or synchrotrons, were the only options available to produce high-intensity x-ray beams. Recently, the development of x-ray optical devices has made it possible to collect the diverging radiation from an x-ray source by focusing the x-rays. A combination of x-ray focusing optics and small, low-power x-ray sources can produce x-ray beams with intensities comparable to those achieved with more expensive devices. As a result, systems based on a combination of small x-ray sources and collection optics have greatly expanded the capabilities of x-ray analysis equipment in, for example, small laboratories. One existing x-ray optical technology is based on diffraction of x-rays on optical crystals, for example, germanium (Ge) or silicon (Si) crystals. Curved crystals can provide deflection of diverging radiation from an x-ray source onto a target, as well as providing monochromatization of photons reaching the target. Two different types of curved crystals exist: singly-curved crystals and doubly-curved crystals (DCC). Using what is known in the art as Rowland circle geometry, singly-curved crystals provide focusing in two dimensions, leaving x-ray radiation unfocused in the third or orthogonal plane. Doubly-curved crystals provide focusing of x-rays from the source to a point target in all three dimensions, for example, as disclosed by Chen and Wittry in the article “Microprobe X-ray Fluorescence with the Use of Point-focusing Diffractors,” which appeared in Applied Physics Letters, 71 (13), 1884 (1997), the disclosure of which is incorporated by reference herein. This three-dimensional focusing is referred to in the art as “point-to-point” focusing. The point-to-point focusing property of doubly-curved crystals has many important applications in, for example, material science structural analysis. Depending on the bending radii of the doubly-curved crystal in the Rowland optic circle plane, curved crystals further divide into Johansson and Johann types. Johansson geometry requires crystals to have a curvature that is equal to the radius of the Rowland circle, while Johann geometry configuration requires a curvature twice the radius of the Rowland circle. One limitation of crystals based on Johann geometry is a low radiation collection angle and, subsequently, reduced deflected beam flux and beam intensity. One way to overcome this limitation, proposed in U.S. Pat. No. 5,127,028, entitled “Diffractor with doubly curved surface steps” of Wittry, is to use more than one diffracting crystal in a stepped geometry. However, the radiation collection angle having stepped geometry, as disclosed in U.S. Pat. No. 5,127,028, still has limitations. For example, such stepped-geometry prior art crystals provide a limited x-ray collection angle are also difficult to manufacture. There exists a need in the art to provide an x-ray focusing device and method which provide a larger collection angle to provide an even higher intensity monochromatic x-ray beam than that provided by the existing art. X-ray sources typically generate diverging radiation. In order to increase x-ray beam flux, diverging radiation is typically collected and focused onto a target. Existing crystal-based focusing devices provide point-to-point focusing by diffracting x-ray radiation. Typically, the radiation collection angle of Johann-type optics is only between 1 degree and 5 degrees, that is, only a small fraction of the radiation emitted by an x-ray source typically reaches the target. Thus, there is a need in the art to provide devices and methods for capturing more of the divergent radiation and provide a high-intensity, x-ray beam focusing devices, systems, and methods with improved x-ray beam utilization. One significant advantage of providing a high-intensity x-ray beam is that the desired sample exposure can typically be achieved in a shorter measurement time. The potential to provide shorter measurement times can be critical in many applications. For example, in some applications, reduced measurement time increases the signal-to-noise ratio of the measurement. In addition, minimizing analysis time increases the sample throughput in, for example, industrial applications, thus improving productivity. There is a clear need in the art to provide devices, systems, and methods that can be used to enhance x-ray analysis methods by reducing experimental measurement time. The present invention provides methods and apparatus which address many of the limitations of prior art methods and apparatus. In the following description, and throughout this specification, the expressions “focus”, “focusing”, and “focused”, among others, may appear, for example, as in “focusing device”, “x-ray focusing device”, “means for focusing”, “focusing optic”, among others. Though according to the present invention these expressions can apply to devices or methods in which x-rays are indeed “focused”, for example, caused to be concentrated, these expressions are not meant to limit the invention to devices that “focus” x-rays. According to the present invention, the term “focus” and related terms are intended to also serve to identify methods and devices which collect x-rays, collimate x-rays, converge x-rays, diverge x-rays, or devices that in any way vary the intensity, direction, path, or shape of x-rays. All these means of handling, manipulating, varying, modifying, or treating x-rays are encompassed in this specification by the term “focus” and its related terms. Also, in the following discussion and throughout this specification, for ease of illustration, the prior art and the various aspects of the present invention will be discussed in terms of their application to the modification of the path of x-ray radiation, but it is understood that the present invention is applicable to the manipulation of other types of radiation, for example, x-rays, gamma rays, and neutrons. Thus, the scope of the present invention is not limited to the manipulation of x-ray beams. One aspect of the invention is an optical device for directing x-rays, the optical device including a plurality of optical crystals positioned with an x-ray source and an x-ray target to define at least one Rowland circle of radius R and a source-to-target line, wherein the optical device provides focusing of x-rays from the source to the target. In one aspect of the invention, the at least one of the plurality of optical crystals may have a surface upon which x-rays are directed, and wherein at least one of the plurality of optical crystals comprises a set of atomic diffraction planes having a Bragg angle θB and oriented at an angle γ with the surface of the at least one of the plurality of optical crystals, and wherein a line drawn from the x-ray source to a point on the surface of the at least one of the plurality of optical crystals makes an angle θB+γ with the source-to-target line. In another aspect of the invention, the line drawn from the x-ray source to a point on the surface of the at least one of the plurality of optical crystals may be a line drawn from the x-ray source to the midpoint of the surface of the at least one of the plurality of optical crystals. In one aspect of the invention, the line drawn from the x-ray source to a point on the surface of the at least one of the plurality of optical crystals may be a line drawn from the x-ray source to about the point of tangency of the surface of the at least one of the plurality of optical crystals and the Rowland circle of the at least one of the plurality of optical crystals. In one aspect of the invention, the plurality of optical crystals may have a radius in the plane of the Rowland circle of about 2R. In one aspect of the invention, at least one of the crystals is a doubly-curved crystal, for example, a toroidal doubly-curved crystal. In one aspect of the invention, the optical device may have a toroidal angle of at least about 30 degrees. In one aspect of the invention, the device may be combined with a source of x-rays. Another aspect of the invention is an optical device for directing x-rays, the optical device including a plurality of optical crystals positioned with an x-ray source and an x-ray target to define at least one Rowland circle of radius R and a source-to-target line, wherein the optical device comprises a toroidal angle about the source-to-target line of at least about 90 degrees. In one aspect of the invention, the optical device may have a toroidal angle about the source-to-target line of at least about 180 degrees, or at least about 270 degrees, or about 360 degrees. In one aspect of the invention, the device provides point-focusing of x-rays. In one aspect of the invention, at least one of the plurality of optical crystals has a surface upon which x-rays are directed, and wherein at least one of the optical crystals comprise a set of atomic diffraction planes having a Bragg angle θB and oriented at an angle γ with the surface of the at least one of the optical crystals and wherein a line drawn from the x-ray source to a point on the surface of the at least one of the optical crystals makes an angle θB+γ with the source-to-target line. In another aspect of the invention, the line drawn from the x-ray source to the point on the surface of the at least one of the optical crystals comprises a line drawn to the midpoint of the at least one of a plurality of optical crystals. In another aspect of the invention, the line drawn from the x-ray source to the point on the surface of the at least one of the optical crystals comprises a line drawn to the point of tangency of the surface of the at least one of the plurality of optical crystals and the Rowland circle of the at least one of the plurality of optical crystals. In one aspect of the invention, the plurality of optical crystals have a radius in the plane of the Rowland circle of about 2R. In another aspect of the invention, the optical device may further include a second plurality of optical crystals positioned with the x-ray source and the x-ray target to define at least one Rowland circle, wherein the second plurality of optical crystals have a radius in the plane of the Rowland circle of about 2R, and wherein the optical device comprises a toroidal angle about the source-to-target line of at least about 90 degrees. Another aspect of the invention is an optical device for directing x-rays, the device including a plurality of optical crystals arranged in a matrix, the matrix being positionable to define at least one Rowland circle with an x-ray source and an x-ray target, and wherein the matrix comprises a plurality of rows, with each row comprising multiple optical crystals of the plurality of optical crystals. In one aspect of this invention, at least one of the crystals is a doubly-curved crystal, for example, a toroidal doubly-curved crystal. In another aspect of the invention, the toroidal doubly-curved crystal defines a toroidal direction and the plurality of rows may be spaced in the toroidal direction or a direction orthogonal to a plane of at least one Rowland circle. In another aspect of the invention, the crystals may have at least one lattice plane and the at least one lattice plane of at least one of the crystals may be parallel to a surface of the crystal; in another aspect of the invention, the at least one lattice plane of at least one of the crystals may be non-parallel to the surface of the crystal. In another aspect of the invention, the at least one toroidal doubly-curved crystal defines a toroidal direction, and wherein an arcuate length of the device in the toroidal direction may be at least about 45 degrees, or at least about 60 degrees, or at least about 90 degrees. The device may also act as a monochromator. In another aspect of the invention, the device may further comprise the device in combination with the source of x-rays. In another aspect of the invention, the source of x-rays may consume less than about 100 Watts, typically less than about 50 Watts, and may even consume less than about 25 Watts or even less than about 10 Watts. Another aspect of this invention comprises a method for directing x-rays, the method including the steps: providing an optical device, the optical device comprising a plurality of optical crystals arranged in a matrix, the matrix being positionable to define at least one Rowland circle with an x-ray source and an x-ray target, and wherein the matrix comprises a plurality of rows, with each row comprising multiple optical crystals of said plurality of optical crystals; and positioning the optical device wherein at least some x-rays from the x-ray source are directed to the x-ray target. In one aspect of the invention of this invention, positioning the optical device may comprise positioning the device wherein at least some x-rays emitted by the source impinge at least some of the crystals of the optical device wherein at least some of the x-rays are diffracted. Another aspect of the invention is a device for directing x-rays, the device including a first curved crystal and at least one second curved crystal spaced from the first crystal, the first and at least one second curved crystal each including at least one lattice plane, and the first curved crystal and the at least one second curved crystal being positionable to define at least one Rowland circle with an x-ray source and an x-ray target, wherein at least some x-rays impinging upon the first curved crystal and the at least one second curved crystal are directed to the target, and wherein the angle of the at least one lattice plane of the first crystal relative to a surface of the first crystal is different from an angle of the at least one lattice plane of the at least one second crystal relative to a surface of the at least one second crystal. In one aspect of the invention, the angle of the lattice planes of the first crystal relative to the surface of the first crystal may be about zero. In one aspect of the invention, the angle of the at least one lattice plane of the at least one second crystal relative to the surface of the at least one second crystal is different from the angle of the lattice planes of the first crystal, for example, the angle of the lattice planes of the at least one second crystal may be different form zero degrees, for instance, about 1 to about 5 degrees. In another aspect of the invention, a line directed from the x-ray source to the center of a surface of the first curved crystal and a line directed from the x-ray source to the center of a surface of the at least one second crystal may define an angle γ. In one aspect of the invention, the angle of the at least one lattice plane of the at least one second crystal relative to the surface of the at least one second crystal may be an angle γ, for example an angle of between about minus 15 degrees and about plus 15 degrees or between about minus 4 degrees and about plus 4 degrees. In another aspect of this invention, the first curved crystal and the at least one second crystal may comprise a first set of crystals, and the device further comprises at least one second set of crystals which are also positioned to define a Rowland circle with the x-ray source and the x-ray target, wherein at least some x-rays which impinge upon the at least one second set of crystals are directed to the x-ray target, the target being common with the first set of crystals, and wherein the second set of crystals is spaced from the first set of crystals in a direction orthogonal to a plane of the Rowland circle of the first set of crystals. In one aspect of the invention, a radius of curvature of a surface of the first curved crystal in the plane of the Rowland circle and a radius of curvature of a surface of the at least one second crystal in the plane of the Rowland circle are about equal to twice the radius of the Rowland circle of the device. In one aspect of the invention, the device provides point focusing of x-rays on the x-ray target, for example, point-to-point focusing from the x-ray source to the x-ray target. In another aspect of the invention, the device further comprises a backing plate onto which the first curved crystal and at least one second curved crystal are mounted. In one aspect of the invention, the device comprises a monochromator. Another aspect of the invention is a device for directing x-rays, comprising a curved crystal optic positionable to define at least one Rowland circle with an x-ray source and an x-ray target, wherein at least some x-rays emitted by the source impinge upon the crystal and are directed to the target, the curved crystal optic comprising at least one lattice plane, wherein the at least one lattice plane of the curved crystal optic is oriented at an angle γ1 relative to a surface of the curved crystal optic. In one aspect of the invention, the curved crystal optic may be a doubly-curved crystal optic and have a curvature in a plane orthogonal to a plane of the Rowland circle, for example, having an arc length of the curved crystal optic in a direction orthogonal to a plane of the Rowland circle of at least about 45 degrees. In one aspect of the invention, the curved crystal optic may comprise a plurality of curved crystals. In one aspect of the invention, the arc length of the curved crystal optic in a direction orthogonal to the plane of the Rowland circle is at least about 90 degrees, or at least about 180 degrees, or about 360 degrees. In one aspect of the invention, the angle of orientation γ1 of the at least one lattice plane relative to the surface of the curved crystal optic may be between about minus 4 degrees and about plus 4 degrees. In one aspect of the invention, the crystal may have a bending radius of between about 20 mm and about 600 mm, for example, in one or more planes or directions. In another aspect of the invention, the device may further include a backing plate onto which the curved crystal optic is mounted. Another aspect of the invention is a circular optic for diffracting x-rays, comprising at least one curved crystal optic positionable to define at least one Rowland circle with an x-ray source and an x-ray target, wherein at least some x-rays impinging upon the curved crystal optic are directed to the target, wherein the at least one curved crystal optic comprises at least one lattice plane and wherein the at least one lattice plane of the at least one curved crystal optic is oriented at an angle γ1 relative to a surface of the at least one curved crystal optic. In one aspect of the invention, the at least one curved crystal optic may comprise at least one doubly-curved crystal. In another aspect of the invention, the at least one curved crystal optic may comprise a plurality of curved crystals. In one aspect of the invention, the angle γ1 may be between about minus 4 degrees and about plus 14 degrees. In one aspect of the invention, the circular optic may have a bending radius between about 20 mm and about 600 mm. In one aspect of the invention, the circular optic provides point focusing on the x-ray target (for example, on a sample), for example, point-to-point focusing from the x-ray source to the x-ray target. In one aspect of the invention, the circular optic may further comprise a backing plate onto which the at least one curved crystal optic is mounted. These and other embodiments and aspects of the present invention will become more apparent upon review of the attached drawings, description below, and attached claims. FIGS. 1 and 2 illustrate a typical prior art x-ray optical device over which the present invention is an improvement. Again, in the following description, the various aspects of the present invention will be discussed in terms of their application to the modification of the path of x-ray radiation, but it is understood that the present invention is applicable to the manipulation of other types of radiation, for example, x-rays, gamma rays, or neutrons, among other types. That is, the scope of the present invention is not limited to the manipulation of x-ray beams. FIG. 1 is a typical isometric view of a prior art x-ray focusing arrangement 10 and FIG. 2 is a cross-sectional view of arrangement 10 as viewed along lines 2—2 in FIG. 1. FIGS. 1 and 2 include the geometry of the corresponding Rowland circle 20 associated with arrangement 10. Arrangement 10 includes a doubly-curved crystal (DCC) optic 12, an x-ray source location 16, and an x-ray target location 18, at which the x-ray image is preferably produced. In FIGS. 1 and 2, and in subsequent aspects of the present invention, x-ray source location 16 represents the source location for a point-like x-ray source. Similarly, in FIGS. 1 and 2 and elsewhere in this specification, target location 18 may be any target at which x-rays or other radiation may be directed. For example, target location 18 may be the location of a chemical specimen undergoing x-ray spectroscopy, human tissue undergoing radiation treatment, or a semiconductor chip undergoing surface analysis, among other things. In addition, the target location 18 may include an x-ray detector for detecting secondary x-rays emitted by the target. As most clearly shown in FIG. 2, the optic 12 has an optic center point 14, and the x-ray source location 16, optic center point 14, and the x-ray target location 18 define a circle 20 known in the art as the Rowland circle or focal circle. Rowland circle 20 has radius RO defined in the art as the Rowland or focal radius. Crystal 12 has a width W and a height H, as shown in FIG. 1. X-ray source location 16 and an x-ray target location 18 are joined by line 22, which is referred to in the art as the “source-to-image connecting line”. The coordinate system of the arrangement shown in FIGS. 1 and 2 has its origin at the point O. In FIGS. 1 and 2, the surface of crystal 12 has a radius R measured from origin O. Crystal 12 typically contains one or more crystal lattice planes (also known as atomic diffraction planes) represented by lines 24. In this typical prior art optic the lattice planes 24 are essentially parallel to the surface of crystal 12. Though prior art optics may be designed for Johan or Johansson geometry, the arrangement shown in FIGS. 1 and 2 has Johan-type geometry in which the radius of curvature R of the surface of crystal 12 is twice the Rowland radius RO, that is, R=2RO. As most clearly shown in FIG. 1, prior art crystal 12 is typically a doubly-curved crystal (DCC), that is, in addition to having a radius of curvature R in the plane of circle 20 (that is, the Rowland plane), crystal 12 also has a radius of curvature r in the plane orthogonal to the plane of circle 20. The direction of curvature r is typically referred to in the art as the toroidal curvature of crystal 12, and r is referred to as the “toroidal rotation radius”. This toroidal direction is indicated by angle φ in FIG. 1. In order to provide essentially point-to-point focusing, DCC 12 typically has a toroidal rotation radius, r, that is equal to the perpendicular distance between crystal center point 14 and source-to-image connecting line 22. The angle θB shown in FIG. 2 is known in the art as the “Bragg angle”, that is, the critical angle of incidence of the x-ray radiation from source location 16 upon the surface of crystal 12 whereby the most radiation is diffracted toward target location 18. At angles of incidence larger and smaller than the Bragg angle less incident radiation is diffracted to the target. The Bragg angle for a system is a function of the crystal used and the frequency of the x-ray radiation used, among other things. In the typical prior art system shown in FIG. 2, system 10 is designed so that the angle of incidence of the x-rays, as indicated by phantom line 26, on center 14 of the surface of crystal 12 relative to source-to-image connecting line 22, is equal to the Bragg angle for the system, typically an angle between about 5 degrees and about 30 degrees. Lines 28 and 30 in FIG. 2 illustrate the divergence of x-ray photons from the source location 16 and lines 32 and 34 illustrate the convergence of x-ray photons to the target location 18 as diffracted by crystal 12. The angle of incidence of the incident x-rays, as indicated by lines 28 and 30, varies from the ideal Bragg angle as the location of the point of impingement varies from center 14. The angle ″ between lines 28 and 30 in the plane of the Rowland circle 20 is referred to in the art as the “crystal collection angle”. In terms of the Bragg angle, the ideal toroidal curvature r is given by the expression 2Rsin2θB. These terms and dimensions used to define the geometry of the prior art shown in FIGS. 1 and 2 will be helpful in describing the present invention. In system 10 of FIGS. 1 and 2, photons emitted from source location 16, which is any conventional x-ray point source, experience Bragg diffraction on crystal 12 and form an image at target location 18. The focus aberration of the image at target location 18 is proportional to the width W of crystal 12 and, consequently, to the crystal collection angle α. Focus aberration considerations typically limit ″ to a value of 1 to 5 degrees, which in turn limits x-ray source radiation utilization. One way to increase the source utilization is to increase the width W of optic 12, but increasing width W increases the focus aberration of the optic due to deviation from the desired Bragg angle of incidence upon the surface of the wider optic. Though the prior art optical system illustrated in FIGS. 1 and 2 can effectively capture x-rays emitted from a divergent source and focus x-rays onto a target, the capacity of this and related prior art systems to utilize as much x-ray energy as possible is limited due to, among other things, their limited ability to capture sufficient x-rays. Another prior art x-ray focusing system is disclosed in U.S. Pat. No. 5,127,028, entitled “Diffractor with doubly curved surface steps”. However, though the optics disclosed in U.S. '028 provides good coverage in the collection angle in the Rowland circle plane, the U.S. '028 optics are limited in their coverage in the plane orthogonal to the Rowland circle plane and source-to-image connecting line, for example, line 22 in FIG. 2. FIGS. 3 and 4 illustrate one aspect of the present invention which overcomes the limitations of the prior art systems, for example, system 10 illustrated in FIGS. 1 and 2 and the art disclosed in U.S. '028. FIG. 3 is a representative isometric view of an x-ray focusing arrangement 120 having a first curved crystal optic 122, a second crystal optic 124, an x-ray source location 126, and an x-ray target location 128. FIG. 4 is a sectional view as viewed along section lines 4—4 shown in FIG. 3. Crystal optics 122 and 124 may comprise doubly-curved crystals and may be mounted on a crystal support frame, which is not shown for ease of illustration, but which is known in the art. According to one aspect of the present invention, first crystal optic 122 has at least one crystal lattice plane 123 and second crystal optic 124 has at least one crystal lattice plane 125. The center point of crystal optics 122 and 124 are identified as points 130 and 132, respectively. As most clearly shown in FIG. 10, the x-ray source location 126; optic center points 130, 132; and x-ray target location 128 define the Rowland circle 129 of radius RO for arrangement 120. X-ray source location 126 and x-ray target location 128 are joined by a source-to-image connecting line 134. θB is the Bragg angle for the first crystal optic 122. Focusing arrangement 120 further includes a first crystal radius 136 having an origin 0 for first crystal optic 122 and a second crystal radius 138 having an origin O′ for second crystal optic 124. First crystal radius 136 and second crystal radius 138 drawn to the counterpoints 130, 132 of their respective crystals make an angle 6 with each other. In one aspect of the invention, radii 136 and 138 are about equal, that is, the length of radii 136, 138 are within about 10% of each other. According to one aspect of the invention shown in FIGS. 3 and 4, the utilization of x-ray energy emitted by a divergent x-ray source positioned at source location 126 is optimized or maximized, compared to prior art arrangements. In one aspect of the invention, this is achieved by varying the orientation of the lattice planes 123, 125 relative to the surfaces of the crystal optics 122, 124, respectively. For instance, in one aspect of the invention, the lattice planes 123 of crystal 122 may be parallel to the surface of the crystal, for example, as in crystal 12 shown in FIGS. 1 and 2. However, according to one aspect of the present invention, the lattice planes 125 of crystal 124 are not parallel to the surface of the crystal but are offset an angle γ relative to the surface of the crystal. In order to compensate for the orientation of crystal 124 relative to the source location 126 (that is, an orientation providing an angle of incidence on crystal 124 which is different, for example, greater, than the desired Bragg angle for the crystal), the orientation of the lattice planes 125 of crystal 124 relative to the surface of crystal 124 is varied to create the desired Bragg angle of incidence on the lattice planes of crystal 124. As shown most clearly in the detail of FIG. 4A, according to this aspect of the invention, lattice planes 125 of crystal 124 make an angle γ with a line 127 tangent to the surface of crystal 124 at the point lattice plane 125 intersects the surface of crystal 124. According to one aspect of the invention, line 140 connecting source location 126 and center point 130 of first optic crystal 122 and line 142 connecting source location 126 and center point 132 of second optic crystal 124 make an angle γ′. In one aspect of the invention, the angle of orientation of the lattice planes 125 of crystal 124 relative to its surface is about equal to γ′, that is, γ˜γ′. In one aspect of the invention, γ and γ′ are essentially identical within the fabrication tolerances of arrangement 120. According to this aspect of the present invention, the diffraction conditions of photons emitted from source location 126 are about equal for both first crystal 122 and second crystal 124. In one aspect of the invention, the value of γ and γ′ typically varies from about minus 15 degrees to about plus 15 degrees, but in one aspect of the invention γ and γ′ are preferably between about minus 10 degrees and about plus 10 degrees. Though in the simplest embodiment of the aspect of the invention shown in FIGS. 3 and 4 only two crystals 122 and 124 may be used, according to another aspect of the invention at least a third crystal 144 or 145 (shown in phantom in FIG. 4) or more crystals may be used. The lattice plane orientation γ of optic crystals 144 and 145 may be oriented to again maximize the Bragg diffraction of x-rays impinging upon the surface of crystal 144 and 145. In another aspect of the invention, further crystals (for example, 5, 7, or more crystals in a row) may be used with appropriate variation in lattice plane orientation to maximize the utilization of the x-rays emitted at source location 126. In addition, in one aspect of the invention, further rows of crystals may be used having appropriate variation in lattice plane orientation. For example, in a fashion similar to the crystal matrix shown in, two or more rows of crystals may be used. For example, rows similar to crystals 122, 124, and 144 or 145 which are offset from each other in a direction orthogonal to the plane of Rowland circle 129 may be used. The orientation of the lattice planes in each of the crystals in these matrices can be varied to effect optimum Bragg diffraction so that x-ray utilization is maximized. In one aspect of the invention, the crystals, 122, 124, 144, 145, and others may be positioned about the same Rowland circle 129. In another aspect of the invention, crystals 122, 124, 144, 145, and others may be positioned about different Rowland circles, for example, Rowland circles lying in a plane oriented obliquely to the plane of Rowland circle 129. FIG. 5 illustrates a representative isometric view of an x-ray focusing arrangement 80 according to one aspect of the present invention. FIG. 5 is similar to FIGS. 1 and 3 and illustrates similar parameters shown earlier, for example, a source location 81, a target location 82, and a source to target line 83 which define a Rowland circle 85. According to this aspect of the invention, arrangement 80 includes a matrix or mosaic 84 comprising a plurality of crystal optics, for example, doubly-curved crystal optics, 86, 88, 90, 92, 94, 96, 98, 100 and 102. FIG. 6 illustrates a projection of the crystals as viewed via line 6—6 shown in FIG. 5. These optics are typically mounted in a rigid support frame, for example, but the frame is omitted from FIGS. 5 and 6 for ease of illustration. FIG. 6A presents a view similar to FIG. 6 but illustrates another aspect of the present invention. In FIG. 6A, matrix 87 is provided by curved crystals 95, 97 and 99 which are longer than the crystals shown in FIG. 6, for example, optic crystals 95, 97, and 99 have an angular extension perpendicular to the plane of Rowland circle 85 that is longer than, for example, for example, optic crystals 86, 88, and 90. According to one aspect of the invention, curved crystals 95 and 99 may also have atomic planes that are not parallel to the surface of their respective crystals. FIG. 7 illustrates a cross-sectional view of optic mosaic 84 as viewed along lines 7—7 shown in FIG. 6 or of optic mosaic 87 along lines 7—7 shown in FIG. 6A. FIG. 8 illustrates a cross-sectional view of optic mosaic 84 as viewed along lines 8—8 shown in FIG. 6. FIG. 8A illustrates a cross-sectional view of optic mosaic 87 as viewed along lines 8A—8A shown in FIG. 6A. As shown in FIG. 7, the middle row of crystals, that is, crystals 86, 88, and 90, having a center line 104 (see FIG. 6), are essentially the same as crystals 144, 123, and 124 shown in FIG. 4 having radii in the Rowland plane equal to about R. The bottom row of crystals in FIG. 6, that is, crystals 92, 94, and 96 having a centerline 106, and the top row of crystals, that is, crystals, 98, 100, and 102 having a centerline 108, may also have a radius R. However, as clearly shown in FIGS. 5 and 6, the top row of crystals and the bottom row of crystals are offset or spaced in the toroidal direction from the middle row of crystals. For example, the centerlines 106 and 108 are typically spaced from centerline 104 by φ degrees, for example, at least about 5 degrees. The angle φ will typically vary depending upon the dimensions of mosaic 84, but is typically between about 30 degrees and about 90 degrees. According to one aspect of the invention, the angular spacing between rows is typically uniform, thought the spacing may be non-uniform. As shown in FIG. 8, the middle column of crystals, that is, crystals 94, 88, and 100, having centerline 110 (see FIG. 6), are each typically similar to crystal 12 shown in FIG. 1 having a toroidal radius r, though crystals with varying values of r may be used. As shown in FIG. 8B, in the aspect of the invention having longer individual crystals, as shown in FIG. 6A, the longer crystals 85, 97, and 99 may also have a toroidal radius r. According to another aspect of the invention, as shown in FIG. 8B, optic crystal 97′ may also be a singly-bent crystal, for example, a crystal curved in the dispersive plane and not curved on the non-dispersive plane. In one aspect of the invention, similar singly-bent crystals 95′ and 99′ (not shown) which are similar to crystals 95 and 99 shown in FIG. 6A may have atomic planes that are not parallel to the surface of their respective crystal. As shown in FIG. 6, the right-hand column of crystals, that is, crystals 86, 92, and 98 having a centerline 112, and the left-hand column of crystals, that is, crystals, 90, 96, and 102, having a centerline 114, may have similar toroidal radii in a direction orthogonal to their respective Rowland circles. As shown in FIG. 5 and 6, the right column of crystals and the left-hand column of crystals may be offset or spaced in the circumferential direction from the middle column of crystals. For example, the centerlines 112 and 114 may be spaced from centerline 110 by an angle φ′, for example, an angle of at least about 5 degrees. The angle φ′ may typically vary depending upon the dimensions of the mosaic 84, but may be between about 30 degrees and about 90 degrees. According to one aspect of the invention, the angular spacing between columns may be uniform, though the spacing may be non-uniform. In operation, each row of crystals in matrix 84 performs like multi-crystal focusing system 40 shown in FIGS. 3 and 4. Therefore, a focusing system based on multi-crystal focusing assembly 82 shown in FIGS. 5 and 6 can typically triple the spatial coverage of multi-crystal focusing system 40. In this approach, a larger number of optical elements can be used to provide an additional improvement in x-ray source utilization. Though the aspect of the invention shown in FIGS. 5 and 6 illustrates a crystal matrix having 3 rows of crystals each row having 3 crystals (or 9 crystals in the matrix), according to one aspect of the invention, at least 2 rows of 2 crystals (that is, at least 4 crystals) may be used. Similarly, other matrices of crystals may be used according to the invention, for example, a 2×3 matrix, a 4×4 matrix, an 8×8 matrix, or a 10×12 matrix, among others, may be used. Regardless of the number of crystals in matrix 84, the arcuate length of matrix 84 in the toroidal direction or in the circumferential direction (that is, the arcuate direction orthogonal to the toroidal direction) may both vary from about 10 degrees to about 360 degrees, but the arcuate length in the toroidal direction is typically at least about 5 degrees and the arcuate length in the dispersive direction is typically at least about 5 degrees. According to one aspect of the invention, the crystals in matrix 84 may be comprised of the same or similar materials, for example, silicon or germanium. However, in another aspect of the invention, the material composition of the crystals in matrix 84 may vary. In one aspect of the invention, the crystals in matrix 84 are doubly-curved crystals. According to one aspect of the invention, the lattice planes of the crystals in matrix 84 are parallel to the surface of the crystals. However, in another aspect of the invention, the lattice planes may not be parallel to the surface of the crystal. For example, the orientation of the lattice planes in the crystals of matrix 84 may vary, for example, in a linear or non-linear fashion, to maximize the focusing of the x-rays on the target location 82. Another aspect of the present invention is illustrated in FIGS. 9, 10 and 11. FIG. 9 is a representative isometric view of an x-ray focusing arrangement 150 having a curved optic crystal 152, an x-ray source location 154, and an x-ray target location 156, which define a Rowland circle 155. X-ray source location 154 and x-ray target location 156 define a source-to-target line 162. In one aspect of the invention, optic crystal 152 is axi-symmetric about source-to-target line 162. According to this aspect of the invention, optic crystal 152 may include at least one optic crystal 164, and typically may include a plurality of individual optic crystals 164. Optic crystal 152 may have an arc length about source-to-target line 162 of at least about 45 degrees, typically, at least 90 degrees, and can be at least 180 degrees. For example, in the embodiment of this aspect of the invention shown in FIG. 9, optic crystal 152 comprises an arc length of about 360 degrees, that is, optic 152 comprises essentially a complete circle. Again, the one or more optic crystals 164 are typically one or more doubly-curved optic crystals. Also, optic 152 may be mounted in a support frame which is again not shown for ease of illustration. FIG. 10 is a cross-sectional view taken along section lines 10—10 shown in FIG. 9. FIG. 11 illustrates a cross section of the crystal optic 152 as viewed through section 11—11 shown in FIG. 10. X-ray source location 154, x-ray target location 156, and source-to-target line 162 shown in FIG. 9 also appear in FIG. 10. As shown in FIG. 10, according to one aspect of the invention, the surface of optic 152, x-ray source location 154, and x-ray target location 156 define one or more Rowland (or focal) circles 160 and 161 of radius R for optic crystal 152. Those of skill in the art will recognize that the number and orientation of the Rowland circles associated with crystal optic 152, or individual crystals 164, will vary with the position of the surface of optic crystal 152, for example, the variation of the toroidal position on optic crystal 152, and that Rowland circles 160 and 162 are only representative of two of many similar circles associated with crystal optic 152. According to one aspect of the invention, focal circles 160 and 161 may be concentric and have the same focal radius R. In another aspect of the invention, as shown in FIG. 10, focal circles 160 and 161 may not be concentric, but have the same focal radius R. According to another aspect of the invention, the focal radii of optic circles 160, 161, and others may vary. According to one aspect of the invention shown in FIGS. 9 and 10, the internal atomic diffraction planes of optic crystal 152 are not parallel to the surface of optic crystal 152 wherein the Bragg diffraction provides optimized focusing of x-rays on target location 156. For example, as shown in FIG. 10, the atomic diffraction planes of crystal 152 make an angle γ1 with the surface of the crystal optic 152 upon which x-rays are directed. According to one aspect of the invention, the atomic diffraction planes of crystal 152 make an angle γ1 with the surface of the crystal at the point of tangency of the surface of the crystal optic 152 and its corresponding optic circle 161 or 162. For example, as shown in FIG. 10, the point of tangency of optic circle 161 and crystal optic 152 is identified as point 158, which may be the geometric midpoint of the surface of crystal optic 152. As shown in FIG. 10, x-ray source location 154, point of tangency 158, and x-ray target location 156 define the Rowland circle 161 of radius R and x-ray source location 154 and x-ray target location 156 define the source-to-image line 162. Again, θB is the Bragg angle for crystal optic 152. Again, though optic 152 may comprise a single crystal, optic 152 typically comprises a plurality of individual crystals 164, for example, 2 or more. Each crystal 164 may be essentially identical, for example, identical to crystal 124 in FIGS. 3 and 4. In one aspect of the invention, when optic 152 comprises a plurality of individual crystals 164, the angle of the atomic diffraction planes, γ1, in each crystal 164 are oriented to satisfy Bragg diffraction conditions, typically to optimize the transmission of x-ray energy to target location 156. According to one aspect of the invention, as shown in FIG. 10, crystal optic 152 is fashioned wherein a line 159 from source location 154 and point 158 on the surface of crystal optic 152 makes and angle of about θB+γ1 with respect to source-to-image line 162. This angular relationship between the source location 154, target location 156, and crystal optic 152 satisfies the Bragg conditions for the atomic diffraction planes of optic 152 wherein the transmission of x-ray radiation from source location 154 to target location is optimized, for example, maximized. Correspondingly, the line 163 directed from target location 165 to point 158 makes an angle θB−γ1 with source-to-target line 162. In one aspect of the invention, this angular relationship applies to the entire surface of crystal optic 152 to which x-rays are exposed; however, according to one aspect of the invention, optic crystal 152 is fashioned wherein this relationship holds for at least one point on the surface of optic crystal 152. According to one aspect of the invention, optic crystal 152 is fashioned wherein this relationship applies to at least one of the individual optic crystals 164 from which crystal optic 152 is fashioned, typically, it holds for a plurality of optic crystals 164 from which crystal optic 152 is fashioned. According to one aspect of the invention, the arrangement of individual crystals 164 shown in FIGS. 10 and 11 provides full coverage in a plane orthogonal to source-to-image connecting line 162. In one aspect of the invention, crystals 164 have a common bending radius D which in one aspect of the invention is selected to provide point-to-point focusing properties. Though the aspect of the invention shown in FIGS. 10 and 11 comprises a complete circular optic 152, in one aspect of the invention, the optic 152 is less than a complete circle. For example, in one aspect of the invention, the circumferential arc length η (see FIG. 11) of optic 152 is at least about 45 degrees. In another aspect of the invention arc length η may be at least 90 degrees, or at least 180 degrees. According to one aspect of the invention, optic crystal 152 is fabricated so it is easily handled during manufacture, for example, during manufacture using the process outlined in U.S. Pat. No. 6,317,483 (the disclosure of which is incorporated by reference herein). According to one aspect of the invention, the radius D of optic crystal 152 varies along the axis of optic crystal 52, for example, along source-to-image line 152, wherein optic crystal can be more readily removed from the mold from which it is manufactured. In addition to providing optimum x-ray collection, crystal 152 can be fabricated without destroying the tooling when removing crystal 152 from a mold, for example, in a fashion similar to the method disclosed in U.S. Pat. No. 6,285,506, entitled “Curved Optical Device and Method of Fabrication”. FIG. 12 is a cross-sectional view similar to the cross-sectional view shown in FIG. 10 of another aspect of the present invention. According to one aspect of the invention, two or more crystal optics are used to capture x-rays, for example, from a common source, and direct them to a common target. FIG. 12 illustrates a cross-sectional view of x-ray optic arrangement 220, having an x-ray source location 254, an x-ray target location 256, an source-to-image line 262, and optic circles 260 and 261. According to one aspect of the invention, arrangement 220 includes at least one optic crystal 152, that is, a crystal optic 152 as shown in FIGS. 9 through 11, and a second crystal optic 252. Crystal optic 252 may be similar to crystal optic 152, for example, crystal optic 252 may have one or more of the physical attributes of crystal optic shown and described with respect to FIGS. 9 through 11, but crystal optic 252 may be smaller or larger in diameter than crystal optic 152. According to one aspect of the invention where crystal optic 152 comprises one or more individual crystal optics 164 having atomic diffraction planes at an angle (1, crystal optic 252 comprises one or more individual crystal optics 264 having atomic diffraction planes at an angle (2, that is, at an angle different from γ1. Crystal optics 152 and 252 may have similar or different Bragg angles θB. According to one aspect of the invention crystal optics 158 and 258 provide point focusing, for example, point-to-point focusing, of x-rays on target location 256. According to one aspect of the invention, optic crystals 152 and 252 are fashioned wherein lines drawn from source location 254 to points on their respective surfaces, for example, points 158 and 258 shown in FIG. 12, make angles θB+γ1 and θB+γ2, respectively, with source-to-image line 262. Again, in one aspect of the invention, the points 158 and 258 may be the midpoints of the surfaces of crystal optics 152 and 252, or points 158 and 258 may correspond to the point of tangency of the surfaces of crystal optics 152 and 252 and their respective optic circles 260 and 261. Again, as described with respect to crystal optic 152, crystal optics, 152 and 252 may comprise complete circular optics; however, in one aspect of the invention, the crystal optics 152 and 252 may be less than a complete circle. For example, in one aspect of the invention, the circumferential arc length η (see FIG. 11) of optics 152 and 252 may be at least about 45 degrees. In another aspect of the invention arc length η may be at least 90 degrees, or at least 180 degrees. One or more aspects of the present invention are exemplified by the following examples. One specific example of an optic fabricated according to the aspect of the invention shown in FIGS. 3 and 4 is a Germanium (Ge) crystal optic for focusing Chromium (Cr) Kα radiation. The Ge crystal fabricated according to the present invention included reflection crystal planes Ge(111) and a Bragg angle for Cr Ka radiation of about 20.5°. According to one aspect of the invention, five Ge crystals pieces with inclined atomic diffraction planes of Ge(111) of −8 degrees, −4 degrees, η degrees, 4 degrees, and 8 degrees respectively were used. The Ge crystal device provided point focusing CrKα beam with a collection solid angle of 0.1 sr. for a 50° revolving angle, φ, see FIG. 1. This optic according to this aspect of the invention produced an x-ray image of about 3×1010 photons/sec at the target location using a 50 Watt, point x-ray source with Cr anode. This intense x-ray beam according to this aspect of the invention can have important applications, for example, in high sensitivity XRF analysis for measuring low Z elements. An example of the aspect of the invention shown in FIGS. 9, 10, and 11 was fabricated from Silicon (Si) crystal for focusing Molybdenum (Mo) Kα radiation. In this aspect of the invention, the atomic reflection crystal planes were Si(220) and the Bragg angle was about 10.6°. The inclined angle of Si(220) was about 1 degree for 16 pieces of crystals that were formed into a ring. The Si optic according to this aspect of the invention had a collection solid angle of about 0.04 sr. and provided about 1×109 Mo Kα photons/sec at the target focal spot. This extremely intense x-ray beam according to this aspect of the invention can be used, for example, for high speed or high sensitivity monochromatic XRF applications. The crystal optics disclosed in FIGS. 3–12 are applicable for use with any kind of x-ray sources, for example, x-ray tubes or synchrotrons. The optics disclosed in FIGS. 3–1 may provide point focusing, for example, point-to-point focusing, line-focusing, for example, point-to-line focusing, or any other type of image focusing depending upon the shape of image desired by the operator. However, regardless of the shape of the source or shape of the focused image, in one aspect of the invention, due to the increased capturing and focusing of x-ray energy by the optics according to the present invention, the x-ray sources can typically consume less power than conventional x-ray sources while still providing sufficient energy flux to the target for many applications. For example, one aspect of the inventions disclosed in FIGS. 5–13 can be used with x-ray sources which consume less than 100 Watts of power during operation. In other aspects of these inventions, the x-ray source can consume less than 50 Watts, less than 25 Watts, or even less than 10 Watts and still provide sufficient energy flux to the target. The present invention provides devices that can be used to dramatically improve the utilization of x-rays in a myriad of analytical and research applications, by among other things, increasing the radiation beam collection angle compared to the prior art. This increased utilization of x-ray beam energy according to the present invention provides the potential to reduce the size of high-energy radiation focusing systems while also reducing required measuring or exposure times in experimental and industrial processes. While the invention has been particularly shown and described with reference to preferred embodiment, it will be understood by those skilled in the art that various changes in form and details may be made to the invention without departing from the spirit and scope of the invention described in the following claims. |
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description | The present invention claims priority to pending Chinese Patent Application No. 201710432639.0, filed on Jun. 9, 2017, and incorporated herein by reference. The present invention relates to nuclear reactor apparatus technology, and more particularly to a containment cooling apparatus. A nuclear reactor is capable of starting, controlling and maintaining nuclear fission or fusion chain reactions. In the nuclear reactor, the reaction rate could be controlled precisely so as to release energy slowly for people consumption. The nuclear reactor comprises various uses, and the most important one is replacing other fuel to generate heat as steam electric power or driving power of apparatus, for example, aircraft carrier. As important energy source, any possible faults occurred during operation of the nuclear reactor cannot be overlooked. A containment spray system is a protective apparatus to lower pressure and temperature inside the containment in the event that the primary circuit has an accident. In the prior art, passive containment spray system is a protective apparatus that can operate during a power outage, guaranteeing safety of the nuclear reactor. Therefore further optimizing the structural design of the passive containment spray system to improve utilization of coolant in passive condition is a main research direction for the people skilled in the art. The present invention provides a containment cooling apparatus aiming to have higher utilization of coolant. According to one embodiment of the present invention, the containment cooling apparatus includes a cooling water tank disposed above a containment; a spray header connected to the cooling water tank via a first communicating pipe, wherein the spray header is disposed on an outside of the containment for spraying cooling water to an outer wall of the containment; a bell shaped shield covering the containment, wherein the cooling water tank is disposed on a top portion of the shield; a space formed between an inner wall of the shield and the outer wall of the containment, wherein the spray header is disposed in the space; an exhaust hole disposed on the top portion of the shield; and a water separator disposed in the exhaust hole and/or the space. Specifically, the cooling water tank stores the cooling water used as the containment coolant. The cooling water tank disposed above the containment allows the cooling water to flow into the spray header via the first communicating pipe under gravity, and then spray from the spray header to realize passive cooling of the containment. In such embodiment, the spray header is disposed in the space, resulting in sufficient contact between the cooling water and the outer wall of the containment, facilitating cooling effect on the containment and collection of the heated cooling water, which the latter benefits the cyclic utilization of the cooling water. Meanwhile, as the existing spray header has a good atomization effect on the cooling water, the heated water vapor comprises a large amount of cooling water absorbing heat insufficiently, which can condense into large water droplets on the water separator and then falls into the space so as to cool the containment again, achieving the purpose of improving the utilization of cooling water. According to another embodiment of the present invention, the containment cooling apparatus further includes a gas tank disposed in the containment, wherein the gas tank is connected to an upper portion of the cooling water tank via a second communicating pipe, and the cooling water tank is a closed container. In such embodiment, the gas tank disposed in the containment is used to store gas. When the water in the primary circuit of the containment leaks, the internal temperature of the containment rises, causing the gas stored in the gas tank to expand. The expanding gas increases the internal pressure of the cooling water tank, and the increased internal pressure affects the cooling water, increasing the flow rate of the cooling water spraying from the spray header. Namely, the faster the temperature rises, the larger the cooling water flow rate is. Thus, the containment cooling apparatus has a strengthened cooling effect on the containment, that is to say the present cooling apparatus has higher cooling reliability. In another embodiment, the cooling water tank is annular, and an axis of the cooling water tank is collinear with an axis of the shield to provide strong stability for supporting. In some embodiments, a cooling water outlet is disposed on a bottom portion of the shield for timely discharging the overheated cooling water remaining in the space, enabling the containment cooling apparatus to have an effective cooling effect on the containment. As the spray header comprises a large number of communicating pipes and nozzles, in some embodiments, the spray header is axisymmetrically disposed above the containment to cool the containment evenly as well as have a strong structural stability. When in use, the temperature of the containment fluctuates slightly, resulting in a change of internal pressure in the gas tank, so that the cooling water stored in the cooling water tank can spray out in normal condition. In order to avoid the aforementioned situation, in some embodiments, the containment cooling apparatus further comprises a rupture disk disposed in the gas tank and/or the second communicating pipe, wherein the rupture disk is ruptured during an increase of pressure in the gas tank, and the rupture disk in an intact state is capable of isolating a space on both sides thereof. In such scheme, the rupture disk is preferably fixed in a pipe section detachable from the second communicating pipe. Meanwhile, the pipe section is located on the outside of the containment to conveniently replace the rupture disk or the overall assembly of the rupture disk and the pipe section. When the pressure applied on the side of the rupture disk adjacent to the gas tank increases to a certain extent, the rupture disk is ruptured. At the same time, the cooling water stored in the cooling water tank cools the containment. In sum, the present invention has at least the following advantages and beneficial effects: The cooling water tank stores the cooling water used as the containment coolant. The cooling water tank disposed above the containment allows the cooling water to flow into the spray header via the first communicating pipe under gravity, and then spray from the spray header to realize passive cooling of the containment. In such embodiment, the spray header is disposed in the space, resulting in sufficient contact between the cooling water and the outer wall of the containment, facilitating cooling effect on the containment and collection of the heated cooling water, which the latter benefits the cyclic utilization of the cooling water. Meanwhile, as the existing spray header has a good atomization effect on the cooling water, the heated water vapor comprises a large amount of cooling water absorbing heat insufficiently, which can condense into large water droplets on the water separator and then falls into the space so as to cool the containment again, achieving the purpose of improving the utilization of cooling water. For making the above and other purposes, features and benefits become more readily apparent to those ordinarily skilled in the art, the preferred embodiments and the detailed descriptions with accompanying drawings will be put forward in the following descriptions. The present invention will now be described more specifically with reference to the following embodiments. It is to be noted that the following descriptions of preferred embodiments of this invention are presented herein for purpose of illustration and description only. It is not intended to be exhaustive or to be limited to the precise form disclosed. As shown in the FIGURE, a containment cooling apparatus includes a cooling water tank 3 disposed above a containment 1; a spray header 4 connected to the cooling water tank 3 via a first communicating pipe 5, wherein the spray header 4 is disposed on an outside of the containment 1 for spraying cooling water to an outer wall of the containment 1; a bell shaped shield 2 covering the containment 1, wherein the cooling water tank 3 is disposed on a top portion of the shield 2; a space 9 formed between an inner wall of the shield 2 and the outer wall of the containment 1, wherein the spray header 4 is disposed in the space 9; an exhaust hole 11 disposed on the top portion of the shield 2; and a water separator 12 disposed in the exhaust hole 11 and/or the space 9. Specifically, the cooling water tank 3 stores the cooling water used as the containment coolant. The cooling water tank 3 disposed above the containment 1 allows the cooling water to flow into the spray header 4 via the first communicating pipe 5 under gravity, and then spray from the spray header 4 to realize passive cooling of the containment 1. In such embodiment, the spray header 4 is disposed in the space 9, resulting in sufficient contact between the cooling water and the outer wall of the containment 1, facilitating cooling effect on the containment 1 and collection of the heated cooling water, which the latter benefits the cyclic utilization of the cooling water. Meanwhile, as the existing spray header 4 has a good atomization effect on the cooling water, the heated water vapor comprises a large amount of cooling water absorbing heat insufficiently, which can condense into large water droplets on the water separator 12 and then falls into the space 9 so as to cool the containment 1 again, achieving the purpose of improving the utilization of cooling water. As shown in the FIGURE, on the basis of Embodiment 1, the present embodiment further includes a gas tank 8 disposed in the containment 1, wherein the gas tank 8 is connected to an upper portion of the cooling water tank 3 via a second communicating pipe 6, and the cooling water tank 3 is a closed container. In such embodiment, the gas tank 8 disposed in the containment 1 is used to store gas. When the water in the primary circuit of the containment 1 leaks, the internal temperature of the containment 1 rises, causing the gas stored in the gas tank 8 to expand. The expanding gas increases the internal pressure of the cooling water tank 3, and the increased internal pressure affects the cooling water, increasing the flow rate of the cooling water spraying from the spray header 4. Namely, the faster the temperature rises, the larger the cooling water flow rate is. Thus, the containment cooling apparatus has a strengthened cooling effect on the containment 1, that is to say the present cooling apparatus has higher cooling reliability. When in use, the temperature of the containment 1 fluctuates slightly, resulting in a change of internal pressure in the gas tank 8, so that the cooling water stored in the cooling water tank 3 can spray out in normal condition. In order to avoid the aforementioned situation, the containment cooling apparatus further comprises a rupture disk 7 disposed in the gas tank 8 and/or the second communicating pipe 6, wherein the rupture disk 7 is ruptured during an increase of pressure in the gas tank 8, and the rupture disk 7 in an intact state is capable of isolating a space on both sides thereof. In such scheme, the rupture disk 7 is preferably fixed in a pipe section detachable from the second communicating pipe 6. Meanwhile, the pipe section is located on the outside of the containment 1 to conveniently replace the rupture disk 7 or the overall assembly of the rupture disk 7 and the pipe section. When the pressure applied on the side of the rupture disk 7 adjacent to the gas tank 8 increases to a certain extent, the rupture disk 7 is ruptured. At the same time, the cooling water stored in the cooling water tank 3 cools the containment 1. As shown in the FIGURE, on the basis of Embodiment 1, the cooling water tank 3 is annular, and an axis of the cooling water tank 3 is collinear with an axis of the shield 2 to provide strong stability for supporting. A cooling water outlet 10 is disposed on a bottom portion of the shield 2 for timely discharging the overheated cooling water remaining in the space 9, enabling the containment cooling apparatus to have an effective cooling effect on the containment 1. As the spray header 4 comprises a large number of communicating pipes and nozzles, the spray header 4 is axisymmetrically disposed above the containment 1 to cool the containment 1 evenly as well as have a strong structural stability. While the invention has been described in terms of what is presently considered to be the most practical and preferred embodiments, it is to be understood that the invention needs not be limited to the disclosed embodiment. On the contrary, it is intended to cover various modifications and similar arrangements included within the spirit and scope of the appended claims which are to be accorded with the broadest interpretation so as to encompass all such modifications and similar structures. |
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claims | 1. A disposal container of high-level radioactive waste using multiple barriers, comprising:an inner wall, made of carbon steel, for being cylindrical in shape; a middle wall, made of Inconel™, for being cylindrical in shape and bonded to an outer surface of the inner wall; and an outer wall, made of copper, for being bonded to a lateral surface of the middle wall. 2. The disposal container of high-level radioactive waste using multiple barriers, according to claim 1,wherein the outer wall further installs a heat sink made of aluminum or copper and separated from a lateral surface for releasing heat which is generated inside and transferred to the outer wall, and a radiation fin made of materials same as the heat sink is combined between the heat sink and the outer wall. 3. The disposal container of high-level radioactive waste using multiple barriers, according to claim 2,wherein the radiation fin is combined to the lateral surface of the heat sink. 4. The disposal container of high-level radioactive waste using multiple barriers, according to claim 1,wherein a siphon pipe for storing refrigerants at a certain level is further installed between the heat sink and the outer wall, thereby making the inside in a vacuum state for releasing heat generated from the inside to the outer wall. 5. A barrier system using the disposal container of high-level radioactive waste using multiple barriers in claim 1, comprising:a disposal tunnel which is formed by digging rock formation;a deposition hole which is vertically or horizontally perforated, thereby storing the disposal container;and a buffer which is filled with the deposition hole and the disposal container. 6. The barrier system using the disposal container of high-level radioactive waste using multiple barriers according to claim 5,wherein the buffer is composed of Na-Bentonite. |
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047175338 | summary | FIELD OF THE INVENTION The invention relates to nuclear reactor fuel assemblies of the type incorporating a cluster of fuel rods (also known as fuel elements) arranged in a regular network and held in position by grids distributed and spaced apart along the assembly. The invention relates more particularly to grids for such assemblies comprising a peripheral girdle frame and at least two series of parallel elements defining openings for fuel rods therethrough and elements substituted therefor at certain points of the network. Although the invention is general in scope, it is particularly suitable for use in fuel assemblies in which the fuel rods are disposed in a triangular network with a pitch which is only slightly greater than the diameter of the fuel rods. The use of such a triangular network is desirable for an undermoderated reactor core since the amount of moderating coolant between adjacent rods must be small, while the required coolant flow should be passed and the coolant streams should mix without an excessive pressure loss. PRIOR ART The grids for nuclear fuel assemblies used in water cooled and moderated reactors are generally provided for retaining the rods at the nodes of a square lattice. They comprise a girdle and two sets of mutually orthogonal plates. A fuel assembly described in U.S. Pat. No. 3,068,163 comprises grids for holding the rods at the nodes of a square lattice. Each grid comprises a straight or undulating strip passing between the rods. The strips interengage at the crossing points. If used in a fuel assembly with a "close" pitch, that design would not ensure passage of the coolant flow under satisfactory conditions, mixture of the different streams and an acceptable pressure loss. French patent No. 2,509,078 discloses a spacing grid with a square network which differs from that described in the above-mentioned U.S. patent in that the orthogonal strips for holding the rods are superimposed and connected together by split tongues. A close pitch cannot be adopted with this solution either. There is disclosed in U.S. Pat. No. 3,158,549 (Fowler) a fuel assembly having fuel rods retained by grids each having cross wires so located that each gridwork provides three of more wires in contact with each fuel elements. However, that arrangement is not sufficient for safely and accurately maintaining the fuel elements in position without impressing an excessive head loss to the coolant flow. SUMMARY OF THE INVENTION It is an object of the invention to provide a nuclear fuel assembly grid for resiliently and efficiently holding the rods in position which however causes a relatively low pressure loss, which ensures good mixing of the fluid streams and which represents a small amount of neutron absorbing material. For that purpose, there is provided a spacing grid comprising a peripheral girdle or frame and at least two series of parallel wires defining passage pockets for the rods. The parallel wires are distributed into at least two beds spaced apart in the longitudinal direction of the assembly and each comprising at least two series of intersecting wires whose ends are fixed to the girdle. When a grid of the invention is used in a fuel assembly with a triangular lattice, each side of the triangular lattice will be parallel to the wires of at least one series. The wires of each bed may provide either two supporting points per rod, the wires of two successive beds providing to a same rod support points at diametrically opposite positions, or four support points per rod. For mixing the coolant streams, each bed may typically comprise two series of wires, the wires of one of the series of each bed forming an angle of 60.degree. with the wires of the two series of the preceding or following bed. Thus swirls are created. If the beds have a distribution which is reproduced cyclically, an overall flow having an helical shape is induced. The term "wire" should be interpreted as meaning an elongated element, whose cross section has in all its directions a dimension of the same order of magnitude less by several orders of magnitude than its length so that this element is deformable in all directions. A wire will generally have a circular section. However, ovalized or flattened sections may also be used. The invention will be better understood from the following description of particular embodiments given by way of examples. |
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050376039 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates generally to fuel assemblies for nuclear reactors and, more particularly, is concerned with an improved hand held tool for removing and replacing a locking tube in a removable top nozzle of a reconstitutable fuel assembly. 2. Description of the Prior Art In most nuclear reactors, the reactor core is comprised of a large number of elongated fuel assemblies. Conventional designs of these fuel assemblies include a plurality of fuel rods and control rod guide thimbles held in an organized array by grids spaced along the fuel assembly length and attached to the control rod guide thimbles. Top and bottom nozzles on opposite ends of the fuel assembly are secured to the guide thimbles which extend slightly above and below the ends of the fuel rods. At the top of the fuel assembly, the guide thimbles are attached in passageways provided in the adapter plate of the top nozzle. The guide thimbles may each include an upper sleeve for attachment to the top nozzle. During operation of such fuel assembly in a nuclear reactor, a few of the fuel rods may occasionally develop cracks along their lengths resulting primarily from internal stresses, thus establishing the possibility that fission products having radioactive characteristics may seep or otherwise pass into the primary coolant of the reactor. Such products may also be released into a flooded reactor cavity during refueling operations or into the coolant circulated through pools where the spent fuel assemblies are stored. Since the fuel rods are supported by the grids in a spaced array with the guide thimbles between the top and bottom nozzles of the fuel assembly, it is difficult to detect and remove the failed fuel rods. Until recently to gain access to these rods it was necessary to remove the affected fuel assembly from the nuclear reactor core and then break the welds securing the nozzles to the guide thimbles. In so doing, the destructive action often rendered the fuel assembly unfit for further use in the reactor because of the damage done to both the guide thimbles and the nozzle which prohibited rewelding. In view of the high costs associated with replacing fuel assemblies, reconstitutable fuel assemblies were developed to minimize operating and maintenance expenses. The general approach to making a fuel assembly reconstitutable is to provide it with a removable top nozzle. One reconstitutable fuel assembly construction is illustrated and described in U.S. Pat. No. 4,631,168 to Shallenberger et al, which is assigned to the assignee of the present invention. It incorporates an attaching structure for removably mounting the top nozzle on the upper ends of the control rod guide thimbles. The attaching structure includes a plurality of outer sockets defined in the adapter plate of the top nozzle, a plurality of inner sockets each formed on the upper end of one of the guide thimbles, and a plurality of removable locking tubes inserted in the inner sockets to maintain them in locking engagement with the outer sockets. Each outer socket is in the form of a passageway through the adapter plate which has an annular groove. Each inner socket is in the form of a hollow upper end portion of the guide thimble having an annular bulge which seats in the annular groove when the guide thimble end portion is inserted in the adapter plate passageway. A plurality of elongated axial slots are provided in the guide thimble upper end portion to permit inward elastic collapse of the slotted portion so as to allow the larger bulge diameter to be inserted within and removed from the annular circumferential groove in the passageway of the adapter plate. In such manner, the inner socket of the guide thimble is inserted into and withdrawn from locking engagement with the outer socket. The locking tube is inserted from above the top nozzle into a locking position in the hollow upper end portion of the guide thimble forming the inner socket. When inserted in its locking position, the locking tube retains the bulge of the inner socket in its expanded locking engagement with the annular groove and prevents the inner socket from being moved to a compressed releasing position in which it could be withdrawn from the outer socket. In such manner, the locking tubes maintain the inner sockets in locking engagement with the outer sockets, and thereby the attachment of the top nozzle on the upper ends of the guide thimbles. Furthermore, to prevent inadvertent escape due to vibration forces and the like, heretofore the locking tubes have been secured in their locking positions. In the one construction of the locking tubes disclosed in the above-cited patent, after insertion of the locking tubes into their locking positions within the inner sockets of the hollow upper end portions of the guide thimbles, a pair of bulges are formed in the upper portion of each locking tube. These bulges fit into the circumferential bulge in the upper end portion of the guide thimble and provide an interference fit therewith. In another construction of the locking tubes disclosed in U.S. Pat. No. 4,699,758 to Shallenberger et al, which is also assigned to the assignee of the present invention, several small dimples are preformed on the exterior surface of the upper end portion of the locking tube circumferentially spaced from one another and projecting from the locking tube exterior surface. The use of the preformed dimples eliminates the necessity to form bulges in the locking tube after insertion into the locking position. Also the preformed dimples allow the locking tube to be reused, whereas the bulged locking tubes were discarded after each use. The reconstitutable fuel assembly construction briefly described above has proven to be an improvement by which domestic and foreign utilities can minimize both operating and maintenance expenses. A fixture developed for removing and replacing the top nozzle of the reconstitutable fuel assembly is disclosed in U.S. Pat. No. 4,664,874 to Shallenberger et al, also assigned to the assignee of the present invention. A locking tube removal and replacement tool provided for use in conjunction with the fixture is disclosed in U.S. Pat. No. 4,639,998 to Shallenberger et al, also assigned to the assignee of the present invention. The tool used both for removal and replacement of locking tubes one at a time basically has an inner tubular member, an actuatable shaft assembly, and an outer tubular member. The inner tubular member of the tool is attached at its upper end to a bail assembly for gripping by a user. At its lower end, the inner tubular member has an extension with an expandable and collapsible lower end in the form of a outwardly-turned annular segmented rim. The actuatable shaft assembly of the tool is mounted for axial movement within the inner tubular member and the extension thereof. The shaft assembly has an actuating knob coupled to its upper end and a conical nose disposed on its lower end. The lower conical nose extends beyond the segmented rim on the lower end of the tubular member extension. To use the tool, the inner tubular member and the actuatable shaft assembly are inserted from above the top nozzle downwardly through one passageway of the adapter plate and the hollow interior of the locking tube and guide thimble until the segmented rim and the conical nose are disposed below the lower edge of the locking tube. Then, the knob on the shaft assembly is rotated in one direction to force insertion of the conical nose into the lower end of the tubular extension and thereby expand the segmented rim such that it will underlie and engage the lower edge of the locking tube when the inner tubular member and shaft assembly of the tool are pulled in an upward direction. The outer tubular member of the tool is disposed about the upper end of the inner tubular member and mounted for slidable movement therealong. The outer tubular member can be manually moved in a reciprocating manner to deliver a series of forceful impacts to the bail assembly of the tool which impacts are, in turn, transmitted via the inner tubular member to the lower annular segment rim of its tubular extension. With the segmented rim expanded to underlie and engage the lower edge of the locking tube, the force of each impact will be transmitted to the lower edge of the locking tube causing displacement of the locking tube in an upward axial direction and removal of the locking tube from its locking position such that it can then be easily carried from the top nozzle by withdrawal of the inner tubular member and shaft assembly of the tool from the top nozzle adapter plate. One major problem with the design of the above-described tool is the difficulty of initially aligning and inserting the lower end of inner tubular member and actuatable shaft assembly of the tool into the hollow interior of the locking tube. The annular segmented rim on the inner tubular member extension, by surrounding and projecting outwardly from the conical nose on the shaft assembly, provides a feature which tends to catch on the upper edge of the locking tube and obstruct insertion therein. Consequently, a need still remains for an alternative design of a tool to use to effectively and efficiently carry out removal and replacement of the locking tubes. SUMMARY OF THE INVENTION The present invention provides an improved hand held tool designed to satisfy the aforementioned needs. The tool has an external configuration with smooth transitions between exterior surfaces of assembled parts for ease of alignment and insertion into the hollow locking tube. The tool is manually held and actuated by a user and capable of precise operation in removing the locking tube from and replacing it in the locking position in the removable top nozzle of the reconstitutable fuel assembly. Accordingly, the present invention sets forth a tool for removing and replacing a hollow locking tube from its locking position in reconstituting a fuel assembly. The fuel assembly includes a removable top nozzle with an adapter plate having at least one passageway, at least one guide thimble with an upper end portion, and an attaching structure having the hollow locking tube for releasably locking the upper end portion of the guide thimble within the passageway of the top nozzle adapter plate. The tool of the present invention comprises: (a) an elongated hollow tubular assembly having upper and lower opposite end portions with the lower end portion insertable in a hollow locking tube, the lower end portion including an outer tubular element having a circumferential guide wall with a plurality of circumferentially spaced apertures, the lower end portion also including a plurality of expandable and contractible lifting members disposed within the tubular element and having catch elements extendable through and retractable from the apertures of the guide wall of the tubular element for engagement with and disengagement from a lower edge of the locking tube; and (b) an actuator assembly mounted through the tubular assembly for axial movement therealong and having upper and lower end portions with the lower end portion for actuating the lifting members of the tubular assembly lower end portion between expanded and contracted conditions to extend and retract their catch elements through and from the apertures in the guide wall of the tubular assembly lower end portion for engaging with and disengaging from the lower edge of the locking tube. More particularly, the tubular assembly lower end portion includes a guide element interfitting the tubular element at an open lower end thereof and having a body portion projecting therefrom. The end of the tubular element and the body portion of the guide element have substantially the same outside diameter so as to provide a continuous smooth transition from the guide element body to the tubular element end for facilitating insertion of the guide and tubular elements of the tubular assembly lower end portion into the hollow locking tube without catching on an upper edge of the locking tube at the transition. Also, the lower end portion of the actuator assembly includes an elongated shaft member extending between the lifting members of the tubular assembly and having upper and lower tandemly-arranged segments, the upper segment being larger in outside diameter than the lower segment such that downward movement of the shaft member removes the lower segment from between the lifting members and inserts the upper segment between the lifting members causing engagement therewith and expansion of the finger elements from contracted to expanded condition, whereas upward movement of the shaft member removes the upper segment from between the lifting members and inserts the lower segment therebetween permitting contraction of the finger elements from the expanded to contracted condition. Further, the lifting members of the tubular assembly have tapered tips. The elongated shaft member extending between and past the lifting members mounts a retractor member at a lower end of the shaft member. The retractor member has a tapered portion for engaging the tapered tips of the lifting members and ensuring that the lifting members move from the expanded to contracted condition as the shaft member is moved upwardly. These and other features and advantages of the present invention will become apparent to those skilled in the art upon a reading of the following detailed description when taken in conjunction with the drawings wherein there is shown and described an illustrative embodiment of the invention. |
041682432 | abstract | A system for disposing of radioactive waste material from nuclear reactors by solidifying the liquid components to produce an encapsulated mass adapted for disposal by burial. The method contemplates mixing of radioactive waste materials, with or without contained solids, with a setting agent capable of solidifying the waste liquids into a free standing hardened mass, placing the resulting liquid mixture in a container with a proportionate amount of a curing agent to effect solidification under controlled conditions, and thereafter burying the container and contained solidified mixture. The setting agent is a water-extendable polymer consisting of a suspension of partially polymerized particles of urea formaldehyde in water, and the curing agent is sodium bisulfate. Methods are disclosed for dewatering slurry-like mixtures of liquid and particulate radioactive waste materials, such as spent ion exchange resin beads, and for effecting desired distribution of non-liquid radioactive materials in the central area of the container prior to solidification, so that the surrounding mass of lower specific radioactivity acts as a partial shield against higher radioactivity of the non-liquid radioactive materials. The methods also provide for addition of non-radioactive filler materials to dilute the mixture and lower the overall radioactivity of the hardened mixture to desired Lowest Specific Activity counts. An inhibiting agent is added to the liquid mixture to adjust the solidification time, and provision is made for adding additional amounts of setting agent and curing agent to take up any free water and further encapsulate the hardened material within the container. |
claims | 1. An antiradiation concrete, comprising: a metallic aggregate having a grain size of up to 7 mm; and at least 5.0% by weight of a boron-containing aggregate having a grain size of up to 1 mm and being finer-grained than said metallic aggregate; said metallic aggregate being larger than said boron-containing aggregate. 2. The antiradiation concrete according to claim 1 , wherein said boron-containing aggregate is at least 7.8% by weight of the antiradiation concrete. claim 1 3. An antiradiation concrete, comprising: a boron-containing aggregate having a grain size of up to 1 mm; and between 80 and 90% by weight of a metallic aggregate having a grain size of up to 7 mm, said metallic aggregate being larger than said boron-containing aggregate. 4. The antiradiation concrete according to claim 3 , wherein said boron-containing aggregate is between 1.0 and 1.5% by weight of the antiradiation concrete. claim 3 5. The antiradiation concrete according to claim 3 , wherein said metallic aggregate is between 85 and 89% by weight of the antiradiation concrete. claim 3 6. The antiradiation concrete according to claim 1 , wherein said boron-containing aggregate is a boron-containing mineral. claim 1 7. The antiradiation concrete according to claim 6 , wherein said boron-containing mineral is colemanite. claim 6 8. The antiradiation concrete according to claim 3 , wherein said boron-containing aggregate is a boron-containing mineral. claim 3 9. The antiradiation concrete according to claim 8 , wherein said boron-containing mineral is colemanite. claim 8 10. The antiradiation concrete according to claim 1 , wherein said metallic aggregate is at least one of granulated iron and granulated steel. claim 1 11. The antiradiation concrete according to claim 3 , wherein said metallic aggregate is at least one of granulated iron and granulated steel. claim 3 12. The antiradiation concrete according to claim 1 , wherein the antiradiation concrete has a minimum bulk density of approximately 3000 kg/m 3 . claim 1 13. The antiradiation concrete according to claim 3 , wherein the antiradiation concrete has a minimum bulk density of approximately 3000 kg/m 3 . claim 3 14. The antiradiation concrete according to claim 1 , wherein the antiradiation concrete has a bulk density of approximately 6000 kg/m 3 . claim 1 15. The antiradiation concrete according to claim 3 , wherein the antiradiation concrete has a bulk density of approximately 6000 kg/m 3 . claim 3 16. The antiradiation concrete according to claim 1 , including another metallic aggregate having a grain size of up to 1 mm. claim 1 17. The antiradiation concrete according to claim 16 , wherein said another metallic aggregate is barite sand. claim 16 18. The antiradiation concrete according to claim 3 , including another metallic aggregate having a grain size of up to 1 mm. claim 3 19. The antiradiation concrete according to claim 18 , wherein said another metallic aggregate is barite sand. claim 18 20. The antiradiation concrete according to claim 1 , including a mineral-containing aggregate with a grain size of up to 7 mm. claim 1 21. The antiradiation concrete according to claim 20 , wherein said mineral-containing aggregate is serpentine. claim 20 22. The antiradiation concrete according to claim 3 , including a mineral-containing aggregate with a grain size of up to 7 mm. claim 3 23. The antiradiation concrete according to claim 22 , wherein said mineral-containing aggregate is serpentine. claim 22 24. An antiradiation shell, comprising: at least one wall region made from an antiradiation concrete including: a metallic aggregate having a grain size of up to 7 mm; and at least 5.0% by weight of a boron-containing aggregate having a grain size of up to 1 mm and being finer-grained than said metallic aggregate; said metallic aggregate being larger than said boron-containing aggregate. 25. An antiradiation shell, comprising: at least one wall region made from an antiradiation concrete including: a boron-containing aggregate having a grain size of up to 1 mm; and between 80 and 90% by weight of a metallic aggregate having a grain size of up to 7 mm; said metallic aggregate being larger than said boron-containing aggregate. 26. In an X-ray device, an antiradiation shell for shielding a radiation source, comprising: at least one wall region made from an antiradiation concrete including: a metallic aggregate having a grain size of up to 7 mm; and at least 5.0% by weight of a boron-containing aggregate having a grain size of up to 1 mm and being finer-grained than said metallic aggregate; said metallic aggregate being larger than said boron-containing aggregate. 27. In a room having a radiation source, an antiradiation shell for shielding a radiation source, comprising: at least one wall region made from an antiradiation concrete including: a metallic aggregate having a grain size of up to 7 mm; and at least 5.0% by weight of a boron-containing aggregate having a grain size of up to 1 mm and being finer-grained than said metallic aggregate; said metallic aggregate being larger than said boron-containing aggregate. 28. In a beam tube in a reactor plant, an antiradiation shell for shielding a radiation source, comprising: at least one wall region made from an antiradiation concrete including: a metallic aggregate having a grain size of up to 7 mm; and at least 5.0% by weight of a boron-containing aggregate having a grain size of up to 1 mm and being finer-grained than said metallic aggregate; said metallic aggregate being larger than said boron-containing aggregate. 29. In an X-ray device, an antiradiation shell for shielding a radiation source, comprising: at least one wall region made from an antiradiation concrete including: a boron-containing aggregate having a grain size of up to 1 mm; and between 80 and 90% by weight of a metallic aggregate having a grain size of up to 7 mm; said metallic aggregate being larger than said boron-containing aggregate. 30. In a room having a radiation source, an antiradiation shell for shielding a radiation source, comprising: at least one wall region made from an antiradiation concrete including: a boron-containing aggregate having a grain size of up to 1 mm; and between 80 and 90% by weight of a metallic aggregate having a grain size of up to 7 mm; said metallic aggregate being larger than said boron-containing aggregate. 31. In a beam tube in a reactor plant, an antiradiation shell for shielding a radiation source, comprising: at least one wall region made from an antiradiation concrete including: a boron-containing aggregate having a grain size of up to 1 mm; and between 80 and 90% by weight of a metallic aggregate having a grain size of up to 1 mm; said metallic aggregate being larger than said boron-containing aggregate. |
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abstract | The present disclosure relates to an extreme ultraviolet (EUV) radiation source having a collector mirror oriented to reduce contamination of fuel droplet debris. In some embodiments, the EUV radiation source has a fuel droplet generator that provides a plurality of fuel droplets to an EUV source vessel. A primary laser is configured to generate a primary laser beam directed towards the plurality of fuel droplets. The primary laser beam has a sufficient energy to ignite a plasma from the plurality of fuel droplets, which emits extreme ultraviolet radiation. A collector mirror, configured to focus the extreme ultraviolet radiation to an exit aperture of the EUV source vessel, which is oriented so that a normal vector extending outward from a vertex of the collector mirror intersects a direction of a gravitation force by an angle that is less than 90°. |
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abstract | A collimator 100 for use in a radiation imaging system 10, and a method for making such collimators, are provided, wherein the collimator 100 is capable of collimating radiation in two orthogonal planes. The collimator in one embodiment includes a block 101 of radiation absorbing material having a plurality of focally aligned channels 102 extending therethrough; in a second embodiment, the collimator includes first and second collimation 204, 212 sections having a respective first plurality of focally aligned plate sets 201 and a respective second plurality of focally aligned plate sets 203 disposed orthogonally to the first plurality of plate sets. The method for making the collimator includes generating a CAD drawing, generating from the CAD drawing one or more stereo-lithographic files, and using the stereo-lithographic files to control an electro-deposition machining machine which creates the channels in the block. |
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054897351 | summary | FIELD OF THE INVENTION This invention relates to a decontamination composition, and more particularly to a composition suitable for decontaminating surfaces contaminated with naturally occurring radioactive materials (NORMs). BACKGROUND OF THE INVENTION Naturally occurring radioactive material (hereinafter "NORM") is present in varying concentrations in groundwater and the like, in water supply wells, oil production wells, gas production wells, and as byproducts in mining operations. In the oil field, NORM is the result of material that has been extracted from the producing zone and is deposited on the equipment in the form of solids, films, pipe scale, sediment, and the like. The radioactive material is typically radium 226, radium 228, radon 222, thorium 232, uranium 235, uranium 238, lead 210, polonium 210, and other naturally occurring radionuclides. Typically these radionuclides are .alpha., .beta. and often .gamma. emitters which have a long half life. Such radionuclides are believed to be associated with toxic and carcinogenic effects. Strict health-based limits thereon have been enacted or are under consideration. For example, the process equipment used in various petrochemical plants, refineries, and the like, and associated piping is exposed to high levels of NORM. The disposal of equipment having a high level of NORM has come under increased scrutiny, particularly in oil-producing states such as Louisiana and Texas. Thus, many companies are stockpiling equipment which will need to either be cleaned for reuse or decontaminated for disposal. Thus, currently the most common practice other than stockpiling is to ship the equipment to a radioactive waste facility which have their own environmental problems. There are also several costly mechanical methods on the market for removing NORM. These include ice, sponge, or carbon dioxide blasting. These methods have limitations in that these methods are more applicable to pipe scale and other solid forms of NORM as compared to NORM deposited in solution or as a film which adheres to metal surfaces and is difficult to remove. Methods are known to remove radioactive materials from surfaces such as those found in nuclear reactors. For example, U.S. Pat. No. 4,537,666 to Murray et al. describes the typical system as treating the surfaces with an oxidizing solution, such as one containing an alkaline permanganate. This is followed by treatment with a decontamination solution which is an aqueous solution of a chelate, such as ethylenediaminetetraacetic acid (EDTA), and a solubilizing agent, such as a mixture of oxalic acid and citric acid. The chelate forms a complex with the metal ions from the deposits and solubilizes them, and, thus prevents them from precipitating out of the solution at another location in the cooling system. The decontamination solution is circulated between the cooling system and a cation exchange resin. The chelated metal ions are deposited on the cation exchange resin, freeing the chelate to solubilize additional metal ions in the deposit. The difficulty with this decontamination process, according to Murray et al., is that both the chelates and the cation exchange resin complete for the metal ions. As a result, the metal ions do not readily leave the chelate and attach themselves to the ion exchange column. This means that long resin contact times are required, and that the ion exchange column effluent may contain relatively high metal ion concentrations. Murray et al. proposes to remove the metal ions by passing the decontamination solution through a porous DC electrode. Other exemplary methods for removing nonnaturally occurring radioactive materials are proposed in U.S. Pat. Nos. 4,704,235 to Arvesen; 4,729,855 to Murray et al.; 4,792,385 to Snyder et al.; and 5,111,887 to Morris et al. Despite the general availability of methods of removing naturally occurring and nonnaturally occurring radioactive materials, there continues to be a need for removing NORMs from surfaces exposed to the same, and particularly NORM deposited as a solution or film and adhered to surfaces. SUMMARY OF THE INVENTION With the foregoing in mind, it is an object of this invention to provide a decontamination composition and a method for decontaminating a surface contaminated with naturally occurring radioactive material (NORM). There are generally three types of NORM contaminants. One is radioactive scale which contains uranium, thorium, radium, and associated decay products from the production of oil and associated brines contaminated with NORM. The radioactivity in the scale originates principally from radium, which coprecipitates with barium and strontium sulfate. Another type is NORM-contaminated film, coating, or plating which can form from natural gas production or processing. Another type is NORM-contaminated sludge in pipelines, processing plants, storage tanks and delivery facilities, pigging operations, and gas lines and other filter assemblies. These films often contain radon and its decay products (i.e., polonium 210, bismuth 210, and lead 210). The film, coating, and plating forms are often more difficult to remove as compared to scale, and moreover the above-described mechanical methods are typically ineffective. These and other objects, features, and advantages of the invention are provided by the decontamination composition of the present invention. The composition comprises 40 to 60 percent of a compound selected from the group consisting of oxalic acid, alkali metal and ammonium salts of oxalic acid and mixtures thereof; 5 to 20 percent of a compound selected from the group consisting of citric acid, alkali metal and ammonium salts of citric acid and mixtures thereof; 20 to 40 percent of a compound selected from the group consisting of polyaminocarboxylic acid, alkali metal and ammonium salts of polyaminocarboxylic acid, and the combination of a polyaminocarboxylic acid with a neutralizing compound, and mixtures thereof; 0 to 2 percent of a nonionic surfactant, 0 to 2 percent of a dispersant; and 0 to 2 percent of a corrosion inhibitor. The present invention also relates to a method of decontaminating a surface whereby contaminants in the form of NORMs are removed therefrom. The method comprises contacting the surface (e.g., a metal surface) with a decontamination composition comprising about 40 to 60 percent of a compound selected from the group consisting of oxalic acid, alkali metal and ammonium salts of oxalic acid and mixtures thereof; about 5 to 20 percent of a compound selected from the group consisting of citric acid, alkali metal and ammonium salts of citric acid and mixtures thereof; about 20 to 40 percent of a compound selected from the group consisting of polyaminocarboxylic acid, alkali metal and ammonium salts of polyaminocarboxylic acid, and the combination of a polyaminocarboxylic acid with a neutralizing compound, and mixtures thereof; about 0 to 2 percent of a nonionic surfactant; about 0 to 2 percent of a dispersant, and 0 to 2 percent of a corrosion inhibitor. |
description | This invention was made with government support under Contract Number N00019-11-G-0001 awarded by The United States Department of Defense. The government has certain rights in the invention. Field of the Disclosure The present disclosure relates to non-destructive inspection systems and techniques, and more specifically, to radiation backscatter inspection. Technical Background Non-destructive inspection systems may be used during and after a product or sub-assembly has been created to ensure reliable and safe operation to specification. In this regard, these systems may detect irregularities which may prematurely reduce the useful lifespan of products. Examples of irregularities include wear, corrosion, foreign objects, and stress cracks. Some irregularities are more serious than others. Non-destructive inspection systems, for example conventional backscatter detectors, have been used to identify irregularities in various locations of products. However, depending upon the location and type of irregularity, there may be difficult cases when it may be impractical or inefficient for conventional backscatter inspection systems to be utilized because of an inability to easily distinguish irregularities associated with various materials of the product or sub-assembly. In these cases, alternative and more expensive inspections may be performed such as disassembly and statistical sampling using destructive testing. What is needed is a more effective approach to inspect products and subsystems to identify and distinguish irregularities associated with various materials. Embodiments enclosed herein include inspection systems employing radiation filters with different attenuation characteristics to determine specimen irregularities, and related methods. An inspection system includes a radiation emitter configured to emit a radiation beam along a radiation trajectory. Some of the radiation may be reflected by the specimen as backscatter and received by at least one radiation detector of the inspection system along the radiation trajectory. Irregularities and various materials of the specimen may produce backscatter radiation at different energies and/or scatter angles which may be identified by employing radiation filters having different attenuation characteristics. By employing these filters in communication with the radiation emitter and the radiation detector, the backscatter radiation passed through the filters may be measured and integrated at different positions of the radiation beam to produce a composite image of the specimen. In this manner, irregularities and associated materials within the specimen may be more easily identified. In one embodiment, an inspection system is disclosed. The inspection system includes a radiation scanner configured to emit a radiation beam along a radiation trajectory. The inspection system also includes a plurality of filters comprising at least two filters selectably positionable into the radiation trajectory, so that at least one of the at least two filters receives at least a portion of the radiation of the radiation beam and passes attenuated radiation. The at least two filters respectively have different attenuation characteristics. The inspection system also includes a radiation detector configured to receive the attenuated radiation and configured to produce detection data associated with an energy intensity of the attenuated radiation, wherein the received attenuated radiation is backscattered. The inspection system also includes a rendering system configured to create a composite image of a specimen disposed along the radiation trajectory using the detection data from the attenuated radiation passed through the at least two filters. In this manner, irregularities of the specimen may be efficiently identified. In another embodiment, a method of inspecting a specimen is disclosed. The method includes emitting a radiation beam from a radiation scanner of a backscatter inspection system into a radiation trajectory. The method also includes selectively positioning at least two filters of a plurality of filters of the backscatter inspection system into the radiation trajectory so that at least one of the at least two filters receives at least a portion of the radiation of the radiation beam and passes attenuated radiation. The at least two filters respectively have different attenuation characteristics. The method also includes receiving the attenuated radiation with a radiation detector of the backscatter inspection system and producing detection data associated with an energy intensity of the attenuated radiation. The received attenuated radiation is backscattered from the specimen. The method also includes creating a composite image of the specimen with a rendering system of the backscatter inspection system using the detection data produced from the attenuated radiation passed through the at least two filters. In this manner, the composite image of the specimen may be created with improved contrast to better detect irregularities in the specimen. In another embodiment, a computer program product is disclosed. The computer program product includes a computer-readable storage medium having computer-readable program code embodied therewith. The computer-readable program code includes computer-readable program code configured to instruct a radiation scanner to emit a radiation beam and along a radiation trajectory. The computer-readable program code also includes computer-readable program code to selectively position at least two filters of a plurality of filters of the backscatter inspection system into the radiation trajectory, so that at least one of the at least two filters receives at least a portion of the radiation of the radiation beam and passes attenuated radiation. The at least two filters respectively have different attenuation characteristics. The computer-readable code also includes computer-readable program code configured to receive detection data produced from a radiation detector of the backscatter inspection system. The radiation detector producing the detection data based on the energy intensity of the attenuated radiation received by the radiation detector, and the received attenuated radiation is backscattered. The computer-readable code also includes computer-readable program code configured to render a composite image of the specimen at a rendering system of the backscatter inspection system using the detection data passed through the at least two filters. In this manner, the specimen may be inspected to distinguish more serious irregularities from more innocuous irregularities of the specimen. Embodiments enclosed herein include inspection systems employing radiation filters with different attenuation characteristics to determine specimen irregularities, and related methods. An inspection system includes a radiation emitter configured to emit a radiation beam along a radiation trajectory. Some of the radiation may be reflected by the specimen as backscatter and received by at least one radiation detector of the inspection system along the radiation trajectory. Irregularities and various materials of the specimen may produce backscatter radiation at different energies and/or scatter angles which may be identified by employing radiation filters having different attenuation characteristics. By employing these filters in communication with the radiation emitter and the radiation detector, the backscatter radiation passed through the filters may be measured and integrated at different positions of the radiation beam to produce a composite image of the specimen. In this manner, irregularities and associated materials within the specimen may be more easily identified. In this regard, FIG. 1A is a schematic diagram of an exemplary inspection system 10(1) including a radiation scanner 12 emitting a radiation beam 14 along a radiation trajectory 15 at a specimen 16. The radiation beam 14 may comprise, for example, x-ray radiation or gamma rays. The radiation beam 14 may be incident upon the specimen 16 which may include a first component 18A and a second component 18B. The first component 18A may comprise a first material 20A, and the second component 18B may comprise a second material 20B having a different atomic number than the first material 20A. For example, the first material 20A may comprise carbon fiber having an atomic number of six (6) and the second material 20B may comprise aluminum having an atomic number of thirteen (13). A portion 22 of the radiation beam 14 may pass through the first component 18A before reaching the second component 18B. In this manner, the first component 18A and the second component 18B may at least partially reflect the backscatter radiations 24(1), 24(2), respectively, at reflection angles theta1 (θ1), theta2 (θ2) towards a radiation filter 26(1) of the inspection system 10(1) along the radiation trajectory 15. It is noted that the radiation trajectory 15 may widen as the backscatter radiations 24(1), 24(2) may have different reflection angles theta1 (θ1), theta2 (θ2). The compositional and directional differences between the backscatter radiations 24(1), 24(2) may determine whether the backscatter radiations 24(1), 24(2) pass through the backscatter filter 26(1) and reach a radiation detector 28. Specifically, the reflection angles theta1 (θ1), theta2 (θ2) may or may not be the same size and the backscatter radiations 24(1), 24(2) may be reflected from different positions within the specimen 16 along a propagation path of the radiation beam 14. Also, the backscatter radiations 24(1), 24(2) may or may not comprise the same energy level distribution or energy flux. In this regard, the radiation filter 26(1) may comprise a filter material 30(1), for example comprising aluminum, which attenuates at least a portion of the backscatter radiations 24(1), 24(2). The radiation filter 26(1) may have a thickness D1 which may determine how much of the backscatter radiations 24(1), 24(2) may pass through the radiation filter 26(1) to reach the radiation detector 28. The thickness D1 of the radiation filter 26(1) may be, for example, in a range from two-hundred fifty (250) microns to six (6) millimeters. In this manner, a portion 32(1) of the backscatter radiation 24(1) may pass through the radiation filter 26(1) to reach the radiation detector 28 while the backscatter radiation 24(2) may not, as depicted in FIG. 1A. Accordingly, information, in the form of energy intensity and corresponding energy level, contained within the backscatter radiation 24(1) regarding the first component 18A of the specimen 16 may be provided to the radiation detector 28 along the radiation trajectory 15. FIG. 1B is a schematic diagram of the inspection system 10(1) of FIG. 1B, wherein the backscatter radiation 24(1), 24(2) from the specimen 16 may be received instead by a second radiation filter 26(2). The backscatter radiation 24(1), 24(2) may be attenuated though the second radiation filter 26(2) of the inspection system 10(1). The radiation filter 26(2) may have a thickness D2 which may determine how much of the backscatter radiation 24(1), 24(2) may pass through the radiation filter 26(2) to reach the radiation detector 28. The thickness D2 of the radiation filter 26(2) may be, for example, in a range from one-hundred fifty (150) microns to four (4) millimeters. In this manner, a portion 32(2) of the backscatter radiation 24(2) may pass through the radiation filter 26(2) to reach the radiation detector 28 while the backscatter radiation 24(2) may not, as depicted in FIG. 1B. Accordingly, information contained within the backscatter radiation 24(2) about the second component 18B of the specimen 16 may be provided to the radiation detector 28 along the radiation trajectory 15. When the portions 32(1), 32(2) of the backscatter radiation 24(1), 24(2) are received by the radiation detector 28, the radiation detector 28 may measure respective energy flux amounts of the portions 32(1), 32(2). The measured energy flux amounts may be transferred as detection data 34(1), 34(2) to a rendering system 36 of the inspection system 10(1). The rendering system 36 may be adapted to create a composite image 38 of the specimen 16 using the detection data 34(1), 34(2) from the portions 32(1), 32(2) of the attenuated radiation 24(1), 24(2) passed through the radiation filters 26(1), 26(2), respectively. The rendering system 36 may include an electronic assembly 40 comprising a processor 41, memory 44, and a storage device 46. The rendering system 36 may also include a monitor 42 for displaying the composite image 38. Once the composite image 38 is analyzed, characteristics of the specimen 16 are determined, including irregularities and material differences of the specimen 16. In this manner, the composite image 38 for identifying irregularities with the specimen 16 may be created and displayed using the attenuated radiation 24(1), 24(2) passed through the radiation filters 26(1), 26(2). Another embodiment of an inspection system 10′(1) is provided in FIGS. 2A through 2E. In this regard, FIGS. 2A through 2D are a perspective view, a front view, left side view, and a top view, respectively, of the inspection system 10′(1) which is a different embodiment of the inspection system 10(1). The inspection system 10′(1) includes a first filter 26A(1), 26B(1) of a plurality of radiation filters 26A(1)-26A(N), 26B(1)-26B(N) receiving the backscatter radiation 24 from the specimen 16 and attenuating different energy level ranges of the backscatter radiation 24. In this manner, a composite image 38 of the specimen 16 may be created. The inspection system 10′(1) may include the radiation scanner 12, the plurality of radiation filters 26A(1)-26A(N), 26B(1)-26B(N), the at least one radiation detector 28A, 28B, and the rendering system 36. Each of these components is discussed sequentially below. With continued reference to FIGS. 2A through 2E, the radiation scanner 12 may be used to emit a radiation beam 14 to be absorbed, transmitted, and/or reflected by the specimen 16. A portion of the radiation beam 14 reflected from the specimen 16 may be the backscatter radiation 24. The radiation scanner 12 may include a radiation source 48 which may produce, for example, x-ray radiation or gamma ray radiation. The radiation source 48 may be, for example, an x-ray tube manufactured by Yxlon International GmbH of Hamburg, Germany. The radiation source 48 may be disposed within an enclosure 50 having an outer surface 52 providing shielding for the radiation emitted by the radiation source 48. The enclosure 50 may also include inner surfaces 54 connected to the outer surface 52 and forming at least one opening 56 for the radiation produced by the radiation source 48 and emitted from the enclosure 50 as the radiation beam 14. Each of the at least one opening 56 may be of a circular shape and may have a width in a range from 100 microns to two (2) millimeters. In this manner, the radiation beam 14 may be emitted from the radiation scanner 12. The radiation scanner 12 contributes to the creation of the composite image 38 providing information about irregularities and material of the specimen 16 by moving the radiation beam 14. The composite image 38 may be formed from the backscatter radiation 24 reflected from the specimen 16 as the radiation beam 14 is moved to different positions upon the specimen 16. In this regard, the enclosure 50 may move to direct the radiation beam 14 in a trajectory upon the specimen 16 in a form of a plurality of scans 58 (FIG. 2A) upon the specimen 16 which the radiation beam 14 follows. Each of the scans 58 may be orientated along the z-direction and separated by a separation distance Ds. In one embodiment, the separation distance DS is in a range from one-hundred fifty (150) to one-thousand (1,000) microns. Movement of the enclosure 50 may be facilitated by a track stage 59 supporting the enclosure 50. The track stage 59 may be movable in a y-direction upon at least one rail 60A, 60B with power provided by, for example, a worm gear (not shown). The track stage 59 may move at a velocity Vy, for example, in an adjustable range from fifty (50) microns per second to one-thousand (1,000) microns per second. The track stage 59 may also include a pivot mechanism 62 (FIG. 2D) to facilitate a rotation R1 of the enclosure 50 about an axis A1. The rotation R1 about the axis A1 enables movement of the at least one opening 56 as well as the radiation beam 14 in the z-direction. The rotation R1 may be, for example, in a range from one-hundred (100) revolutions per second to one (1) revolution per second. The angular position of the rotation R1 of the enclosure 50 and a y-position of the track stage 59 may be forwarded to the rendering system 36 as beam position data 63 to associate a position of the radiation beam 14 to the detection data 34(1), 34(2). In this manner, the radiation beam 14 may move along the scans 58 in the x-direction and the z-direction across the specimen 16 to enable the backscatter radiation 24(1), 24(2) to be produced at different positions of the specimen 16. Next, and with continued reference to FIGS. 2A-2E, the radiation filters 26A(1)-26A(N), 26B(1)-26B(N) also contribute to the creation of the composite image 38 by receiving a portion of the radiation beam 14 reflected by the specimen 16 as the backscatter radiation 24 and respectively passing the attenuated radiation 32A(1)-32A(N), 32B(1)-32B(N) to the radiation detectors 28A, 28B. The plurality of filters 26A(1)-26A(N) include at least two filters 26A(1), 26A(2) respectively adapted to attenuate different energy ranges of the backscatter radiation 24(1), 24(2). The at least two filters 26A(1), 26A(2) may be mounted on one or more movable filter mounts 64A, 64B. As shown, the movable filter mounts 64A, 64B are circular-shaped elements disposed side-by-side, one on either side of the radiation scanner 12. The movable filter mounts 64A, 64B may rotate about respective central axes A2A, A2B, respectively, in order to position respective ones of the radiation filters 26A(1)-26A(N), 26B(1)-26B(N) between the at least one radiation detector 28A, 28B and the specimen 16 at different times to produce the attenuated radiation 26A(1)-26A(N), 26B(1)-26B(N). For example, FIGS. 2B and 2D depicts the filters 26A(1), 26B(1) disposed between the radiation detectors 28A, 28B and specimen 16 to pass the attenuated radiation 32A(1), 32B(1). The attenuated radiation 32A(1), 32B(1) received and converted to detection data 34A(1), 34B(1) by the radiation detectors 28A, 28B may be used to create a portion of the composite image 38 associated with that position of the radiation beam 14 upon the specimen 16 as shown in FIG. 2E. The movable filter mounts 64A, 64B may move, for example with a respective rotations R2A, R2B, to dispose the radiation filter 64A, 64B between the respective radiation detectors 28A, 28B and the backscatter radiation 24. The rotations R2A, R2B may occur continuously or intermittently. For example, the rotations R2A, R2B may be adapted so that the backscatter radiation 24 from each of the scans 58 may pass through respective ones of the radiation filters 26A(1)-26A(N), 26B(1)-26B(N). In this manner, the attenuated radiation 32A(1)-32A(N), 32B(1)-32B(N) received at each of the radiation detectors 28A, 28B may be more easily associated with respective ones of the radiation filters 26A(1)-26A(N), 26B(1)-26B(N) to simplify the analysis of the radiation by the rendering system 36. It is noted that the various ones of the radiation filters 26A(1)-26A(N), 26B(1)-26B(N) associated with the attenuated radiation 32A(1)-32A(N), 32B(1)-32B(N) received at each of the radiation detectors 28A, 28B may monitored by the rendering system 36. In some cases different ones of the radiation filters 26A(1)-26A(N), 26B(1)-26B(N) may be used during the same scan 58 to minimize redundant sweeping of the radiation beam 14 over portions of the specimen 16 to increase inspection speed. In another embodiment, only a single one of the radiation filters 26A(1)-26A(N), 26B(1)-26B(N) may be associated with each of the scans 58. In these embodiments, the radiation detectors 28A, 28B may avoid saturation issues that slow the inspection process by changing the energy flux received when different ones of the radiation filters 26A(1)-26A(N), 26B(1)-26B(N) are used. The radiation detectors 28A, 28B receive the attenuated radiation 32A(1)-32A(2), 32B(1)-32B(2) and produce the detection data 34A(1)-34A(2), 34B(1)-34B(2), respectively, which may be sent to the rendering system 36. The radiation detectors 28A, 28B may each be, for example, a sodium iodide (NaI) scintillation detector as manufactured by Horiba Instruments, Inc. of Kyoto, Japan. Other embodiments of the radiation detectors 28A, 28B may comprise at least one plastic scintillation detector. According to particular embodiments, the radiation detector 28A, 28B may have a width in a range from two (2) centimeters to twenty-four (24) centimeters. The radiation detector 28A, 28B may be compatible with attenuated radiation 32(1), 32(2) having an energy level in a range from two (2) keV to two-hundred (200) keV. In one embodiment, the radiation detectors 28A, 28B and the radiation scanner 12 are disposed on the track stage 59. In this way, the radiation detectors 28A, 28B and the radiation scanner 12 may remain stationary with respect each other as the track stage 59 moves with velocity Vy. In this manner, the radiation detectors 28A, 28B may be positioned to receive the attenuated radiation 32A(1)-32A(N), 32B(1)-32B(N), and then send the detection data 34A(1)-34A(N), 34B(1)-34B(N), respectively, to the rendering system 36. With continued reference to FIGS. 2A-2E, the rendering system 36 creates the composite image 38 from the detection data 34A, 34B and the beam position data 63. For example, trajectories of the radiation beam 14 may be associated with energy flux at respective distributions of wavelengths of the attenuated radiation 32(1), 32(2) and positions of the radiation filters 26A, 26B included in the detection data 34A, 34B. The rendering system 36 may include the electronic assembly 40 including the processor 41, the memory 44, and the storage device 46. The processor 41 may execute computer software code as part of a software program 66, to associate the detection data 34A(1)-34A(N), 34B(1)-34B(N) with positions of the radiation beam 14. The processor 41 may also serve as a controller configured to operate and coordinate the various configurable and movable components of the inspection system 10′(1), e.g., the track stage 59, the enclosure 50, the plurality of radiation filters 26A(1)-26A(N), 26B(1)-26B(N), and the rendering system 36. In this manner, the composition of the specimen 16 and irregularities associated with the composition may be determined according to the detection data 34A(1)-34A(N), 34B(1)-34B(N) received by the rendering system 36. An exemplary method for inspecting the specimen 16 with the inspection system 10′(1) is now discussed. In this regard, FIG. 3 is a flow chart diagram of the method 100 for inspecting the specimen 16 with the inspection system 10′(1) of FIG. 1A. The method 100 may be discussed using the terminology introduced above for consistency and clarity. The method 100 includes emitting the radiation beam 14 from the radiation scanner 12 into the radiation trajectory 15 (operation 102A of FIG. 3). The method 100 also includes selectively positioning the at least two filters 26A(1), 26A(2) of the plurality of filters 26A(1)-26A(N), 26B(1)-26B(N) into the radiation trajectory 15 so that the at least two filters 26A(1), 26A(2) receive the at least the portion of the radiation of the radiation beam 14 and passes the attenuated radiation 32A(1), 32A(2) (operation 102B of FIG. 3). According to one embodiment, the radiation filters 26A(1), 26A(2) of the radiation filters 26A(1)-26A(N), 26B(1)-26B(N) respectively have different attenuation characteristics. Some of the radiation filters 26A(1)-26A(N), 26B(1)-26B(N) may have attenuation characteristics which minimally attenuates the portion of the radiation beam 14. The method 100 also includes receiving the attenuated radiation 32A(1), 32A(2) with the at least one radiation detector 28 of the inspection system 10′(1) and producing detection data 34A(1), 34A(2) associated with the energy intensity of the attenuated radiation 32A(1), 32A(2) (operation 102C of FIG. 3). The method 100 also includes determining whether the emitting of the radiation beam 14 upon the portion of the specimen 16 is complete (operation 102D of FIG. 3). The method 100 may direct the inspection system 10′(1) to either operation 102E 102F, or 102A based on the determined answer from operation 102D. If operation 102D determines that the emitting may be complete for inspection of the portion of the specimen 16, then the composite image 38 may be created of the portion of the specimen 16 by the rendering system 36 using the detection data 34(1), 34(2) produced from the attenuated radiation passed through the at least two filters 26A(1), 26A(2) (operation 102E of FIG. 3). Otherwise, the inspection system 10′(1) may emit the radiation beam 14 upon a radiation trajectory 15 over the portion of the specimen 16 (operation 102A) or may also move the filters 26A, 26B with respect to the radiation detector 28 to enable selectable ones of the plurality of filters 26A(1)-26A(N), 26B(1)-26B(N) to pass attenuated radiation to the radiation detector 28 (operation 102F of FIG. 3). In this manner, a wide variety of information included as part of the backscattered radiation 24 of the specimen 16 may be associated with different ones of the radiation filters 26A(1)-26A(N), 26B(1)-26B(N) and irregularities may be more easily identified by comparing the received values of the attenuated radiation 32A(1)-32A(N), 32B(1)-32B(N) as the radiation beam 14 sweeps over the specimen 16. It is noted that the method 100 may also include determining whether other portions of the specimen 16 are to be inspected (operation 102G of FIG. 3). If additional portions of the specimen 16 are to be inspected, then the inspection system 10′(1) may emit and sweep the radiation beam 14 across the other portions of the specimen 16 (operation 102A of FIG. 3). Otherwise, the method 100 may end. In this manner, the portions of the specimen 16 to be inspected may be evaluated by the inspection system 10′(1) to determine irregularities and related material compositions of the specimen 16. Referring now to FIG. 4A, a schematic view of another embodiment of an inspection system 10(2) is shown. The embodiment of FIG. 4A is similar to the inspection systems 10(1), 10′(1), so only the differences will be discussed for clarity and conciseness. The inspection system 10(2) includes at least one filter 106 at a different angular position (theta) relative to the radiation beam 14 emitted from the inspection system 10(2) compared to the filters 26A, 26B. The inspection system 10(2) also includes a radiation detector 104 which is disposed to receive attenuated radiation 32(3) which is passed through the filter 106. The radiation detector 104 may provide detection data 34(3) to the rendering system 36, where the information provided by the attenuated radiation 32(3) may be analyzed to better determine irregularities of the specimen 16. FIG. 4B is a top view of the inspection system 10(2) of FIG. 4A depicting backscatter radiation 24(1)-24(N) reflected from the specimen 16. The rendering system 36 is hidden from view to emphasize other features of the inspection system 10(2). In this regard, the backscatter radiation 24(1)-24(N) is received by the filters 26A, 26B, 106A, 106B at the different angular positions (theta) relative to the radiation beam 14. The filters 106A, 106B pass attenuated radiation 32A(3), 32B(3), respectively, to the radiation detectors 104A, 104B. In this manner, backscatter radiation 24A(3), 24B(3) directed at a reflection angle (theta) more flared from the radiation beam 14 may be received by the inspection system 10(2). The reflection angle (theta) of the backscatter radiation 24(1)-24(N) may provide information regarding the presence of different characteristics of the specimen 16. In this regard, FIG. 4C is a graph depicting two (2) distributions of energy backscattered from the specimen 16 of FIG. 4B, wherein the two (2) distributions include respectively a higher energy portion 112A including a 511 keV energy portion of the backscattered radiation 24(1)-24(N) and a lower energy portion 112B including a 2.75 keV energy portion of the backscattered radiation 24(1)-24(N). The lower energy portion 112B may preferentially backscatter with a reflection angle (theta) near 180 degrees, as opposed to the side, for example, near 90 degrees. In contrast, the higher energy portion 112A scatter more uniformly over a wide range of reflection angles (theta). In this manner, the radiation detectors 104A, 104B may be disposed in a manner to preferentially receive energy portions of the backscatter radiation 24(1)-24(N) that reflect from the specimen 16 at various predetermined reflection angles (theta) to isolate energy portions of the backscatter radiation which contain specific information to the irregularities and associated material compositions of the specimen 16. In this regard, a practical use for having the radiation detectors at different angles (theta) may be to selectively filter different types of irregularities of the specimen 16. When the specimen 16 includes multiple material types, each of the material types may reflect different energies and at different reflection angles (theta) relative to the radiation beam 14. For example, the specimen 16 may include the inner portion 18B including a metal material which may be covered with the outer portion 18A of composite materials. The outer portion 18A including the composite materials may reflect the lower energy portion 112B narrowly near the radiation beam 14 and this lower energy portion 112B may be selectively received by the radiation detectors 28A, 28B in a range of reflection angles (theta) from 135 degrees to 225 degrees. Any irregularities related to the portion 18A of the specimen 16 may be discerned from the backscatter radiation received from the portion 18B of the specimen 16. In this regard, one or more of the radiation filters 26A, 26B may be configured to attenuate the higher energy portion 112A to focus on information provided by the lower energy portion 112B. The inspection system 10(2) includes other features to discern irregularities and material compositions of the inner portion 18B of the specimen 16. The higher energy portion 112A of the radiation beam 14 may mostly pass through the outer portion 18A of the specimen 16 to be incident upon the inner portion 18B of the specimen 16 as represented by the portion 22 of the radiation beam 14 in FIG. 4A. Unlike the outer portion 18A, the inner portion 18B of the specimen 16, including the metal material, may be more reflective to the higher energy portion 112A of the radiation beam 14 and the higher energy portion 112A may be reflected from the inner portion 18B with a relatively uniform distribution along reflection angles as depicted in FIG. 4C. Specifically, the higher energy portion 112A may also preferentially reflect with a wide range of reflection angles (theta), including values less than 135 degrees and more than 225 degrees where the higher energy portion 112A is more predominantly reflected in comparison to the lower energy portion 112B. The radiation filters 106A, 106B may be configured to attenuate the lower energy portion 112B to focus on information provided by the higher energy portion 112A. In this manner, as the radiation beam 14 may be emitted and swept across portions of the specimen, changes in the lower energy portion 112B received at the radiation detectors 28A, 28B may indicate irregularities in the outer portion 18A of the specimen whereas changes in the higher energy portion 112A received at the radiation detectors 104A, 104B may indicate irregularities in the inner portion 18B of the specimen 16. FIGS. 5A and 5B are a schematic view and a top view, respectively, of an inspection system 10(3). The inspection system 10(3) is similar to the inspection system 10(1) and so only the differences will be discussed for clarity and conciseness. The inspection system 10(3) may include radiation detectors 126, 128, 130 in a layered arrangement relative to the direction of the backscattered radiation 24(4). The radiation detectors 126, 128, 130 may serve as layered filters which are selective, so that the backscattered radiation 24(4) may pass through various ones of the radiation detectors 126, 128, 130 depending upon respective energy distribution of the backscatter radiation 24(4). For example, a higher energy portion 132 of the backscatter radiation 24(4) may pass through the radiation detectors 126, 128 to be captured and measured at the radiation detector 130. A medial energy portion 134 of the backscatter radiation 24(4) may pass through the radiation detector 126 to be captured and measured at the radiation detector 128 and a lower energy portion 136 of the backscatter radiation 24(4) may be captured and measured at the radiation detector 126. In this manner, the inspection system 10(3) may facilitate selective filtering of the backscatter radiation 24(4) to discriminate between various materials of the specimen 16 and/or irregularities which may scatter back at different energies. It is noted that in FIG. 5B the inspection system 10(3) may include the track stage 59, with the radiation detectors 126, 128, 130 disposed thereon. In this way, the track stage 59 can move the radiation detectors along the y-axis. In one embodiment, the radiation detector 126 may be made up of a plurality of radiation detectors. For example, as illustrated in FIG. 5B, the radiation detector 126 includes two radiation detectors 126A, 126B located on opposite sides of the radiation scanner 12. Likewise, the radiation detector 128 may include radiation detectors 128A, 128B and the radiation detector 130 may include radiation detectors 130A, 130B, where each of the constituent radiation detectors are located on opposite sides of the radiation scanner 12. In this manner, the inspection system 10(3) may identify irregularities of the specimen 16 using the backscatter radiation 24(4) reflected on opposite sides of the radiation beam 14 of the radiation scanner 12. FIGS. 6A and 6B are schematic views of an inspection system 10(4). The inspection system 10(4) is similar to the inspection system 10(1) and so only the differences will be discussed for clarity and conciseness. Instead of including the radiation filters 26A(1)-26A(N), 26B(1)-26B(N) attenuating the backscatter radiation 24, the inspection system 10(4) includes radiation filters 150(1), 150(2) which selectively attenuate the radiation beam 14 prior to being incident upon the specimen 16. In this regard, FIG. 6A depicts the radiation filter 150(1) being used to attenuate the radiation beam 14 to pass attenuated radiation 152(1) which may be reflected from the specimen 16 as backscattered attenuated radiation 32(1) (compare to FIG. 1A). FIG. 6B depicts the radiation filter 150(2) displacing the radiation filter 150(1) by, for example, translation or rotation about an axis of rotation A3 so that the radiation filter 150(2) is positioned to attenuate the radiation beam 14, resulting in attenuated radiation 152(2) being propagated to specimen 16. Attenuated radiation 152(2) is then reflected as backscatter from the specimen 16 as the attenuated radiation 32(2) (compare to FIG. 1B). The this manner, the attenuated radiation 32(1), 32(2) may be alternatively received by the radiation detector 28 to be analyzed by the rendering system 36 to determine irregularities and associated material compositions of the specimen 16. The descriptions of the various embodiments of the present invention have been presented for purposes of illustration, but are not intended to be exhaustive or limited to the embodiments disclosed. Many modifications and variations will be apparent to those of ordinary skill in the art without departing from the scope and spirit of the described embodiments. In one example, the specimen 16 may be an aircraft wing having the outer portion 18A be an aircraft skin made of composite and the inner portion 18B being an aircraft structural member (or “spar”) made of aluminum or other metal. In some embodiments, it is recognized that the inspection system could include optical equipment like beam steering components (e.g., reflective mirrors or refractive lenses), focusing lenses, collimators, filters, and/or others to steer the radiation along a radiation trajectory. The terminology used herein was chosen to best explain the principles of the embodiments, the practical application or technical improvement over technologies found in the marketplace, or to enable others of ordinary skill in the art to understand the embodiments disclosed herein. The present invention may be a system, a method, and/or a computer program product. The computer program product may include a computer-readable storage medium (or media) having computer-readable program instructions thereon for causing a processor to carry out aspects of the present invention. The computer-readable storage medium can be a tangible device that can retain and store instructions for use by an instruction execution device. The computer-readable storage medium may be, for example, but is not limited to, an electronic storage device, a magnetic storage device, an optical storage device, an electromagnetic storage device, a semiconductor storage device, or any suitable combination of the foregoing. A non-exhaustive list of more specific examples of the computer-readable storage medium includes the following: a portable computer diskette, a hard disk, a random access memory (RAM), a read-only memory (ROM), an erasable programmable read-only memory (EPROM or Flash memory), a static random access memory (SRAM), a portable compact disc read-only memory (CD-ROM), a digital versatile disk (DVD), a memory stick, and any suitable combination of the foregoing. A computer-readable storage medium, as used herein, is not to be construed as being transitory signals per se, such as radio waves or other freely propagating electromagnetic waves, electromagnetic waves propagating through a waveguide or other transmission media (e.g., light pulses passing through a fiber-optic cable), or electrical signals transmitted through a wire. Computer-readable program instructions described herein can be downloaded to respective computing/processing devices from a computer-readable storage medium or to an external computer or external storage device via a network, for example, the Internet, a local area network, a wide area network and/or a wireless network. The network may comprise copper transmission cables, optical transmission fibers, wireless transmission, routers, firewalls, switches, gateway computers and/or edge servers. A network adapter card or network interface in each computing/processing device receives computer-readable program instructions from the network and forwards the computer-readable program instructions for storage in a computer-readable storage medium within the respective computing/processing device. Computer-readable program instructions for carrying out operations of the present invention may be assembler instructions, instruction-set-architecture (ISA) instructions, machine instructions, machine dependent instructions, microcode, firmware instructions, state-setting data, or either source code or object code written in any combination of one or more programming languages, including an object oriented programming language such as Java, Smalltalk, C++ or the like, and conventional procedural programming languages, such as the “C” programming language or similar programming languages. The computer-readable program instructions may execute entirely on the user's computer, partly on the user's computer, as a stand-alone software package, partly on the user's computer and partly on a remote computer or entirely on the remote computer or server. In the latter scenario, the remote computer may be connected to the user's computer through any type of network, including a local area network (LAN) or a wide area network (WAN), or the connection may be made to an external computer (for example, through the Internet using an Internet Service Provider). In some embodiments, electronic circuitry including, for example, programmable logic circuitry, field-programmable gate arrays (FPGA), or programmable logic arrays (PLA) may execute the computer-readable program instructions by utilizing state information of the computer-readable program instructions to personalize the electronic circuitry, in order to perform aspects of the present invention. These computer-readable program instructions may be provided to a processor of a general purpose computer, special purpose computer, or other programmable data processing apparatus to produce a machine, such that the instructions, which execute via the processor of the computer or other programmable data processing apparatus, create means for implementing the functions/acts specified in the flowchart and/or block diagram block or blocks. These computer-readable program instructions may also be stored in a computer-readable storage medium that can direct a computer, a programmable data processing apparatus, and/or other devices to function in a particular manner, such that the computer-readable storage medium having instructions stored therein comprises an article of manufacture including instructions which implement aspects of the function/act specified in the flowchart and/or block diagram block or blocks. The computer-readable program instructions may also be loaded onto a computer, other programmable data processing apparatus, or other device to cause a series of operational steps to be performed on the computer, other programmable apparatus or other device to produce a computer implemented process, such that the instructions which execute on the computer, other programmable apparatus, or other device implement the functions/acts specified in the flowchart and/or block diagram block or blocks. The flowchart and block diagrams in the Figures illustrate the architecture, functionality, and operation of possible implementations of systems, methods, and computer program products according to various embodiments of the present invention. In this regard, each block in the flowchart or block diagrams may represent a module, segment, or portion of instructions, which comprises one or more executable instructions for implementing the specified logical function(s). In some alternative implementations, the functions noted in the block may occur out of the order noted in the figures. For example, two blocks shown in succession may, in fact, be executed substantially concurrently, or the blocks may sometimes be executed in the reverse order, depending upon the functionality involved. It will also be noted that each block of the block diagrams and/or flowchart illustration, and combinations of blocks in the block diagrams and/or flowchart illustration, can be implemented by special purpose hardware-based systems that perform the specified functions or acts or carry out combinations of special purpose hardware and computer instructions. While the foregoing is directed to embodiments of the present invention, other and further embodiments of the invention may be devised without departing from the basic scope thereof, and the scope thereof is determined by the claims that follow. |
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description | 1. Field of the Invention The present invention relates to a radiation detector using a scintillator, and more particularly, to a two-dimensional radiation detector which is referred to as a flat panel detector. 2. Description of the Related Art In radiography which acquires an image by applying radiation to an object and detecting radiation which passes through the object, digital radiography (DR) which acquires an image by converting the detected radiation into an electric signal is popular. Generally, in DR, a flat panel detector (FPD) is used which includes a light receiving element having two-dimensionally arranged pixels and a scintillator layer placed on a light receiving surface of the light receiving element. Depending on the application, in most cases, a wide imaging area of several tens of centimeters or more per side is required for the FPD, and thus, the scintillator layer to be formed is required to have a large area. Therefore, the scintillator layer is formed by using vacuum deposition which enables formation of a large-area layer or an applying method of applying a binding agent having scintillator particles dispersed therein. In particular, a scintillator layer formed by vapor depositing cesium iodide has an advantage that, because cesium iodide is grown as needle crystals and so-called crosstalks are suppressed by a light guiding effect in the needle crystals, a high position resolution can be obtained. However, actually, adjacent needle crystals adhere to each other in places, and thus, in order to obtain a still higher position resolution, it is effective to cause the scintillator layer to have a greater extent of anisotropy in light propagation by causing the scintillator layer to have a structure in which two crystal phases having different refractive indices are completely separate from each other. In order to manufacture such a scintillator layer including a structure in which two crystal phases having different refractive indices are completely separate from each other (phase separation structure), it is conceivable to employ a technology of micromachining a scintillator crystal, a technology of separating two phases of eutectic composition in one axial direction and growing the two phases, or the like. However, it is technically difficult to obtain by these technologies a phase separation structure having a large area of several tens of centimeters or more per side. In order to use a phase separation structure as a scintillator layer of an FPD, it is necessary to spread (tile) multiple phase separation structures processed to have a predetermined shape all over a surface of a light receiving element in order to secure a large imaging area. In this case, a problem is newly found that slight clearance which appears between adjacent phase separation structures due to limitations on the processing accuracy has a nonnegligible effect on a taken image. Specifically, a medium in the clearance which appears between phase separation structures is typically air (having a refractive index of 1.0), and thus, due to an effect of total reflection at an interface with a phase separation structure (tiling interface), the propagation characteristics of scintillation light generated in the phase separation structure locally change greatly. As a result, in pixels corresponding to the clearance between adjacent phase separation structures, the amount of incident scintillation light considerably reduces, and thus, the clearance between adjacent phase separation structures appears in the taken image as defects. Such defects are conspicuous when a large amount of scintillation light is generated at a location near a tiling interface. On the other hand, when a large amount of scintillation light is generated at a location far from the tiling interface, due to the great extent of anisotropy in light propagation of the phase separation structure, the amount of scintillation light which reaches the tiling interface is small, and thus, the defects are not conspicuous. In summary, when am object is actually imaged, bright portions and dark portions differ depending on each object, and thus, the effect of defects on the taken image differs accordingly. In order to perform calibration of such defects by image correction, a sophisticated correction technology is required with regard to each object. In view of the above-mentioned problem, the present invention provides a radiation detector which is capable of suppressing, in an FPD having a scintillator layer formed therein by tiling phase separation structures having anisotropy in light propagation, the effect of clearance between adjacent phase separation structures on a taken image without making sophisticated image correction. In order to solve the above-mentioned problem, according to one aspect of the present invention, there is provided a radiation detector, including: a two-dimensional light receiving element including a plurality of pixels; and a scintillator layer having multiple scintillator crystals two-dimensionally arranged on a light receiving surface of the two-dimensional light receiving element, in which: the multiple scintillator crystals each include two crystal phases, which are a first crystal phase including a material including a plurality of columnar crystals extending in a direction perpendicular to the light receiving surface of the two-dimensional light receiving element and having a refractive index n1, and a second crystal phase including a material existing between the plurality of columnar crystals and having a refractive index n2; and a material having a refractive index n3 is placed between adjacent scintillator crystals of the multiple scintillator crystals, the refractive index n3 satisfying a relationship of one of n1≦n3≦n2 and n2≦n3≦n1. Further, according to another aspect of the present invention, there is provided a radiation detector, in which a crystal phase including a material having a higher refractive index of the two crystal phases, which are the first crystal phase and the second crystal phase together constituting the scintillator crystal functions as a scintillator. Further, according to still another aspect of the present invention, there is provided a radiation detector, in which the first crystal phase is a crystal phase including NaCl as a main component and the second crystal phase is a crystal phase including CsI as a main component. According to the present invention, in the FPD having the scintillator layer formed therein by tiling the phase separation structures having large anisotropy in light propagation, it is possible to suppress the effect of clearance between the adjacent phase separation structures on the taken image. Further features of the present invention will become apparent from the following description of exemplary embodiments with reference to the attached drawings. An embodiment for carrying out the present invention is described in the following with reference to the attached drawings. Note that, there are various embodiments for carrying out the present invention (various structures and various materials), but a point common to all the embodiments is that a scintillator crystal having a phase separation structure including two crystal phases, one crystal phase having a refractive index lower than that of the other crystal phase, has a first principal surface and a second principal surface which are not located on a same surface, the other crystal phase being exposed on part of the first principal surface and on part of the second principal surface, the part of the other crystal phase exposed on the first principal surface and the part of the other crystal phase exposed on the second principal surface being connected to each other. This causes light in a higher refractive index crystal phase to be totally reflected by a lower refractive index crystal phase located around the higher refractive index crystal phase, and as a result, to be guided and propagate through the higher refractive index crystal phase. During the propagation, the higher refractive index crystal phase is exposed on the first principal surface and the second principal surface, and these exposed portions are connected to each other, and thus, waveguiding (light guiding) is achieved toward the first principal surface or the second principal surface. In other words, it can be said that light generated in the scintillator crystal travels toward the first principal surface or the second principal surface under a state in which the light is confined in the other crystal phase having the higher refractive index (that is, without diffusion of the light). In this way, in all the embodiments of the present invention, the scintillator crystal itself has a waveguiding function (light guiding function). Note that, in this case, as illustrated in FIG. 8, for example, a first principal surface 81 is a surface which faces photodetectors 84 provided on a substrate 83, and a second principal surface 82 is a surface through which radiation such as X-rays enters. This enables waveguiding (light guiding) of light generated in a scintillator crystal 85 toward the photodetectors, and a scintillator crystal with an excellent efficiency in the use of light can be provided, and a radiation detector using the same having high luminance and high resolution can be provided. Note that, in the embodiment described below, a structure is preferred in which the one crystal phase as the lower refractive index crystal phase also has a part exposed on the first principal surface and a part exposed on the second principal surface and the exposed parts are connected to each other. This enables waveguiding (light guiding) of light in the other crystal phase as the higher refractive index crystal phase to the first principal surface or the second principal surface to be achieved with more reliability and without diffusion of the light. Further, a structure is preferred in which the one crystal phase as the lower refractive index crystal phase is located within the other crystal phase as the higher refractive index crystal phase. This can suppress the ratio occupied by the one crystal phase as the lower refractive index crystal phase in the scintillator crystal, and still, a sufficient waveguiding function (light guiding function) can be acquired. The embodiment of the present invention is described in the following. As illustrated in FIGS. 1A and 1B, in the radiation detector according to the present invention, a scintillator layer is formed by two-dimensionally arranging and tiling multiple scintillator crystals 12, which each have a phase separation structure, on a light receiving surface of a two-dimensional light receiving element 11 which includes a plurality of pixels. FIG. 2 illustrates an exemplary structure of a scintillator crystal having a phase separation structure according to the present invention. FIG. 2 illustrates a scintillator crystal having a phase separation structure in which a first crystal phase (cylinders) 21 including a plurality of columnar crystals having unidirectionality and having a refractive index n1 is formed in a second crystal phase (matrix) 22 having a refractive index n2. Further, any one of the first crystal phase and the second crystal phase having the higher refractive index is formed of a scintillator material which emits light when being excited by radiation. Therefore, when n1>n2, scintillation light is generated from the cylinders, and light is confined within and propagated through the cylinders by total reflection at the interfaces between the cylinders and the matrix, and thus, a great extent of anisotropy in light propagation develops in the direction of the cylinders. On the other hand, when n1<n2, scintillation light is generated from the matrix, and total reflection occurs at the interfaces between the matrix and the cylinders. As a result, horizontal spread of light within the matrix is suppressed by the plurality of cylinders extending vertically when the light is propagated, and thus, anisotropy in light propagation develops although the extent of the anisotropy is smaller than that when light is confined within the cylinders. The scintillator crystal including the phase separation structure as illustrated in FIG. 2 is manufactured by a technology of micromachining a scintillator crystal, a technology of separating two phases of eutectic composition in one axial direction and growing the two phases, or the like. It is technically difficult to obtain, by any one of these technologies, a scintillator crystal having a large area. Therefore, the radiation detector according to the present invention secures a large imaging area by tiling scintillator crystals processed to have a predetermined shape as illustrated in FIGS. 1A and 1B. In this case, by tiling the scintillator crystals 12 with a resin 13 having a refractive index n3 provided between adjacent scintillator crystals 12, slight clearance which appears between the adjacent scintillator crystals 12 is filled with the resin 13 having the refractive index n3. The medium to be provided between adjacent scintillator crystals 12 may be other than a resin insofar as the refractive index n3 of the medium is between the refractive index n1 of the first crystal phase and the refractive index n2 of the second crystal phase and the medium does not absorb the scintillation light. A medium which does not absorb the scintillation light specifically means a medium having a transmittance that is similar to that of the crystal phase functioning as the scintillator of the first crystal phase and the second crystal phase. As illustrated in FIGS. 1A and 1B, by providing the resin 13 having the refractive index n3 between the adjacent scintillator crystals 12, compared with a case in which the resin is not provided therebetween, the difference in refractive index at the tiling interfaces becomes smaller. As a result, total reflection which occurs at the tiling interfaces reduces to suppress a local great change in propagation characteristics of the scintillation light. Specifically, an effect of the tiling on a taken image can be suppressed. In this example, an effect of clearance between adjacent scintillator crystals on a taken image when scintillator crystals having a phase separation structure are tiled was estimated from a geometrical optic simulation. In this example, calculation was performed with regard to an NaCl—CsI phase separation scintillator crystal including two crystal phases: a first crystal phase of NaCl (having a refractive index n1 of 1.55); and a second crystal phase of CsI (having a refractive index of 1.78). Specifically, as illustrated in FIG. 3, in the structure of the scintillator crystal, NaCl portions 31 having a diameter of 2 μm were arranged to form a triangle lattice with a period 32 of 4 μm, and scintillation light was caused to be generated from CsI 33 as a matrix portion. Further, as illustrated in FIGS. 4A and 4B, two scintillator crystals 41 whose dimensions were W×D×H=500 μm×1,000 μm×500 μm were arranged with clearance 42 of 40 μm provided therebetween. Radiation entered from upper surfaces 43 of the scintillator crystals 41, and light receiving surfaces 45 provided on lower surfaces 44 were adapted to detect scintillation light. Further, a reflection plane 46 for reflecting the scintillation light was provided on the upper surfaces 43. The scintillation light was caused to be isotropically generated from the scintillator crystals 41. The light flux of the scintillation light was caused to have intensity distribution which decreases exponentially from the upper surfaces toward the lower surfaces taking into consideration of absorption of the radiation. FIG. 5 shows the result of calculation of illuminance distribution on the light receiving surfaces when the refractive index of the clearance 42 was 1.0 based on geometrical optics. FIG. 5 shows linear illuminance distribution when Y=0 (−150 μm≦×≦150 μm), in which the X axis and the Y axis were taken with the center of the light receiving surface 45 of 1,040 μm×1,000 μm being an origin 61 as shown in FIG. 6. In this case, the illuminance was calculated and plotted with regard to every region of 10 μm×10 μm when Y=0, and the data were normalized with respect to the maximum illuminance. From FIG. 5, it can be confirmed that, while the illuminance in regions other than the clearance 42 was almost constant, the illuminance of a region corresponding to the clearance 42 was considerably lowered, and that the amount of incident scintillation light was lowered by the effect of total reflection which occurred at the tiling interfaces. Further, (1-α), where a is the minimum illuminance in FIG. 5, is thought to be an index of the effect of the clearance between the scintillator crystals on the taken image (propagation loss). In the case shown in FIG. 5, (1-α) is 0.525=52.5%, and thus, the region corresponding to the clearance causes a propagation loss of 52.5% compared with the regions without the clearance. Next, the refractive index n3 of the medium of the clearance 42 illustrated in FIGS. 4A and 4B was changed and calculation similar to the one described above was performed. FIG. 7 is a graph showing a change in propagation loss in accordance with the refractive index n3. From FIG. 7, it can be confirmed that, while the propagation loss greatly changed in accordance with the refractive index n3 of the medium of the clearance 42, generally when 1.55≦n3≦1.78, the propagation loss was suppressed to be 20% or lower, and a great change in propagation loss did not occur in the region. Specifically, by tiling the NaCl—CsI phase separation scintillator crystals with a medium having the refractive index n3 which satisfies 1.55≦n3≦1.78 provided between adjacent NaCl—CsI phase separation scintillator crystals, the effect of the tiling on a taken image can be suppressed. As a material having the refractive index n3, for example, an epoxy resin (having a refractive index of 1.55 to 1.61), a melamine resin (having a refractive index of 1.6), a polystyrene resin (having a refractive index of 1.6), a vinylidene chloride resin (having a refractive index of 1.61), and a polycarbonate resin (having a refractive index of 1.59) can be used. Further, in this example, calculation was performed with regard to an NaCl—CsI phase separation scintillator crystal, but the present invention is not limited to this example. Specifically, insofar as the medium having the refractive index n3, the first crystal phase having the refractive index n1, and the second crystal phase having the refractive index n2 satisfy n1≦n3≦n2 or n2≦n3≦n1, even when a phase separation structure other than the NaCl—CsI phase separation scintillator is used, a similar effect can be obtained. This example relates to imaging using the radiation detector according to the present invention. First, an NaCl—CsI phase separation scintillator is manufactured which includes the first crystal phase whose main component is NaCl and the second crystal phase whose main component is CsI, the first crystal phase being columnar crystals having a diameter of 2 μm and an average period of 4 μm, the second crystal phase existing between the columnar crystals. Specifically, NaCl and CsI are mixed in composition at a eutectic point. After the mixture is heated and melted at 500° C., the mixture is cooled so as to solidify and to have unidirectionality. The acquired NaCl—CsI phase separation scintillator is cut and polished using a lapping sheet to prepare two NaCl—CsI phase separation scintillator crystals whose dimensions are W×D×H=5 mm×5 mm×500 μm. Then, the scintillator crystals are arranged side by side on a CCD sensor with 20-μm pixel pitches. Aluminum foil is placed on a surface which is opposite to a surface in contact with the CCD sensor. X-rays are applied from above the aluminum foil, and an image acquired from the CCD sensor is evaluated. Specifically, a taken image when an epoxy resin having a refractive index of 1.6 is provided between the scintillator crystals and a taken image when the epoxy resin is not provided between the scintillator crystals in a case where the two NaCl—CsI phase separation scintillator crystals are arranged side by side on the CCD sensor are compared with each other. When the two taken images are compared, it is confirmed that, in the case of the taken image when the epoxy resin is not provided, luminance is considerably lowered in a linear region corresponding to the clearance between the two scintillator crystals. In other words, the clearance between the scintillator crystals which appears when the scintillator crystals are arranged side by side is thought to affect the taken image. On the other hand, when the epoxy resin is provided, a linear region in which the luminance is lowered is less conspicuous. It follows that, by filling the clearance between the scintillator crystals with the epoxy resin, the effect on the taken image is thought to be suppressed. While the present invention has been described with reference to exemplary embodiments, it is to be understood that the invention is not limited to the disclosed exemplary embodiments. The scope of the following claims is to be accorded the broadest interpretation so as to encompass all such modifications and equivalent structures and functions. This application claims the benefit of Japanese Patent Application No. 2011-163111, filed Jul. 26, 2011, which is hereby incorporated by reference herein in its entirety. |
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abstract | A method for displaying a changing combustor condition including: sensing the combustor condition in real time using a sensor array in a gas path of the combustor; generating data from the sensor array representative of the combustor condition at a plurality of positions in the gas path; transmitting the generated data to a computer system proximate to a controller for the combustor; generating a graphical representation of the real time showing combustor conditions in the gas path, and displaying the graphical representation in real time on the computer system. |
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051805430 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention This invention relates to the field of nuclear reactors, particularly of the pressurized water type, and is concerned with the fluid systems which operate following postulated events to provide required safety functions which includes providing emergency water addition to the reactor core following pipe breaks, providing a source of assured water addition for small leaks, removing reactor core decay heat from the reactor core, and assuring that the reactor core is subcritical. 2. Description of the Related Art Present pressurized water reactor (PWR) designs have proven to be sound, safe performers. Recently, however, there has been wide spread interest in simplifying safety features of PWR designs so as to eliminate and/or reduce pumps, piping, instrumentation, etc., which tend to escalate the cost of building and maintaining plants. An important manifestation of new development activities in the area of nuclear plant design is the use of passive rather than active safety features as well as simplified systems design. One such PWR design is described in U.S. Pat. No. 4,753,771 to Conway and Schultz (co-inventors herein). This patent describes a passive safety injection system (PSIS) which once aligned relies on natural forces such as gravity and natural circulation of water and air to provide all required safety functions. Portions of this PSIS is shown in FIG. 1, including a reactor vessel and core 44, and one or more reactor coolant system hot legs, cold legs, steam generator (not shown) and reactor coolant pumps (not shown) which are all of essentially conventional design. A pressurizer 48 is connected to one of the hot legs. This previously patented pressure safety injection system (PSIS) is comprised of the following essential components: 1) A single passive residual heat removal heat exchanger 34 which is located above the reactor coolant system hot leg HL and cold leg CL and is connected at the top by a pipe attached to the hot leg HL and is connected at its bottom to the cold leg CL. This heat exchanger can remove reactor core decay heat when either normally closed valves 38 are opened. It transfers heat to water stored in the in-containment refueling water storage tank (IRWST) 36. 2) Two core makeup tanks 40 and their associated piping (one of two shown) which are located above the reactor coolant system hot leg HL and cold leg CL. This tank is completely filed with water and will drain by gravity into the reactor vessel 44 when either normally closed valves are opened and a) pressurizer 18 water level is below the top of the core makeup tank and the reactor coolant pump(s) is shut off or b) when the water inventory in the reactor coolant system is greatly reduced such that the cold leg(s) CL contains steam. The core makeup tank(s) 40 provide assured inventory makeup to the reactor and can provide sufficient flow to maintain core cooling following postulated ruptures of the reactor coolant system pressure boundary. 3) Two sets of depressurization valves 49 (one set shown) are provided at the top of the pressurizer. Each set of depressurization valves may contain multiple parallel flowpaths depending on the required flowrate vs. optimum/desired depressurization valve sizes. These valves are normally closed but are opened when the water level in the core makeup tank(s) 40 is reduced significantly. This action assures that the reactor coolant system is depressurized sufficiently so that water from the in-containment RWST 36 will begin draining by gravity into the reactor vessel 44 before the core makeup tank(s) 40 have completely drained. 4) The in-containment refueling water storage tank 36 (IRWST) is located above the reactor coolant system hot leg(s) HL and cold leg(s) CL and contains water which acts as a heat sink for operation of the passive RHR heat exchanger 34, quenches steam released during depressurization of the reactor coolant system, provides a longer term source of water injection by gravity into the reactor vessel in the event of a pipe break, and which floods the lower portions of the containment in which the reactor is housed such that the reactor coolant system is flooded above the hot leg(s) and cold leg(s). When the lower portion of the containment is flooded a long term (indefinite time) source of water makeup to the reactor vessel 44 is established from the flooded containment through piping conduits 37 and 39 which is driven by gravity. These features in conjunction with a passive containment cooling system (not shown) described in the above patent, which transfer heat from the steel containment shell to the environment by natural convection results in condensing steam on the inside steel surface of the containment. This condensed steam (water) drains back to the lower portions of the containment and thus replenishes and maintains the water available for gravity drain into the reactor vessel storage tank 36 which serves as a heat sink. The bottom of the heat exchanger is located about 8 feet above the loops. The heat exchanger 34 is actuated by opening either of the air operated valves 38 which fail open on loss of power or signal. If the reactor coolant pumps are operating, the flow through the passive residual heat removal heat exchanger 34 will force circulation from the higher pressure cold leg through the heat exchanger to the hot leg. In case the reactor coolant pumps are not available, the flow will be by natural circulation from the hot leg to the top of the passive residual heat removal heat exchanger 34 to the cold leg. The air operated control valves give the operator a means of controlling the reactor coolant system temperature to a constant value or if desired, to cool down the reactor coolant system. The in-containment refueling water storage tank 36 will absorb decay heat for several hours before the water becomes saturated. However, it will take days to boil off sufficient water from the in-containment refueling water storage tank 36 before the heat removal capability degrades. This provides ample time to recover main or start feed water or to align the normal residual heat removal cooling equipment which is part of the spent fuel cooling system. The passive heat exchanger 34 is made up of headers to which tubes are welded. The tubes are oriented vertically and are about 20 feet long. There are four headers which are arranged in parallel, separated by several feet to promote good mixing of the steam generated on the surface of the tubes with the water in the in-containment refueling water storage tank 36. The passive residual heat removal heat exchanger replaces the safety grade auxiliary feed water system used in the past and does not rely on pumps, AC power or air/water cooling systems. The function of the passive heat exchanger is also not affected by failure of a steam generator pressure boundary, such as steam or feed line breaks or steam generator tube ruptures. With respect to the passive safety injection function, passive reactor coolant makeup is provided to accommodate small leaks when the normal makeup system is unavailable and to accommodate larger leaks resulting from loss of coolant accidents (LOCA). Safety grade reactor coolant makeup and safety injection are provided by a set of water tanks: two core makeup tanks 40 (only one of which is shown in FIG. 1), two accumulators 42 (only one of which is shown in FIG. 1) and an in-containment refueling water storage tank 36. The core makeup tanks 40 are designed to provide makeup for small reactor coolant system leaks at any pressure and to provide safety injection for small LOCA. These tanks utilize gravity for their injection force. They are located above the reactor coolant loops and have a pressure balance line connected to the top of the tank to equalize pressures. Each of the core makeup tanks is full of borated water, and are designed for the same pressure as a reactor coolant system. The discharge from the core makeup tanks is from the bottom of each tank to a separate safety injection nozzel on the reactor vessel. The injection water enters the cold leg downcomer region 44. The discharge line is normally isolated by two parallel air operated valves 46 that fail open on loss of air pressure or control signal. Two separate pressure balancing lines are provided for each core makeup tank 40. One line is from the top of the pressurizer 48 and another line is from a reactor coolant cold leg pipe. The line from the pressurizer is a small line that provides reactor coolant makeup following transients or whenever normal makeup is not available. This line is normally open and contains a check valve to prevent possible back flow or leakage from the cold legs which are at a higher pressure when the reactor coolant pumps are operating. In order to allow core makeup tank injection, the reactor coolant pumps are tripped when the pressurizer level reaches a low-low level. The line from the cold legs to the core makeup tanks is a larger line that provides reactor coolant makeup capability as required for LOCA. This line is normally isolated by two parallel air operated valves 50 that fail open on loss of air pressure or control signal. If the cold legs become voided as they do during a LOCA, this line provides a greater flow of steam to the top of the core makeup tanks which allows for a greater flow of water to the reactor coolant system. The accumulators 42 are required for large LOCAs because of the need for very high makeup flows to refill the reactor vessel downcomer and lower plenum. The accumulator tanks contain borated water with an over pressure of nitrogen. Because there are limited volumes of water in the core makeup tanks and in the accumulators, additional sources of water are required in the longer term. The in-containment refueling water storage tank 36 is thus relied on as the longer term source of makeup water. However, in order to get injection from the in-containment refueling water storage tank, the reactor coolant system pressure must be reduced to about 10 PSIG above containment pressure. An automatic depressurization system is provided to accomplish this function. A series of valves connected to the pressurizer provide a phased depressurization capability. The discharge from these valves is sparged into the in-containment refueling water storage tank to minimize the consequences of a spurious opening of one of the depressurization valves. These valves are arranged in three stages with the first stage being smaller. The staging reduces the peak flow rates and the resulting load on the discharge pipes, spargers, and the in-containment refueling water storage tank. After about 10 hours, the in-containment refueling water storage tank will also be empty. However, by that time the containment will be flooded above up to above the reactor coolant loop level and the water in the containment will drain by gravity back into the reactor coolant system. A stable long term core cooling/makeup to the reactor cooling system is thus established. Boron or borated water is generally known as a means of reducing or controlling nuclear reactor power due to boron's ability to absorb neutrons. However, the introduction of boron into the passive safety system concept presents a number of difficult problems. In order to assure the reactor remains subcritical after any postulated event, all the sources of water from the PSIS to the reactor vessel must contain boric acid solution. Referring to FIG. 2, the long term core cooling mode of the passive safety systems consist of boiling water in the reactor vessel and steam produced is vented to the containment were it is cooled/condensed and drained back to the flooded lower elevations of the containment building. Due to the continued boiling of water in the core region, the boron concentration can eventually become high enough in the core region to impede heat transfer. As water boils, it leaves the boron behind when the steam is vented to the containment through the pressurizer, and the boron thus becomes concentrated in the reactor vessel. The various water sources from the core makeup tanks and the water storage tank are fed into the reactor vessel through a line 11. Water in the containment is illustrated by the line A. Water drains from within the containment into the reactor vessel through a sump screen 13 by the difference in water head between the maximum water level A in the containment and the water level in the reactor vessel. The core makeup tank 40 shown in FIG. 1 (only one of two shown) provides a source of water that can drain by gravity into the reactor at any prevailing pressure to make up water lost from the reactor coolant system due to small leaks or even the postulated rupture of the largest pipe. These tanks are designed to operate at full reactor pressure. In the referenced patent, the two CMT's and their associated piping are sized to provide, by themselves, sufficient water to provide acceptable core cooling for even the largest postulated pipe rupture. In a larger reactor, a proportionally larger flowrate is required to rapidly refill the reactor and reflood the reactor core following a postulated severance of the largest reactor coolant system pipe. This would require that the core makeup tanks and their associated piping become proportionately larger in volume and area respectively. This direct scale-up approach would not be the most cost effective means of achieving higher injection flowrates and may become impractical. The depressurization valves and associated piping shown in FIG. 1 assure that the reactor coolant system pressure can be reduced sufficiently and in a timely manner such that the reactor pressure is less than the elevational head of water stored in the in-containment refueling water storage tank before the core makeup tank(s) have completely drained. In a larger reactor application, the size (areas) of the depressurization valves and associated piping must be increased proportionately to achieve a similar depressurization rate and pressure. For large reactors with the depressurization arrangement shown in FIG. 1, the number of depressurization valves and/or valve and piping size may become impractical and not the most cost-effective solution. The preferred embodiment of the above patent was for a "small" nuclear reactor (45 million watt thermal output). In order to apply the passive safety system concept in the above referenced patent to a "large" commercial sized nuclear reactor (500 to 4,000 million watt thermal output) in the most economical fashion some specific modifications are preferred, namely the preferred embodiment of the previously referenced patent was based on the fact that only control rods (inserted by gravity into the reactor core region) were used to effect nuclear shutdown of the core. These rods were also mechanically positioned to control and/or change reactor power. In larger power reactors, boric acid dissolved in the reactor coolant water is used in conjunction with control rods to control reactor power, to compensate for fuel depletion, to compensate for increase water density when the reactor is at cold conditions, and to ensure past-accident nuclear shutdown. The combined use of boric acid solution with mechanical control rods reduces the number of control rods required, simplifies the mechanical design of the reactor, and promotes a more even level of power generation in individual fuel rods. These result in significant reductions in the initial cost of the plant and permits a higher power generation level to be achieved. SUMMARY OF THE INVENTION An object of the present invention is to provide a passive safety injection system which uses borated water and prevents boron from concentrating in the reactor vessel and possibly preventing heat transfer in the reactor core. Another object of the present invention is to improve a passive safety injection system by creating natural circulatory flow patterns. Another object of the present invention is to improve a passive safety injection system for application to larger sized nuclear reactors by providing an alternate method of supplying water at high flowrates following the postulated severance of a large reactor coolant system pipe. Another object of the present invention is to improve a passive safety injection system by improving the means to depressurize the reactor coolant system in order to be able to drain water by gravity into the reactor vessel from an elevated source of water. These and other objects of the invention are met by providing a passive safety injection system for a nuclear power plant including a containment, a reactor vessel having a core, a hot leg, a cold leg, and a borated water supply for injecting into the reactor core during a loss of coolant accident in which the containment if flooded, the system including a first flow path coupled to the hot leg below the flood up level of water in the containment, and being in communication with the containment, and a second flow path coupled to the reactor vessel and being in communication with the containment, the first flow path inducing a natural circulatory flow of water from within the containment through the reactor core based on differences in water density produced by the reactor core to thereby preventing concentration of boron in the reactor vessel. In another aspect of the invention, a passive safety injection system for a nuclear power plant includes a reactor vessel having a core, a hot leg and a cold leg coupled to the reactor vessel, a pressurizer coupled to the hot leg, a core makeup tank having a discharge line connected to the reactor vessel and a pressure balance coupled to the pressurizer, wherein the pressurizer has a normal water level and the core makeup tank is located substantially below the pressurizer normal water level, and wherein the core makeup tank contains borated water having a higher concentration of boric acid then water in the reactor coolant system, and wherein after cool down the reactor coolant system water volume shrinks and thereby automatically allows the core makeup tank to drain into the reactor vessel and increase the reactor coolant boric acid concentration. In another aspect of the present invention, a passive safety injection system for a nuclear power plant including a reactor vessel having a core, a hot leg and a cold leg, the system including a core makeup tank which is filled with borated water at a predetermined concentration of boric acid and having a drain line for draining the contents thereof into the reactor vessel and a pressure balance line connecting a top of the core makeup tank to the cold leg, a vent coupling the pressure balance line of the core makeup tank, the vent inducing a natural circulatory flow of borated water from within the core makeup tank to the reactor vessel by means of hot water rising into the pressure balance line and flowing into the core makeup tank through the vent. The foregoing and other objects and advantages of the passive safety injection system in accordance with the present invention will become more apparent from the following detailed description, taken in conjunction with the drawings. |
claims | 1. A method of providing a nuclear fuel, comprising:processing molybdenum to deplete the isotope 95-Mo; andforming an alloy of metallic uranium and the processed molybdenum, wherein the uranium is enriched in the isotope 235-U. 2. A method according to claim 1, wherein the fuel contains more than 3 grams/cm3 of uranium. 3. A method according to claim 1, wherein the fuel contains more than 4 grams/cm3 of uranium. 4. A method according to claim 1, wherein the fuel contains more than 5 grams/cm3 of uranium. 5. A method according to claim 1, wherein the fuel contains more than 7.5 grams/cm3 of uranium. 6. A method according to claim 1, wherein the depleted molybdenum contains less than 15% by weight of the molybdenum isotope 95-Mo. 7. A method according to claim 1, wherein the depleted molybdenum contains less than approximately 5% by weight of the molybdenum isotope 95-Mo. 8. A method according to claim 1, wherein processing molybdenum comprises enriching molybdenum in the isotope 92-Mo, 94-Mo, 96-Mo, 97-Mo, 98-Mo, 100-Mo, or any combination thereof. 9. A method of providing a nuclear fuel, comprising: forming an alloy of metallic uranium and molybdenum having uranium enriched in the isotope 235-U and molybdenum depleted in the isotope 95-Mo, and the content of molybdenum in the uranium-molybdenum alloy is in the range of 2–20% by weight. 10. A method according to claim 1, wherein the content of molybdenum in the uranium-molybdenum alloy is in the range of 5–10% by weight. 11. A method according to claim 1, wherein processing molybdenum comprises using ultracentrifuges. 12. A method according to claim 1, wherein the uranium-molybdenum alloy is dispersed in aluminum. 13. A method according to claim 1, wherein the enriched uranium is obtained by mixing highly enriched uranium with lowly enriched or natural uranium. 14. A method according to claim 1, wherein the enriched uranium contains 2–40% by weight of the isotope 235-U. 15. A method according to claim 1, wherein the enriched uranium contains 10–20% by weight of the isotope 235-U. 16. A fuel element including a nuclear fuel made by the method according to claim 1. 17. A method according to claim 1, wherein processing molybdenum to deplete the isotope 95-Mo comprises processing natural molybdenum. 18. A method according to claim 1, wherein processing molybdenum to deplete the isotope 95-Mo comprises processing molybdenum in a non-fission reaction. 19. A method according to claim 9, further comprising processing molybdenum to deplete the isotope 95-Mo. 20. A method according to claim 19, wherein processing molybdenum comprises processing natural molybdenum using ultracentrifuges. 21. A method according to claim 19, wherein processing molybdenum comprises enriching natural molybdenum in the isotope 92-Mo, 94-Mo, 96-Mo, 97-Mo, 98- Mo, 100-Mo, or any combination thereof. |
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claims | 1. A method of predicting the lifetime reliability of an integrated circuit device with respect to one or more defined failure mechanisms, the method comprising:breaking down the integrated circuit device into microarchitecture structures;further breaking down each structure into one or more of elements and devices, with a device comprising a sub-component of an element;determining, for each vulnerable device, the impact of a failure of the device on the functionality of the specific element associated therewith, and classifying the failure into one of a fatal failure and a non-fatal failure, wherein a fatal failure of a given device is one in which the failure causes the element employing the given device to fail;determining, for those devices whose failures are classified as fatal, one or more of an effective stress degree and an effective stress time based on one or more architecture-level events and states;determining one or more of a failure rate and a probability of fatal failure for the devices, using the one or more of the associated effective stress degree and effective stress time; andaggregating the one or more of the failure rate of the devices and the probability of fatal failures of the devices, across the structures for the one or more defined failure mechanisms. 2. The method of claim 1, wherein the structures include one or more of: register files, arrays, control logic, data paths, multiplexers, latches, repeated wires, and logic gates. 3. The method of claim 1, wherein the elements include one or more of: array/register file bitlines, array/register file wordlines, memory cells, gates of transistors, and wire repeaters. 4. The method of claim 1, wherein the devices include one or more of: metal lines, vias, PFET devices, and NFET devices. 5. The method of claim 1, wherein the failure mechanisms include one or more of: electromigration (EM), negative bias temperature instability (NBTI) and time dependent dielectric breakdown (TDDB). 6. The method of claim 5, further comprising evaluating each device to determine whether the device is vulnerable to the one or more defined failure mechanisms and eliminating from consideration those devices determined not to be vulnerable. 7. The method of claim 6, wherein a device vulnerable to electromigration comprises at least one of a metal line and a via dominated by unidirectional current flow therethrough. 8. The method of claim 6, wherein a device vulnerable to NBTI comprises a PFET device having a negative gate bias applied thereto. 9. The method of claim 6, wherein a device vulnerable to TDDB comprises one or more of: a PFET device having a logic low gate voltage and one of a logic high source and drain, and an NFET device having a logic high gate voltage and one of a logic low source and drain. 10. The method of claim 6, wherein a fatal failure of a device due to electromigration comprises a condition in which a failure of one of a metal line and a via leads to one or more of a short circuit, an open circuit, and a timing violation due to increased wire resistance. 11. The method of claim 6, wherein a fatal failure of a device due to NBTI comprises a condition in which a failure of a PFET device along a critical path leads to a timing violation. 12. The method of claim 6, wherein a fatal failure of a device due to TDDB comprises a condition in which one or more of a PFET device and an NFET device has leakage current through a gate oxide thereof exceeds a value that is able to be tolerated by logic driving the same. 13. The method of claim 6, wherein a device under stress for the electromigration failure mechanism comprises a via having current generated therethrough during one of a logical one-to-zero and a logical zero-to-one value transition of metal lines. 14. The method of claim 6, wherein a device under stress for the NBTI failure mechanism comprises a PFET device having a gate coupled to a logic low voltage and a source coupled to a logic high voltage. 15. The method of claim 6, wherein a device under stress for the TDDB failure mechanism comprises one or more of a PFET device having a gate coupled to a logic low voltage and a source coupled to a logic high voltage, and an NFET device having a gate coupled to a logic high voltage and a source coupled to a logic low voltage. 16. The method of claim 1, wherein architecture-level states comprise one or more of: a number of accesses to the device, a number of access patterns to the device, and data patterns of inputs and outputs of the device. 17. The method of claim 1, wherein architectural configuration parameters include one or more of: a number of cells in an array, a number of read ports, a number of write ports, and a number of data paths. 18. The method of claim 1, wherein defect density is calculated as the ratio of the number of fatal failures of the devices of each structure to the area of the structure. 19. The method of claim 18, wherein defect density for the electromigration failure mechanism is calculated by counting the number of vias having unidirectional current of each structure and dividing the total number of vias by the area of the structure. 20. The method of claim 18, wherein defect density for the NBTI failure mechanism is calculated by counting the number of PFETs along the critical paths of each structure and dividing the total number of PFETs by the area of the structure. 21. The method of claim 18, wherein defect density for the TDDB failure mechanism is calculated by counting the number of gate oxide breakdowns of both PFET and NFET devices of each structure, and dividing the total number of breakdowns by the area of the structure. 22. The method of claim 18, wherein the aggregating one or more of the failure rate of the devices and the probability of fatal failures of the devices is implemented by one or more of summation and weighted summation. 23. The method of claim 6, wherein the failure rates are computed in terms of a technology and environment independent failures-in-time (FIT) of a reference circuit (FORC) defined for each of the failure mechanisms. 24. The method of claim 23, wherein the failure rates are further computed in absolute values utilizing power and temperature maps along with technology and implementation parameters, by calculating the value of FORC for each component and multiplying the calculated FORC values by the technology/environment-independent values of the failure rates for each of the components. |
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048184705 | claims | 1. In a nuclear reactor system having a hold down bolt attached to a steam separator within reactor water for the prevention of radioactive exposure to testing personnel, said bolt having an inner, lower tension member with a rectangular engaging lug, and an outer, coaxial compression bolt member, an apparatus for testing the tension member at the lug, comprising: a shoe; means for attachment of a pole to the top of said shoe to permit submersed manipulation of said shoe; a piezoelectric device mounted to the bottom of said shoe and exposed upwardly to and toward said tension member of said bolt for contacting said tension member at said lug for nondestructive piezoelectric testing of said tension member from said lug at the bottom of said bolt to and toward the top of said bolt; a clamp member for reciprocal movement towards and away from said shoe member; means mounting said clamp member for reciprocal movement towards and away from said shoe for releasably clamping said lug against said shoe at said piezoelectric device for ultrasonic testing of said inner tension member of said bolt with ultrasonic sound from said piezoelectric device. maintaining said steam separator under water; moving said bolts to unlatch said bolts from brackets on said shroud adjacent said steam separator; providing a shoe having a piezoelectric device mounted to the bottom of said shoe and exposed upwardly; providing a remotely actuated clamp attached to said shoe overlying said piezoelectric device; providing a mount to said shoe for manipulating said shoe underwater in a depending relationship at the bottom end of a pole; providing a pole and attaching said pole to said shoe; manipulating said shoe to the bottom of said bolt; clamping said shoe to said bolt; and testing said bolt with said piezoelectric device. 2. The invention of claim 1 and wherein said shoe defines a concavity for receiving the bottom of said bolt. 3. The invention of claim 2 and wherein said shoe has an end wall and two side walls to define said concavity. 4. The invention of claim 1 and wherein said bolt has a shaft, and said clamp member defines an open shaft receiving aperture for sliding movement on the shaft of said bolt. 5. A process of testing hold down bolts depending from the sides of a steam separator within a nuclear reactor, said process comprising the steps of: |
description | This patent application is a U.S. National Stage of PCT/US2012/020906, filed Jan. 11, 2012, which claims priority to UK Patent Application No. 1100504.6, filed Jan. 12, 2011, each of is incorporated herein in its entirety. This invention relates to a process. In particular, it relates to a process for treating hydrogen gas liberated from the acid or alkaline dissolution of a metal. It also relates to a heating apparatus. Acid or alkaline dissolution of a metal liberates hydrogen gas. At standard temperature and pressure, hydrogen gas is a colourless, odourless, tasteless and highly combustible diatomic gas. It reacts with any oxidizing agent. Hydrogen gas reacts vigorously with oxygen to produce water in a highly exothermic reaction. It also reacts spontaneously and violently at room temperature with chlorine and fluorine to form the corresponding hydrogen halides, which are potentially dangerous acids. The highly flammable and explosive properties of hydrogen gas make it a hazardous by-product in many processes. Technetium-99m is the most widely used radiometal for medical diagnostic and therapeutic applications. Tc-99m is prepared by decay of Mo-99 in so-called Tc-99m generators. Such a generator typically comprises an aqueous solution of Mo-99 loaded onto an adsorbent (usually alumina). Following decay of the Mo-99 to Tc-99m, which has a lower affinity for the alumina, the Tc-99m may be eluted, typically using a saline solution. For the preparation of Tc-99m generators, a high purity source of Mo-99 is therefore essential. In order to obtain Mo-99 of high specific activity, it is commonly prepared by the neutron-induced fission of a U-235 target. U-235 is typically present in a target form of U-metal foil, or constructs of U and Al (e.g., a uranium-aluminium alloy). The fission reaction leads to a proportion of the U-235 being converted to Mo-99, but also leads to a number of impurities in the reactor output. Most known processes for Mo-99 production employ acid or alkaline dissolution of the irradiated target, followed by purification of the Mo-99 product. Apart from the solid and liquid impurities, which include Cs, Sr, Ru, Zr, Te, Ba, Al and alkaline and alkaline earth metals, the reaction also produces hydrogen gas. Due to the highly inflammable and explosive properties of hydrogen gas, one of the most important off-gas treatments in the Mo-99 production process is the oxidation of hydrogen gas to form water.2H2(g)+O2(g)→2H2O(l) In most known Mo-99 production processes, the oxidation of hydrogen gas is carried out in the presence of copper (II) oxide (CuO) in the following irreversible reaction:CuO+H2→Cu0+H2O This process, first developed by Sameh and Ache in 1987 (Sameh and Ache, 1987 Radiochim. Acta 41, 65), is performed in a so-called “CuO oven”, which is a fixed-bed chemical reactor. After the dissolution of the irradiated targets, the evolved hydrogen is passed over hot CuO in the CuO oven to oxidise the hydrogen to water. The water vapour is then condensed. This reaction is a typical gas/solid reaction, during which the reaction front moves through the reactor until all CuO is consumed. A typical CuO reactor bed weighs significantly more than 10 kg and runs at a temperature range of 360-400° C. (targeting a reaction temperature of 385° C.). The heating of the CuO bed is conventionally done by means of a heating plate, located beneath the CuO oven. It is not optimal, taking 24 hours to reach steady state conditions. The CuO reactor bed is projected to be completely spent after the dissolution of a certain number of targets (or number of production runs). In a larger Mo-99 production facility, e.g., which enables the processing of more than 6 targets per run, the life time of this CuO reactor is reduced to a smaller number of runs, for example 8 runs. Therefore, there is a need for the development of a reactor with a prolonged life time for the treatment of hydrogen gas. US2005/0220689 A1 discloses a method of purifying helium gas by extracting hydrogen and other impurities from a helium gas stream. The method comprises 1) passing the gas stream over a first catalytic adsorber module containing a Cu—CuO mixture, in which hydrogen and carbon monoxide are oxidised into water and carbon dioxide, respectively, and CuO is reduced to Cu, 2) passing the gas stream resulting from step 1), along with oxygen gas, into an oxidation catalyst to convert methane and/or tritium into carbon dioxide and/or water, respectively, and 3) passing the gas stream resulting from step 2), which contains excess oxygen, into a second catalytic adsorber module containing a Cu—CuO mixture, in which the oxygen gas is used to oxidise Cu into CuO. Once the CuO in the first catalytic adsorber is consumed, the order in which the first and the second catalytic adsorbers are connected in the flow path of the gas stream is switched round such that the CuO generated in the second adsorber is used for the oxidation of hydrogen and carbon monoxide, and the Cu in the first adsorber is used to remove the excess oxygen from the purified helium gas stream. However, this process relates to extraction of hydrogen from a gas stream comprising a number of other components and requires the use of two separate reactors containing Cu—CuO mixtures, which adds to the cost of the process. In addition, similar to hydrogen (i.e., 1H) gas, tritium is a highly explosive gas. The addition of oxygen to a gas stream containing tritium can potentially be hazardous. Therefore, there is a need for the development of a simple method for the treatment of hydrogen using an oxidising agent, during which the oxidising agent is regenerated so as to prolong the use thereof. In accordance with a first aspect of the present invention, there is provided a process of treating hydrogen gas liberated from the acid or alkaline dissolution of a metal, the process comprising a step of passing the liberated hydrogen gas through a reactor containing an oxidising agent for oxidation of the hydrogen gas into water, followed by a step of regenerating the oxidising agent. In a preferred embodiment, a step of regenerating the oxidising agent is carried out after each oxidation step. By carrying out a step of regenerating the oxidising agent frequently, such as after each oxidation step, the amount of the oxidising agent initially contained in the reactor can be reduced. Accordingly, the reactor can be scaled down, for example, to around 11 or 12 kg for oxidising the same amount of hydrogen gas. The reduction in the size of the reactor not only allows an easier handling charge/discharge operation but also reduces the time for the reactor to heat up or reach steady state conditions from presently 24 hours to about 3 hours, thereby reducing the cost of the process. A further advantage of the process according to the present invention is that the oxidation reaction of hydrogen gas can potentially be conducted at lower temperatures, such as at around 200° C., with regeneration also at around 200° C. This is particularly the case when a finely dispersed oxidising agent, such as the BASF catalyst materials mentioned herein, is used. Such finely dispersed systems are more active than those used in known processes. The oxidation reaction used for the regeneration of the oxidising agent is, in certain embodiments, highly exothermic. Therefore, the heat given off from the regeneration reaction can be used to heat or maintain the temperature of the reactor. This further reduces the cost of the process. In some embodiments, the oxidising agent is a metal oxide, such as copper oxide, such as copper (II) oxide, which is converted to copper metal during the process. The copper oxide, such as copper (II) oxide, is present either in a bulk form or finely dispersed on the surface of an inert support, such as in the Puristar® R3-11G and R3-17 catalysts from BASF (BASF SE, Ludwigshafen, Germany). In addition, the oxidising agent, such as copper oxide, may be diluted (either in bulk form or in a finely dispersed form) with an inert, thermally-conductive diluent, such as stainless steel pellets. The diluent helps to prevent uncontrolled heating of the bed of oxidising agent as a result of the heat given out by the exothermic reaction. A further advantage of using the finely dispersed oxidising agents, such as the BASF materials mentioned above, is that higher yields are achieved during the regeneration process than with bulk metal oxide (e.g., CuO) material. In certain embodiments, the step of regenerating the oxidising agent comprises passing a gas containing oxygen through the reactor containing the oxidising agent to be regenerated. For example, air or air in combination with nitrogen gas can be used. Since the gas containing oxygen used for the regeneration step is not mixed with the hydrogen gas, the present invention ensures that potential hazards caused by the highly explosive nature of the hydrogen gas are kept to a minimum. In some embodiments, the metal, the dissolution of which liberates the hydrogen gas, comprises uranium, optionally in combination with one or more other metals, for example a uranium-aluminium alloy. In some embodiments, the reactor containing the oxidising agent is at least partially immersed in an alumina bath. Optionally, the reactor is completely immersed in an alumina bath. The term “alumina bath” refers to a vessel containing a quantity of alumina (Al2O3), which may be used as a means for regulating the temperature of a reactor immersed therein. The bath may take the form of a substantially cylindrical or cuboidal container. The outer walls of the bath are preferably formed of metal, e.g., stainless steel or aluminium. The alumina may form a lining within the container and define a cavity into which a reactor may be placed. In a particular embodiment, the alumina bath is supplied with one or more optionally external heating elements. The relatively high thermal conductivity of alumina allows heat from the heating elements to be efficiently passed to a reactor immersed in the bath. Equally, during the highly exothermic process of the H2 conversion and reactor regeneration steps, the alumina helps to conduct heat away from the reactor, thereby preventing it from overheating. Whether or not an alumina bath is employed as described above, the reactor may also (or alternatively) be heated by means of one or more heating elements positioned in contact with the reactor. The heating elements (those in contact with the reactor, or those associated with the alumina bath) may conveniently be electrical heating elements. In accordance with a second aspect of the present invention, there is provided an apparatus for carrying out a process according to the first aspect, the apparatus comprising a reactor containing an oxidising agent for the oxidation of hydrogen gas into water, wherein the reactor is at least partially immersed in an alumina bath. In a preferred embodiment, the alumina bath is supplied with one or more heating elements. The heating elements are preferably external to the alumina bath. In some embodiments, the oxidising agent is copper oxide, such as copper (II) oxide, either in a bulk form or finely dispersed on the surface of an inert support, such as in the Puristar® R3-11G and R3-17 catalysts from BASF. The oxidising agent may be diluted (either in bulk form or in a finely dispersed form) with an inert, thermally-conductive diluent, such as stainless steel pellets. In accordance with a third aspect of the present invention, there is provided a heating apparatus comprising an alumina bath supplied with one or more optionally external heating elements, wherein the alumina bath defines a cavity into which a vessel to be heated may be placed in use. In accordance with a fourth aspect of the present invention, there is provided a process of treating hydrogen gas liberated from the acid or alkaline dissolution of a metal, the process comprising a step of passing the liberated hydrogen gas through a reactor containing an oxidising agent for oxidation of the hydrogen gas into water, the oxidising agent comprising a metal oxide finely dispersed on an inert carrier and/or diluted with an inert, thermally-conductive diluent. In a preferred embodiment, the metal oxide comprises copper oxide, such as copper (II) oxide. More preferably, the copper oxide finely dispersed on an inert carrier comprises the Puristar® R3-11G and/or R3-17 catalysts from BASF. The inert diluent may comprise stainless steel pellets. As mentioned above, the use of a finely dispersed oxidising agent, such as the specified BASF materials, gives the advantage that a lower temperature (around 200° C.) can be used for the oxidation reaction. While R3-11 is known to be usable at temperatures of 200° C. or more, R3-17 is indicated by the manufacturer for use (in different applications) at temperatures not exceeding 100° C. It has been found that R3-17 is capable of use in the processes of the present invention at around 200° C. with no deleterious effects to the material. In some embodiments, the process further comprises a step of regenerating the oxidising agent in accordance with the first aspect of the present invention. A further advantage of using the finely dispersed oxidising agents is that higher yields are achieved during the regeneration process than with bulk metal oxide (e.g., CuO) material. As shown in FIG. 1, following the dissolution of an irradiated uranium-aluminium target, the hydrogen gas liberated in the dissolver is passed into a CuO reactor, in which the hydrogen gas is oxidised into water while the CuO is converted into Cu. After each oxidation step (and before the next round of oxidation begins), a stream of air-containing nitrogen gas (N2/air) is fed into the CuO reactor to oxidise Cu in order to regenerate CuO. The CuO reactor can be heated using a heating apparatus of the present invention as shown in FIG. 2. Such a heating apparatus can improve heat transfer during the operation of the reactor. The CuO reactor can be immersed in an alumina bath fitted with one or more external heating elements (see FIG. 2). The heating elements may be present as one or more collars or jackets around the bath, or as a heating coil. The heating elements are preferably electrically heated. The alumina bath works as a heat exchanger. Firstly, the bath heats up the reactor to the desired reaction temperature (see “Start-up” in FIG. 2). Since the H2 conversion and Cu oxidation reactions are highly exothermic, however, the bath also works as a cooler during these processes (charge exhaustion and regeneration), preventing the reactor from overheating (see “Process” in FIG. 2). When the H2 conversion reaction is still taking place, but not sufficiently to heat up the reactor to its optimal working temperature, the bath resumes its heating function to keep the temperature of the reactor in the desired range (see “Start-up” in FIG. 2). As an alternative to the use of an alumina bath as shown in FIG. 2, it is possible to hear the reactor directly using one or more heated clamps or bands positioned in contact with the reactor. FIG. 3 shows such an arrangement, with three heating bands displayed for illustrative purposes. It will be appreciated that the bands may be in the form of a single helical band which runs along at least part of the length of the reactor. In the left part of FIG. 3 (H2 oxidation), H2 gas from the dissolution of a metal is introduced via a first conduit which passes the gas to the bottom of the reactor (as shown). The gas passes through the bed of CuO, and reacts therewith leading to the production of gaseous water. The gaseous water exits through a venting conduit positioned towards the top of the reactor (as shown). In the right part of the Figure (Cu oxidation), a mixture of air and nitrogen is introduced through the first conduit. The oxygen in the air reacts with the partially- or fully-spent CuO bed, so as to reoxidise the Cu present therein. The waste nitrogen gas exits the reactor through the venting conduit. A number of cycles of H2 conversion and oxidant regeneration were performed using a process according to the invention. The cycles were performed using solid aluminium ‘targets’ as the metal for dissolution. Twenty cycles were performed, with the dissolution of a total amount of Al equivalent to more than 200 U—Al targets. The experimental set-up mimicked the current process line in a Mo-99 production facility. In each cycle, the number of Al ‘targets’ dissolved was equivalent to the maximum amount of targets allowed in the production facility employed. The initial temperature of the CuO reactor was 200° C. The maximum temperature in the CuO reactor during the conversion and regeneration reactions was within limits which are considered acceptable in current processes for H2 removal. The reactor was heated directly using heating elements in the form of bands surrounding the reactor. The amount of CuO material was around 11 kg. In addition, the lower 1.5 kg part of the CuO bed was diluted 50% by weight with stainless steel pellets The average H2 conversion during the 20 cycles was >95%. During the regeneration phase of the cycles, the average CuO bed regeneration was >90%. |
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047724315 | abstract | The invention relates to a process for the immobilization of nuclear waste in a borosilicate glass.. In the process, the following are mixed simultaneously:. a silica-based gel precursor, PA0 a concentrated aqueous solution of a boron compound, and concentrated aqueous solutions of the other constituents of the final glass, i.e. a solution (solutions) of the waste to be treated and a solution of the vitrification adjuvant, PA0 with vigorous stirring, mixing taking place at between 20.degree. and 80.degree. C., preferably at 65.degree.-70.degree. C., in proportions corresponding to the desired composition of the glass, the said mixture having an acid pH, preferably a pH of between 2.5 and 3.5, and the said mixture is dried, calcined at between 300.degree. and 500.degree. C. and then melted.. The invention is applied to the treatment of solutions of nuclear waste, especially to solutions of fission products. |
claims | 1. A dodecahedron neutron spectrometer monitor, comprising: a plurality of neutron detectors; said monitor is placed in proximity to a suspected concentration of neutron radiation; each of said plurality of neutron detectors further comprising a detector means stacked on an absorbing layer, said absorbing layer, being composed of a first material that absorbs protons, and each of said absorbing layers is stacked on a separate hydrogenous surface facet of a dodecahedron assembly; said hydrogenous surface facet being composed of a second material having hydrogen atoms, said hydrogen atoms interacting with said suspected concentration of neutron radiation, converting said neutrons to recoil protons, each of said detector means detecting recoil protons passing through said absorbing layer; each of said absorbing layers having a different thickness to absorb neutron energies from 1 to 250 MeV; said dodecahedron assembly being housed concentrically in a spherical chamber; each of said detector means being coupled to a means for data processing; said data processing means providing a count of recoil protons to a means for proton distribution; and said means for proton distribution determines a proton distribution pattern to construct a neutron pattern indicating the spectrum of neutrons from said suspected concentration of neutron radiation. 2. The dodecahedron neutron spectrometer monitor, as recited in claim 1 , further comprising said first material that absorbs protons being aluminum. claim 1 3. The dodecahedron neutron spectrometer monitor, as recited in claim 2 , further comprising said second material being solid polyethylene. claim 2 4. The dodecahedron neutron spectrometer monitor, as recited in claim 3 , further comprising said dodecahedron assembly having at least 12 surface facets. claim 3 5. The dodecahedron neutron spectrometer monitor, as recited in claim 4 , further comprising said spherical chamber being composed of titanium. claim 4 6. The dodecahedron neutron spectrometer monitor, as recited in claim 5 , further comprising said spherical chamber serves as an outer shield. claim 5 7. The dodecahedron neutron spectrometer monitor, as recited in claim 6 , wherein said surface facets are pentagon-shaped. claim 6 8. The dodecahedron neutron spectrometer monitor, as recited in claim 7 , wherein each of said plurality of neutron detectors is pentagon-shaped. claim 7 9. The dodecahedron neutron spectrometer monitor, as recited in claim 8 , further comprising each of said detector means being a solid state detector. claim 8 10. The dodecahedron neutron spectrometer monitor, as recited in claim 9 , further comprising each of said detector means being a depleted n/p diode. claim 9 11. The dodecahedron neutron spectrometer monitor, as recited in claim 10 , further comprising said dodecahedron assembly having 12 surface facets. claim 10 12. The dodecahedron neutron spectrometer monitor, as recited in claim 11 , further comprising said monitor being employed in an aircraft. claim 11 |
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abstract | An air distribution system for supplying filtered air to isolator working volumes includes an inlet including a HEPA filter and an outlet including a slidably mounted sintered panel. Methods for supplying filtered air to an isolator working volume are also disclosed. |
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claims | 1. A nuclear reactor instrumentation system for monitoring a nuclear power system having a reactor core within a reactor vessel, the reactor vessel having a plurality of sensors for monitoring parameters of the nuclear power system, comprising:a computer having a processor configured to be powered by a normal power source and a backup power source;a one-way wireless transmitter operable under the control of the processor and housed in a reactor building with the nuclear power system;one or more remote monitoring computers positioned external to the reactor building and communicably coupled to the computer through the one-way wireless transmitter;at least one sensor communicably coupled to the computer and operable to switch power to the computer from the normal power source to the backup power source upon a loss of normal power form the normal power source; andat least one memory coupled to the processor and containing stored programming instructions which, when executed by the processor, cause the processor to:receive data from the sensors;identify the loss of normal power from the normal power source; andin response to identifying the loss of normal power, cause the one-way wireless transmitter to transmit an encrypted version of the received data from the computer to the one or more remote monitoring computers in a single direction wireless transmission from the one-way wireless transmitter to the one or more remote monitoring computers. 2. The nuclear reactor instrumentation system of claim 1, wherein the stored programming instructions further cause the processor to switch operation of the computer from the normal power source to the backup power source. 3. The nuclear reactor instrumentation system of claim 2, wherein the identified loss of normal power comprises a loss of power to a post-accident monitoring system. 4. The nuclear reactor instrumentation system of claim 2, wherein the identified loss of normal power comprises a loss of power to a nuclear power system control room. 5. The nuclear reactor instrumentation system of claim 2, wherein the identified loss of normal power comprises a loss of power to one or more nuclear power system monitoring panels. 6. The nuclear reactor instrumentation system of claim 1, wherein the one or more remote monitoring computers are configured to receive the encrypted version of the data transmitted from the one-way wireless transmitter and decrypt the encrypted version of the data. 7. The nuclear reactor instrumentation system of claim 1, wherein the one or more sensors comprises one or more of a valve position indicator, a temperature gauge, and a pressure gauge. 8. The nuclear reactor instrumentation system of claim 1, wherein the reactor building is positioned in a protected area that encompasses the nuclear power system and the nuclear reactor instrumentation system. 9. The nuclear reactor instrumentation system of claim 8, wherein the one or more remote monitoring computers are positioned outside of the protected area. 10. The nuclear reactor instrumentation system of claim 1, wherein the at least one sensor comprises an electro-mechanical switch. 11. The nuclear reactor instrumentation system of claim 1, wherein the backup power source comprises a qualified battery system of VLA batteries. 12. The nuclear reactor instrumentation system of claim 11, wherein the backup power source is electrically coupled to an external power source that comprises a generator. 13. The nuclear reactor instrumentation system of claim 2, wherein the identified loss of normal power comprises a loss of power to a post-accident monitoring system, a loss of power to a nuclear power system control room, and a loss of power to one or more nuclear power system monitoring panels. 14. The nuclear reactor instrumentation system of claim 13, wherein the one or more remote monitoring computers are configured to receive the encrypted version of the data transmitted from the one-way wireless transmitter and decrypt the encrypted version of the data. 15. The nuclear reactor instrumentation system of claim 14, wherein the one or more sensors comprises one or more of a valve position indicator, a temperature gauge, and a pressure gauge. 16. The nuclear reactor instrumentation system of claim 15, wherein the reactor building is positioned in a protected area that encompasses the nuclear power system and the nuclear reactor instrumentation system, and the one or more remote monitoring computers are positioned outside of the protected area. 17. The nuclear reactor instrumentation system of claim 16, wherein the at least one sensor comprises an electro-mechanical switch. 18. The nuclear reactor instrumentation system of claim 17, wherein the backup power source comprises a qualified battery system of VLA batteries, and the backup power source is electrically coupled to an external power source that comprises a generator. 19. The nuclear reactor instrumentation system of claim 1, wherein the stored programming instructions further cause the processor to request authentication credentials from at least one of the one or more remote monitoring computers. 20. The nuclear reactor instrumentation system of claim 1, wherein the stored programming instructions further cause the processor to provide power from the backup power source to the plurality of sensors for monitoring parameters of the nuclear power system upon the loss of normal power. |
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description | 1. Field of the Invention The present invention relates to an X-ray waveguide, in particular, an X-ray waveguide used in an X-ray optical system, for example, in an X-ray analysis technology, an X-ray imaging technology, or an X-ray exposure technology. 2. Description of the Related Art When an electromagnetic wave having a short wavelength of several tens of nanometers or less is dealt with, a difference in refractive index for any such electromagnetic wave between different materials is extremely small, specifically, 10−4 or less, and thus, for example, a critical angle for total reflection becomes extremely smaller. In view of the foregoing, a large-scale spatial optical system is usually used for controlling such an electromagnetic wave including an X-ray. Among the main components of which the spatial optical system is formed, multilayer mirrors obtained by alternately laminating materials having different refractive indices are playing various roles such as beam shaping, spot size conversion, and wavelength selection. A conventional X-ray waveguide such as a polycapillary propagates, in contrast to such mainstream spatial optical system, an X-ray by confining the X-ray in itself. Researches have been recently conducted on X-ray waveguides propagating X-ray by confining the X-ray in a thin film or a multilayer film, with a view to reducing the size and improving the performance of an optical system. Specifically, researches have been conducted on, for example, thin-film waveguides of such a constitution that a waveguiding layer is interposed between two layers of one-dimensional periodic structures (see Physical Review B, Volume 67, Issue 23, p. 233303 (2003)), and an X-ray fiber of such a constitution that a vacuum core is surrounded with a cladding obtained by alternately laminating a heavy element and a light element (see Japanese Patent Application Laid-Open No. H03-146909). In Japanese Patent Application Laid-Open No. H03-146909, however, it is difficult to produce a core region having such a small diameter that an independent waveguide mode of an X-ray is formed or to uniformly form a laminated film of a cladding on the outer periphery of the fiber. In addition, since an inorganic material that absorbs the X-ray to a large extent is used in the cladding, the propagation loss of the X-ray increases. Further, the oxidation of the inorganic material may cause the deterioration or structural change of the waveguide. Further, Physical Review B, Volume 67, Issue 23, p. 233303 (2003), which is an investigation based on calculation, adopts such a model that Ni and C are used as components of a multilayer film serving as the cladding. Accordingly, the propagation loss of the X-ray due to Ni is large. In addition, a labor and time is required for the step of laminating those materials to show sufficient reflectance. The present invention has been made in view of such conventional problems as described above, and an object of the present invention is to provide an X-ray waveguide which: shows a small propagation loss of an X-ray; does not deteriorate owing to oxidation; and can be easily produced. In one aspect of the present invention, an X-ray waveguide is provided, which includes: a core for guiding an X-ray in such a wavelength band that the real part of refractive index of materials is 1 or less; and the cladding for confining the X-ray in the core, in which: the cladding has a one-dimensional periodic structure consisting of at least two different materials having different real parts of refractive index; one of the materials is inorganic one, and another one of materials is any of an organic material, a gas, or vacuum. material; and the core and the cladding are formed so that a critical angle for total reflection at the interface between the core and the cladding is smaller than a Bragg angle depending on the periodicity of the one-dimensional periodic structure. Further features of the present invention will become apparent from the following description of exemplary embodiments with reference to the attached drawings. Hereinafter, the present invention is described in detail. The term “X-ray” as used in the present invention refers to electromagnetic waves in such a wavelength band that the refractive index real part of a material is 1 or less. Specifically, the term “X-ray” as used in the present invention refers to electromagnetic waves, each of which having a wavelength of 100 nm or less including extreme ultraviolet light (EUV light). Regarding the electromagnetic wave having such short wavelength, the following fact has been known. Since the electromagnetic wave has so high a frequency that an electron in the outermost shell of a material cannot respond to the frequency, the real part of the refractive index of the material for an X-ray is smaller than 1 unlike the frequency band of an electromagnetic wave (visible light or infrared light) having a wavelength longer than that of ultraviolet light. As represented in the following formula (1), such refractive index n of a material for an X-ray is generally represented by using a decrement δ of a real part from 1 and an imaginary part β′ related to absorption.[Math. 1]n=1−δ−1β′=n′−1β′ (1) Since the δ is proportional to an electron density ρe of the material, the real part of the refractive index reduces as the electron density of the material increases. In addition, the refractive index real part n′ is 1−δ. Further, the ρe is proportional to an atomic density ρa and an atomic number Z. As described above, the refractive index of a material for an X-ray is represented in terms of a complex number. In the specification, the real part of the complex number is referred to as a “refractive index real part” or a “real part of the refractive index,” and the imaginary part of the complex number is referred to as a “refractive index imaginary part” or an “imaginary part of the refractive index.” A case where a refractive index real part for an X-ray becomes maximum is the case where the X-ray propagates in a vacuum. Under a general environment, however, the refractive index real parts of air for nearly all materials except gases become maximum. In the specification, the term “material” is applied to a vacuum as well. In the present invention, at least two materials having different refractive index real parts can be interpreted as two or more kinds of materials having different electron densities in many cases. The minimum unit structure that forms a periodic structure is referred to as a “structure element” in the specification. The present invention is characterized in that at least one kind of material out of all materials of which a one-dimensional periodic structure that forms a cladding is formed is an organic material, and at least another one kind of material out of all materials is a continuous inorganic material. The propagation loss of an X-ray due to absorption can be reduced because one kind of the materials of which the one-dimensional periodic structure is formed is an organic material that absorbs the X-ray to a small extent. In addition, the term “continuous inorganic material” refers to a material in which inorganic elements such as Si and Ti are bonded in a film fashion by a covalent bond through O or the like such as SiO2, TiO2, or SnO2, or a material in which atoms are bonded in a film fashion by a metal bond such as Au or Pt. Such continuous inorganic material can improve the strength of the one-dimensional periodic structure itself. Further, the use of an oxide such as SiO2 or TiO2 as the inorganic material can obviate the deterioration or structural change of an X-ray waveguide of the present invention due to oxidation, thereby resulting in improved durability. A one-dimensional periodic structure of such a constitution that the inorganic material and the organic material are alternately laminated in a one-dimensional direction is a multilayer film in the present invention. When the one-dimensional periodic structure is a multilayer film, while the inorganic material in the structure elements that form the periodic structure is continuously formed in the in-plane direction of the multilayer film, it is not needed to be continuous between the respective structure elements. A method of laminating the inorganic material that is an oxide is, for example, vapor deposition or a sputtering method. Further, the multilayer film formed of such material is, for example, a lamellar film as a meso structured film of a lamellar structure produced by employing a sol-gel process. The term “lamellar film” as used herein refers to a mesostructured film having a lamellar structure. The term “mesostructured film” refers to an organic-inorganic hybrid material film formed by the self-assembly of a surfactant. Mesostructured films having various meso-scale period of the periodic structureicities are available, and among those, a mesostructured film of such a lamellar structure that sheets (thin films) of an organic material and an inorganic material are laminated is suitably used in the present invention. Representative examples of the inorganic material of such mesostructured film include oxides such as SiO2, TiO2, SnO2, and ZrO2. Such mesostructured film of the lamellar structure can be formed on a substrate by an approach such as the sol-gel process. The period of the periodic structure of the mesostructured film of the lamellar structure can be appropriately adjusted to a desired value depending on, for example, the kind and concentration of the surfactant to be used, and reaction conditions. Since the mesostructured film of the lamellar structure forms the one-dimensional periodic structure in a self-organizing fashion in one step, the time and labor of a production step for the film can be markedly curtailed. In the present invention, the cladding is preferably formed of a multilayer film formed of a mesostructured film of the lamellar structure serving as a meso structure. In addition, the number of periods of the multilayer film obtained by periodically laminating the films is preferably 20 or more. In the present invention, the mesostructured film is a film of periodic structure having a period of the periodic structure of 2 nm or more and 50 nm or less. The lamellar structure is a layered structure formed of two different kinds of materials, and the two kinds of materials are formed of an inorganic material mainly formed of an inorganic component and an organic material mainly formed of an organic component. The inorganic material mainly formed of the inorganic component and the organic material mainly formed of the organic component may be bonded to each other as required. A product in which the materials are bonded to each other is specifically, for example, a mesostructure prepared from a siloxane compound to which an alkyl group is bonded. A material for the inorganic material mainly formed of the inorganic component, which is not particularly limited, is, for example, an inorganic oxide from the viewpoint of produceability and such a viewpoint that the periodic structure is formed of materials having different refractive index real parts. Examples of the inorganic oxide include silicon oxide, tin oxide, zirconia oxide, titanium oxide, niobium oxide, tantalum oxide, aluminum oxide, tungsten oxide, hafnium oxide, and zinc oxide. The surface of a wall part may be modified as necessary. For example, the surface of the wall part may be modified with a hydrophobic molecule for inhibiting the adsorption of water. The organic material mainly formed of the organic component, which is not particularly limited, is, for example, a surfactant or a material in which a site having a function of forming a molecular assembly is bonded to a material of which a wall part is formed or a precursor for the material of which a wall part is formed. Examples of the surfactant include ionic and nonionic surfactants. The ionic surfactant is, for example, a halide salt of a trimethylalkylammonium ion. The chain length of the alkyl chain is, for example, 10 to 22 in terms of a carbon number. Examples of the nonionic surfactant include surfactants, each of which is containing polyethylene glycol as a hydrophilic group. Specific examples of the surfactants, each of which is containing polyethylene glycol as a hydrophilic group include a polyethylene glycol alkyl ether and a polyethylene glycol-polypropylene glycol-polyethylene glycol block copolymer. The chain length of the alkyl chain of the polyethylene glycol alkyl ether is, for example, 10 to 22 in terms of a carbon number, and the number of repetitions of the polyethylene glycol is, for example, 2 to 50. The period of the periodic structure can be changed by changing the hydrophobic group or hydrophilic group. In general, the period of the periodic structure can be extended by making a hydrophobic group or hydrophilic group large. The material mainly formed of the organic component may contain water, an organic solvent, a salt, or the like as required, or as a result of a material to be used or a step. Examples of the organic solvent include an alcohol, ether, and a hydrocarbon. Next, a method of producing the mesostructured film is described. Although the method of producing the mesostructured film is not particularly limited, the film is produced by, for example, adding a precursor for an inorganic oxide to a solution of an amphipathic material (especially a surfactant) that functions as an assembly to perform film formation so that a reaction for producing the inorganic oxide may be advanced. In addition, an additive for adjusting a period of the periodic structure as well as the surfactant may be added. The additive for adjusting a period of the periodic structure is, for example, a hydrophobic material. Examples of the hydrophobic material include alkanes and aromatic compounds free of hydrophilic groups. The hydrophobic material is specifically, for example, octane. Examples of the precursor for the inorganic oxide include an alkoxide and a chloride of silicon or a metal element. More specific examples thereof include an alkoxide and a chloride of Si, Sn, Zr, Ti, Nb, Ta, Al, W, Hf, or Zn. Examples of the alkoxide include a methoxide, an ethoxide, a propoxide. The alkoxides may partly be substituted with an alkyl group. Examples of the film-forming method include a dip coating method, a spin coating method, and a hydrothermal synthesis method. Further, special examples of the one-dimensional periodic structure in the present invention include a mesoporous film and a mesoporous film whose pores are filled with an organic material and the like. Each of those mesoporous films is such that pores or pores filled with the organic material are placed in an inorganic material in a two- or three-dimensional direction. However, any such film serves as the one-dimensional periodic structure when attention is paid to a refractive index or density at which an average refractive index has a periodic distribution in a one-dimensional direction depending on a material out of which, and a condition under which, the film is produced. In particular, when the inside of each pore is filled with a gas such as air or is evacuated to a vacuum, a difference in refractive index between the materials of which any such film is formed can be increased. In addition, the propagation loss of an X-ray can be reduced. Hereinafter, such mesoporous films are described. (A) Mesoporous Film whose Pores are Hollow The mesoporous film is a porous material having a pore diameter of 2 to 50 nm, and a material for a wall part, which is not particularly limited, is, for example, an inorganic oxide in terms of produceability. Examples of the inorganic oxide include silicon oxide, tin oxide, zirconia oxide, titanium oxide, niobium oxide, tantalum oxide, aluminum oxide, tungsten oxide, hafnium oxide, and zinc oxide. The surface of the wall part may be modified as necessary. For example, the surface may be modified with a hydrophobic molecule for inhibiting adsorption of water. Although a method of preparing the mesoporous film is not particularly limited, the film can be prepared by, for example, the following method. A precursor for the inorganic oxide is added to a solution of an amphipathic material whose assembly functions as a template to perform film formation so that a reaction for producing the inorganic oxide may be advanced. After that, template molecules are removed so that the porous material may be obtained. The amphipathic material, which is not particularly limited, is suitably a surfactant. Examples of the surfactant molecule include ionic and nonionic surfactants. The ionic surfactant is, for example, a halide salt of a trimethylalkylammonium ion. The chain length of the alkyl chain is, for example, 10 to 22 in terms of a carbon number. Examples of the nonionic surfactant include surfactants, each of which is containing polyethylene glycol as a hydrophilic group. Specific examples of the surfactants, each of which is containing polyethylene glycol as a hydrophilic group include a polyethylene glycol alkyl ether and a polyethylene glycol-polypropylene glycol-polyethylene glycol block copolymer. The chain length of the alkyl chain of the polyethylene glycol alkyl ether is, for example, 10 to 22 in terms of a carbon number, and the number of repetitions of the polyethylene glycol is, for example, 2 to 50. The period of the periodic structure can be changed by changing the hydrophobic group or hydrophilic group. In general, a pore diameter can be extended by making a hydrophobic group or hydrophilic group large. In addition, an additive for adjusting a period of the periodic structure as well as the surfactant may be added. The additive for adjusting a period of the periodic structure is, for example, a hydrophobic material. Examples of the hydrophobic material include alkanes and aromatic compounds free of hydrophilic groups. The hydrophobic material is specifically, for example, octane. Examples of the precursor for the inorganic oxide include an alkoxide and a chloride of silicon or a metal element. More specific examples thereof include an alkoxide and a chloride of Si, Sn, Zr, Ti, Nb, Ta, Al, W, Hf, or Zn. Examples of the alkoxide include a methoxide, an ethoxide, a propoxide. The alkoxides may partly be substituted with an alkyl group. Examples of the film-forming method include a dip coating method, a spin coating method, and a hydrothermal synthesis method. Examples of the method of removing the template molecules include calcination, extraction, ultraviolet irradiation, and ozonation. (B) Mesoporous Film whose Pores are Mainly Filled with Organic Compound Any one of the same materials as those described in the section (A) can be used as a material for a wall part. The material with which each pore is filled is not particularly limited as long as the material is mainly formed of an organic compound. The term “mainly” means that a volume ratio of the organic compound to the material is 50% or more. The organic compound is, for example, a surfactant or a material in which a site having a function of forming a molecular assembly is bonded to the material of which a wall part is formed or a precursor for the material of which a wall part is formed. Examples of the surfactant include the surfactants described in the section (A). In addition, examples of the material in which the site having a function of forming a molecular assembly is bonded to the material of which a wall part is formed or the precursor for the material of which a wall part is formed include an alkoxysilane having an alkyl group and an oligosiloxane compound having an alkyl group. The chain length of the alkyl chain is, for example, 10 to 22 in terms of a carbon number. The inside of each pore may contain water, an organic solvent, a salt, or the like as required, or as a result of a material to be used or a step. Examples of the organic solvent include an alcohol, ether, and a hydrocarbon. A method of preparing the mesoporous film whose pores are mainly filled with the organic compound, which is not particularly limited, is, for example, a step before the template removal of the method of preparing the mesoporous film described in the section (A). A material of which a core is formed is desirably air, an organic material such as a polymethyl methacrylate resin (PMMA) or polydimethylsiloxane (PDMS), or a material having a light electron density, in other words, a large refractive index real part and a low absorption loss, such as carbon (C) or boron (B). In addition, the core is not limited to one formed of one kind of uniform material, and even a material whose core itself has a periodic structure such as a multilayer film can be used to form the X-ray waveguide of the present invention. Since the behavior of the X-ray basically follows Maxwell's equations, the reflection and refraction of an X-ray occur at an interface between two materials having different refractive index real parts. As long as the propagation angle of the X-ray at the interface between the two materials having different refractive index real parts measured from the interface is larger than a critical angle for total reflection at the interface, the reflection and refraction repeatedly occur at an interface between the respective layers even when a periodic structure is formed of these materials. As a result, multiple interference occurs in the periodic structure. In particular, in the present invention, Bragg reflection occurs as a result of multiple interference in the cladding. An X-ray is confined in the core by the Bragg reflection. When a relationship represented by the following formula (2) is satisfied between a critical angle for total reflection θc at an interface between the cladding and the core, and a Bragg angle θB of the periodic structure, an X-ray that propagates at an angle around the Bragg angle undergoes Bragg reflection as a result of multiple interference in the periodic structure, and hence the X-ray can be confined in the core.[Math. 2]θC<θB (2) In the formula, θc represents an angle measured from an interface between a material having a relatively large refractive index real part (a relatively low electron density) and a material having a relatively small refractive index real part (a relatively high electron density) upon incidence of an X-ray from the former material on the interface. The Bragg angle θB of the one-dimensional periodic structure is a Bragg angle depending on the periodicity of the multilayer film. In addition, the Bragg angle θB of the one-dimensional periodic structure, i.e., the multilayer film is roughly represented by the following formula (3) when the period of the one-dimensional periodic structure as the cladding is represented by d and the average refractive index real part of the one-dimensional periodic structure as the cladding is represented by navg. [ Math . 3 ] θ B ≈ 180 π arcsin ( 1 n avg m λ 2 d ) ( 3 ) In the formula, m represents a natural number, d represents the period of the periodic structure of the one-dimensional periodic structure, and λ represents the wavelength of an X-ray. The Bragg reflection holds completely true when the periodic structure infinitely continues. Accordingly, the left and right sides of the formula (3) are not connected to each other with a complete equal sign, and as the number of periods of the one-dimensional periodic structure as an actual periodic structure increases, the angle for high reflectance caused by the multiple interference in the one-dimensional periodic structure approaches the right side of the formula (3). The number of periods of the one-dimensional periodic structure in the X-ray waveguide of the present invention is set to about 20 periods or more. In the specification, reflection caused by the multiple interference of an X-ray in such realistic periodic structure that does not infinitely continue is also referred to as “Bragg reflection,” and the angle at which the X-ray is reflected from a direction parallel to a film interface is referred to as a “Bragg angle.” And it's preferable that the Bragg angle should be determined by the X-ray diffraction measurement or the like in reality. As a result, the Bragg reflection, and since the one-dimensional periodic structure of the present invention is regarded as a one-dimensional photonic crystal, a photonic band gap (stop band) effect exists. In the case of the one-dimensional periodic structure, a lowest-order photonic band gap is identical to first-order (m=1 in math [3]) Bragg reflection caused by the periodicity of the periodic structure in the direction in which the periodicity exists. In other words, reflection caused by the lowest-order photonic band gap is the first-order Bragg reflection. For example, FIG. 3 is a schematic view illustrating an X-ray 304 having a specific single wavelength, and the X-ray is incident upon a cladding 301 as the one-dimensional periodic structure of which the X-ray waveguide of the present invention is formed at an incidence angle θ(°) and then reflected. The incidence angle is an angle measured from a direction parallel to each interface of the one-dimensional periodic structure. The cladding 301 as the one-dimensional periodic structure is a one-dimensional periodic structure formed as a multilayer film by alternately laminating 50 layers each of an organic material 303 and an inorganic material 302. FIG. 1 illustrates the incidence angle dependence of reflectance of X-ray base on the ratio of the intensity of the reflected X-ray 305 to that of the incident X-ray 304. In FIG. 1, a angle range for total reflection 101, a critical angle for total reflection 102, and a peak 103 are shown. In FIG. 1, θC represents a critical angle for total reflection at an interface between the cladding and air. When an X-ray is incident at an incidence angle equal to or smaller than the θC, a high reflectance 101 is obtained by virtue of total reflection. When the incidence angle is equal to or larger than the θC, the X-ray undergoes nearly no reflection. However, a high reflectance is obtained at the Bragg angle θB, based on the periodicity of the cladding, and appears as the peak 103 on the graph. The peak 103 has a width in terms of an angle, and the width corresponds to a photonic band gap in angle presented by the periodic structure that forms the cladding. The cladding of the X-ray waveguide of the present invention is formed of such one-dimensional periodic structure as described above, and an X-ray is confined in the core by the cladding. As illustrated in FIG. 2, in the case of a constitution in which a core is interposed between two claddings, each of which is serving as such one-dimensional periodic structure, an X-ray having a specific wavelength introduced into the core causes multiple reflection between both claddings. As a result, the X-ray undergoes Bragg reflection (is reflected by a photonic band gap) to form a waveguide mode in the core. The propagation constant of the waveguide mode depends on the periodicity of the one-dimensional periodic structure as each cladding. The phase of the waveguide mode is also affected by the periodicity of the one-dimensional periodic structure, and is hence matched in the direction perpendicular to a guiding direction of X-ray of waveguide mode and having a high periodicity. The phrase “phase of the waveguide mode is matched” as used in the present invention refers not only to that a phase difference of the electromagnetic field of waveguide mode over the area of core in a plane perpendicular to the guiding direction of X-ray of waveguide mode is zero but also to that the phase difference of the electromagnetic field of the waveguide mode periodically changes between −n and +n over the area in correspondence with the spatial refractive index distribution of the periodic structure. When the Bragg angle θB of the one-dimensional periodic structure is larger than the critical angle for total reflection θC at the interface between the cladding and the core, only a waveguide mode resulting only from the vicinity of the Bragg angle of the one-dimensional periodic structure (angle corresponding to the photonic band gap) can be independently formed in the core as illustrated in FIG. 1. In contrast, when the Bragg angle of the one-dimensional periodic structure is smaller than the critical angle for total reflection at the interface between the cladding and the core, multiple waveguide modes resulting from a wide angle range equal to or smaller than the critical angle for total reflection at the interface between the cladding and the core are formed. Accordingly, the waveguide mode resulting from the Bragg angle is hardly discriminated from those waveguide modes. The critical angle for total reflection at the interface between the core and the cladding in the present invention is smaller than the Bragg angle of the one-dimensional periodic structure, provided that the Bragg angle θB exists on the basis of a structure parameter and a physical property parameter in correspondence with the positive integer m as represented by the formula (3). In the case of the constitution of the X-ray waveguide of the present invention, a Bragg angle that contributes to an actual waveguide mode is determined in relation to the physical properties and thickness of the core, and furthermore, to the structures and characteristics of the cladding and substrate. Accordingly, the effective propagation angle of the lowest-order waveguide mode confined in the core does not necessarily coincide with the vicinity of the Bragg angle for m=1. Therefore, in the present invention, a Bragg angle under such a condition that the critical angle for total reflection at the interface between the core and the cladding is smaller than the Bragg angle affects the effective propagation angle of the lowest-order waveguide mode confined in the core. FIG. 2 is a view illustrating Example 1 of the X-ray waveguide of the present invention. A lamellar film 202 formed as a multilayer film to serve as a cladding is formed on an Si substrate 204. Then, PDMS having a thickness of about 32 nm is formed as a core 201 on the film by spin coating. Further, a lamellar film 203 is formed on the core. The inorganic material of the lamellar films is SiO2. The lamellar films 202 and 203 of this example are produced by the following method. (a) Preparation of Solution A mesostructured film having a lamellar structure is prepared by a spin coating method. A precursor solution is prepared by dissolving n-decyltrimethoxysilane, tetramethoxysilane, water, and hydrochloric acid in a tetrahydrofuran solvent, and stirring the resultant at 25° C. for 3 hours. The mixing ratio (molar ratio) of n-decyltrimethoxysilane, tetramethoxysilane, water, hydrochloric acid, and tetrahydrofuran is set to 1:4:19:0.01:20. (b) Film Formation After the substrate has been washed, coating is performed with a spin coating apparatus under the conditions of 3,000 rpm and 10 seconds. At this time, a temperature is 25° C. and a relative humidity is 40%. After having been formed, a film is held in a thermo-hygrostat at 25° C. and a relative humidity of 50% for 4 weeks. (c) Evaluation The mesostructured film is subjected to X-ray diffraction analysis in a Bragg-Brentano geometry. As a result, it is confirmed that the mesostructured film has high order in the normal direction of the substrate surface and its plane spacing is 3.56 nm. The X-ray waveguide of this example is of such a structure that an X-ray is confined in the core 201 by Bragg reflection, in other words, a photonic band gap effect exerted as a result of the multiple interference of the X-ray in the lamellar films. The direction in which the X-ray propagates is a z direction in the figure. FIG. 5 is a graph illustrating the intensity distribution of electric field of the lowest-order waveguide mode present in the X-ray waveguide in this example. The energy of the X-ray by calculation is 12.4 keV. Here, an effective propagation angle θ′ (°) in the specification is an angle from a direction parallel to a film surface, and is represented by the following formula (4) with a wave vector (propagation constant) kz in the propagation direction of a waveguide mode and a wave vector k0 in a vacuum of the waveguide mode. [ Math . 4 ] θ ~ = 180 π arccos ( k z k 0 ) ( 4 ) Since the kz is constant at an interface between the respective layers by virtue of a condition of continuity, as illustrated in FIG. 6, the effective propagation angle θ′ (°) represents an angle defined between the propagation constant kz of the fundamental wave of the waveguide mode and the wave vector k0 in a vacuum of the fundamental wave at which the fundamental wave of the waveguide mode propagates in a vacuum. The angle can be considered to approximately represent the angle at which the fundamental wave of the waveguide mode propagates in the core. In the graph of FIG. 5, a region 501 corresponds to the core in the X-ray waveguide of this example, and regions 502 and 503 correspond to the claddings in the X-ray waveguide of this example. The effective propagation angle of the waveguide mode corresponds to the vicinity of a Bragg angle of about 0.82°. Only a waveguide mode resulting from the Bragg angle can exist by being confined in an additionally wide angle region, and hence no other waveguide mode can be present in a specific angle range. As a result, the waveguide mode is an independent waveguide mode in a wide angle range, and can propagate an X-ray without mixing with any other mode. In addition, the phase of the waveguide mode is basically matched because the phase is affected by the Bragg reflection at each lamellar film. In addition, as can be seen from the graph, the waveguide mode is strongly confined in the region of the core 501. FIG. 4 is a view illustrating Example 2 of the X-ray waveguide of the present invention. The claddings of this example are the films of a mesoporous silica (mesoporous material) whose inorganic material is silica (SiO2), and the inside of each pore of the film is filled with air. The X-ray of waveguide mode propagates in a z direction in the figure. The X-ray waveguide of this example is such that the film of the mesoporous silica is formed as a cladding 402 or 403. However, a core 401 is formed of air. The mesoporous silica film is of a structure having a one-dimensional periodicity in a y direction in the figure, and the period of the structure is about 6 nm. A lamellar film is formed on each of Si substrates 404 and 405, and Au pattern of about 50 nm height is made on the surface of each Si substrates by sputtering, lithography, and etching process. After that, both the Si substrates are attached to each other so that the respective mesoporous silica films may be opposite to each other. The presence of the Au film 406 provides a 50-nm air gap between the two claddings of mesoporous silica film, which is the core 401. The Au part of this example exists only for forming the air gap serving as the core. Formed in the X-ray waveguide of this example is a waveguide mode having an effective propagation angle that coincides with the vicinity of a Bragg angle of about 0.47° of Bragg reflection resulting from the periodicity of each mesoporous silica film. The X-ray waveguide of the present invention can be utilized in the field of X-ray optics technologies such as X-ray optical systems for controlling X-ray from, for example, a synchrotron, or a part for use in an X-ray imaging technologies, an X-ray exposure technologies, or the like. While the present invention has been described with reference to exemplary embodiments, it is to be understood that the invention is not limited to the disclosed exemplary embodiments. The scope of the following claims is to be accorded the broadest interpretation so as to encompass all such modifications and equivalent structures and functions. This application claims the benefit of Japanese Patent Applications No. 2010-127338, filed Jun. 2, 2010, and No. 2011-101307, filed Apr. 28, 2011, which are hereby incorporated by reference herein in their entirety. |
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abstract | A shield structure configured to protect a head and/or neck of a patient during a radiologic procedure comprises a bottom wall, a side wall, and an opening. The bottom wall includes radiation attenuating material and is configured to be positioned between the head and/or neck of the patient and a radiation source so as to shield the patient from radiation directed toward the bottom of the patient. The bottom wall is of a general size to shield the head and/or neck of the patient. The side wall includes radiation attenuating material and is configured to extend upward from the bottom wall so as to shield the patient from radiation directed toward a side of the patient. The opening is configured to receive the head and/or neck of the patient. |
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061817627 | summary | TECHNICAL FIELD The present invention relates to a nuclear fuel bundle for a nuclear reactor and particularly relates to a nuclear fuel bundle for a boiling water reactor having fuel rods with different peak power limits dependent upon lattice location in the bundle. BACKGROUND OF THE INVENTION Fuel bundles for nuclear reactors typically include a plurality of nuclear fuel rods extending in generally parallel relation one to the other and arranged in a rectilinear matrix of fuel rods, e.g., 8.times.8, 9.times.9, 10.times.10 arrays, with peripheral or edge fuel rods surrounding interior fuel rods, as well as one or more interior water rods. Characteristic operating conditions of a BWR fuel bundle lattice are very heterogeneous. For example, large quantities of non-voided water (moderator) lie between the fuel rods adjacent lower portions of the bundle, while boiling water (voided) lies adjacent the upper end of the bundle. This results in reduced average water density. The power in any fuel rod is proportional to the low energy neutron density within the fuel rod. When the neutrons are liberated in the fission process, the neutron energy is transferred to the water within the reactor core through elastic and inelastic collisions, resulting in a shift in the neutron energy spectrum towards the low end, with the highest population of low energy neutrons existing near regions of high water density. It has been observed that in regions of the reactor that contain high water density and exhibit large thermal neutron densities, the fuel rods near those regions exhibit relatively high powers. Particularly, it has been observed that fuel rods about the periphery or edge of the fuel bundle typically operate at powers that are substantially, e.g., on the order of 20%, higher than the majority of the interior rods. Interior fuel rods adjacent one or more water rods also exhibit somewhat elevated powers. Peak power limit for each fuel rod in a nuclear fuel bundle is defined as the maximum power limit at which each rod may operate, i.e., a maximum power output per unit length of fuel rod during steady state operation. Peak power limit is evidenced by a thermomechanical curve that basically identifies the maximum peak power output at which each rod can operate as a function of time. This limiting curve is the same for all fuel rods in the lattice and all fuel rod positions independent of the number of pellets within the fuel rod, their enrichment, column length, fission gas plenum volume and the like. All rods within the fuel bundle, regardless of type, e.g., fuel rods only, rods having a mixture of fuel with poisons such as gadolinium or part-length fuel rods, must operate below the peak power limit. Because the natural power peaking in a BWR fuel bundle is dependent upon the position of the fuel rod relative to the neutron moderator, i.e., water, it is common to have a subset of fuel rods that dictate the maximum power that is achievable in the fuel bundle. Thus, the fuel rods adjacent the periphery or edge of the fuel bundle typically define the maximum power peaking in the BWR fuel bundle. Stated differently, the interior fuel rods typically operate with a greater margin relative to the peak power limit than do the peripheral or edge rods and, in essence, are under-utilized. To offset that, fissile enrichment in these high-power peripheral fuel rods is often depressed, hence increasing the operating margins for the rods while disadvantageously limiting the power output that can be generated from the fuel bundle. In known prior designs, all of the fuel rods of a bundle have the same peak power limit and all fuel rods of that bundle operate below the peak power limit with different margins. For example, while a majority of fuel rods in a BWR fuel bundle are uranium rods that do not contain poisons, even those rods which do contain poisons such as gadolinium, as well as part-length fuel rods, must operate below the peak power limit. These rods are typically located within the interior of the bundle. Consequently, when the peak power limit is established and the fuel rods are designed to balance power producing capability and fuel bundle weight, the resulting fuel bundle is fundamentally unbalanced because the fuel rod power behavior is very dependent upon the lattice position of the rods within the bundle such that some rods operate near the peak power limit and others have significant margins. BRIEF SUMMARY OF THE INVENTION In accordance with the present invention, there is provided a fuel bundle having two or more peak power limits for two or more sets of fuel rods dependent upon lattice location. For example, if the peak power limit for the edge or peripheral rods of the bundle is raised relative to the peak power limit of the interior rods, the edge rods are provided with greater margins, notwithstanding their higher power locations in the lattice. By differentiating between the peak power limits for two or more sets of fuel rods within the same fuel bundle, e.g., a higher peak power limit for edge rods than the peak power limit for the interior rods, bundle power can be increased while optimizing bundle uranium weight, thereby improving operating margin and fuel cycle costs. Thus, the present invention provides a fuel bundle wherein one set of rods, e.g., the edge rods, have a higher peak power limit than the peak power limit of another set of rods, e.g., the interior rods. Further peak power differentiation can also be provided. For example, interior fuel rods adjacent one or more of the water rods may have a higher peak power limit than other interior rods but lower than the increased peak power limit of the edge rods. Peak power limit differentiation within the interior rods, however, is of lower order value than as between the edge and interior rods. To provide a bundle with differentiated peak power limits optimized for rod lattice position, the peak power limit of a first set of rods, for example, the peripheral or edge rods, is raised. This is manifested in a number of ways. For example, the length of the nuclear fuel column within one or more of the rods forming the peripheral or edge rods can be shortened leaving an increased gas plenum volume at the upper region on top of the fuel rod. Stated differently, the present invention provides for the removal of fuel from the peripheral or edge rods in those regions of low power generation, i.e., near the top or bottom of the fuel rod. For example, a fuel pellet predominantly of natural uranium and therefore of low power generation capacity can be removed at the top of the fuel rod, increasing the gas plenum volume without significantly affecting power output. By increasing the fission gas volume/fuel volume ratio, the power generated from the higher power generating portion of the fuel rod, e.g., approximately 85% of the fuel rod length excluding about the upper and lower 20 inches each of the fuel rod, can be increased, for example, by fuel enrichment in that higher power generating region. Significantly, by increasing the gas plenum/fuel volume ratio in this manner, e.g., by removing fuel from low power regions of the edge fuel rods, the peak power limit of the edge rods can be increased prior to the first fission chain reaction of the fuel bundle in the reactor relative to the peak power limit of the interior rods. This is distinguished from maintaining the same peak power limit for all rods while flattening the power distribution curve by uniformly reducing fuel in the edge rods to improve margin and hence increase actual power output (while still maintaining all fuel rods below the same initial peak power limit prior to any fission chain reaction). Thus, if the typical fuel rod at the edge has a conventional nuclear fuel column 150 inches long, the length of the fuel column can be shortened, e.g., to approximately 146 inches increasing the available gas plenum volume within the fuel rod. Alternatively, or conjunctively, the fuel pellet density or pellet diameter can be changed. For example, the pellet diameter of the edge rods can be reduced, increasing the gap between the outer surface of the pellet and the cladding. In general, the magnitude of the nuclear fuel within the fuel rods at higher power output lattice locations within the fuel bundle, e.g., the edge rods, can be reduced. This enables the reactor to be operated at higher power output with closer margins for the majority of the rods while avoiding exceeding the peak power limit of those rods in higher power lattice locations. Consequently, the edge fuel rods would be operated at a higher power level, even though there is less fuel in the rods. Also, increased power output can be accomplished by increasing the amount of enrichment in the fuel in the edge fuel rods in comparison with their enrichment in a conventional bundle with a single peak power limit. In a preferred embodiment according to the present invention, there is provided a nuclear fuel bundle comprising a plurality of elongated, generally parallel nuclear fuel rods containing nuclear fuel and arranged in a matrix thereof, at least a first rod of the plurality of rods having a first peak power limit higher than a second peak power limit of at least a second rod of the plurality of rods prior to a first fission chain reaction of the nuclear fuel in the bundle in a nuclear reactor for power generation. In a further preferred embodiment according to the present invention, there is provided a nuclear fuel bundle comprising a plurality of elongated, generally parallel nuclear fuel rods containing nuclear fuel and arranged in a rectilinear array thereof having edge fuel rods about the periphery of the bundle and fuel rods interior of the edge rods, the edge and interior rods having discrete peak power limits, the peak power limit for the edge rods being higher than the peak power limit for the interior rods prior to a first fission chain reaction of the nuclear fuel in the bundle in a nuclear reactor. In a still further preferred embodiment according to the present invention, there is provided in a nuclear fuel bundle for disposition in a nuclear reactor, a method of arranging a plurality of elongated, generally parallel nuclear fuel rods in a matrix thereof, comprising the steps of disposing at least first and second fuel rods of the matrix thereof at locations in the fuel bundle affording first and second power outputs, respectively, the first power output being higher than the second power output, determining a different peak power limit for the first and second fuel rods prior to a first fission chain reaction of the nuclear fuel bundle in a nuclear reactor and initially operating the reactor with the first rods having a higher peak power limit than a second peak power limit of the second rods. In a still further preferred embodiment according to the present invention, there is provided a method of operating a nuclear reactor having a plurality of fuel bundles arranged in a nuclear core with each bundle having a plurality of elongated generally parallel nuclear fuel rods arranged in an orthogonal array thereof comprising the steps of, for a selected fuel bundle, identifying a first set of rods thereof which are anticipated to have a higher power output than a second set of rods thereof, before a first fission chain reaction in the bundle, providing the first set of rods with a first peak power limit and the second set of rods with a second peak power limit lower than the peak power limit for the first set of rods and operating the nuclear reactor to increase the power output of the selected bundle by decreasing power output margins of each of the first and second sets of rods relative to the first and second peak power limits, respectively. |
060751762 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to processes of low-level mixed waste solidification. 2. Related Art Low-level mixed waste contains hazardous chemical and low-level radioactive species. The chemical contaminants are often volatile compounds or pyrophorics and cannot be disposed of by conventional high-temperature methods. Portland cement grouting (PCG) is one of the prior art radioactive waste solidification methods. PCG results in hydration-induced hardening. In addition to hydration-induced hardening cements, there are also chemical hardening cements. Various phosphate compositions belong to the class of chemical hardening cements. The solidification of phosphate compositions results from a number of chemical reactions of metal oxides and orthophosphoric acid at room temperature, thereby causing generation of a hard solid phosphate form with a low solubility in water. These phosphate forms are very efficient for immobilization of rare earth and transuranic elements. Radioactive and toxic incinerator ash has also been immobilized by incorporating the ash into cements based on zirconium orthophosphate and dual magnesium-sodium and magnesium-ammonia orthophosphate. Immobilization in chemical hardening cements is caused by both physical isolation of the dispersed hazardous elements and their structural integration into the phosphate matrix upon its formation. Phosphate binders are heterogenous systems consisting of a powder with basic properties (metal oxide or hydroxide) and phosphoric acid. Chemical reactions between the two leads to self-setting of such a system. The products of this reaction are hydrated salts of orthophosphoric acid that can be characterized as inorganic polymers. These polymers, also defined in some literature as phosphate ceramics, have several desirable characteristics including: high compression strength, adhesion to inert surfaces, insolubility in water and the ability to withstand very high temperatures. The relevant prior art processes concern incinerator ash immobilization in a magnesium-phosphate cement matrix. The prior art process is implemented by the following operations: 1. Preparation of cement powder by mixing magnesium oxide powders calcined at 1,000.degree. C., and 15 mass % boric acid (reaction moderator); PA1 1. Longer hardening period, while the hardening moderator, i.e. iron oxide (3+), also functions as the matrix material; PA1 2. Low cost of the input materials, i.e. iron oxides; PA1 3. Metallurgical waste (cinder) and iron-containing natural minerals (magnetite) are used as the input matrix material; PA1 4. Ferro-magnetic properties of the matrix provide for remote transfer of the radiation hazardous compounds by means of the electromagnetic equipment; and PA1 5. Capability to control the setting time by varying the concentration of Fe.sub.2 O.sub.3 in the system. PA1 (a) forming a mixture of iron oxide powder having ratios in mass % of FeO:Fe.sub.2 O.sub.3 :Fe.sub.3 O.sub.4 =25-40:40-10:35-50; PA1 (b) forming a powder phase of waste powder and the mixture of iron oxide powder; PA1 (c) forming a solution of orthophosphoric acid H.sub.3 PO.sub.4 ; PA1 (d) mixing the solution of orthophosphoric acid with the powder phase of waste powder and the mixture of iron oxide powder in mass % of waste powder:iron oxide powder:acid solution=30-60:15-10:55-30 to form a slurry; PA1 (e) blending the slurry to form a homogeneous mixture; PA1 (f) setting the slurry to form monolithic specimens; and PA1 (g) curing the homogeneous mixture at room temperature to form a final product. PA1 (a) preparing a predetermined amount of magnetite powder; PA1 (b) forming a powder phase of waste powder and the magnetite powder; PA1 (c) forming a solution of orthophosphoric acid and ferric oxide; PA1 (d) mixing the solution of orthophosphoric acid and ferric oxide with the powder phase of waste powder and magnetite in mass % of waste powder:magnetite powder:acid solution=30-60:15-10:55-30 to form a slurry; PA1 (e) blending the slurry to form a homogeneous mixture; and PA1 (f) curing the homogeneous mixture at room temperature to form a final product. 2. Mixing of the generated cement powder and the ash powder; 3. Mixing of the generated powder and 50% orthophosphoric acid solution; and 4. Molding and setting of the samples. The generated magnesium-phosphate cement samples incorporate 35 mass % of the incinerator ash. The compression strength of these samples is 275 kg/cm.sub.2. The leach rate data for toxic and radioactive metals have been obtained by using Environmental Protection Agency Method 1311, Toxicity Characteristic Leaching Procedure (TCLP) and American Nuclear Society, American National Standards Institute Measurement of the Leachability in Solidified Low-Level Radioactive Wastes by a Short-Term Test Procedure, Method (ANSI 16.1). The leach rate values for the toxic metals obtained by TCLP method do not exceed the established limits, while the determined leachability indices for various elements, estimated by ANSI 16.1, range from 15 to 22, thereby exceeding the passing criterion of 6 set by the Nuclear Regulatory Commission (NRC). The sample mass loss during long-lasting leaching tests does not exceed 1%. Therefore, the characteristics of the magnesium-phosphate materials that incorporate incinerator ash indicate their high chemical stability and compression strength. The prior art magnesium-phosphate cement used for incinerator ash immobilization suffers from a number of disadvantages. Magnesium oxide is expensive and requires high-temperature annealing to slow down its reaction speed. In addition, the composition is difficult to mix because of how fast it hardens. The prior art magnesium-phosphate cements also require the application of additives that function as moderators, e.g. boric acid, to slow down hardening. SUMMARY OF THE INVENTION An object of the invention is immobilization of incinerator ash into the phosphate compositions that will provide the optimal hardening speed and leaching resistance. In accordance with a first embodiment of the invention, a method of immobilizing mixed low-level waste is provided. The method includes: (a) forming a mixture of iron oxide powder having ratios in mass % of FeO:Fe.sub.2 O.sub.3 :Fe3O.sub.4 =25-40:40-10:35-50; (b) forming a powder phase of waste powder and the mixture of iron oxide powder; (c) forming a solution of orthophosphoric acid; (d) mixing the acid solution with the powder phase of waste powder and the mixture of iron oxide powder in mass % of waste powder:iron oxide powder:solution=30-60:15-10:55-30 to form a slurry; (e) blending the slurry to form a homogeneous mixture; and (f) curing the homogeneous mixture at room temperature to form a final product. In a preferred embodiment, the solution is added to the powder phase of waste powder and the mixture of iron oxide powder in step (d). In an alternative preferred embodiment, the waste powder is mixed with the orthophosphoric acid, then the mixture of iron oxide powder is added to the acid/waste mixture. In one alternative embodiment, the iron oxide powder comprises metallurgical cinder. If the ratio of FeO:Fe.sub.2 O.sub.3 :Fe.sub.3 O.sub.4 in the cinder is not in the ranges of (25-40):(40-10):(35-50), then the appropriate amount of iron oxides should be added. Another method of immobilizing mixed low-level waste is provided in accordance with a second embodiment of the invention, wherein the naturally occurring ore magnetite is used as the source of iron oxides. Magnetite is an iron oxide with the theoretical formula Fe.sub.3 O.sub.4 .dbd.FeO.Fe.sub.2 O.sub.3. The method includes: (a) preparing a predetermined amount of magnetite powder; (b) forming a powder phase of waste powder and the magnetite powder; (c) forming a solution of ferric oxide in orthophosphoric acid (70-90 g. of ferric oxide per liter, in a preferred embodiment); (d) combining the acid solution with the powder phase in mass % of waste powder:magnetite powder:acid solution=30-60:15-10:55-30 to form a slurry; (e) blending the slurry to form a homogeneous mixture; and (f) curing the homogeneous mixture at room temperature to form a final product. In a preferred embodiment, the acid solution is added to the powder phase of waste powder and the magnetite powder in step (d). In an alternative preferred embodiment, the powder phase of waste powder and the magnetite powder are added to the solution in step (d). The amount of the low-level waste in the final product can be up to about 60 mass % for some types of wastes such as solid deposits or soil with phosphogypsum. In a preferred embodiment the amount of low-level waste in the final product is 30-40 mass %. Advantageously, the pure phosphoric acid (without water) content of the slurry is about 25-35 mass %. The present process has many advantages over the processes of the prior art, including: Other features and advantages of the invention will be set forth in, or apparent from, the following detailed description of the preferred embodiments of the invention. DESCRIPTION OF THE PREFERRED EMBODIMENTS In a preferred embodiment of the invention iron oxides, i.e. FeO, Fe.sub.2 O.sub.3 and Fe.sub.3 O.sub.4, are used as feed powder matrix materials. Waste metallurgical cinder may be used as the source of the iron oxides provided that the ratio of the three oxides is adjusted to be mass % FeO:Fe.sub.2 O.sub.3 :Fe.sub.3 O.sub.4 =(25-40):(40-10):(35-50). The specific correlations of iron oxide quantities with different iron valences makes it possible to optimize the slurry mixing and hardening time, thereby controlling the quality of the final product. Solidification of surrogate incinerator ash waste is performed according to the following procedure: In an alternative preferred embodiment, the orthophosphoric acid solution is added to the waste powder and then the iron oxide powder is combined with the mixture of orthophosphoric acid and waste. The amounts of the iron oxides with different iron valences used in the preferred embodiment optimize conditions for handling the slurry and provide good mixing, molding and hardening of the mixture 3-4 hours after the start of the process. The blending time must be long enough to allow a good mixing of the slurry, so that the waste is evenly distributed in the forming matrix. The preferred curing time is obtained by using oxides containing a cation with ionic potential of between 2.5 and 4.5, which corresponds to the oxide-phosphate systems that cure at low temperature. In an alternative preferred embodiment, magnetite powder is used as the feed powder matrix material. Solidification of surrogate incinerator ash waste is performed according to the following procedure: The final products of the processes contain 30-60 mass % of the incinerator ash and have high compression strength values. The elements in the matrix meet the TCLP, ANSI 16.1, Material Characterization Center, test #1(MCC-1) and Product Consistency Test (PCT) leaching test requirements. The waste loading is preferably not greater than about 60%. Excessive waste loading leads to deficiencies in the final product. Furthermore, the concentration of the orthophosphoric acid solution used to prepare the slurry is preferably no less than 50 mass %, more preferably from 50-83%; and, the orthophosphoric acid content (i.e. pure H.sub.3 PO.sub.4, without water) of the final slurry is preferably 25 to 35 mass %, to produce a firm ceramic matrix. |
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description | The present application is a divisional of pending U.S. application Ser. No. 10/657,302, filed Sep. 9, 2003, which in turn in based on and claims priority of U.S. provisional Application No. 60/409,359, filed Sep. 9, 2002, both of which are incorporated herein by reference. This invention was supported by funds from the U.S. Government (DARPA Grant No. N66001-97-1-8911 and DARPA Grant No. N66001-01-C-8056). The U.S. Government may therefore have certain rights in the invention. The invention relates to controlled, magnetohydrodynamically-driven, fluidic networks suitable for use in devices for processing and analyzing biological and chemical samples such as laboratories on chips and micro-total analysis systems. Placed within a temperature gradient, the fluidic networks of the present invention can further act as thermal cyclers, particularly of the type used for polymerase chain reactions (PCR). The invention also relates to magnetohydrodynamic stirrers that are capable of generating chaotic advection within a microfluidic conduit or chamber. In recent years, there has been a growing interest in developing minute chemical and biological laboratories, analytical devices, and reactors known collectively as laboratories on chips. The ability to perform chemical and biochemical reactions in such devices offers many benefits including reduced reactant and media volumes for safety and economy, and improved performance from increased thermal and mass transfer. In such devices, a spatially defined and controlled environment permits precise flow of reactants through the network. The flow of a fluid from one part of the device to another, and the efficient mixing of fluids are tasks that are far from trivial. In a micro-scale device such as a laboratory on a chip, mixing of fluids is a particular challenge as flows are at very low Reynolds numbers, turbulence is not available to promote mixing, and the insertion of moving components into these devices is difficult. Electrostatic forces have been used to move liquids around such devices. These forces usually induce only very low flow rates, require the use of high electrical potentials, and can often cause significant heating of the solution which may be inappropriate for the materials being used or the reactions to be performed. The use of electromagnetic forces offers a means for manipulating at least slightly conductive liquids in microfluidic devices and systems. The application of electromagnetic forces to pump and/or confine fluids is not new. It is known that magnetohydrodynamic (MHD) systems are capable of converting electromagnetic energy into mechanical work in fluid media. To date, MHD systems have mostly been used to pump highly conducting fluids such as liquid metals and ionized gases, to study ionospheric/astrophysical plasmas, and to control magnetic fusion devices. Recently, however, MHD micro-pumps in silicon and in ceramic substrates have been constructed demonstrating the ability of such pumps to move liquids through microscale conduits. These efforts, however, have addressed individual pumping devices and have not provided an effective means for either the controlled movement of liquids through a microfluidic network or the efficient mixing of liquids in such microscale environments. Although it is envisioned that this invention will be mostly used in the context of minute devices, the concepts are not limited for small devices and can be applied for large devices as well. Relevant publications, each of which are incorporated herein in their entirety, are identified as follows: Bau, H. H., 2001, A Case for Magnetohydrodynamics, Proceedings of the 2001 ASME International Mechanical Engineering Congress and Exhibition, New York, N.Y. 2001, November 11-16. CD. Vol 2. Bau, H. H., Zhong, J., and Yi, M., 2001, A Minute Magneto Hydro Dynamic (MHD) Mixer, Sensors and Actuators B, 79/2-3, 205-213. Bau, H., H., Zhu, J., Qian, S., and Xiang, Y., 2003, A Magneto-Hydrodynamically Controlled Fluidic Network, Sensors and Actuators B, 88, 205-216 Bau. H., H., Zhu, J., Qian, S., Xiang, Y., 2002, A Magneto-Hydrodynamic Micro Fluidic Network, IMECE 2002-33559, Proceedings of IMECE'02, 2002 ASME International Mechanical Engineering Congress & Exposition, New Orleans, La., Nov. 17-22, 2002. Jang, V., and Lee, S. S., 2000, Theoretical and Experimental Study of MHD (Magneto-hydrodynamic) Micropump, Sensors and Actuators A, 80, 84-89. Lemoff, A. V., and Lee, A. P., 2000, An AC Magnetohydrodynamic Micropump, Sensors and Actuators B, 63, 178-185. Lee, A. P. and Lemoff, A., V., Micromachined Magnetohydrodynamic Actuators and Sensors, U.S. Pat. No. 6,146,103. Qian, S., Zhu, J., and Bau, H. H., 2002, A Stirrer for Magneto-Hydrodynamically Controlled Micro Fluidic Networks, Physics of Fluids, 14 (10): 3584-3592. Yi, M., Qian, S., and Bau, H. H., A Magneto-hydrodynamic (MHD) Chaotic Stirrer, J. Fluid Mechanics, 468, 153-177. Xiang, Y. and Bau, H. H., 2003, Complex Magneto Hydrodynamic, Low Reynolds Number Flows, Physical Review Letters E, 68, 016312-1-016312-11. Zhong, J., Yi, M., and Bau, H. H., 2002, A Magneto Hydrodynamic Pump Fabricated with Low Temperature Co-fired Ceramic Tapes, Sensors and Actuators A: Physical, 96, 1, 59-66. One aspect of the present invention is a controlled, magnetohydrodynamically-driven, fluidic network comprising a plurality of connected and individually controlled conduits each having at least one pair of opposing walls and at least one pair of electrodes disposed along the opposing walls, and at least one electrode controller in operational engagement with the electrodes for implementing an activation sequence comprising a current or potential across electrode pairs. In a preferred embodiment, the network further comprises an algorithm for determining the activation sequence. In operation, the fluidic network is provided with an at least slightly conductive fluid, is placed at least partially within a suitable magnetic field, and an electric field of a specific current or potential is applied for a predetermined period of time across electrode pairs and through the fluid within the network. In accordance with MHD principles, the magnetic field is oriented approximately perpendicular both to the orientation of the electric field and to the axis of flow along the conduit. Interaction of the electric and magnetic fields generate volumetric forces, called Lorentz body forces that propel the fluid through the network. By the selective application of electric fields of specific currents or potentials at various points in the network for specific time intervals, the fluid may be directed with precision through the network in any desired pattern without the need for moving parts such as mechanical pumps or valves. In this manner, the control and flow of fluid through the network is similar to the control and flow of an electrical current in an electronic circuit. The fluidic network is controlled by means of at least one electrode controller in operational engagement with the electrodes positioned on the sidewalls of the conduits. The controller or controllers implement an activation sequence of currents or potentials that are applied across the electrode pairs of the network. In one embodiment, the activation sequence is determined in accordance with an algorithm. The algorithm can be in any form that is capable of determining the magnitude, polarity and duration of current or potential across various electrode pairs throughout the network associated with generating a precise pattern of Lorentz forces for propelling the fluid in a controlled manner along any desired path through the network including, for example, an equation, a series of equations, a series of iterative steps, or software. The activation sequence may be entirely pre-determined by the algorithm or determined with the use of feedback generated by the operation of the network. Another aspect of the invention is a controlled, magnetohydrodynamically-driven thermal cycler comprising the fluidic network of the present invention positioned at least partially within a temperature gradient. The imposition of a temperature gradient across the network allows fluid to move through one or more zones of differing temperatures as it circulates within the network. By circulating materials through different temperature zones, chemical and biochemical reactions such as, for example, PCR may be readily accomplished. The thermal cycler described herein may be used singly or in combination, and may operate either as part of an MHD-driven fluidic network, as a component in a non-MHD fluidic system, or as a stand-alone device. Yet another aspect of the invention is a MHD stirrer for use in microfluidic networks. The MHD stirrer comprises a conduit or cavity having at least two electrodes disposed therein such that complex secondary flows including flows characterized by chaotic advection are generated upon application of a current or potential across electrode pairs in a magnetic field. If two electrodes are used, at least one electrode must be movable between at least two positions to allow for the alternating application of current or potential across the electrodes in at least two different positions. If three or more electrodes are used, then the electrodes may be movable or fixed provided that a current or potential is alternately applied between electrode pairs in at least two different positions. In either embodiment, the application of a current or potential across the electrodes induces at least two different alternating flow patterns which in turn induces chaotic advection. The MHD stirrer may be used singly or in combination, and may operate either as part of an MHD-driven microfluidic network, as a component in a non-MHD microfluidic system, or as a stand-alone device. A further aspect of the invention is a method for controlling the flow of fluid through a MHD-driven fluidic network comprising a plurality of connected and individually controlled conduits each having a pair of opposing walls and at least one pair of electrodes disposed along the opposing walls, comprising the step of implementing an activation sequence of electrical currents or potentials across the electrode pairs by means of at least one electrode controller governed by an algorithm and in operable engagement with the electrode pairs. Yet another aspect of the present invention is a method for generating chaotic advection within a conduit or chamber of a microfluidic network having at least two electrodes disposed therein comprising the step of applying an electrical current or potential across the electrodes to generate at least two different alternating flow patterns and induce chaotic advection. In one embodiment, the current or potential is alternately applied between a stirring electrode and at least two different electrodes disposed along the internal walls of the conduit. The two electrodes may be disposed on the same wall, on adjacent walls or on opposing walls of the conduit. In an alternative embodiment, the current or potential is alternately applied between one electrode disposed along an internal wall of the conduit and at least two stirring electrodes positioned within the conduit and away from the internal wall. In another embodiment, the polarity of the electric field between the one or more stirring electrodes and the one or more electrodes disposed along the internal walls is repeatedly reversed. The Lorentz forces generated by the configuration of electrodes and applied currents or potentials within the conduit result in secondary flows, and in particular flows characterized by chaotic advection, which are effective in mixing laminar fluids such as fluids present within a microfluidic network. Controlled-flow MHD fluidic networks, thermal cyclers, and chaotic advection stirrers for use in microfluidic devices for processing and analyzing biological and chemical samples such as laboratories on chips and micro-total analysis systems are described. The controlled-flow MHD fluidic network comprises a plurality of connected and individually controlled conduits for the transmission of fluid each conduit having a pair of opposing walls, at least one pair of electrodes disposed along the opposing walls of the conduits, and at least one electrode controller in operational engagement with the electrodes for implementing an activation sequence of currents or potentials across the electrode pairs. In preferred form, the network further comprises an algorithm for determining the activation sequence. Movement of fluid through the network is accomplished in accordance with principles of magnetohydrodynamics which utilize the interaction of approximately perpendicularly oriented electric and magnetic fields to generate Lorentz forces within the network. The pattern of Lorenz forces moving fluid through the network is governed by the activation sequence which defines the particular magnitude, polarity and timing of current or potential to be applied across individual electrode pairs. The activation sequence itself is determined by an algorithm which may produce a predetermined activation sequence or a sequence which uses information about the state of the fluidic network or of the fluid circulating within the network during operation of the network. In this manner, fluid within the network that is at least slightly conductive may be directed with precision and control through the network along any desired path and without the need for mechanical valves or pumps. The basic building block of the controlled-flow MHD fluidic network is the conduit. Described generally with reference to FIGS. 1a and 1b, the individual conduit of the fluidic network has length L, width W, and height h. The conduit may be capped, as depicted in FIGS. 1a, 1b, 2a and 2b, or open from above as shown in FIG. 6. Moreover, conduits comprising a network of conduits may have the same or different shapes, lengths and sizes provided that the conduits are capable of bearing electrodes positioned suitably for the generation of Lorentz forces upon the application of a current or potential within a magnetic field. Suitable configurations include, for example, rectangular, as shown in FIGS. 2a and 2b. Alternatively, the conduits may comprise straight, curved or slanted walls that in cross-section are square, trapezoidal, circular, oval, or any other such suitable shape or combination of shapes. The network of conduits may be simple or complex comprising any combination of curved or straight conduits with few or many interconnections arrayed in either two or three dimensions. A network comprising solely of straight conduits is shown in FIG. 7, and FIG. 5 depicts an example of a network comprising a combination of straight and curved conduits. Further, FIGS. 4, 5, and 7 provide schematic depictions of embodiments of relatively simple, two-dimensional networks. In FIGS. 4 and 5, the individual conduits, similar in structure to the conduit depicted in FIGS. 1a and 1b, are denoted with numbers. The network can be connected to external supplies and drains denoted as a, c, and e in FIG. 4. Alternatively, fluid can circulate inside the network without external links as shown in FIG. 5. The MHD-controlled networks can be fabricated from a variety of substrate materials including, for example, silicon, monomers, prepolymers, polymers, elastomers, glass, plastics, metals (in combination with dielectric materials) and ceramic materials such as, for example, low temperature, co-fired ceramic tapes. Ceramic tapes are a convenient substrate material as they are dielectric and amenable to layered manufacturing techniques. Individual tapes may be machined and electrodes, conductors, and resistors may be printed or otherwise applied on the tapes with metallic pastes or inks in their green (pre-fired) state when the tapes are soft and pliable. A plurality of individually processed tapes can then be stacked, aligned, laminated, and co-fired to form a monolithic device that integrates hydraulic conduits and conductive paths arrayed in a two or three dimensions. Such manufacturing techniques also provide a means for inexpensive and rapid prototyping. Ceramic tapes may further include magnetic materials such as magnetic particles thus integrating the magnetic field source into the substrate and eliminating the need for the use of external magnets to generate Lorentz forces within the network. In the pre-fired (green) state, ceramic tapes may comprise oxide particles such as alumina and/or silica, glass frit, and an organic binder that can be made from photo-resist. The tapes are available in a variety of thicknesses, typically from about 100 μm±7% to a few hustred microns, although thinner tapes of about 40 μm can be casted. The tapes in their green state are soft and pliable, and can be readily machined by a variety of known techniques including laser, milling, and photolithography when the binder is photo-resist. Conductive paths such as metallic circuits may be either printed, processed photolithographically, or otherwise applied to the tapes to form electrical circuits and components such as electrodes, resistors, conductors, and thermistors. Conduit sizes may be any size suitable for use in MHD-driven fluidic networks, and in one embodiment may range from about 10 μm to several millimeters. Individual tapes may be stacked, aligned, laminated, and co-fired to form sintered, monolithic structures having complex, either two- or three-dimensional networks of fluidic conduits, electronic circuits, and electrodes. Glass or other materials can be attached to or incorporated in the tapes to facilitate optical paths. Further, the tapes may include a magnetic material such as magnetic particles and/or a single or multiple layers of coils may be embedded in the tapes to generate a magnetic field. The conduits of the fluidic network are provided with at least one pair of electrodes (denoted in FIGS. 1a and 1b as 11a and 11b of length Le) positioned on opposing internal surfaces of the conduit. These electrodes, referred to as driving electrodes, define the region along the conduit in which an electric current or potential is applied. The driving electrodes may be positioned on the internal surfaces of the conduits in a variety of ways all of which are considered within the scope of the invention. Two such electrode configurations are depicted in FIGS. 2a and 2b. FIG. 2a depicts an arrangement of electrodes comprising four individual electrodes 12a, 12b, 12c, and 12d positioned along the corners of a conduit as shown incross-section in operational engagement with electrode controller 8. FIG. 2b depicts an arrangement of electrodes comprising a pair of individual electrodes 14a and 14b each covering the entire area of opposing sidewalls of a conduit in operational engagement with electrode controller 8. The arrangement of electrodes as shown in FIG. 2b is a configuration that can provide a nearly uniform current density in a fluid within the conduit. Electrodes may also be used to control the shape of the velocity profile and, depending on the specific application, arrangements other than those shown in FIGS. 2a and 2b may be preferable. It is further aspect of the invention that not every individual conduit within the network need be provided with driving electrodes provided that conduits that are not so equipped are in communication with at least one conduit that is so equipped. In this manner, the propulsion of fluid through the conduit having driving electrodes is capable of driving fluid through the conduit lacking such electrodes. In preferred form, the driving electrodes terminate some distance away from the ends of the conduit so as to minimize current leakage (cross-talk) between or among adjacent conduits comprising the network. The driving electrodes themselves may be used to form virtual conduits, that is, conduits which lack physical walls for the containment of the fluid. Flow of fluid through a network comprising virtual conduits is spatially defined by the configuration of the electrodes on the substrate and controlled by the current or voltage applied across electrode pairs. In such networks, the electrodes may either protrude from, be flush with, and/or terminate beneath the surface of the substrate. FIGS. 3a and 3b depict an example of a toroidal virtual conduit 16 and a straight virtual conduit 18 in which the “walls” of the conduits are the electrodes 17a, 17b, 19a, and 19b themselves. FIGS. 3a and 3b correspond, respectively, to a top view and a view in cross-section of the virtual conduits. Complicated patterns of electrodes may be readily manufactured using a variety of printing and lithographic techniques for applying the electrodes to the substrate. Each pair of driving electrodes is in operable engagement with an electrode controller that acts to control the magnitude, polarity and timing of the current or potential applied across pairs of driving electrodes. The network may comprise a single electrode controller in operable engagement with each of the driving electrodes of the network. Alternatively, the network may comprise a plurality of electrode controllers each of which controls one or more driving electrodes of the network. In one embodiment, each pair of driving electrodes is controlled by a separate electrode controller. By controlling the current and/or potential applied across each electrode pair, the one or more electrode controllers regulate in a precise pattern and with precise timing the generation of Lorentz forces that propel the fluid through the conduits of the network. An implementation of an exemplary electrode controller and algorithm are described in greater detail in Bau, H., H., Zhu, J., Qian, S., and Xiang, 2003, Y., A Magneto-Hydrodynamically Controlled Fluidic Network, Sensors and Actuators B: Chemical, 88, 205-216, which is incorporated herein in its entirety. The one or more electrode controllers of the network are governed by an activation sequence that coordinates and controls the flow and combination of fluids within the network. In one embodiment, the activation sequence is determined in accordance with an algorithm which computes and defines the magnitudes, polarities and timing of currents or potential differences applied across the various driving electrode pairs of the network that are necessary to achieve the desired control of flow paths and flow rates throughout the network. The algorithm can be in any form that is capable of determining the specific current or potential across various electrode pairs throughout the network associated with generating a precise pattern of Lorentz forces for propelling the fluid in a controlled manner along any desired path through the network including, for example, an equation, a series of equations, a series of iterative steps, or software. In one embodiment, the user specifies the desired flow path and the flow rates associated with the various conduits. The algorithm then computes the magnitudes, polarities and timing of currents or the voltages that are needed to implement the desired conditions. The algorithm may also compute the magnitudes, polarities and timing of currents or voltages while minimizing an objective function such as, for example, the total power dissipation of the device. The sequence of specific magnitudes, polarities and timing of currents or voltages across particular electrode pairs comprises the activation sequence that is used by the electrode controllers to generate the Lorentz forces necessary to propel fluid in the network along the desired flow path. Preferably, the algorithm is in the form of a software program capable of calculating specific magnetic and/or electric field strengths associated with flow rates within conduits of a known size. As a software program, the algorithm may be resident on the one or more electrode controllers or located remote from the controllers provided the activation sequence generated by the algorithm is capable of communication with and implementation by the one or more electrode controllers. The algorithm may determine the magnitude, polarity and timing of current or potential in a predetermined mode or in a mode that uses feedback generated by the operation of the network in determining the activation sequence. In embodiments in which the activation sequence is determined at least partially with the use of feedback, the network further comprises a sensor assembly capable of continuously or periodically collecting information about the state of the network or the fluid circulating within it during operation and inputting this information into the algorithm. MHD-driven fluidic networks in which movement through the network can be controlled by an activation sequence generated by an algorithm are suitable for a variety of applications including point-of-care medical diagnosis; laboratory diagnosis; drug discovery; air, food, and water quality monitoring; and detection of pathogens and chemical agents associated with biological and chemical warfare agents. In accordance with MHD principles, the orientation of the magnetic field need not be vertical with respect to a conduit oriented in a horizontal plane. For example, if a pair of driving electrodes were positioned on the top and bottom walls of the conduit oriented in a horizontal plane, MHD principles would require the magnetic field to be oriented also horizontally but transverse to the direction of flow of the conduit. Thus, the controlled-flow MHD fluidic network of the present invention may accommodate any combination of electrical and magnetic fields that are approximately perpendicular to each other and in any orientation with respect to the conduit provided both fields are approximately perpendicular with respect to the axis of flow through the conduit. In one embodiment, the conduits comprising the network are arranged in a planar configuration and the magnetic field is oriented approximately perpendicular to the plane in which the conduits are arrayed. As shown in FIGS. 1a and 1b, a three-dimensional Cartesian coordinate system can be represented with respect to an exemplary conduit of the network by axes x1, x2, and x3. The vertical arrow denoted with the letter (B) indicates the orientation of the magnetic field. In operation, the entire device is subjected to a magnetic field of a specific intensity. The magnetic field may be generated by a permanent magnet or an electromagnet. Alternatively, the entire network may be fabricated with the inclusion of a magnetic material, thereby eliminating the need for an external magnetic field source. The most suitable source for the magnetic field will depend on a variety of factors including the particular application for which the network is intended. In other embodiments, synchronized, alternating electric and magnetic fields may also be used. In such embodiments, the fields may be synchronized such that the resulting Lorentz forces remains essentially steady. Fluid is transmitted from one region of the MHD network to another by currents Ii or potential differences Vi applied across the driving electrode pairs within the conduits of the network. The potential difference Vi in a given conduit (i) induces an electric current of density Ji˜siVi/Wi where Wi is the conduit's width and si is the specific electric conductivity of the fluid. This current, in turn, interacts with the magnetic field to produce a Lorentz body force of density (JiB) directed along the axis of the conduit. The magnitude of the force and its direction may readily be controlled by respectively controlling the magnitude and polarity of either the potential difference Vi or the current Ii. Since the relationship between the flow rate and the current is linear over the domain of interest, the electric current typically will be the preferred control variable. To a first approximation, the flow rate (Qi) in the conduit is given as a function of the potential difference Vi (or the current Ii), and the pressure drop across the length of the branch (DPi) by the constitutive relationships of the type: Qi=HiDPi+MivVi or Qi=HiDPi+MiIIi where Hi and Mi are, respectively, the hydraulic and MHD conductivities. Preferably, the conduits comprising the network are sufficiently long so that fringe effects can be neglected, and the current flow is essentially one-dimensional. An exemplary MHD-controlled fluidic network is shown in FIG. 4. The individual conduits comprising the network are denoted by the numbers 1 through 6 and the nodes are denoted with the letters a through e. Nodes a, c, and e communicate beyond the network or with reagent reservoirs and serve as sinks and sources. In an alternative embodiment, the network is not provided with any sinks or sources. Conduits which do not contain driving electrodes have hydro-magnetic conductivity set to zero. For the network depicted in FIG. 4, six equations relate the flow rate in a conduit to the pressure drop along that conduit's length and the potential difference across the electrodes. Additionally, mass continuity (Kirchhoff s law) requires that all the flow rates arriving at each node sum up to zero. When the potential differences across all the conduits and the pressures at the sources and sinks are given, these equations can be solved to obtain the flow rates in all the conduits. An embodiment of the network as shown in FIG. 5 is manufactured with low temperature, co-fired ceramic tapes. The device has planar architecture, that is, all the conduits of the network are arrayed in a single layer. While the single conduit layer shown in FIG. 5 consists of a plurality of tapes, an individual layer may be formed from a single tape or from a series of tapes depending on the thickness of the tapes and the desired layer thickness. In an alternative embodiment, conduits may be fabricated in multiple layers and interconnected through one or more vertical wells to form a network comprising a three-dimensional array of conduits. In one embodiment, a planar, MHD-controlled fluidic network is fabricated with LTCC 951AX co-fired ceramic tapes supplied by DuPont that have a nominal (pre-fired) thickness of ˜250 μm. FIG. 6 provides an exploded view of the elements of the network as shown schematically in FIG. 5. The fabrication process consists of blanking rectangular segments of tapes to a desired size. A few layers of tapes are laminated to form a part. The various parts are machined individually using a numerically-controlled milling machine. Subsequently, electrodes and conductor paths are printed on the various parts. In one embodiment as shown in FIG. 6, layer 40 is the top layer that contains the flow conduits 42 and includes 1.1 mm widex 1.7 mm deep flow conduits 42 and soldering pads 44 using DuPont 6134 solderable conductors. The soldering pads 44 are connected through vertical vias 45 filled with DuPont 6141 via fill paste (not shown) to the various electrodes. While relatively large conduits are fabricated in this embodiment to facilitate easy flow visualization, similar networks may be fabricated having much smaller dimensions. Layer 46 comprises the bottom wall of the conduits 48 and contains the electrodes 50 and some of the electrical leads 52 connecting to the electrodes. A more detailed layout of the electrodes shown in layer 46 is provided in insert E. About 20 μm thick×2 mm wide electrodes 50 made from DuPont 5734 gold paste are printed on the surface of layer 46. The gold electrodes 50 are aligned with the edges of the conduits 42 such that when layers 40 and 46 are attached, about 0.1 mm of the widths of the electrodes 50 along each side of the conduits' vertical walls are exposed to the conduits 42. Each conduit 42 is provided with a pair of driving electrodes 50. A gap separates the driving electrodes in adjacent conduits. Silver conductors 56 made from DuPont 6145 conductor paste are printed on both layers 46 and 54 to facilitate the connection of each electrode to the soldering pads located on the surface of layer 40. All the leads are connected through vertical vias 45 to terminals located on the surface of layer 40. Layer 58 the bottom layer, contains additional leads (not shown). Subsequent to machining and printing, the individual parts are stacked, aligned, laminated, and co-fired to form a sintered, monolithic block. The device may capped with a cover plate or left uncapped to facilitate easy access to the channels and to enable dye injection for flow visualization. In one embodiment, the electrodes of the network may be controlled by an electrode controller comprising computer-controlled relay actuators and a D/I card. The relays are programmed to switch on and off in such a way that any one or combination of electrode pairs in the network can be active at any given time and for any given interval. Additionally, the relays allow for the switching of the polarity of any given pair of electrodes and the supply of power either in controlled-voltage or controlled-current modes. In operation, the conduits are filled with a fluid that is at least slightly conductive such as, for example, saline solution. While 0.1M and 0.3M solutions are suitable, MHD-driven networks can operate with ion concentrations as low as about 50 mM. The device is placed on top of a neodymium (NdFeB) permanent magnet of approximate intensity B=0.4 T (Edmund Scientific). Dye (Cole Parmer Instrument Co.) is injected at various locations to achieve flow visualization. The fluidic network may be analyzed using linear graph theory methodology, and the potentials Vi or currents Ii may be determined so as to direct the fluid to follow any desired path. In one set of experiments utilizing a network configured as shown in FIG. 5, a trace of dye was inserted into the fluid at conduit 1. The electrodes of all the conduits were activated, and the network was programmed to pump the fluid around the large circuit (conduits 1, 2/3, 4, 5/6, and 7) with the flow divided between conduits 2 and 3 and between conduits 5 and 6. By appropriate choice of current or potential differences, the flow can be split between conduits 2 and 3, and between conduits 5 and 6 in any desired proportion. When the dye entered the torus comprising conduits 2 and 3, the electrodes within all the conduits but 2 and 3 were switched off. The polarity of conduit 3's electrodes was reversed, and the fluid was forced to circulate around the torus. In one embodiment, the torus can be positioned across a temperature gradient having two or more different temperature zones such as may be needed for DNA amplification reactions. Subsequently, the polarity of electrodes in conduit 3 was reset to its original setting, all the other electrodes were turned on, and the dye was pumped out of the torus into conduit 4. The dye then split between conduits 5 and 6 and recombined in conduit 7. By appropriate choice of current or potential differences, the flow can be made to circulate around the loop consisting of conduits 6 and 7. In an alternative network, liquids are pumped from the wells at the end of conduits 8 and 9, the fluids are mixed in conduit 1 equipped with stirring electrodes as described herein, and the electrodes are programmed to pump the liquid into the torus defined by conduits 2 and 3. FIG. 7 depicts schematically a more complicated MHD network consisting of a plurality of wells 62 and conduits 10 through which reagents, analytes, or chemicals may be pumped along any desired path and stirred, causing various interactions and/or reactions. Each of the conduits shown in FIG. 7 has a structure similar to the conduit depicted in FIG. 1. Analytes and reagents may be pumped from any of the wells, brought together, and mixed to interact and/or react with reagents pumped from other wells. The network may also facilitate combinatorial screening in which many processes are carried out in parallel. Moreover, reaction and interaction products may be used in subsequent reactions or interactions in either pre-determined or feedback modes. The embodiment depicted in FIG. 7 can readily be expanded to a three-dimensional network allowing a much larger number of connections. These examples illustrate that MHD-controlled networks provide an easy, effective and inexpensive way of circulating fluids through microfluidic laboratory on a chip conduits. MHD-controlled networks can operate with a wide variety of electrolyte and buffer solutions such as, for example, solutions containing NaCl, KCl, NH4Cl, CuSO4, FeCl2/FeCl3, NaH2PO2, and Hydroquinone among many others. The performance of the device, however, may be affected by the particular solution and electrode materials that are used. For example, the use of NaCl solutions may lead to bubble production at relatively high current densities and electrode corrosion. To the extent MHD-driven devices are used as disposable devices, electrode corrosion may not be an issue of significance. Moreover, the MHD-driven devices, depending on the application, can operate either open or capped. With open conduits, bubble formation may not present a problem. In closed conduits, however, bubble generation must be addressed and preferably limited. In one embodiment, the use of redox species such as FeCl2/FeCl3 solution with platinum electrodes may sustain higher current densities than a NaCl solution without bubble formation and without electrode corrosion. Ultimately, though, the choice of the electrolyte or buffer is dictated by, among other things, the compatibility of the electrolyte or buffer with the specific processes to be performed in the system. In addition to the MHD forces, the fluid within the network may also be subject to a pressure gradient that is either flow assisting or flow adverse. Another aspect of the present invention is an MHD stirrer. Chemical reactions and biological interactions in a microfluidic device often involve mixing or stirring fluids in order to bring various molecules together. Mixing by diffusion alone in a microfluidic device is often not efficient. The diffusion time of macromolecules may be prohibitively large even when the lengths are measured in hundreds of microns. Moreover, since flows are often laminar and corresponding Reynolds numbers in microdevices are usually very small, one is also denied the benefits of turbulence as an efficient mixer. In one embodiment, the MHD stirrer of the present invention comprises a conduit or chamber having at least two electrodes disposed therein such that the application of a current or potential across the electrodes within a magnetic field generates secondary flows such as flows characterized by chaotic advection. In embodiments in which two electrodes are used to induce chaotic advection, at least one electrode must be movable so that the current or potential may be applied alternately across electrode pairs in at least two positions. In embodiments in which at least three electrodes are used, the electrodes may be movable or fixed and disposed along and/or away from the internal walls of the conduit or chamber. In either embodiment, a current or potential is alternately applied across electrodes occupying at least three positions to induce at least two alternating flow patterns which generates chaotic advection. In one embodiment, the MHD stirrer of the present invention comprises a conduit having at least one electrode disposed along the wall of the conduit, and at least two electrodes positioned within the conduit and away from the wall. In another embodiment, the MHD stirrer comprises at least two electrodes disposed along at least one wall, and at least one electrode positioned within the conduit and away from the wall. In a further embodiment, the stirrer has at least two electrodes aligned along at least one wall, and at least one electrode disposed along another wall. In yet another embodiment as shown in FIGS. 1a and 1b, the MHD stirrer comprises a pair of electrodes 11a and 11b disposed along the opposing walls 10a and 10b, and at least two electrodes 11ci positioned within the conduit and away from the opposing walls. In this embodiment, the electrodes 11ci positioned within the conduit and away from the opposing walls 10a and 10b may be aligned as shown in FIG. 1a along the centerline of the conduit's bottom. In still another embodiment, the MHD stirrer comprises a cylindrical chamber with an electrode disposed around its internal periphery and at least two electrodes positioned eccentrically inside the chamber. The placement of the one or more electrodes permits the generation of complex secondary flows including flows characterized by chaotic advection that is beneficial for mixing or stirring within a fluidic conduit or chamber. The conduits as described in all of these embodiments may comprise a conduit of the MHD-driven fluidic network or thermal cycler of the present invention. MHD stirrers that generate chaotic advection may operate either by varying the current or potential applied across electrode pairs between zero and a prescribed value (either positive or negative) or by repeatedly reversing the polarity of each electrode by varying the current or potential between negative and positive values. Depending on the particular electrolyte used, reversal of polarity may be advantageous in certain cases since by reducing electrode corrosion and bubble accumulation on electrode surfaces. Furthermore, in applications in which analyte migration in the electric field is a problem, reversing polarity is likely to reduce or eliminate such migration. In the embodiment shown in FIGS. 1a and 1b, a conduit 10 of a controlled, MHD-driven microfluidic network is provided with a pair of electrodes 11a and 11b disposed on opposing walls 10a and 10b of the conduit and a series of electrodes denoted 11ci disposed along the centerline of the conduit and away from the opposing walls. Electrodes 11ci (where i=1, 2, 3, . . . ) are referred to as stirring electrodes. This particular implementation of the stirrer is described in greater detail in Qian, S., Zhu, J., and Bau, H. H., 2002, A Stirrer for Magneto-Hydrodynamically Controlled Micro Fluidic Networks, Physics of Fluids, 14 (10): 3584-3592 which is incorporated herein in its entirety. In order to operate a MHD conduit as a stirrer, the electrodes intended for use in creating secondary flows are in operable engagement with at least one electrode controller such as, for example, a computer-controlled relay actuator. In one embodiment, relay-actuators combine both driving electrodes 11a and 11b into a single electrode C. When a potential difference is applied across the electrode pair C-11ci, circulatory motion of the fluid within the conduit is generated, with the fluid circulating around electrode 11ci. When the electrode pair C-11c1 is activated for a time interval T1, electrode pair C-11c2 for another time interval T2, then electrode pair C-11c1 once again, and so on in a periodic fashion, chaotic advection is generated. As the magnitude of the period (T=T1 T2) increases, the chaotic region increases in size and complexity. In some circumstances, it may be advantageous to alternate the electrode potentials in anon-periodic fashion. In demonstrating this effect, flow visualization experiments of the stretching and deformations of a dye blob were performed. FIG. 8 depicts computational and experimental results when a blob of dye 70 was inserted into the conduit 72 and the evolution of the dye was tracked over time. Both in experiment and theory, a rapid spread of the dye was observed indicating efficient stirring. By engaging a larger number of electrode pairs C-11ci, one can further extend the fraction of the conduit that participates in the mixing process. As electrodes may be readily patterned into various shapes, electric fields may be induced in different directions. The interaction of such electric fields with the magnetic field can be used to induce secondary complex flows that may be beneficial for stirring and mixing. Stirring electrodes such as electrodes Ai shown in the embodiment depicted in FIGS. 1a and 1b may be located singly or in combination anywhere within a conduit provided they are away from either of the opposing walls. The electrodes need not to be aligned along the conduit's center. Although it is convenient to print the electrodes on the device's floor to avoid intrusion, one can also use other arrangements such as, for example, electrodes in the form of pins that protrude into the conduit. The MHD stirrer may comprise either an open or a closed cavity of any suitable shape. With reference to FIG. 9, an embodiment of the stirrer is described comprising a circular cavity with an electrode C deposited along its periphery. Two additional electrodes, A and B, are deposited on the cavity's bottom. The cavity is filled with a conducting liquid such as, for example, a saline solution, and it is positioned in a uniform magnetic field oriented parallel to the cavity's axis. When a potential difference is imposed across the two electrodes A and C, an electric current flows between the two electrodes and the interaction between this current and the magnetic field results in Lorentz forces that induce, for example, counterclockwise flow circulation in the cavity centered next to the location of electrode A. Subsequently, when the potential difference is switched from electrode pair A/C to electrode pair B/C, depending on the polarity of the electrodes, either a counterclockwise or a clockwise circulatory pattern may be induced centered next to the location of electrode B. The device is operated by alternately engaging electrode pairs A/C and B/C with a period T. In some circumstances, it may be advantageous to alternate the electrode potentials in a non-periodic fashion. FIGS. 10a through 10i depict Poincaré sections (stroboscopic images) of the trajectories of passive tracers (FIGS. 10a through 10c), actual passive tracers' trajectories (FIGS. 10d through 10f), and flow visualizations (FIGS. 10g through 10i) of traces of a drop of dye inserted in the cavity as functions of the normalized (dimensionless) alternations period (T). The columns of FIGS. (10a, 10d and 10g), (10b, 10e and 10h) and (10c, 10f and 10i) correspond, respectively, to T=2, 6 and 8. When T=2, the motion depicted in the Poincaré section shown in 10a is mostly regular and consists of two sets of closed orbits, one set encircling one electrode and the other set encircling the other electrode. FIG. 10d depicts the actual motion of the tracer which shows the presence of jitters resulting from the tracer being trapped (at different times) by the flow fields induced by the two electrode pairs. The presence of two families of periodic orbits is well supported by the flow visualization experiments as shown in FIG. 10g. As the period T increases, chaotic islands become visible. FIG. 10e illustrates that as the period increases, so does the magnitude of the jitters. The presence of the global structure consisting of two counter-rotating circulations is visible in FIG. 10h. When T=8, FIG. 10c depicts the trajectory of a single tracer. The irregular chaotic region appears to have spread to cover almost the entire cavity. Similar to FIG. 10c, the flow visualization experiments depicted in FIG. 10i illustrate the presence of counter-rotating eddies through the existence of an unmixed zone at the ends of the diagonal that is perpendicular to the line connecting the two electrodes. The operation of the device is described in greater detail in Yi, M., Qian, S., and Bau, H. H., A Magneto-hydrodynamic (MHD) Chaotic Stirrer, J. Fluid Mechanics, 468, 153-177 (2002) which is incorporated herein in its entirety. FIG. 11 depicts a schematic representation of another embodiment of a MHD stirrer. As shown in FIG. 11, the stirring electrodes 82a, 82b, 82c, 82d, and 82e are aligned perpendicular to the conduit's walls 80a and 80b. By subjecting these electrodes to varying potential differences in the presence of a magnetic field, forces are generated that drive fluid flow in various directions in “virtual” conduits whose geometry is dictated by the positioning of the electrodes. An implementation of the stirrer is described in greater detail in Bau, H. H., Zhong, J., and Yi, M., 2001, A Minute Magneto Hydro Dynamic (MHD) Mixer, Sensors and Actuators B, 79/2-3, 205-213; and Xiang, Y. and Bau, H. H., 2003, Complex Magneto Hydrodynamic, Low Reynolds Number Flows, Physical Review Letters B, 68, 016312-1-016312-11, which are incorporated herein in their entirety. FIG. 12 depicts the deformation of an initially straight dye line resulting from the application of Lorentz forces by means of a MHD stirrer of the present invention. A thin trace of dye 90 (Water Soluble Fluorescent Liquid Dye, Model 298-16-Red, Cole Palmer Instrument Company, Niles, Ill., USA) is applied by means of a syringe across the cavity 92 and then a potential difference is applied across adjacent electrodes 94a, 92b, 92c, 92d, and 92e. As a result of the application of the potential difference, fluid flow is induced in the cavity. The motion consists of rotating cells with the fluid moving up in one interval between two electrodes and down in the adjacent interval. Frame A in FIG. 12 depicts the line of dye initially inserted into the device. Depending on the polarity of the electrodes, the dye either moves upwards or downwards as shown in frame B of FIG. 12. After a few seconds, the polarity of the electrodes is reversed. Since diffusion is relatively slow and the flow is at a relatively low Reynolds number, the dye retracts its steps in almost a reverse fashion as shown in frame C of FIG. 12 and then starts deforming in the opposite direction as shown in frame D of FIG. 12. When the process is allowed to continue for some time, the dye traces the convective cells as shown in frame E of FIG. 12 in good qualitative agreement with theoretical predictions. The electrodes may be patterned in many different ways to induce various flow patterns. The embodiments described above are just a few examples of numerous possible variants of MHD stirrers. FIG. 13 compares theoretical predictions shown in the left column and experimental results shown in the right column obtained in another implementation. FIG. 13 depicts the deformation of an initially straight line of dye under various operating conditions. The top row, FIGS. 13a and 13b, depicts the flow structure when only the odd-numbered electrodes are active. By alternating the potential difference across non-adjacent pairs of electrodes, it is possible to induce chaotic motion in the cavity 100. For example, electrodes 102a, 102c, and 102e may be engaged for the time interval T1, and then electrodes 102b and 102d for the time interval T2. By repeating this mode of operation, fairly complicated flows are generated and effective stirring is provided. The results of this mode of operation are depicted in the bottom row of FIG. 13, FIGS. 13c and 13d. Another aspect of the invention is a controlled, MHD-driven thermal cycler comprising the fluidic network of the present invention positioned at least partially within a temperature gradient. Magnetohydrodynamics provides the means to circulate fluids continuously in a closed loop. Different parts of the loop may be maintained at different temperatures, enabling the cycling of the liquid to subject the liquid to different temperature zones. FIG. 14 depicts one embodiment of the thermal cycler of the present invention. The cycler comprises a closed conduit loop 110 with electrodes aligned along opposing walls. Electrodes 112 and 114 are aligned along the inner wall of the loop 110, and electrodes 116 and 118 are aligned along the outer wall of the loop 110. An entry port 120 with electrodes 122 and 124 aligned along its opposing walls leads into the loop 110 and an exit port 126 with electrodes 128 and 130 aligned along its opposing walls leads out of the loop 110. While the device shown in FIG. 14 has one inlet and a separate exit port, the cycler may be equipped with a single port or larger number of inlet and exit ports. In order to utilize the conduit loop depicted in FIG. 14 as a thermal cycler, different parts of the loop are maintained at different temperatures. The various temperature zones may be maintained, for example, with the use of electrical resistors or thermoelectric units (not shown). FIG. 14 depicts three thermal zones. It is contemplated as within the scope of the invention that a larger or a fewer number of thermal zones may be used as is suitable with reference to the particular application. At the beginning of operation, an electrical potential is applied to the electrodes such that the material is drawn into the loop. The polarities of either electrode pair 112 and 116 or electrode pair 114 and 118 are then reversed so that the material within the conduit loop is forced to circulate continuously around the loop. The particular choice of polarity will determine whether the motion is in the counterclockwise or clockwise direction. If necessary, the polarities and the magnitudes of the potentials applied to electrodes 110, 124, 128, and 130 may be adjusted so as to prevent the material within the conduit from leaving the loop. Also, if necessary, the direction of the flow in the thermal cycler may be periodically changed to minimize analyte migration in the electric field. As the material within the conduit cycles around the loop, it is exposed to different temperatures. In certain embodiments, this cycling between or among different temperature zones facilitates biological interactions such as, for example, those needed for PCR. After the material within the conduit has completed the desired number of cycles around the loop, electrical potentials are supplied to the various electrodes so as to pump the reaction products out of the loop. The reaction products may be pumped either through the exit port 110 defined by electrodes 128 and 130 back through the inlet port 120 defined by electrodes 122 and 124 or split among any number of exit ports (not shown in the figure) so as to transport parts of the sample to different subsequent analysis paths. The embodiment of the MHD thermal cycler depicted in FIG. 14 may be readily integrated into a magneto-hydrodynamic network such as the one depicted in FIG. 6, integrated into a network in which fluids are propelled by other means than MHD, or used as a stand-alone device. In all three implementations, MHD stirrers of the present invention may be integrated into the MHD thermal cycler of the present invention to enhance efficiency. One application of the MHD thermal cycler is for PCR. The MHD thermal cycler has the advantage over other continuous flow devices in that the number of cycles may be readily adjusted in a predetermined mode according to the characteristics of the analyte to be amplified or in a feedback mode with the use of a sensor capable of detecting the amplification rate. Since it is not necessary to cycle the substrate temperature as is done in conventional PCRs, the MHD thermal cycler is capable of facilitating rapid amplification of DNA. |
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abstract | An installation for treating articles with radiation, the installation comprising a structure having pivotally mounted thereon an inlet starwheel and an outlet starwheel respectively arranged facing an inlet and an outlet of a shielded enclosure in which there are mounted at least one pivotal treatment starwheel and at least one electron emitter in the vicinity of the treatment starwheel. The installation comprises a linear inlet conveyor and a linear outlet conveyor extending inside the shielded enclosure respectively facing the inlet and the outlet, the linear conveyors each comprising a respective transporter surrounding a shielded wall forming a baffle. |
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047939621 | claims | 1. A process for placing in a case a bundle of rods arranged in n parallel rows of at least r rods, in accordance with a square pitch lattice in a nuclear fuel assembly, wherein it comprises placing the bundle of rods horizontally in accordance with said lattice in a waiting position above a cassette having p recesses separated by partitions, p being an integer at least equal to n, so that each row of rods is located above a recess in the cassette, lowering the rods into the cassette recesses, placing the cassette in the horizontal extension of a case, by interposing between cassette and case a transformation member, whose width decreases progressively from cassette to case by a value equal to the cumulative thickness of the cassette partitions and transferring the rods from the cassette into the case through the transformation member by simultaneously pushing all the rods. 2. A process according to claim 1, wherein the number of rods in certain of the rows exceeds r,so that use is made of a cassette whereof each recess can contain at the most s rods, s being and integer at the most equal to r and comprising a number p of recesses exceeding n and such that the product p. s is at least equal to the number of rods in the bundle and wherein, after lowering s rods of each row into each of the n recesses located below the n rows of rods, there is a relative horizontal displacement perpendicular to the rods between the cassette and the rods of the bundle not introduced into the n aforementioned recesses, called "remaining rods" in order to bring the (p-n) remaining recesses successively below each row of rods and to lower said remaining rods into said (p-n) recesses. 3. A process according to claim l, wherein the bundle of rods is positioned horizontally in the waiting position, whilst maintaining the rods in accordance with said square pitch lattice by at least one set of horizontal combs and one set of vertical combs and wherein the set of horizontal combs is retracted prior to introducing the rods into the cassette recesses. 4. A process according to claim 3, wherein, prior to placing the bundle of rods horizontally above the cassette, part of the assembly is eliminated in order to free one of the ends of the rods of the bundle, the assembly being placed in the extension of said waiting position and the bundle of rods is extracted from the assembly to bring it directly into said waiting position, the horizontal and vertical combs being put into place as the displacement of the bundle towards this position advances. 5. A process according to claim 1, wherein, prior to introducing the rods into the case, at the entrance of the latter is placed a base, in which are located the ends of the rods turned towards the case, said base being able to slide in the case with th rods during the introduction of the latter. 6. A process according to claim 1, wherein, prior to introducing the rods into the recesses of the cassette, the bottom of at least certain of th recesses is displaced upwards, so as to reduce the number of rods which can be received in each recess, whose bottom has been displaced. 7. A process according to claim 6, wherein, after introducing the rods into the cassette recesses, the bottom of at least some of the recesses whose bottom has been displaced upwards beforehand is displaced downwards, so that the rods contained in these recesses are at the same level as that part of the case in which they are to be introduced. 8. An installation in ocmbination with a case and cassette for placing in the case a bundle of rods arranged in n parallel rows of at least r rods, according to a square pitch lattice in a nuclear fuel assembly, wherein it comprises means for maintaining the bundle of rods horizontally in accordance with said lattice in a waiting position above the cassette having p recesses separated by partitions, p being an integer at least equal to n, in such a way that each row of rods is positioned above a cassette recess, means for lowering the rods into the cassette recesses and means for simultaneously pushing all the rods contained in the cassette so as to introduce them into a case positioned horizontally in the extension of the cassette, via a transformation member interposed between cassette and case and whose width decreases progressively from the cassette towards the case by a value equal to the cumulative thickness of the cassette partitions. 9. An installation according to claim 8, wherein the means for introducing the rods into the cassette recesses comprise at least two horizontal support members extending perpendicularly with respect to the rods and on which the latter can rest by gravity and means for displacing said members vertically between a top position in which said members are located above the cassette and a bottom position in which said members are located below the base of the recesses in vertical notches formed in the cassette to the right of the support members. 10. An installation according to claim 8, wherein the number of rods of certain of the rows exceeds r, the recesses of the cassette have a depth able to house s rods, s being an integer at the most equal to r, the cassette comprising p recesses, p being an integer greater than n, so that the product p.s is at least equal to the number of rods in the bundle and the means for introducing the rods into the recesses comprise means for effecting a relative horizontal displacement perpendicular to the rods between the cassette and said means in order to horizontally maintain the bundle of rods. 11. An installation according to claim 8, wherein the means for maintaining the bundle of rods horizontally comprise at least one set of retractable horizontal combs and one set of retractable vertical combs. 12. An installation according to claim 11, wherein said installation also comprises means for maintaining the assembly horizontally in the extension of the waiting position, means for extracting the bundle of rods from said assembly and for bringing the bundle into the waiting position and means for permitting the automatic putting into place of the combs in proportion to the displacement of the bundle towards said position. 13. An installation according to claim 8, wherein the bottom of at least some of the recesses of the cassette is partly formed on mobile members making it possible to vary the height of these recesses. 14. An installation according to claim 13, wherein means are provided for controlling a vertical displacement of the mobile members. |
abstract | A forged upper shroud section which may be machined from a single piece rectangular cross-section ring forging and includes a circular flange and a cylindrical shell is described. Openings and slots are machined into the flange to align and support the shroud head. A groove is machined along an inside surface of the cylinder section, and the groove may be used to support top guide grid (not shown). An end of the cylinder section is machined with a weld prep for attachment to the core section of the shroud. |
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claims | 1. A storage system for radioactive nuclear waste comprising:a longitudinal axis;a cask comprising a hermetically sealable internal cavity configured to hold an inventory of water sufficient to submerge the nuclear waste therein; anda pressure surge capacitor disposed in the cask, the pressure surge capacitor comprising a vacuum cavity evacuated to sub-atmospheric conditions;wherein the pressure surge capacitor is configured to suppress a pressure surge in the internal cavity of the cask. 2. The system according to claim 1, wherein the pressure surge capacitor further comprises at least one pressure relief device constructed to burst at a predetermined pressure level inside the cask, the pressure relief device when burst placing the vacuum chamber of the pressure surge capacitor in fluid communication with the internal cavity to reduce pressure inside the cask. 3. The system according to claim 1, wherein the pressure surge capacitor has a longitudinally elongated tubular body having a height extending for at least a majority of a height of the internal cavity of the cask. 4. The system according to claim 2, wherein the pressure relief device comprises a rupture disk which seals the vacuum cavity of the pressure surge capacitor. 5. The system according to claim 4, wherein the rupture disk is disposed in a first nd cap of the pressure surge capacitor. 6. The system according to claim 5, further comprising a second pressure relief device comprising a second rupture disk disposed in a second end cap of the pressure surge capacitor which seals the vacuum cavity of the pressure surge capacitor. 7. The system according to claim 4, wherein the rupture disk is recessed in the first end cap and disposed in a flow inlet opening formed through the first end cap into the vacuum chamber. 8. The system according to claim 7, further comprising a disk retaining ring comprising central opening exposing the rupture disk, the disk retaining ring trapping the rupture disk in the inlet opening of the pressure surge capacitor. 9. The system according to claim 1, wherein the pressure surge capacitor comprises an elongated cylindrical sidewall shell extending between opposite ends of the pressure surge capacitor. 10. The system according to claim 1, wherein the cask comprises a sealable lid assembly, a cylindrical circumferential wall comprising radiation shielding material, and a base. 11. The system according to claim 1, wherein the internal cavity of the cask further comprises a fuel basket comprising a plurality of longitudinally elongated fuel storage cells each configured to hold a spent nuclear fuel assembly. 12. The system according to claim 11, wherein the pressure surge capacitor is disposed in a peripheral region of the internal cavity of the cask formed between the fuel basket and the circumferential wall of the cask. 13. The system according to claim 11, wherein the pressure surge capacitor is fixedly attached to the fuel basket. 14. The system according to claim 10, wherein the pressure surge capacitor is disposed between an inner shell and an outer shell of the circumferential wall of the cask. 15. The system according to claim 10, wherein the pressure surge capacitor is affixed to an underside of the lid assembly in the internal cavity of the cask. 16. The system according to claim 1, further comprising a second pressure surge capacitor disposed in the internal cavity of the cask. 17. A cask with overpressurization protection for storing nuclear waste fuel comprising:a longitudinal axis;a cask body comprising a removable lid assembly, a base, and a circumferential wall including radiation shielding, the cask body forming a hermetically sealed internal cavity configured for holding spent nuclear fuel submerged in an inventory of water in the internal cavity;a pressure surge capacitor disposed in the cask, the pressure surge capacitor comprising a vacuum cavity evacuated to sub-atmospheric conditions; andthe pressure surge capacitor further comprising at least one rupture disk constructed to burst at a predetermined pressure level inside the cask associated with a cask overpressurization condition;wherein the rupture disk when burst allows a portion of the water to fill the vacuum chamber to reduce pressure inside the cask. 18. The system according to claim 17, wherein the pressure surge capacitor has a longitudinally elongated cylindrical body having a height extending for at least a majority of a height of the internal cavity of the cask. 19. The system according to claim 18, wherein the rupture disk is disposed in a first end cap of the pressure surge capacitor. 20. The system according to claim 19, further comprising a second rupture disk disposed in a second end cap of the pressure surge capacitor opposite the first end cap. 21. The system according to claim 17, wherein the internal cavity of the cask has a fuel basket comprising a plurality of longitudinally elongated fuel storage cells of rectilinear cross-sectional shape each configured to hold a spent nuclear fuel assembly; and wherein the pressure surge capacitor is disposed adjacent to the circumferential wall in one of a plurality of peripheral regions of the internal cavity of the cask formed between the fuel basket and the circumferential wall of the cask. 22. The system according to claim 21, wherein the one of the plurality of peripheral regions has a non-rectilinear cross-sectional shape. 23. The system according to claim 17, wherein the pressure surge capacitor is disposed between an inner shell and an outer shell of the circumferential wall of the cask. 24. The system according to claim 17, wherein the pressure surge capacitor is affixed to an underside of the lid assembly in the internal cavity of the cask. |
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051456354 | description | DETAILED OF THE PREFERRED EMBODIMENTS Referring firstly to FIG. 1, a fuel assembly 1 for a nuclear reactor core, comprises 151 fuel rods 3 arranged axially vertically in a channel box 2 having a hexagonal outer shape. The rods 3 are packed extending side-by-side in the channel box 2 in a close-packed hexagonal lattice. They are mounted by conventional mounting means. Eighteen lattice sites are occupied by control rod guide tubes 4, again in a conventional manner. The outer diameter of each fuel rod is 11.8 mm and the fuel rods are spaced apart by 1.3 mm, giving a geometrical water:fuel volume ratio of 0.5. The reactor is a boiling water reactor, and the aim is to achieve an effective water:fuel volume ratio preferably not more than 0.4, with a view to achieving a conversion ratio approaching unity. With the configuration described, a void fraction of 20% will give an effective water:fuel volume ratio of about 0.4, while this latter ratio falls to about 0.24 when the void fraction rises to 55%. The part of the fuel assembly charged with uranium-plutonium mixed fuel, or enriched uranium fuel, is referred to hereinafter as the "effective fuel region". FIG. 2 shows schematically the axial extent of this region for a first embodiment. The effective fuel region is divided into a top half and a lower half. The uranium-plutonium mixed oxide fuel 15 making up the lower half has an enrichment of 7.0% (enrichment is given here and hereinafter as weight %) of fissile plutonium. The fuel 11 making up the upper half of the effective fuel region has a fissile plutonium enrichment of 6.0%. Thus, the average enrichment of the fuel assembly is 6.5% fissile plutonium, but the upper half has a lower average enrichment than the lower half. In this embodiment, the conceptual boundary between upper and lower halves in fact coincides with an actual boundary between two regions of different plutonium enrichment. That is, the average enrichment throughout the upper half region is 6.0% and throughout the lower half region is 7.0%. In this embodiment, natural uranium enriched with plutonium is used. However, it will be appreciated that recovered uranium from spent fuel reprocessing, depleted uranium from enriching operations, slightly enriched uranium, or a mixture of any of these might be used for enrichment with plutonium. Furthermore, oxide sintered bodies of natural or depleted uranium may be placed at the end peripheries of the fuel region ends to decrease neutron leakage from the reactor core and to provide neutron shielding, in a known manner. FIG. 3(a) shows a nuclear reactor core made up from six hundred and one of the fuel assemblies 1. In the reactor core 5, the fuel assemblies are positioned axially vertically and coolant flows through them axially upwardly from the bottom (inlet) end to the top (outlet) end. FIG. 3(b) shows how, in a nuclear reactor, the core 5 is arranged horizontally, with the assemblies 1 axially vertical, between coolant supply 25 and coolant take-off 26 of the reactor (shown schematically). In an embodiment, the reactor core had the following specification. TABLE 1 ______________________________________ Thermal output 2700 MW Electrical output 900 MW Number of fuel assemblies 601 Height of reactor core 2.00 m Coolant flow rate 2.25 .times. 10.sup.4 t/h Core outlet quality 27% Specific power 17.5 kW/kg Power density 85.2 kW/l ______________________________________ Under reactor conditions of low water:fuel volume ratio e.g. 0.24, maintained with a view to achieving high conversion ratio, a relationship between plutonium enrichment, burn-up, and void coefficient is as shown in FIG. 4. FIG. 4 shows how a fuel having a lower enrichment of plutonium gives a smaller void coefficient than a highly enriched fuel at the same degree of burn-up. The difference becomes smaller as burning of fuel proceeds, but in a normal burn-up range for a light water reactor fuel, i.e. about 45 GWd/t, there is always a difference as shown and the lines do not cross. FIG. 5 shows a relation between burning average void fraction (an average void fraction value during the fuel burning period) and void coefficient. The curve assumes a constant burn-up value and water:fuel geometrical ratio of 0.5. It is seen that, even when two fuels have the same enrichment of plutonium, a larger void coefficient is achieved when a fuel has a higher burning average void fraction, owing to a larger rate of change in neutron infinite multiplication factor with change in void fraction. In operation of the reactor, coolant at the fuel assembly boils, with heat generated from nuclear fission, particularly at the upper region of the effective fuel region so that coolant void in this upper region is relatively large. Accordingly, this upper region--considered from halfway up the longitudinal axis--is a region with a high burning average void fraction and hence a large void coefficient. Use of fuel assemblies as shown in FIG. 2 reduces the relative enrichment of fissile plutonium in this upper region, and hence suppresses its contribution to void coefficient. Because the burning average void fraction is high here, this measure is particularly effective. Conversely, the relatively high enrichment of the fuel in the lower region increases reactivity in this region, where burning average void fraction is relatively small and hence the effective water:fuel volume ratio is larger than in the upper region. Thus, burning of fuel proceeds easily in this region and compensates for any lowering of upper-region reactivity owed to the relatively reduced fissile plutonium enrichment in the upper region. The void coefficient contributions from a FIG. 2-type assembly are indicated schematically by the bold portions of the curves in FIG. 5. The left-hand end of the graph corresponds to the low average void fraction, i.e. the bottom of the core, and this has the high plutonium enrichment as shown by the bold line. Half-way up, there is a transition to lower plutonium enrichment for the upper half which has high void fraction. It should be noted that as void fraction increases, the difference between the void coefficients of high and low enrichment fuels gradually increases too. The reduction in void coefficient by a transition to lower enrichment is therefore especially marked. To consider the advantage provided by the present embodiment, reactor conditions may be compared with a reactor core otherwise similar but loaded with fuel assemblies having a uniform 6.5% plutonium enrichment, all the way down the effective fuel region. Compared with this comparative example, the void coefficient suppression effect and reactivity increment effect of the fuel assemblies embodying the invention provided, when integrated and consolidated in the reactor core, a void coefficient reduced by about 0.4.times.10.sup.-4 .DELTA. k/k/% void and a reactivity increased by about 0.1% .DELTA. k/k. Thus the void coefficient is relatively low while reactivity is maintained. FIG. 6 shows a second embodiment of fuel assembly. This corresponds to the first embodiment except as regards the distribution of plutonium enrichment in the effective fuel region. FIG. 6 shows extreme top and bottom regions of the assembly, each taking up one tenth of the total axial length of the effective fuel region. These comprise uranium-plutonium mixed oxide fuel 9 having 5.1% average fissile plutonium enrichment. The four-tenths of length up from the bottom end tenth 9, i.e. the segment up to the half-way line, consists of fuel 16 having 7.1% enrichment. The remaining four-tenths, i.e. the region from the 5/10 to the 9/10 line (considering the bottom of the assembly as a base line) comprises uranium-plutonium mixed oxide fuel 12 having 6.1% enrichment. The average enrichment over the effective fuel region of this assembly is 6.3%. In the upper half, the average enrichment is 5.9%. In the lower half, the average enrichment is 6.7%. The low-enrichment portions 9 at the extreme ends of the effective fuel region give small power output and slow burning, while the relatively more enriched fuel with large neutron infinite multiplication factor is in the central region where the contribution to reactivity is large. Overall, this gives increased reactivity. It is found that a reactor core (as previously described), loaded with such fuel assemblies has a reactivity larger by 0.5% .DELTA.k/k than a comparative reactor core having the same average enrichment (6.3%) but distributed uniformly along the axial length of the assembly. Indeed, this embodiment with average 6.3% enrichment achieves the same reactivity as a core using 6.5% average enrichment distributed uniformly. Also, we find that in this embodiment the void coefficient is less by about 0.4.times.10.sup.-4 .DELTA.k/k/% void than that achieved with a core using a uniformly distributed 6.5% enrichment, i.e. of the same reactivity. FIG. 7 shows schematically a third embodiment. Apart from the distribution of enrichment, the make-up of the fuel assembly is the same as in the first embodiment. From the base line up to 1/10, and from the top down to 8/10, the effective fuel region has end regions containing 4.8% enriched fuel 6. Thus, the low-enrichment region at the top is twice as long as at the bottom. Between these regions 6, the region from 1/10 to 8/10 comprises uniformly 6.8% enriched fuel 13. Accordingly the average plutonium enrichment overall is 6.2%, that in the upper half of the effective fuel region is 6.0% and in the lower half 6.4%. This embodiment differs from the first two embodiments in that the half-way line does not correspond with any actual boundary between regions of different fuel enrichment. We find that a reactor core loaded with fuel assemblies according to this embodiment has a void coefficient less by about 0.2.times.10.sup.-4 .DELTA.k/k/% void than that of a corresponding core loaded with fuel assemblies having the 6.5% enrichment uniformly axially distributed through the effective fuel region. The advantage is therefore clear. Furthermore, the reactivity is greater by about 0.7% .DELTA.k/k relative to the comparative case (6.2% uniformly), and is comparable with that achievable using 6.5% enriched fuel with a uniform axial distribution. A particular advantage of this embodiment is that the fuel assembly comprises fuels of only two different degrees of enrichment. In particular, the aim is to improve the void coefficient, while forming the downstream (upper) part of the effective fuel region (where the fuel has high burning average void fraction and a large reactivity change with change in void fraction) with as few different enrichment grades as possible. FIG. 8 shows a fourth embodiment. As in the third embodiment, the half-way position at 5/10 does not coincide with any actual boundary between regions of different enrichment. As in the second embodiment the assembly has one-tenth end regions of relatively low enrichment, in this case of 4.85% enrichment fuel 7. A region from 2/10 to 7/10, i.e. most of the bottom half and part of the top half, is of 6.85% enriched uranium-plutonium mixed oxide fuel 14, while a region from 7/10 to 9/10 is 5.85% enrichment fuel 10. The average fissile plutonium enrichment over the effective fuel region is 6.25%, in the upper half 6.05% and in the lower half 6.45%. While the effective fuel region is divided into four different enrichment sections, as in the second embodiment, the present embodiment contains a relatively larger region of high enrichment fuel with more fissile substance in that region. Consequently, the power load per unit weight of high enrichment fuel is less in this embodiment than in the second embodiment for the same whole-core operation power. Thus the burn-up of high enrichment fuel is relatively lower than in the second embodiment so that the increment of void coefficient due to increased burn-up (see FIG. 4) is relatively suppressed. Compared with a loaded reactor core of similar fuel assemblies having an overall average enrichment of 6.5% distributed axially uniformly, the present embodiment gives a void coefficient smaller by about 0.6.times.10.sup.-4 .DELTA.k/k/% void. This is a substantial improvement. As before, the low enrichment fuels positioned at the ends of the effective fuel region lead to an increased overall reactivity--a relative increase of about 0.6% .DELTA.k/k relative to the comparative example (6.25% uniformly). This is the same reactivity as requires 6.5% average enrichment when that enrichment is uniformly axially distributed. Finally, FIG. 9 shows a fifth embodiment of fuel assembly. As in the third and fourth embodiments, the half-way mark of the effective fuel region does not coincide with any actual boundary between two regions of different plutonium enrichment. As before, the general construction of the fuel assembly is similar to that described for the first embodiment. The enrichment distribution, however, is as follows. As in the second embodiment the assembly has end regions of relatively low enrichment fuel 17, in this case 5.9% enrichment, each occupying one tenth of the axial extent of the fuel region. From 2/10 up to 4/10 i.e. a minor proportion of the central region at the bottom, the assembly comprises uniformly 6.9% enrichment uranium-plutonium mixed oxide fuel 18 while the remaining, upper, region from 5/10 to 9/10 i.e. a major proportion of the central axial length comprises uniformly 6.4% enrichment fuel 19. The average enrichment over the effective fuel region is 6.45%, in the upper half 6.3% and in the lower half 6.6%. This is a relatively small difference in enrichment between the two halves. Consequently the burn-up of higher enrichment fuel is less than in the second embodiment above and there is relatively a suppression of the increment of void coefficient, based on the relation between burn-up and void coefficient shown in FIG. 4. A reactor core loaded with fuel assemblies of this embodiment obtains the same reactivity as a reactor core using assemblies of 6.5% enriched fuel distributed uniformly, even though the average enrichment is 6.45% in the embodiment, and this reactivity is largely due to the placement of lower enrichment fuels at the ends of the effective fuel region. Furthermore, the reactor core has a void coefficient smaller by about 1.4.times.10.sup.-4 .DELTA.k/k/% void than a core of equal reactivity having uniformly-distributed enrichment (i.e. of 6.5%). The improved effect is easy to perceive. From the description above it will be seen that the assemblies described enable a reduction of a positive void coefficient in a high conversion ratio reactor core, without any decrease in reactivity, and hence enable improvement of power coefficient and safety margin values in such a reactor. |
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