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051065703 | abstract | A method for generating a beam of negatively charged hydrogen ions is described which comprises the steps of providing a source of metal hydride, heating the hydride to extract either atomic hydrogen or negative hydrogen ions directly therefrom, directing an electron beam onto the hydride or applying electrical charge to the hydride in order to ionize the hydrogen atoms or to prevent the ions from losing charge, and electrically accelerating the negative hydrogen ions so produced as a directed beam. |
RE0315834 | abstract | The invention provides a hold-down device which maintains a net downward force on a fuel assembly in a nuclear reactor under all thermal and hydraulic conditions in the reactor core region. The hold-down device is particularly applicable in reactors having stainless steel support barrels and Zircaloy guide tubes.. The hold-down device of the preferred embodiment of the invention is comprised of coil springs coaxially disposed about alignment posts which extend upwardly from the top end plate of each fuel assembly. A hold-down plate is slideably mounted on the alignment posts and the coil springs are interposed, in compression, between the hold-down plate and the top end plate. In use, the coil springs bias the hold-down plate upwardly against a core alignment plate to provide a downward force on the fuel assembly. The hold-down plate may also serve as a lifting surface for the fuel assembly during fuel handling. |
claims | 1. An anti-scatter grid useful for X-ray imaging, comprising:an anti-scatter layer having a plurality of partitions; andat least one plate of expanded polymer material fixed to one surface of the anti-scatter layer, wherein thickness of one of the at least one plate is equal to or greater than thickness of the anti-scatter layer, each plate thickness being defined by a distance between two opposing surfaces of the respective plate that are oriented such that an incident X-ray passes first through one of the surfaces and then through the other surface;the anti-scatter layer being a polymer substrate with open cells formed therein, inner walls of the open cells layered with an X-ray absorbing metal layer, thereby defining the plurality of partitions;the anti-scatter layer being disposed such that a part of the incident X-ray passes through the anti-scatter layer via the polymer substrate, while another part of the X-ray passes through the anti-scatter layer via the open cells;wherein each plate imparts a rigidity to the anti-scatter layer;wherein each plate has a low surface density that does not substantially attenuate an X-ray that passes therethrough, and has sufficient homogeneity so as not to disturb an X-ray image through artifacts; andwherein the at least one plate consists of two plates of expanded polymer material arranged on opposing surfaces of the anti-scatter layer. 2. The grid according to claim 1 wherein at least one of the plates is in polymethacrylimide foam or expanded polyetherimide. 3. The grid according to claim 1 wherein at least one of the two plates is formed of a material having a density of between 20 and 70 kg/m3. 4. The grid according to claim 2 wherein at least one of the two plates is formed of a material having a density of between 20 and 70 kg/m3. 5. The grid according to claim 1 wherein at least one of the two plates has a thickness of between 2 and 6 mm. 6. The grid according to claim 2 wherein at least one of the two plates has a thickness of between 2 and 6 mm. 7. The grid according to claim 3 wherein at least one of the two plates has a thickness of between 2 and6 mm. 8. The grid according to claim 4 wherein at least one of the two plates has a thickness of between 2 and6 mm. 9. The grid according to claim 1 wherein at least one of the two plates is bonded to the anti-scatter layer. 10. The grid according to claim 2 wherein at least one of the two plates is bonded to the anti-scatter layer. 11. The grid according to claim 3 wherein at least one of the two plates is bonded to the anti-scatter layer. 12. The grid according to claim 4 wherein at least one of the two plates is bonded to the anti-scatter layer. 13. The grid according to claim 5 wherein at least one of the two plates is bonded to the anti-scatter layer. 14. The grid according to claim 6 wherein at least one of the two plates is bonded to the anti-scatter layer. 15. The grid according to claim 7 wherein at least one of the two plates is bonded to the anti-scatter layer. 16. The grid according to claim 8 wherein at least one of the two plates is bonded to the anti-scatter layer. 17. The grid according to claim 9 wherein the bonding is an adhesive arranged on a peripheral area of the anti-scatter layer. 18. The grid according to claim 9 wherein the bonding is an adhesive that forms a thin film extending over an entire surface of the anti-scatter layer. 19. The grid according to claim 18 wherein the adhesive is an aerosol adhesive sprayed to form a film. 20. The grid according to claim 18 wherein the adhesive is in film form. 21. The grid according to claim 1 wherein the two plates are of the same thickness. 22. The grid according to claim 1 further comprising:a protection layer for one of the plates. 23. The grid according to claim 22 wherein the protection layer is a polymer material formed of a composite material containing carbon fibers, a lacquer, or a varnish. 24. The grid according to claim 22 wherein the protection layer has a thickness in the order of 0.1 mm. 25. The grid according to claim 22 wherein the protection layer is arranged on a surface of at least one of the two plates oriented in a direction opposite to a means for providing a source of radiation. 26. The grid according to claim 1 wherein the partitions form a plurality of focalized cells. 27. The grid according to claim 1 wherein the partitions form a plurality of cells, inner walls of the cells being coated with a layer that absorbs radiation. 28. The grid according to claim 1 wherein the grid is positioned within means for protecting the grid. 29. A method for fabricating an anti-scatter grid useful for X-ray imaging, comprising:forming an anti-scatter layer having a plurality of partitions, the plurality of partitions defined by forming a polymer substrate with open cells therein and metallizing inner walls of the open cells with an X-ray absorbing metal layer;disposing the anti-scatter layer such that a part of an incident X-ray passes through the anti-scatter layer via the polymer substrate, while another part of the X-ray passes through the anti-scatter layer via the open cells; andfixing at least one plate of expanded polymer material to at least one surface of the anti-scatter layer, thereby imparting a rigidity to the anti-scatter layer, the at least one plate having a low surface density that does not substantially attenuate an X-ray that passes therethrough, and having sufficient homogeneity so as not to disturb an X-ray image through artifacts;wherein thickness of each of the at least one plate is equal to or greater than thickness of the anti-scatter layer, the at least one plate thickness being defined by a distance between two opposing plate surfaces of the at least one plate that are oriented such that an incident X-ray passes first through one of the surfaces and then through the other surface;wherein the at least one plate consists of two plates of expanded polymer material arranged on opposing surfaces of the anti-scatter layer. 30. The method according to claim 29 wherein each of the two plates is bonded to the anti-scatter layer. 31. The method according to claim 29 further comprising:forming a protection layer for at least one of the two plates. 32. The method according to claim 30 further comprising:forming a protection layer for at least one of the two plates. 33. The method according claim 31 wherein the protection layer is arranged on a surface of at least one of the two plates oriented in a direction opposite to a means for providing a source of radiation. 34. The method according claim 32 wherein the protection layer is arranged on a surface of at least one of the two plates oriented in a direction opposite to a means for providing a source of radiation. 35. The method to claim 29 further comprising:positioning the grid with means for protecting the grid. 36. An anti-scatter grid useful for X-ray imaging, comprising:an anti-scatter layer having a plurality of partitions;the anti-scatter layer comprising a polymer substrate with open cells formed therein, inner walls of the open cells comprising an X-ray absorbing metal layer, thereby defining the plurality of partitions;the anti-scatter layer being disposed such that a part of an incident X-ray passes through the anti-scatter layer via the polymer substrate, while another part of the X-ray passes through the anti-scatter layer via the open cells;at least one plate of expanded polymer material fixed to at least one surface of the anti-scatter grid, the at least one plate imparting a rigidity to the anti-scatter layer, having a low surface density that does not substantially attenuate an X-ray that passes therethrough, and having sufficient homogeneity so as not to disturb an X-ray image through artifacts;a cross-piece positioned on one side of an assembly formed by the layer and the at least one plate;respective U-shaped sections positioned on two opposite sides of the assembly; anda layer on another side of the assembly;wherein the cross-piece, the sections, and the layer on another side form a frame in which the grid is positioned;wherein thickness of each of the at least one plate is equal to or greater than thickness of the anti-scatter layer, each plate thickness being defined by a distance between two opposing plate surfaces of the respective plate that are oriented such that an incident X-ray passes first through one of the plate surfaces and then through the other plate surface; andwherein the at least one plate consists of two plates of expanded polymer material arranged on opposing surfaces of the anti-scatter layer. 37. An anti-scatter grid useful for X-ray imaging, comprising:an anti-scatter layer having a plurality of partitions;the anti-scatter layer comprising a polymer substrate with open cells formed therein, inner walls of the open cells comprising an X-ray absorbing metal layer, thereby defining the plurality of partitions;the anti-scatter layer being disposed such that a part of an incident X-ray passes through the anti-scatter layer via the polymer substrate, while another part of the X-ray passes through the anti-scatter layer via the open cells;at least one plate of expanded polymer material fixed to one surface of the anti-scatter grid, the at least one plate imparting a rigidity to the anti-scatter layer, having a low surface density that does not substantially attenuate an X-ray that passes therethrough, and having sufficient homogeneity so as not to disturb an X-ray image through artifacts; anda cross-piece positioned on one side of an assembly formed by the layer and the at least one plate;wherein thickness of each of the at least one plate is equal to or greater than thickness of the anti-scatter layer, each the plate thickness being defined by a distance between two opposing plate surfaces of the respective plate that are oriented such that an incident X-ray passes first through one of the surfaces and then through the other plate surface; andwherein the at least one plate consists of two plates of expanded polymer material arranged on opposing surfaces of the anti-scatter layer. 38. An anti-scatter grid useful for X-ray imaging, comprising:an anti-scatter layer having a plurality of partitions;the anti-scatter layer comprising a polymer substrate with open cells formed therein, inner walls of the open cells comprising an X-ray absorbing metal layer, thereby defining the plurality of partitions;the anti-scatter layer being disposed such that a part of an incident X-ray passes through the anti-scatter layer via the polymer substrate, while another part of the X-ray passes through the anti-scatter layer via the open cells;at least one plate of expanded polymer material fixed to at least one surface of the anti-scatter grid, the at least one plate imparting a rigidity to the anti-scatter layer, having a low surface density that does not substantially attenuate an X-ray that passes therethrough, and having sufficient homogeneity so as not to disturb an X-ray image through artifacts; andmeans for protecting an assembly formed by the layer and the plate;wherein thickness of each of the at least one plate is equal to or greater than thickness of the anti-scatter layer, each plate thickness being defined by a distance between two opposing surfaces of the respective plate that are oriented such that an incident X-ray passes first through one of the plate surfaces and then through the other surface; andwherein the at least one plate consists of two plates of expanded polymer material arranged on opposing surfaces of the anti-scatter layer. |
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abstract | A system for externally cooling a radiation shielded cask containing heat-emitting high level radioactive waste such as spent nuclear fuel. The system includes the cask defining an internal cavity configured to hold an unshielded canister containing the spent nuclear fuel. An annular cooling water header extends circumferentially around the entire circumference of the cylindrical sidewall of the cask. The header comprises plural dispensing outlets which direct cooling water onto the cask, thereby wetting the entire sidewall of the cask. The cooling water provides an external heat sink for absorbing the heat emitted through the external wall surface of the cask generated by the spent nuclear fuel. In various embodiments, the cooling water header may have a continuous annular structure, or be formed by two or more header segments. The header may be supported directly from the cask by detachably mounted brackets. |
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description | This is a continuation, under 35 U.S.C. §120, of copending international application PCT/EP2005/011499, filed Oct. 27, 2005, which designated the United States; this application also claims the priority, under 35 U.S.C. §119, of German patent application DE 10 2004 054 461.1, filed Nov. 11, 2004; the prior applications are herewith incorporated by reference in their entirety. The invention relates to a method for testing whether the fuel rods of fuel assemblies of a boiling water reactor leak. In a boiling water reactor, the fuel rods of the fuel assemblies need to be tested regularly for leaks, so that defective fuel rods can be replaced in time and radioactive contamination of the cooling water is avoided. The aim in this case is to test all the fuel assemblies of the core within the shortest possible time during routine maintenance work in order to avoid unnecessary downtimes. A commonly used technique for finding leaking fuel assemblies is the so-called sipping method, which is based on the concept of detecting any existing leak by taking a liquid sample (wet sipping) or a gaseous sample (dry sipping) from the area surrounding the fuel rod and testing it for the presence of radioactive fission products. In order to increase the detection sensitivity, suitable measures are used to expel the radioactive fission products located inside the irradiated fuel rods through any leaks which may be present in the fuel rod cladding tubes with the result that, if a leak is present, they can accumulate outside the fuel rod and can be more easily detected in the sample which was taken. A particularly suitable technique is the so-called hood sipping, as is known for example from the commonly assigned international PCT publication WO 00/74071, and its counterpart U.S. Pat. No. 6,570,949 B2. There, a plurality of fuel assemblies, for example 16 fuel assemblies, are covered by a hood and a gas cushion is produced above these fuel assemblies. The exchange of cooling water is stopped because of said gas cushion which is located under the hood. The fuel assemblies heat up on account of their afterheat and radioactive fission products are increasingly expelled from any defective fuel rods which may be present. Some of these fission products are substances which are dissolved in water and gaseous under standard conditions, in particular Kr-85 and Xe-133. In order to detect these fission products, water samples are continuously taken (wet sipping) and continuously degassed. The gas which is produced during degassing is analyzed continuously using a radiation detector during the taking of the samples. In the known method, the time necessary to examine the entire core is now shortened by, in a first step, combining the water samples from the four fuel assemblies which are located in one cell of the core grid, and feeding them to a respective analyzing device which includes a degassing apparatus and a detector arrangement. In other words, four analysis devices can be used to simultaneously test 4×4 fuel assemblies. If a result is positive in one of these cells, the analysis devices are now switched over, and the four water samples from the four fuel assemblies of this cell are fed separately to the analysis devices, in order that one or more defective fuel assemblies of this cell is clearly identified in this manner. Since generally only a few fuel assemblies actually have a defect, this known sipping strategy provides a test time which is clearly reduced with respect to other conventional sipping techniques. In the known method, however, it is an essential prerequisite for clearly identifying a defective fuel assembly that the water level is sufficiently reduced by means of the gas cushion produced in the sipping hood for an exchange of cooling water between the fuel assemblies of such a cell in the intake region to be virtually ruled out, since in this case sound fuel assemblies arranged in the same cell next to a defective fuel assembly could be found positive. In other words, the water level must be sufficiently lowered in such a cell for the top edges of the fuel assembly channels which surround the respective fuel rods to lie above the water level. The top edges of the fuel assembly channels can, however, project beyond the top edge of the upper core grid by different amounts or even lie below this top edge firstly because of design differences between the fuel assemblies of different manufacturers and secondly because of plant-specific different installation conditions. In the known method it is therefore necessary for the water level in every cell to be adjusted to the respective prevailing installation conditions in order to avoid an exchange of cooling water between neighboring fuel assemblies. This requires an additional inspection effort before the samples are taken, which is associated with an increased time requirement. Moreover, it is possible for installation conditions to be such that a simultaneous and isolated sampling from all the fuel assemblies in a core cell is not possible in any event. It is accordingly an object of the invention to provide a method for testing whether or not the fuel rods of fuel assemblies of a boiling water reactor are subject to leakage which overcomes the above-mentioned disadvantages of the heretofore-known devices and methods of this general type and which provides for a short test time. With the foregoing and other objects in view there is provided, in accordance with the invention, a leak detection method for testing fuel rods of fuel assemblies of a boiling water reactor for leaks, wherein a plurality of fuel assemblies are arranged adjacent one another in a cell of an upper core grid of the boiling water reactor. The method comprises the following steps: a) placing a hood above the fuel assemblies of a plurality of cells forming a division for simultaneously heating the fuel assemblies of the division; b) taking at least one water sample from each cell of the division; c) combining the water samples from a plurality of cells forming a group and testing the water samples for the presence of radioactive fission products contained therein; d) simultaneously, and independently of one another, analyzing a plurality of groups in an equal number of groups of measurement channels of an analysis device; e) on registering a positive result in a given group, separately feeding the water samples from the cells located in the given group to the analysis device and analyzing the water samples separately in a corresponding number of measurement channels; and f) individually testing fuel assemblies of a cell that have been analyzed as positive outside the hood. In other words, the objects of the invention are achieved with the leak testing method in which the fuel assemblies of a plurality of cells forming a division are heated simultaneously by placing a hood above this division. At least one water sample is taken from each cell of this division and the water samples from a plurality of cells forming a group are combined and tested for the presence of radioactive fission products contained in the water sample, with a plurality of groups being analyzed simultaneously and independently of one another in a number (which corresponds to said plurality of groups) of measurement channels of an analysis device. If a result in a group is positive, the water samples from the cells which are located in this group are fed separately to the analysis device and are analyzed separately in a corresponding number of measurement channels, and the fuel assemblies of a cell analyzed as positive are tested individually outside the hood. The invention here proceeds from the consideration that the fuel assemblies located in a cell cannot be reliably separated from one another by producing a gas cushion under a hood on account of differing installation conditions, with the result that, when the fuel assemblies located under the hood are analyzed individually, it is difficult to interpret the analysis results and reliably identify a defective fuel assembly. The procedure according to the invention, i.e. the use of the hood sipping method merely for the identification of a cell occupied by a defective fuel assembly, means it is possible to test n×n cells in only two steps using a small number n of measurement channels. Since generally only a small number of fuel assemblies are actually defective, only the fuel assemblies of a small number of cells need to be subjected to an individual analysis which is carried out when the fuel assembly is no longer under the hood, with the result that a high reliability is achieved when identifying defective fuel assemblies even under complicated plant-specific installation conditions, while the total test time remains short. Preferably, the cells located under the hood are separated hydraulically from one another in the region of the hood and from the cells located outside the hood. This separation prevents the convective cooling of the fuel assemblies located under the hood on the one hand and the exchange of water between the cells located under the hood on the other hand. The total test time is shortened in an especially efficient manner if, once one division has been analyzed, the hood is relocated onto another division and if the test of the individual fuel assemblies of a cell in the event of a cell being found positive is carried out at the same time as the test of the other division. The samples from a cell are taken in particular using a number of intake lines, which number corresponds to the number of the fuel assemblies located in said cell, with each of the removal locations thereof being assigned to one fuel assembly. This increases the measurement sensitivity. The fuel assemblies of a cell analyzed as positive are preferably lifted up successively from the core using a telescopic mast of a fuel assembly handling machine and water samples are taken in this lifted-up position and analyzed. A mast sipping method of this type results in a further shortening of the total test time. Other features which are considered as characteristic for the invention are set forth in the appended claims. Although the invention is illustrated and described herein as embodied in method for removing particulates from exhaust gases, and corresponding fiber layer and particulate filter, it is nevertheless not intended to be limited to the details shown, since various modifications and structural changes may be made therein without departing from the spirit of the invention and within the scope and range of equivalents of the claims. The construction and method of operation of the invention, however, together with additional objects and advantages thereof will be best understood from the following description of specific embodiments when read in connection with the accompanying drawings. Referring now to the figures of the drawing in detail and first, particularly, to FIG. 1 thereof, the reactor pressure vessel 1 of a boiling water reactor is open for the purposes of testing the fuel assemblies arranged in its core 2. The reactor well, wherein the reactor pressure vessel 1 is located, is flooded up to a level 3. A division 5, which comprises a plurality of cells 4 of the core 2, which are arranged adjacent to one another and each contain four fuel assemblies, is covered by a hood 6 which is seated on the upper boundary 8 of an upper core grid 10. Water samples are taken using a multichannel sampling system 11, merely indicated schematically in the figure by a removal line, and fed to a multichannel analysis device 14, wherein the water samples are tested for the presence of radioactive fission products either directly or after they have been degassed (so-called hood sipping). In a preceding work step the hood 6 was placed over the cells 4 to be tested using the telescope mast 16 of a fuel assembly handling machine 18. Once the hood 6 is in position, it is uncoupled from the telescope mast 16 so that the latter can be moved into a different position above the core 2. The figure illustrates a situation where a sampling device 20 mounted on the free end of the telescope mast 16 is located above a cell 4 which was found positive in a preceding test using the hood 6. The fuel assemblies located in this cell 4 are successively lifted out of the core 2 using the telescope mast 16 and, in the lifted-up position, are individually, i.e. successively, tested for leaks in the analysis device 14 by taking a water sample. This method, which is referred to as mast sipping, takes advantage of the fact that fission products are increasingly expelled from any leaks which may be present on account of the pressure relief which follows the lifting operation. While the fuel assemblies are individually tested, the cells 4 located under the hood 6 are tested in the multichannel analysis device 14. Alternatively it is also possible to relocate the fuel assemblies of the cell 4 analyzed as positive into the fuel assembly storage pool and to analyze them there in a separate measuring station. According to FIG. 2, the hood 6 covers the division 5, which contains 3×3 cells 4 in the example, of which only three can be seen in the schematic sectional illustration. Partition walls 60 arranged inside the hood 6 are used to separate the individual cells 4 located under the hood 6. Seals 62 of the hood 6 and the partition walls 60 bear on the webs 22 of the upper core grid 10. Four fuel assemblies 40 are located in each cell 4. The hood 6 separates the fuel assemblies 40 of one cell 4 located under the hood 6 from the fuel assemblies 40 of the other cells 4 located under the hood 6. Water then no longer flows through the fuel assemblies 40 located under the hood 6, with the result that the afterheat heats the now still water, which is located in the fuel assemblies, to a temperature which is about 10° C. to 15° C. higher than the water flowing outside the hood 6. Blowing in compressed gas lowers the water level 24 under the hood 6, with the result that a gas cushion is produced. The figure illustrates an installation condition wherein the top edges 41 of the fuel assembly channels 42 of the fuel assemblies 40 are located below the upper boundary 8 of the core grid 10. Both the spacing a between the top edge 41 of a fuel assembly channel and the upper boundary 8 of the core grid 10 and also the spacing b between an upper rod-holding plate 43 and the top edge 41, depicted in the Fig. only for the left-hand fuel assembly 40, can vary from fuel assembly to fuel assembly on account of different radiation-induced increases in length such that an installation condition may occur as is depicted in the figure. An intake line 12 of the sampling system 11 is assigned to each of the cells 4, which intake line has at least one removal location 120 which is used to take the water samples just below the water level 24 which lies at least above the highest fuel assembly channel 42. One intake line 12 with only one removal location 120 is provided for each of the left-hand and right-hand cells 4. An alternative refinement of the sampling process is illustrated in the middle cell 4, wherein four partial branches 122 branch off from the intake line 12, with each of the removal locations 120 thereof being assigned to one fuel assembly 40. The aim in this case is, as far as it is possible in technical terms on account of the respective plant-specific installation condition, to arrange the removal locations 120 inside the fuel assembly channel 42, if possible. The installation condition illustrated in the figure, which is the same for all the cells 4 in the simplified illustration, but in reality differs from cell 4 to cell 4, now is such that, under the hood 6, a simultaneous individual measurement of the fuel assemblies 40 located in a cell 4 is not fundamentally possible using the method described in the introduction and known from the prior art. The reason is that for the purposes of an isolated sampling of water from the left-hand fuel assembly 40 in a cell 4, the water level would have to be lowered to a level 240 which is below the upper rod-holding plate 43 of the right-hand fuel assembly. With such a level 240, however, no water sample could be taken from the right-hand fuel assembly 40 because the removal location 120 cannot be guided through the rod-holding plate 43. Accordingly, samples would have to be taken at differently adjusted water levels in succession in terms of time. Furthermore, it would also be necessary for the position of the removal locations 120 to be matched in terms of height to the respective installation conditions, which differ from cell to cell, by remote control when the hood is in position, as is illustrated by the dashed lines in the left-hand cell 4. These difficulties are avoided in accordance with the invention firstly by omitting an individual analysis of the fuel assemblies 40 while they are located under the hood 6 and secondly by means of a simplified fixed arrangement of the removal locations 120 such that detection of any fission products of each fuel assembly 40 which may emerge from a fuel rod is ensured even if they cannot be assigned to a particular fuel assembly 40. According to FIG. 3, firstly the hood is used to cover and heat a division 5 (emphasized by hatching) of 3×3 cells 4. Three cells 4 of this division 5 are each combined to form a group A, B or C by virtue of the fact that the intake lines 12 of the cells 4 belonging to these groups A,B,C are connected to a shared collection line 13A,B,C assigned to each of these groups. The water samples taken are fed separately to measurement channels 14A, 14B and 14C of the analysis device 14 via the collection lines 13A,B,C and are analyzed there independently of one another. Another measurement channel 14D is used to analyze the water sample which was taken using the sampling device 20 which is arranged on the telescope mast. In the case of a positive result in one of the measurement channels 14A-14C, for example in measurement channel 14A, the water samples taken from the cells 4 of the group A using the intake lines 12 are fed, as is depicted in FIG. 4, separately to the measurement channels 14A-14C in order to identify in this way the cells 4 of the Group A containing a defective fuel assembly. When the analysis of the division 5 is complete, the hood is relocated onto another division 5. The fuel assemblies 40 of the cells 4 which were previously identified as defective (found positive) are tested individually once the hood has been relocated (see FIG. 1), for example using a particularly time-saving mast sipping method, while the other division 5 is analyzed at the same time. |
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summary | ||
claims | 1. A method for fabricating a supermirror for forming a neutron guide having neutrons therein, the method comprising:1) forming a plurality of monochromators comprising the steps of:a) forming a plurality of bilayers of different thickness wherein each bilayer comprises a Ni thin layer and a Ti thin layer wherein the Ni thin layer of each bilayer has a thickness which is equal to the Ti thin layer thickness of that bilayer; andb) alternately stacking a plurality of said bilayers having the same thickness to form the plurality of monochromators of different thickness;2) stacking at least some of the plurality of monochromators of different thickness on each other for forming a supermirror. 2. The method of claim 1, wherein the number of the nickel thin films and titanium thin films stacked in each monochromator is adjusted so as to control reflectivity. 3. The method of claim 1, wherein the reflectivity of each of the monochromators is calculated according to a periodic number-calculating equation and an equation for calculating full width at half maximum (FWHM), and the positions of peaks are determined by the calculated periodic number and FWHM. 4. The method of claim 1, wherein, among the plurality of monochromators, a monochromator reflecting a specific wavelength is removed to extract the specific wavelength. |
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abstract | An apparatus for measuring and controlling a target trajectory within a chamber apparatus for generating extreme ultraviolet light from plasma generated by irradiating a droplet target supplied from a target injection nozzle with a driver laser beam from an external driver laser. The apparatus includes: a nozzle adjustment mechanism for adjusting at least one of a position and an angle of the target injection nozzle; a target trajectory measuring unit for measuring a target trajectory to obtain trajectory information on the target trajectory; a target trajectory angle detecting unit for obtaining a value related to an angle deviation between the target trajectory represented by the trajectory information and a predetermined target trajectory; and a nozzle adjustment controller for controlling the nozzle adjustment mechanism based on the value related to the angle deviation such that the droplet target passes through a predetermined laser beam irradiation position. |
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041860500 | summary | The present invention relates generally to nuclear reactors and more particularly to a liquid-cooled nuclear reactor which incorporates an improved nuclear fuel. The TRIGA research reactor, which was developed and is being marketed by the assignee of this application, is an inherently safe reactor utilizing a uranium-zirconium hydride fuel that has a large prompt negative temperature coefficient of reactivity, which is primarily characteristic of the fuel itself. The reactor is described in more detail in U.S. Pat. No. 3,127,325, issued Mar. 31, 1964, the disclosure of which is incorporated herein by reference. As a result of the inherent safety of the reactor core, a single or multiple control rod can be instantaneously removed from the core without having the resulting power pulse damage the core. In fact, one of the normal operating modes of the TRIGA reactor is termed the pulse mode, wherein such a fast ejection is performed in order to produce a high energy pulse of radiation for experimental purposes. The temperature coefficient of the reactor is prompt because of the intimate mixture of the nuclear fuel with a large portion of solid moderator in the form of zirconium hydride. Thus, the fuel and solid moderator temperatures rise together instantaneously, with no heat transfer delays before the occurrence of moderator-related temperature coefficient effects. The prompt negative temperature coefficient for TRIGA reactors is considered to be a result of the following three contributing components: (1) thermal neutron spectrum hardening effects, (2) Doppler broadening of resonances and (3) neutron leakage from the reactor core. In the standard TRIGA reactors operating throughout the world today, the thermal spectrum hardening effects contribute the largest share to the total prompt negative temperature coefficient. The extensive thermal spectrum hardening is caused by the unique neutron-moderating characteristics of zirconium hydride. As neutrons gain energy from the spectrum hardening, the probability of their escape from the fuel element before being captured in the fuel is significantly increased. As a result, the ratio of neutron absorptions in the fuel to total absorptions in the core unit cell decreases as the temperature is increased, and this manifestation is termed the "cell effect". The standard TRIGA fuel elements employed in these reactors contain a homogeneous mixture of about 8.5 weight percent low-enriched uranium and about 91.5 weight percent zirconium hydride, wherein the uranium contains about 20 percent U-235 and about 80 percent U-238. In such reactors utilizing these standard TRIGA fuel elements, more than 50 percent of the prompt negative coefficient results from the phenomenon of thermal spectrum hardening whereas the remaining amount is contributed about equally by the other two factors. For applications where a long burn-up lifetime of fuel is deemed economically desirable, a fuel referred to as TRIGA-FLIP (Fuel Lifetime Improvement Program) was developed. This fuel is designed to be usable in the standard TRIGA reactors, as well as in other similar pool-type research reactors, and it uses 70 percent enriched uranium (i.e., 70 percent of the atoms were U-235). The FLIP fuel is a homogeneous mixture of about 8.5 weight percent uranium, about 1.6 weight percent erbium and the remainder zirconium hydride. The erbium is a strong contributor to the prompt negative temperature coefficient as a result of the interaction of its low energy resonances and the spectrum hardening effects of the zirconium hydride. It also serves as a burnable poison to compensate for the excess reactivity provided by the high-enriched uranium and thus maintains the reactivity balance of the fuel relatively flat throughout the lifetime of the overall reactor core loading. The TRIGA-FLIP fuel elements also demonstrate inherent safety characteristics, similar to the standard TRIGA fuel elements. In the FLIP fuel core, an even greater percentage of the prompt negative temperature coefficient is contributed by thermal spectrum hardening, i.e., an amount of more than 85 percent. The contribution to the coefficient by the Doppler effect decreases slightly, and the contribution resulting from the increase in thermal neutron leakage decreases by about 75 percent. Because these decreases are more than offset by the increase resulting from thermal spectrum hardening, the overall prompt negative temperature coefficient of a FLIP fuel core is equal to or slightly greater than that of a reactor operating with the standard TRIGA fuel core. The United States Government is presently pressing forward with non-proliferation policies which place limits upon the amount of enrichment that may be included within nuclear reactor fuel. In accord with these policies, 70 percent enriched fuel may not continue to be supplied. Accordingly, it is an object of the invention to provide an improved long-life reactor core for a pool-type reactor, such as the TRIGA, which utilizes low-enriched uranium but which also exhibits the desired prompt negative temperature coefficient. It is a further objective to provide fuel elements which can be employed in existing TRIGA reactors to provide a long-life fuel core without the incorporation of highly enriched uranium. It has been found that a pool-type reactor, such as a TRIGA reactor, can be provided with a long-life reactor core loading by fabricating nuclear fuel elements from a homogeneous mixture including between about 20 and about 50 weight percent of low-enriched uranium, which for purposes of this application is defined as having not more than about 20 percent enrichment. By homogeneously mixing this uranium with zirconium hydride and with a small amount of erbium, preferably between about 0.5 and about 1.5 weight percent, it has been found that the resulting core unexpectedly has a prompt negative temperature coefficient which compares very favorably with that of a reactor fueled with standard TRIGA fuel. Past experience had indicated that a reactor core of this proportion would have a sizeably reduced prompt negative temperature coefficient compared with a reactor incorporating the standard TRIGA fuel because: (1) the large increase in the amount of uranium-235 would drastically reduce the cell effect, i.e., the thermal neutron leakage from the fuel-moderator material to the surrounding coolant, (2) the larger amount of uranium would considerably reduce the total amount of zirconium hydride in the fuel elements and thus the amount of hydrogen, the moderating effect of which is one of the major factors in the mechanism for creating a large prompt negative temperature coefficient, and (3) the decrease of the amount of hydrogen in the fuel mixture also decreases the overall reactivity of the core assembly and thus would substantially lessen the amount of erbium that could be added to contribute to the prompt negative temperature coefficient, thereby offsetting some of the losses incurred by items (1) and (2), and to offset excess reactivity. |
abstract | The reactor core includes at least one module, a solid neutron moderator, and a liquid neutron moderators. Each module contains a casing, at least one heat pipe, one fuel element and thermal insulation. The heat pipe comprises a casing, a wick, and a coolant. The fuel element is made of nuclear fuel, arranged along an evaporation area of the heat pipe, around the heat pipe casing, in thermal contact with the heat pipe casing, and enclosed in a can. Liquid metals are used as the coolant of the heat pipe. Thermal insulation is arranged between the can and the module casing. At least one hole is made in the solid neutron moderator. Each module is arranged within a respective hole of the solid neutron moderator. The space between the module casing and the solid neutron moderator is filled with a liquid neutron moderator. |
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description | The present disclosure relates to a nuclear reactor and, more specifically, to fuel rods and fuel assemblies for a nuclear reactor core. The statements in this section merely provide background information related to the present disclosure and may not constitute prior art. In a nuclear reactor, a fissile fuel atom, such as U-235, absorbs a neutron in its nucleus that results in a nuclear disintegration which produces on the average two fission fragments of lower atomic weight with kinetic energy and several neutrons at high energy. In a typical nuclear reactor, fuel is in the form of fuel rods, each of which contains stacked sintered pellets of a nuclear fuel arranged within an elongated cladding tube. Each fuel rod can be of the same length or a different length. Typically, each fuel rod has a fuel enrichment distribution in the vertical/axial direction of the rod and is often designed for a uniform enrichment across the axial length of the rod. Groups of fuel rods are coupled together and often enclosed within a casing to form fuel bundles (also referred to as fuel assemblies). The fuel assemblies are placed within the reactor core and are supported between upper and lower core plates within the core. A plurality of fuel assemblies are arranged in a matrix to form the nuclear reactor core that is capable of a self-sustained fission reaction. The kinetic energy of the fission products is dissipated as heat in the fuel rods. Energy is also deposited in the fuel assemblies and moderated by the neutrons, gamma rays, and other radiation resulting from the fission process. During operation of the reactor, water (that serves both as a coolant and as neutron moderator) enters the bottom of the fuel assembly and flows upwards through the fuel assembly past the fuel rods. Heat is given off by the fuel rods and is taken up by the water which boils and is transformed into steam. The coolant (liquid and steam) rise upward through the upper portion of the fuel assembly and the steam exits the top of the fuel assembly where it is collected for delivery to a turbine for generating electrical energy. As the water and steam rise, the coolant reduces in liquid content and increases in steam content. At the upper portion of the fuel rods, the coolant is primarily steam content. This results in the fuel at the top of the fuel assembly not being utilized as efficiently in the generation of steam from the liquid as the fuel at the bottom of the fuel assembly. Additionally, the higher steam content at the top of the fuel assembly results in less cooling of the fuel rods by the coolant than at the lower portions where the ratio of liquid to steam is higher. If the heat from the fuel rod becomes excessive as compared to the available coolant at the top of the fuels rods, there is a risk of dryout. Additionally, the higher percentage pressure drop from bottom to top of the fuel assemblies and core increases the instability of the core. When there is a higher percentage of steam, the neutron moderation of the coolant is reduced because steam is an inferior moderator compared to water. During operation, the percentage of steam voids increases towards the top of the reactor, leading to decreased moderation in the top regions of the core and about the fuel rods and assemblies. As such, the power distribution within the reactor core is generally skewed toward the lower regions of the core. It is a known practice to compensate for this by distributing a burnable absorber in an axially inhomogeneous manner and to enrich the uranium in the middle and/or top axial portions of the core. A burnable absorber is a neutron absorber which is converted by neutron absorption into a material of lesser neutron absorbing capability. A number of the fuel rods are often provided with a burnable absorber with a distribution in the fuel rod skewed toward the axial region of hot operating maximum reactivity. A well-known burnable absorber is gadolinium, normally in the form of gadolinia. The burnable absorbers available for use in design have an undesirable end-of-refueling cycle neutron absorption reactivity residual due to residual isotopic neutron absorption by small neutron cross section absorbers. For example, if gadolinium is used as a burnable absorber, the high cross section isotopes (Gd-155 and Gd-157) deplete rapidly but residual absorption remains due to continued neutron capture in the even isotopes (Gd-154, Gd-156, and Gd-158). As such, the use of a burnable absorber is not the most effective or desirable method of reactor core design and results in fuel cycle inefficiencies. Additionally, when the reactor is in the cold shutdown condition, the top of an irradiated boiling water reactor core is more reactive than the bottom due to greater plutonium production at the top and less U-235 destruction in the top during operation (greater conversion ratio and smaller burnup occurs in the top of the core). In the cold shutdown condition, the steam voids in the upper part of the core are eliminated, thus making the top of the core more reactive than the bottom. As noted, axial power shaping within the core and fuel assemblies is traditionally provided by including greater amounts of burnable absorber in the lower portions of the reactor core. However, the optimum burnable absorber shaping for full power optimization to maintain a desired shutdown margin is not adequately maintained during cold shutdown. In order to meet cold shutdown margin objectives, it is typically necessary to design fuel assemblies with excess burnable absorber residual that penalizes the initial enrichment and uranium ore requirements, reduces fuel cycle efficiency and therefore increases the fuel cycle cost of the reactor. A further problem is that available burnable absorbers such as gadolinia reduce the thermal conductivity of the fuel rods and increases fission gas release. Consequently, the gadolinia-containing rods are frequently the most limiting rods in the fuel assembly, and have to be down-rated in power with a correspondingly adverse effect on local power distributions. The amount of power down-rating that is required depends on the gadolinia concentration, but becomes a serious problem in extended burn up fuel bundle designs and/or high energy cycle designs where increased gadolinia concentrations are required in order to provide adequate cold shutdown margins. Thus, the current methods of axial shaping of fuel assemblies by having higher enrichments in the middle or upper portions and using burnable absorbers in the lower portion have significant negative effects on obtaining optimal core configurations, on fuel cycle efficiencies and on operating costs of a nuclear reactor. The inventors hereof have succeeded at designing and producing complex axially varied enriched fuel rods for nuclear reactors. Various embodiments of this disclosure have demonstrated one or more of the following: improved fuel utilization and fuel cycle efficiency, reduced use of non-efficient burnable absorbers, increased plutonium generation at the top of the fuel bundle, reduced top of core leakage, improved fuel loading pattern flexibility, optimize cold shut down margins, increased critical power margins, increased Kw/ft margins, reduced total amounts of uranium and the average enrichment in fuel reloads, and/or reduced overall fuel cycle costs. According to one aspect, a fuel rod for a nuclear reactor includes a first axial zone positioned proximate to a bottom end, a second axial zone positioned adjacent to the first axial zone in an intermediate region, and a third axial zone positioned proximate to a top end. The first axial zone has an enrichment greater than the second axial zone and the second axial zone has an enrichment greater than or equal to the third axial zone. According to another aspect, a method of designing a fuel rod for a nuclear reactor includes specifying an enrichment for a first axial zone positioned proximate to a bottom end of a fuel rod to minimize a local peak power, determining an enrichment for a second axial zone positioned adjacent to the first axial zone in an intermediate region of the fuel rod, and determining an enrichment for a third axial zone positioned proximate to a top end of the fuel rod, wherein determining the enrichments for the second and third axial zones are to minimize an R-factor. According to yet another aspect, a fuel assembly for a nuclear reactor includes a plurality of fuel rods having one or more fuel rods including a first axial zone positioned generally at a bottom end, a second axial zone positioned adjacent to the first axial zone in an intermediate region, and a third axial zone positioned generally at a top end, wherein the first axial zone has an enrichment greater than the second axial zone and the second axial zone has an enrichment greater than or equal to the third axial zone. According to another aspect, a fuel assembly for a nuclear reactor includes a plurality of fuel rods wherein one or more fuel rods includes a first axial zone positioned generally at a bottom end, a second axial zone positioned adjacent to the first axial zone in an intermediate region, and a third axial zone positioned generally at a top end, wherein the first axial zone is configured to optimize a local peak power of the first axial zone and the second and third axial zones are configured to optimize an R-factor for the fuel assembly. According to yet another aspect, a method of designing a fuel assembly for a nuclear reactor includes specifying enrichments for a first axial zone of each of a plurality of fuel rods within the fuel assembly to optimize local peak power of the first axial zone, the first axial zone being positioned proximate to a bottom end of each fuel rod, and determining enrichments for a second axial zone and a third axial zone of each of a plurality of fuel rods within the fuel assembly to optimize an R-factor for the fuel assembly, the second axial zone being positioned adjacent to the first axial zone in an intermediate region of each fuel rod, and the third axial zones being positioned proximate to a top end of each fuel rod. According to still another aspect, a method of manufacturing a fuel rod includes filling a tube with a plurality of first fuel pellets having a first enrichment to a first axial zone length to form a first axial zone positioned generally in a bottom end of the tube and filling the tube with a plurality of second fuel pellets having a second enrichment to a second axial zone length to form a second axial zone with the second enrichment being less than the first enrichment. The method also includes filling the tube with a plurality of third fuel pellets having a third enrichment to a third axial zone length to form a third axial zone with the third enrichment being less than or equal to the second enrichment. According to still another aspect, a method of manufacturing a fuel bundle for a nuclear reactor wherein the fuel bundle includes a plurality of fuel rods, the method including filling a first tube with a plurality of fuel pellets having a first enrichment to a lower zone length to form a lower axial zone positioned generally in a lower portion of the tube, filling the first tube with a plurality of fuel pellets having a second enrichment to an intermediate axial zone length to form an intermediate axial zone, the second enrichment being less than the first enrichment, and filling the first tube with a plurality of fuel pellets having a third enrichment to an upper axial zone length to form an upper axial zone, the third enrichment being less than or equal to the second enrichment; and filling a second tube with a plurality of the fuel pellets having the first enrichment to a second lower axial zone length to form a second lower axial zone positioned generally in a lower portion of the second tube, a plurality of fuel pellets having the second enrichment to a second intermediate axial zone length to form a second intermediate axial zone, and a plurality of fuel pellets having a third enrichment to a second upper axial zone length to form a second upper axial zone, wherein the second intermediate axial zone length is greater than the first intermediate axial zone length. Further aspects of the present invention will be in part apparent and in part pointed out below. It should be understood that various aspects of the disclosure may be implemented individually or in combination with one another. It should also be understood that the detailed description and drawings, while indicating certain exemplary embodiments, are intended for purposes of illustration only and should not be construed as limiting the scope of the disclosure. It should be understood that throughout the drawings, corresponding reference numerals indicate like or corresponding parts and features. The following description is merely exemplary in nature and is not intended to limit the present disclosure or the disclosure's applications or uses. As will be described herein, integrated axial varying enrichment (IAVE) includes fuel rod and fuel bundle design and manufacturing processes where the enrichment distribution changes in the axial direction at multiple elevations or lengths. Traditionally, those skilled in the art have believed that optimal performance of the nuclear reactor was obtained when the top or highest portions of the fuel rods had the highest levels or equivalent levels of enrichment. However, the current inventors have succeeded in designing fuel rods and fuel assemblies with improved operational characteristics through a new integrated axial varying enrichment (IAVE) design having the highest levels of enrichment in the bottom or lower axial portions of the fuel rods and fuel assemblies. The inventors have demonstrated that various embodiments of the present disclosure have provided improved operating characteristics for fuel rods, fuel assemblies and nuclear reactor cores. For example, their efforts have indicated that by utilizing one or more embodiments of the invention as described herein, core enrichment deltas can be obtained when required for reactor designs that have at least 0.2 between the bottom enrichment zone and a middle enrichment zone, and between the middle enrichment zone and a top enrichment zone. The length or height of each of the enrichment zones is varied in designing fuel rods and fuel assemblies for optimizing fuel cycle performance of the reactor and/or a shaping or smoothing of the enrichment across all or a portion of the axial length of the fuel bundle or core. Additionally, by incorporating one or more embodiments of the fuel rods in a fuel bundle and core design, the manufacturing of the fuel rods and fuel bundles can be simplified and therefore result in lowered manufactured costs. This can be achieved through the enabled design improvements that can utilize fuel rods with a fewer number of enrichment zones that have integrated axial varying enrichments. For example, in many embodiments the number of enrichment zones and variety of enrichment zones can be significantly reduced over prior practice. In some cases, the varying of heights or zone lengths in the reduced number of enrichment zones can provide sufficient design variations for most, if not all, fuel bundle and core design requirements. Fuel bundles can be designed and manufactured to have several different effective enrichment zones by varying one or more zone lengths of one or more rods within the fuel bundle and by producing a subset of rod zones. As a result, complex fuel bundle enrichment profiles can be designed and produced with significantly simplified individual fuel rods. In some embodiments, a bottom end axial zone and/or a top axial zone have enrichments equal to the second and/or third axial zones to produce fuel rods or fuel bundles having high enrichment, or can utilize natural uranium to form fuel rods or fuel bundles having a upper or lower section of low enrichment. The fuel rods and/or fuel bundles as described herein can provide, in some embodiments, fuel bundles with improved local peakings and R-factors, relating to improvements in Kw/ft and critical power (CPR) margins. As known, optimization or improvement of local peakings and R-factors generally include minimizing the local peakings and minimizing the R-factors. For example, fuel bundles consistent with one exemplary embodiment as described herein have demonstrated local peakings that are as low as 1.14 in the lower part of the bundle, also known as the Power Shaping Zone (PSZ). Additionally, other exemplary fuel bundles according to other embodiments have demonstrated R-factors as low as 0.93. As known to those skilled in the art, these levels of local peaking and R-factors are lower than generally considered possible. As a result, a core designed with fuel rods and fuel bundles as described in this disclosure are capable of producing improved kilowatt per foot (Kw/ft) margins and critical power ratio (CPR) margins as compared to fuel rods and fuel bundles having equivalent level of enrichment and therefore fuel cost. As a result, one or more embodiments as described herein can provide for improved fuel cycle efficiency for a nuclear reactor. Additionally, in some embodiments an end of cycle shape can be lower and less peaked as compared to previous designs. As a result, lower operating limit minimum critical power ratios (OLMCPR) are produced and therefore additional CPR margin can be obtained. This can provide for additional improved fuel cycle efficiency and operational flexibility. Referring now to FIG. 1, an exemplary embodiment of an operating environment for fuel rods and fuel assemblies of the present disclosure is illustrated in the sectional view, with parts cut away, of a boiling water nuclear reactor pressure vessel (RPV) 10. The reactor pressure vessel 10 has a generally cylindrical shape and is closed at one end by a bottom head 12 and at its other end by a removable top head 14. A side wall 16 extends from bottom head 12 to top head 14. The side wall 16 includes a top flange 18 on which the top head 14 is attached. A cylindrically shaped core shroud 20 surrounds a reactor core 22. The shroud 20 is supported at one end by a shroud support 24 and includes an opposed removable shroud head 26. An annulus 28 is formed between the shroud 20 and the side wall 16. In some embodiments, a pump deck 30, which has a ring shape, extends between the shroud support 24 and the side wall 16. The pump deck 30 includes a plurality of circular openings 32, with each opening housing a jet pump 34. The jet pumps 34 are circumferentially distributed around the core shroud 20. In other embodiment, pumps 34 are not required as the water flow is created non-mechanical methods as are known in the art. Heat is generated within the core 22, which includes fuel cells 40. Each fuel cell 40 includes one or more fuel bundles 42, and in one embodiment contains four fuel bundles 42, by way of example. Each fuel bundle 42 includes a plurality of fuel rods (not shown) of fissionable material and a control rod 44 associated with it. The fuel rods are arranged in a matrix such as a 10×10 matrix. The fuel bundles 42 are typically mounted on top of a core plate 46 in a boiling water reactor from which the fuel bundles 42 receive water and from which the control rods 44 are inserted and withdrawn from the core 22 and between one or more of the fuel bundles 42. The amount of heat generated in the core 22 is regulated by inserting and withdrawing a plurality of control rods 44 of neutron absorbing material, for example, hafnium. To the extent that a control rod 44 is inserted next to a fuel bundle 42, it absorbs neutrons that would otherwise be available to promote the chain reaction which generates heat in core 22. Control rod guide tubes 48 are located beneath the core plate 46 and receive the control rods 44 when withdrawn from the core 22. The control rod guide tubes 48 restrict non-vertical motion of the control rods 44 during insertion and withdrawal. Water is circulated up from beneath the core plate 46 through the fuel bundles 42 and is at least partially converted to steam by heat generated by the nuclear reaction within the fuel rods of the fuel bundles 42. Steam separators 50 separate the steam from the water, and the water is recirculated. Steam dryers 52 remove residual water from the steam. The steam exits reactor pressure vessel 10 through a steam outlet 54 near vessel top head 14. Referring now to FIG. 2, a simplified core 56A is illustrated having a plurality of fuel cells 40, bundles 42 and control rods 44 arranged as a matrix within the core 22. It should be understood that the core 56 is simplified as a typical core 22 can contain more than 700 fuel bundles 42, more than 95,000 fuel rods, and more than 190 control rods. In FIG. 3, a fuel cell 40 is illustrated in a partly cut away isometric view and having four fuel bundles 42 and a control rod 44. Each fuel bundle 42 includes a plurality of fuel rods 58 enclosed within a fuel bundle casing 60. One simplified illustration of a fuel bundle 42 is illustrated in FIG. 4 having a 10×10 matrix or lattice of fuel rods 58 within fuel casing 60. The fuel rods 58 within the fuel bundle 42 can be defined by their matrix or lattice position within the fuel bundle (such as by the lettered columns and numbered rows), by their relative position (such as indicated by those in the matrix having a radial position 61 about the perimeter of the fuel rod matrix), and/or by their configuration. The fuel bundle 42 can also include one or more water channels 59. Referring again to FIG. 3, the fuel bundle 42 includes a handle 62 located on the top of the fuel bundle 42 to aid in fuel bundle 42 insertion and removal from the core 22. Each fuel bundle 42 includes a lower tie plate 64 on which the fuel rods 58 are mounted, and a nose 66 having openings 68 through which coolant is received to flow upward within the fuel bundle casing 60 and around the fuel rods 58 as described above. The control rod 44 includes a plurality of control blades 70 each of which is configured to interpose between four adjacent fuel bundles 42 for controlling nuclear reaction therebetween. As shown in FIG. 3, each fuel rod 58 within a fuel bundle 42 has a length that typically is in the range of about 150 inches, and can include lengths about equal to 120 inches, 133 inches, 138 inches, 139 inches, and 145 inches, by way of examples. As such, each fuel rod has a bottom end mounted on the lower tie plate 64, a top end positioned on the other end of the fuel rod, and an intermediate portion positioned therebetween. The length from the bottom end to the top end defines the axial dimension of the fuel rod 58 that is typically referred to as the axial position relative to the bottom end. Each fuel rod 58 within the fuel bundle 42 can have a different enrichment or gadolinium doping distributed along the axial dimension of the fuel assembly as defined by the stacked sintered fuel pellets arranged within an elongated fuel bundle channel 60 from the bottom end to the top end. Various contents of fuel pellets with various levels of enrichment or other material within the fuel rod 58 are generally referred herein as an axial zone. In one exemplary embodiment, a fuel rod for a nuclear reactor includes a first axial zone positioned proximate to a bottom end, a second axial zone positioned adjacent to the first axial zone in an intermediate region, and a third axial zone positioned proximate to a top end. The first axial zone has an enrichment greater than the second axial zone and the second axial zone has an enrichment greater than or equal to the third axial zone. Such fuel rods can be limited to having only three axial zones in some embodiments (such as illustrated in FIG. 5) and in other embodiments may have four, five, or more axial zones (also shown in FIG. 5, by way of examples), with a combination of enrichments, burnable absorbent or other dopings, and/or natural uranium. As shown in FIG. 5, each of twelve fuel rods has a different height of first axial zone Z1, second axial zone Z2, third axial zone Z4, etc. Fuel rods A-F, I, K and L have enrichments (not shown) while fuel rods G, H, and J are denoted as including a burnable absorbent as indicated by the cross-latching. As noted in the various fuel rod designs of FIG. 5, the total number of axial zones of each fuel rod is limited, but the number of axial zones in each rod and the height of each axial zone can vary. Additionally, while not explicitly shown in FIG. 5, the enrichment in each zone can also vary, consistent with the first axial zone having the highest enrichment of the axial zones within each fuel rod. In a first exemplary embodiment, a first axial zone has a greater enrichment than the second axial zone. One of the advantages of the embodiment is that the axial shape is maintained at the bottom of the core to provide improved fuel efficiency and better pressurization transients. In a second exemplary embodiment, the second axial zone has a greater enrichment than the third axial zone. One advantage of this embodiment is that less leakage occurs at the top of the core for providing improved fuel efficiency. In a third exemplary embodiment, the first axial zone has an enrichment greater than the second and third axial zones. In this design, the fuel assembly provides for improved local peaking factors and R-factors, and therefore, improved fuel cycle efficiency. In other embodiments, the second and third axial zones have substantially equivalent enrichments. By these three examples, it can be seen that by utilizing one or more embodiments of the integrated axially varying enrichment profiles as described in the present disclosure, various design objectives and fuel cycle efficiencies are achievable through cost effective fuel bundles having various profiles. These can be achieved by varying the number of axial zones within a limited number of variations, varying the enrichments of the axial zones with the lower first axial zone having a greater enrichment than higher axial zones and varying the zone lengths in various rods within fuel bundles. As described above, the zone height relative to the bottom end or zero height indication can be varied in addition to the number of zones. It should be understood that the description and recitation of three, four or five axial zones is not intended to limit the existence of one or more additional axial zones within a fuel rod or the location within a fuel rod unless specifically noted within one or more of the described exemplary embodiments. However, as will be noted in further detail below, in some embodiments, a substantial portion or all fuel rods within a fuel bundle have the same number of axial zones, with each axial zone in each rod being filled with pellets having the same enrichment. In such cases, complex fuel bundles are configured only by changes in the zone heights between fuel rods and the position of the various fuel rods within the fuel bundle. In other embodiments, the number of zones per rod and the enrichment and/or burnable absorbent in one or more zones can also be varied to produce complex fuel bundles. For example, in some embodiments a bottom end axial zone is positioned at the bottom of the fuel rod between the bottom end and the first axial zone. This bottom end axial zone can include natural uranium (not enriched) or can include an enrichment equal to, or less than the first axial zone. Also, in some embodiments a top end axial zone is positioned at the top of the fuel rod. This top end axial zone can include natural uranium (not enriched) or can include an enrichment equal to, or less than the first axial zone. As noted above one or more additional axial zones of enrichment, burnable absorber dopings, or natural uranium can be included in one or more positions or locations on the fuel rod. However, in one embodiment a fuel rod has no more than three enrichment zones and in another embodiment a fuel rod has no more than five enrichment zones, with two of the five being the bottom end axial zone and the top end axial zone. Similarly, the enrichment of the axial zones and the length of each zone can also be customized for desired nuclear reactor core designs and operations. For example, in one embodiment the enrichment and/or length of the first axial zone can be configured to optimize a local peak power (typically in Kw/ft) of the fuel rod or a peak power or peaking factor of a fuel bundle in which the fuel rod is assembled. The zone length for the first axial zone can be a minor or substantial portion of the total length of the fuel rod and in one embodiment is about one third of the total length of the fuel rod. In another exemplary embodiment, the enrichment and/or a length (sometime referred herein to as the height) of one or both of the second and third axial enrichment zones are dimensioned to optimize an R-factor (e.g., critical power ratio or CPR), a power distribution profile, and/or a local peak power of the fuel rod or the fuel bundle in which the fuel rod is assembled. The enrichments in each of the first, second, and/or third axial zones can be uniformly distributed axially within each zone, or can be tapered or otherwise shaped for fuel rod and fuel bundle design requirements and performance objectives. In another embodiment, a method of designing a fuel rod for a nuclear reactor includes specifying an enrichment for a first axial zone positioned proximate to a bottom end of a fuel rod to minimize a local peak power, determining an enrichment for a second axial zone positioned adjacent to the first axial zone in an intermediate region of the fuel rod, and determining an enrichment for a third axial zone positioned proximate to a top end of the fuel rod, wherein determining the enrichments for the second and third axial zones are to minimize an R-factor. This can include specifying the enrichment for the first axial zone to specify first enrichment that is greater than the enrichment for the second axial zone and/or the third axial zone. Additionally, this can include determining the enrichment of the third axial zone that is less than or equal to the enrichment of the second axial zone. Also in some embodiments, the zone length for the first axial zone is determined to optimize the local peak power, and one or both of the zone length for the second axial zone and the zone length for the third axial zone are each determined to optimize the R-factor as described above. As noted above, one or more embodiments of the fuel rods as described herein can be assembled into a fuel bundle (also referred herein as a fuel assembly) and one or more fuel bundles can be arranged to form a core in a nuclear reactor. Each fuel bundle includes a plurality of fuel rods with one or more of the fuel rods having a first axial zone positioned generally at a bottom end, a second axial zone positioned adjacent to the first axial zone in an intermediate region, and a third axial zone positioned generally at a top end. As noted above, the first axial zone has an enrichment greater than the second axial zone and the second axial zone has an enrichment greater than or equal to the third axial zone. Also as noted above, one or more of the fuel rods within the fuel assembly can have various enrichments that can be determined and/or customized for various optimizations. For example, one or more axial zones of one or more rods within the fuel bundle can include an enrichment or include zone lengths that provides for an optimized R-factor and/or local peak power of the rod, the fuel bundle and/or the core. Additionally, as noted above the lengths of the axial zones in one or more fuel rods within the fuel bundle can be varied to provide for improved and/or controlled operational characteristics. By way of example, the zone length of the first axial zone can be equal to about one-third of the total length of the fuel rod in one or more of the fuel rods of the fuel assembly. In other fuel rods in the fuel bundle, the length of the first axial zone can be more or less than the one-third length. For example, in some embodiments, a fuel assembly for a nuclear reactor includes a plurality of fuel rods wherein one or more fuel rods includes a first axial zone positioned generally at a bottom end, a second axial zone positioned adjacent to the first axial zone in an intermediate region, and a third axial zone positioned generally at a top end, wherein the first axial zone is configured to optimize a local peak power of the first axial zone and the second and third axial zones are configured to optimize an R-factor for the fuel assembly. Similarly, in some embodiments a method of designing a fuel assembly for a nuclear reactor includes specifying enrichments for a first axial zone of each of a plurality of fuel rods within the fuel assembly to optimize local peak power of the first axial zone, the first axial zone being positioned proximate to a bottom end of each fuel rod, and determining enrichments for a second axial zone and a third axial zone of each of a plurality of fuel rods within the fuel assembly to optimize an R-factor for the fuel assembly, the second axial zone being positioned adjacent to the first axial zone in an intermediate region of each fuel rod, and the third axial zones being positioned proximate to a top end of each fuel rod. As discussed above with regard to a single fuel rod, the enrichments and zone lengths for a plurality of the fuel rods within the fuel assembly can be specified and determined to meet objectives for the fuel assembly. This can include determining an enrichment for each second axial zone and each third axial zone of each fuel rod with the fuel assembly such that both have a lower enrichment for the first axial zone within each corresponding fuel rod and in some cases across all of the fuel rods within the assembly. In other embodiments, the enrichment for each third axial zone of each fuel rod can be determined such that the enrichment of one or both are less than the first fuel rod and such that in some embodiment, the enrichment of the third axial zone is less than or equal to the enrichment for the second axial zone in each corresponding fuel rod and in some cases across all of the fuel rods within the fuel assembly. This can also include determining a zone length for each of the first axial zones of each fuel rod to optimize the local peak power of the first axial zone and determining a zone length for each of the second axial zones and each of the third axial zones of each fuel rod to optimize the R-factor for the fuel assembly. In one embodiment, the enrichments for each second axial zone and each third axial zone are determined after the enrichments for each first axial zone for all of the fuel rods within the fuel assembly. For example, the first axial zone within the fuel assembly can be designed for minimizing the local peak power and then the second and third axial zones of the fuel rods and the fuel assembly are designed to minimize the R-factor for the entire fuel assembly, taking into account the previously minimized local peak power in the first or lower portion of the fuel assembly. This can also include increasing the enrichment of the second axial zones and the third axial zones in some of the fuel rods, such as the fuel rods positioned about the perimeter or edge of the fuel assembly. And in some embodiments, this can include reducing the enrichments for the second and third axial zones of fuel rods positioned within a center portion (other than the edge) of the fuel assembly. As known to those skilled in the art from this discussion, not all fuel rods within a fuel bundle or fuel bundles within a core are required to have the same configuration, such having the same enrichments for each zone, or same length of each zone, or same number of zones per fuel rod. For example, in one fuel bundle all or a substantial portion of the fuel bundles can be limited to the same number of axial zones (such as three, four, or five) which provides manufacturing advantages. However, the enrichment, burnable absorber doping, and/or length of each axial zone between one or more of the fuel rods can be different. In this manner, manufacturing costs can be reduced and reactor core designer's have the ability to design a core to a preferred design or to desired operational characteristics by per fuel rod, per fuel bundle, and core design adjustments. In one exemplary operation, each of a plurality of fuel bundles is configured to have different local peaks and different R-factors. In two or more of the fuel bundles, a substantial portion of the fuel rods have essentially three axial enrichment zones (first second and third axial zones). This may include a nominal bottom end zone at the bottom and a nominal top end zone at the top. A substantial portion of the first zones in each fuel rod has first fuel pellets of essentially equal enrichment, a substantial portion of the second zones in each fuel rod has fuel pellets of essentially equal enrichment, and a substantial portion of the third zones in each fuel rod has fuel pellets of essentially equal enrichment. The enrichment of each first fuel pellets is greater than the second fuel pellets and the third fuel pellets. A plurality of the fuel rods in the fuel bundle can have different zone lengths or heights for the first, second and third zones. By only changing the zone lengths in the plurality of fuel rods with each fuel bundle and the location of each of such axially varying enrichments, two or more of the fuel bundles have different local peak powers and different R-factors. As such, a designer of the fuel matrix can create a desired or optimized core from a variety of complex axially varying enrichment bundles through using axially varying enrichment fuel rods that are otherwise similarly composed and manufactured. Additionally, the fuel rods within the fuel assembly can be optimized based on their position within the fuel assembly. For example, edge rods are typically the rods positioned along or about a perimeter of the matrix of fuel rods within a fuel assembly. In some embodiments, only the first axial zones of the radial fuel rods are optimized for local peak power. The first axial zones of the other fuel rods are separately enriched and/or configured, but not for optimizing local peak power. Similarly, in some embodiments, only the second axial zones and/or third axial zones of the radial rods are optimized for R-factor or critical power ratio (CPR). The second and third axial zones in the other rods within the fuel assembly have different levels of enrichment and different zone lengths, but are not optimized for their R-factor. In another exemplary embodiment, one or more fuel rods of a fuel assembly can include two or more groups of fuel rods each having a different combination of axial zones, enrichments, optimized factors such a local peak power and R-factor, by way of example, and/or axial zone lengths for customizing the design and characteristics of the fuel assembly. Each group can be composed of a subset of the fuel rods and possibly one or more having a substantial portion of the total number of fuel rods within the fuel bundle. Referring now to FIG. 6, in one exemplary embodiment a first bundle (denoted as bundle 1) includes 22 fuel bundle elements, two being water channels W1 and W2 and 20 different types of fuel rods. Each of the 20 different fuel rods in this example includes a bottom axial zone of naturally enriched uranium of 0.71. Additionally, many of the fuel rods include a top axial zone also including naturally enriched uranium of 0.71. In some of the fuel rods, a top zone includes natural uranium as indicated by the designation “0.71.” Each axial zone includes the amount of uranium enrichment as indicated by the number within the zone. In some fuel rods, an amount of gadolinium doping is also indicated by a “G” such as G4.0. As shown, most fuel rods are full length rods of 150 inches. However, two of the fuel rods are partial length rods F9 and F10, each of which includes only a first axial zone with a high enrichment of 4.8 and 5.0, respectively. Note that the top portion of these partial length rods are denoted with a “V” and referred to by those skilled in the art as vanished rods. These 20 types of fuel rods are utilized to form the fuel bundle as shown in FIG. 7. Each fuel rod and its position within the fuel bundle are designed and assembled to achieve the design objectives of the core. Other fuel bundle designs and design objectives can be obtained by using the same 20 fuel rods as illustrated in FIG. 6. However, in some embodiments variations to one or more of the fuel rods and additional fuel rod design may be desired or required for other fuel bundles within a core design in order to achieve the desired objectives. As shown in FIG. 6, a select number of fuel rods (those indicated at the top with the asterisk) can be modified through changes in the number of axial zones, the enrichment or doping of the axial zones, and/or the zone lengths to achieve different bundle designs. FIG. 8 illustrates one set of example fuel rod axial zone variations for seven different fuel bundles (one through seven) for use in an exemplary core 56B as illustrated in quarter core format in FIG. 9. The entire core of FIG. 9 is composed of these seven fuel bundles, each of which is composed of the fuel rods of FIGS. 6 and/or variations as shown in FIG. 8. FIG. 8 illustrates variation in or modifications to the make up and length of the axial zones of fuel rods, X1, X4, X6, X7 and X9, with each variation being indicated by the primes and double primes at the top of each modified fuel rod. The bundles within the core containing each fuel rod makeup are indicated at the bottom of each fuel rod. In FIG. 8, the fuel rod X1 is shown having a first axial zone with an enrichment of 4.8 and a gadolinium doping of 4.0. The second axial zone has a lower enrichment of 4.4 without gadolinium doping. The X1 fuel rod is utilized in fuel bundles 1-3, and 5-7. However, in fuel bundle 4, a modified X1A fuel rod includes a segmentation of the first axial zone to include a higher gadolinium doped portion or zone in the middle of the fuel rod. A similar modification to the fuel rod X4 is shown as fuel rod X4A, but where the gadolinium doping is held constant but the uranium enrichment is reduced from a first axial zone to a second axial zone located in the middle of the fuel rod of bundle 4. As indicated in FIG. 8, the fuel rod X6 (from FIG. 6) is only in fuel bundle 1. The modified fuel rod X6A is used in fuel bundle 4 and the modified fuel rod X6B is in fuel bundles 2, 3, and 5-7. In fuel rod X6A, one of the enrichment zones is eliminated and in fuel rod X6B gadolinium doping is removed from the first axial zone and the enrichment of the second axial zone is reduced. The fuel rod X7 is in fuel bundles 1-3, but is modified as X7A for use in fuel bundle 4 and as X7B for use in fuel bundles 5-7 as shown. The fuel rod X9 is in fuel bundles 1, 2, and 4-6, but is modified for use in fuel bundles 3 and 7 by adding in fuel rod X9A a third axial zone and adjusting the lengths of the first and second axial zones. As noted above, some fuel rods may only be configured for use in a subset of the core such as in FIG. 9. As illustrated in FIG. 8, fuel rod X10 is only in fuel bundle 4, and fuel rods X10A and G15 are only in fuel bundles 3 and 7. Each of these includes variations of enrichment, number of axial zones, and gadolinium doping levels according to various embodiments of this disclosure. The core 56B design as shown in FIG. 9 that is configured from the various fuel rods and fuel bundles as described in FIGS. 6-8 can provide for improved axial power axial nodal performance. For example, as illustrated in the chart 100 of axial power axial nodal to exposure in FIG. 10, line 102 illustrates the axial power at 200 megawatt days per standard ton of uranium (MWD/ST) and line 104 illustrates the axial power at 5,000 MWD/ST. Both of these reflect a peak around axial node 4, or very low within the core. As noted above, improved fuel cycle efficiency results from such an axial power axial nodal distribution low in the core. Additionally, with additional exposure as illustrated by line 106, at 12,500 MWD/ST) the peak of the axial power rises from axial node 4 to about axial node 14, or around the axial middle of the core. This is still considered to be relatively low in the core especially at 12,500 MWD/ST. As known to those skilled in the art, such a lowered axial power peaking over such exposures have not been herebefore attainable and provide a number of signification advantages as noted elsewhere and known to those skilled in the art. The various fuel rods as described by the exemplary embodiments can be manufactured by a variety of methods. For example, in one exemplary method of manufacturing a fuel rod such as a tube having a bottom end, a top end, and an intermediate region located between the bottom end and the top end is filled with a plurality of first fuel pellets having a first enrichment to a first axial zone length to form a first axial zone positioned generally in the bottom end of the tube. The first zone can have any length and in one embodiment is about equal to a third of the total length of the tube. The tube can also be filled with a plurality of second fuel pellets having a second enrichment to a second axial zone length to form a second axial zone with the second enrichment being less than the first enrichment. The method also includes filling the tube with a plurality of third fuel pellets having a third enrichment to a third axial zone length to form a third axial zone with the third enrichment being less than or equal to the second enrichment. In some embodiments, the method includes filling the tube with bottom end fuel pellets before filling the tube with first fuel pellets to form a bottom end axial zone positioned adjacent to the bottom end of the tube and between the bottom end and the first axial zone and having a bottom end zone length. Additionally, in some embodiments the tube can be filled with top end fuel pellets after filling the tube with third fuel pellets to form a top end axial zone positioned proximate or adjacent to the top end of the tube and having a top end zone length. One or more of the fuel bundles described herein can be manufactured by a variety of methods and still be in the scope of the disclosure. For example, in one exemplary method of manufacturing, a fuel bundle for a nuclear reactor is assembled by filling a first tube with a plurality of fuel pellets having a first enrichment to a lower zone length to form a lower axial zone positioned generally in a lower portion of the tube, filling the first tube with a plurality of fuel pellets having a second enrichment to an intermediate axial zone length to form an intermediate axial zone, the second enrichment being less than the first enrichment, and filling the first tube with a plurality of fuel pellets having a third enrichment to a upper axial zone length to form an upper axial zone, the third enrichment being less than or equal to the second enrichment; and filling a second tube with a plurality of the fuel pellets having the first enrichment to a second lower axial zone length to form a second lower axial zone positioned generally in a lower portion of the second tube, a plurality of fuel pellets having the second enrichment to a second intermediate axial zone length to form a second intermediate axial zone, and a plurality of fuel pellets having a third enrichment to a second upper axial zone length to form a second upper axial zone. The second intermediate axial zone length is greater than the first intermediate axial zone length. As noted above, the lower axial zone lengths of the first and second tubes can be dimensioned as a function of optimizing a peak power of the fuel bundle and the axial zone lengths of the intermediate and upper axial zones for the first and second tubes are each dimensioned as a function of optimizing an R-factor of the fuel bundle. In some embodiments, a variety of different zone axial zone lengths for axial zone 1, zone 2 and/or zone 3 can be varied during fuel rod and fuel bundle assembly for a variety of portions of the fuel rods within a fuel assembly. This method of manufacturing can provide for reduced manufacturing costs for fuel rods and fuel assemblies while also providing a core for nuclear reactors having improved fuel cycle efficiency. For example, in one embodiment the inventors were successful at designing fuel bundles having bundle enrichments 0.10 below reference bundle average enrichments. This corresponds to a significant reduction in the cost of uranium per bundle and a large corresponding fuel cycle efficiency improvement. Similar or better efficiency improvements are expected from other embodiments as described herein. Where one or more fuel rods includes the bottom end or bottom axial zone, the method of manufacturing can also include filling the first tube and the second tube with a plurality of bottom fuel pellets to a bottom axial zone length or height before filling the lower axial zones. The bottom fuel pellets can include natural uranium and or enriched bottom fuel pellets having an enrichment about equal to the first enrichment, by way of example. The method can also include filling the first and second tubes with a plurality of top fuel pellets after filling the upper axial zones. In such embodiments, filling with top fuel pellets is to a top end axial zone length or height to form a top axial zone positioned proximate to the top of the tube. The first lower axial zone can have a length about equal to a length of the second lower axial zone, the first intermediate axial zone has a length greater than a length of the second intermediate axial zone, and the first upper axial zone has a length less than the length of the second upper axial zone. In some embodiments, the method further includes filling a third tube with a plurality of the fuel pellets having a bottom end enrichment to a third lower axial zone length to form a third lower axial zone positioned generally in a lower portion of the third tube, a plurality of fuel pellets having a top end enrichment to a third intermediate axial zone length to form a third intermediate axial zone, and a plurality of fuel pellets having a sixth enrichment to a third upper axial zone length to form a third upper axial zone. Various enrichment combinations can include the first enrichment being greater than the bottom end enrichment, the second enrichment being greater than the top end enrichment; and/or the third enrichment being greater than the sixth enrichment, by way of examples. When describing elements or features and/or embodiments thereof, the articles “a”, “an”, “the”, and “said” are intended to mean that there are one or more of the elements or features. The terms “comprising”, “including”, and “having” are intended to be inclusive and mean that there may be additional elements or features beyond those specifically described. Those skilled in the art will recognize that various changes can be made to the exemplary embodiments and implementations described above without departing from the scope of the disclosure. Accordingly, all matter contained in the above description or shown in the accompanying drawings should be interpreted as illustrative and not in a limiting sense. It is further to be understood that the processes or steps described herein are not to be construed as necessarily requiring their performance in the particular order discussed or illustrated. It is also to be understood that additional or alternative processes or steps may be employed. |
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claims | 1. A ventilated system for storing high level radioactive waste comprising:a below-grade storage assembly comprising:an air-intake shell forming an air-intake downcomer cavity and extending along a central axis;a plurality of storage shells surrounding the air-intake shell in a side-by-side relationship, each storage shell forming a storage cavity and extending along a central axis, each of the storage shells comprising a sidewall having a first opening, a second opening, and a third opening each of which provides a passageway into a bottom of the storage cavity;for each storage shell:a primary air-delivery pipe extending from the air-intake shell to the first opening in the sidewall, the primary air-delivery pipe forming a primary air-delivery passageway that extends along a substantially linear axis from a bottom of the air-intake downcomer cavity to the bottom of the respective storage cavity, wherein the entirety of each of the primary air-delivery passageways is distinct from the entireties of all other of the primary air-delivery passageways of the below-grade storage assembly, and wherein the substantially linear axis of the primary air-delivery pipe intersects the central axis of the air-intake shell; anda first secondary air-delivery pipe extending from the second opening in the sidewall of the storage shell to one of the openings in the sidewall of a first adjacent one of the storage shells and a second secondary air-delivery pipe extending from the third opening in the sidewall of the storage shell to one of the openings in the sidewall of a second adjacent one of the storage shells, each of the first and second secondary air-delivery pipes forming a secondary air-delivery passageway that extends along a substantially linear axis between the bottom of the storage cavity of the storage shell and the bottom of the storage cavity of one of the first and second adjacent storage shells;a hermetically sealed container for holding high level radioactive waste positioned in one or more of the storage cavities;a lid positioned atop each of the storage shells and comprising at least one air-outlet passageway; andwherein for each storage cavity in which one of the hermetically sealed containers is positioned, a bottom end of the hermetically sealed container is located at an elevation above an uppermost end of the primary air-delivery passageway for that storage cavity. 2. The ventilated system according to claim 1 wherein the substantially linear axis of each of the primary air-delivery pipes is substantially perpendicular to the central axis of the air-intake shell. 3. The ventilated system according to claim 1 wherein the central axis of the air-intake shell and the central axis of each of the storage shells are substantially vertical, and wherein each of the primary air-delivery passageways are located within the same horizontal plane. 4. The ventilated system according to claim 1 wherein the secondary air-delivery passageways and the storage cavities of the plurality of storage shells collectively form a fluid-circuit loop, wherein the entirety of the fluid-circuit loop is independent of the entirety of all of the primary air-delivery passageways of the below-grade storage assembly. 5. The ventilated system according to claim 1 wherein for each storage cavity, there are at least three air-delivery passageways leading from the air-intake cavity to the storage cavity, wherein the entirety of each of the three air-delivery passageways is distinct from the entireties of the other two air-delivery passageways. 6. The ventilated system according to claim 1 wherein the below-grade storage assembly is hermetically sealed to the ingress of below-grade fluids. 7. The ventilated system according to claim 1 wherein at least two of the hermetically sealed containers are positioned in each of the storage cavities in a stacked arrangement. 8. The ventilated system according to claim 1 wherein each of the storage cavities has a transverse cross-section that accommodates no more than one of the containers. 9. The ventilated system according to claim 1 further comprising an enclosure forming an enclosure cavity, the below-grade storage assembly positioned within the enclosure cavity such that the air-intake shell and the storage shells extend though a roof slab of the enclosure, wherein the enclosure comprises a floor slab, the below-grade storage assembly positioned atop and secured to the floor slab, and the ventilated system further comprising a layer of grout in the enclosure that encases a bottom portion of the air-intake cavity, bottom portions of the storage cavities, and all air-delivery pipes; wherein a remaining volume of the enclosure cavity is filled with low level radioactive waste that provides radiation shielding for the high level radioactive waste within the hermetically sealed containers; and wherein the low level radioactive waste is selected from a group consisting of low specific activity soil, low specific activity crushed concrete, low specific activity gravel, activated metal, and low specific activity debris. 10. The ventilated system according to claim 1 wherein the hermetically sealed containers are positioned within the storage cavities such that no portion of the hermetically sealed container overlaps the openings in the sidewall of the storage shell in which it is positioned. 11. The ventilated system according to claim 10 wherein the bottom end of the hermetically sealed container is located at an elevation above a top end of the openings in the sidewall of the storage shell in which it is positioned. 12. The ventilated system according to claim 10 wherein there is no line of sight through the openings in the storage shells to the hermetically sealed container positioned within that storage shell. 13. A ventilated system for storing high level radioactive waste comprising:a below-grade storage assembly comprising:an air-intake shell forming an air-intake downcomer cavity and extending along a central axis;a plurality of storage shells surrounding the air-intake shell in a side-by side relationship, each storage shell forming a storage cavity and extending along a central axis,for each storage shell:a primary air-delivery pipe that forms a primary air-delivery passageway that extends from a bottom of the air-intake downcomer cavity to a bottom of the respective storage cavity, wherein the entirety of each of the primary air-delivery passageways is distinct from the entireties of all other of the primary air-delivery passageways of the below-grade storage assembly; anda secondary air-delivery pipe extending between each pair of adjacent ones of the storage shells, the secondary air-delivery pipe forming a secondary air-delivery passageway between the bottoms of the storage cavities of the adjacent ones of the storage shells;a hermetically sealed container for holding high level radioactive waste positioned in one or more of the storage cavities; anda lid positioned atop each of the storage shells and comprising at least one air-outlet passageway. 14. The ventilated system according to claim 13 wherein the secondary air-delivery passageways and the storage cavities of the plurality of storage shells collectively form a fluid-circuit loop, wherein the entirety of the fluid-circuit loop is independent of the entirety of all of the primary air-delivery passageways of the below-grade storage assembly. 15. The ventilated system according to claim 13 wherein for each storage cavity in which one of the hermetically sealed containers is positioned, a bottom end of the hermetically sealed container is located at an elevation above an uppermost end of the primary air-delivery passageway for that storage cavity. 16. The ventilated system according to claim 15 wherein each storage shell has a sidewall with an opening therein at which the primary air-delivery pipe for that storage shell terminates, and wherein the hermetically sealed containers are positioned within the storage cavities such that no portion of the hermetically sealed container overlaps the opening in the sidewall of the storage shell in which it is positioned. 17. The ventilated system according to claim 16 wherein the bottom end of the hermetically sealed container is located at an elevation above a top end of the opening in the sidewall of the storage shell in which it is positioned. 18. The ventilated system according to claim 16 wherein there is no line of sight through the opening in the storage shells to the hermetically sealed container positioned within that storage shell. 19. A ventilated system for storing high level radioactive waste comprising:a below-grade storage assembly comprising a plurality of shells arranged in a side-by-side orientation forming a 3×3 array, the plurality of shells comprising:an air-intake shell located in a center of the 3×3 array and forming an air-intake downcomer cavity, the air-intake shell extending along a central axis and comprising a sidewall having eight openings therein, each of the openings forming a passageway into a bottom portion of the air-intake downcomer cavity;eight storage shells collectively surrounding the air-intake shell in a spaced apart manner such that each storage shell is adjacent to two other storage shells, each storage shell forming a storage cavity and comprising a sidewall having a first opening, a second opening, and a third opening, each of the first, second, and third openings forming a passageway into a bottom portion of the storage cavity of the storage shell;a plurality of primary air-delivery pipes extending between the air-intake shell and the storage shells such that one of the primary air-delivery pipes extends from each of the openings in the sidewall of the air-intake shell to the first opening in the sidewall of one of the eight storage shells, each of the primary air-delivery pipes forming a distinct primary air-delivery passageway that extends along a linear axis from the bottom portion of the air-intake downcomer cavity to the bottom portion of the respective storage cavity;for each of the storage shells:a first secondary air-delivery pipe extending from the second opening in the sidewall of the storage shell to one of the openings in the sidewall of a first adjacent one of the storage shells; anda second secondary air-delivery pipe extending from the third opening in the sidewall of the storage shell to one of the openings in the sidewall of a second adjacent one of the storage shells, each of the first and second secondary air-delivery pipes forming a secondary air-delivery passageway that extends along a substantially linear axis between the bottom portion of the storage cavity of the storage shell and the bottom portion of the storage cavity of one of the first and second adjacent storage shells;a hermetically sealed container for holding high level radioactive waste positioned in one or more of the storage cavities;a lid positioned atop each of the storage shells and comprising at least one air-outlet passageway; andwherein the primary air-delivery pipes and the first and second secondary air-delivery pipes form three distinct air-delivery passageways from the air-intake downcomer cavity to the storage cavity of each of the storage shells. 20. The ventilated system according to claim 19 wherein for each of the storage shells, the three distinct air-delivery passageways comprises:a first air-delivery passageway extending from the air-intake downcomer cavity to the storage cavity directly, the first air-delivery passageway comprising a first one of the primary air-delivery passageways extending from the air-intake shell to the first opening in the storage shell;a second air-delivery passageway extending from the air-intake downcomer cavity to the storage cavity, the second air-delivery passageway comprising: a second one of the primary air-delivery passageways extending from the air-intake shell to the first adjacent one of the storage shells, the storage cavity of the first adjacent one of the storage shells, and the first secondary air-delivery pipe extending from the first adjacent one of the storage shells to the second opening in the sidewall of the storage shell; anda third air-delivery passageway extending form the air-intake downcomer cavity to the storage cavity, the third air-delivery passageway comprising: a third one of the primary air-delivery passageways extending from the air-intake shell to the second adjacent one of the storage shells, the storage cavity of the second adjacent one of the storage cavities, and the second secondary air-delivery pipe extending from the second adjacent one of the storage shells to the third opening in the sidewall of the storage shell. |
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050080458 | abstract | A casting containing toxic waste is made by centrifugally casting toxic waste in a pre-formed cage having heat conducting means therein. The casting is formed with suitable shielding materials selected as a function of the waste being cast. The finished casting may contain passageways through which heat is removed. These passageways are oriented such that when two or more castings are stacked in abutment, the passageways of adjacent castings interconnect. Also disclosed is a method and apparatus for forming castings of polygonal exterior cross section having a contaminant barrier wall of substantially uniform thickness so that the waste storage volume of the castings is maximized and so that the castings may be stacked and stored in honeycomb fashion in a respository to minimize dead storage space. |
summary | ||
description | This application claims the benefit of the filing date of U.S. Provisional Patent Application Ser. No. 61/151,816, filed Feb. 11, 2009. Not applicable. Not applicable. Not applicable. Not applicable. 1. Field of the Invention The present invention relates generally to a destructive method of disposing of unwanted legacy nuclear materials: surplus weapons grade plutonium and reactor grade plutonium, specifically. More particularly, the invention involves a thorium-plutonium-hydride fuel used with lead or lead alloy coolants in a fast spectrum reactor. Still more particularly, the invention relates to a lightly hydrided/deuterated metallic plutonium-thorium fuel for use in a fast fission pool-type nuclear reactor cooled with liquid metal coolants, preferably including lithium-7 lead eutectic, lead bismuth eutectic or lead. Plutonium-239 is consumed, and merchantable heat is produced along with fissile uranium-233, which can be denatured with uranium-238 and used in light water reactors as fuel. 2. Discussion of Related Art Including Information Disclosed Under 37 CFR §§1.97, 1.98 The general principles governing epi-thermal and fast spectrum nuclear reactors are well known in the art. In the earliest years of the nuclear era, nuclear physicists, chemists, and engineers noted that fast spectrum reactors have advantages over thermal-spectrum nuclear reactors. The neutron capture cross sections of elements used as structural, coolant and cladding materials for the reactors (generally elements bearing atomic numbers 11 to 83) are significantly smaller in the harder energy spectra than in the thermal spectrum. Further, no neutrons are lost by hydrogen capture to light water. U.S. Pat. No. 2,993,850, to Soodak, et al (issued Jul. 25, 1961) teaches that parasitic neutron capture is significantly reduced in fast reactors, and therefore a much greater neutron economy is achievable when the reactor is designed for the fast neutron spectrum. Linton Lang conceived of a fast breeder reactor intended to produce “clean” uranium-233 (uranium-233 without co-produced uranium-232). U.S. Pat. No. 3,658,644, to Lang, discloses a fast breeder reactor designed to produce clean uranium-233, wherein the fuel production and power functions are separated in the reactor design. The fast reactor shown obtains its power mostly from fast fission of the fissile material in the fuel. The '644 patent teaches a moderator partition demising the power production region of the central core of the reactor from a thorium-containing blanket on the other side of the partition. The purpose was to eliminate most of the energetic fission neutrons and those neutrons having sufficient energy to produce n, 2n reactions with the thorium blanket. The moderators suggested for the inventive partition included zirconium hydride and lithium-7. U.S. Pat. No. 4,393,510, to Lang, et al, discloses a light water reactor and a process to produce uranium-233 with less than 10 parts per million of uranium-232. The patent teaches that the production of uranium-232 in uranium-233 can be suppressed by separating a thorium-232 reactor blanket from the nuclear fuel using a moderator partition that reduces the energy of the incident neutrons below an energy threshold of 6 million electron volts (6 MeV). The importance of the teachings in the '510 patent resides in the fact that uranium-232, when present in concentrations over 10 parts per million, makes fuel unfit for glove-box handling as the gamma radiation from thallium-208 is too high for worker safety. Of equal importance in the teaching is the fact that the amount of co-produced uranium-232 is a function of the number interactions between thorium-232 atoms and neutrons that have energy exceeding 6 MeV. Significantly, however, the '510 patent deals with the use of a water cooled reactor, a thermal spectrum reactor, and not a liquid metal cooled reactor, as employed in the present application. Another early patent, U.S. Pat. No. 2,904,429, to Schonfeld, discusses means and methods to fabricate binary alloys of thorium and plutonium and shows that when thorium atoms constitute 85% or more of the binary thorium-plutonium alloy, the compound possesses a face-centered cubic crystalline structure that is stable at elevated temperature. When the percentage of thorium exceeds 85%, the binary alloy is stable at temperatures exceeding 900 degrees C. The '429 patent shows that a plutonium-thorium compound is an excellent metallic nuclear fuel. The foregoing patents reflect the current state of the art of which the present inventor is aware. Reference to, and discussion of, these patents is intended to aid in discharging Applicant's acknowledged duty of candor in disclosing information that may be relevant to the examination of claims to the present invention. However, it is respectfully submitted that none of the above-indicated patents disclose, teach, suggest, show, or otherwise render obvious, either singly or when considered in combination, the invention described and claimed herein. Well designed and properly managed light water civilian reactors presently produce electricity safely and reliably. However, reactor-grade plutonium, americium, curium and neptunium from spent light water fuel must be secured. Surplus weapons-grade plutonium must be secured to manage geopolitical risks associated with weapons proliferation. Fissile plutonium-239 has a half life of 24,110 years and decays to uranium-235, a fissile material that has a half life of 700 million years. The present invention discloses a method for the destruction of plutonium-239 in weapons grade form and in reactor grade form with its associated transuranics, neptunium, americium and curium. The present invention also provides a method that efficiently destroys both weapons-grade and reactor grade plutonium associated with transuranics. Plutonium is alloyed with thorium and doped with modest but computationally engineered amounts of hydrogen and/or deuterium to make a reactor fuel that destroys undesirable plutonium isotopes and other transuranic isotopes by fission and transmutation. The percentage of the constituents of the thorium-plutonium-hydride fuel alloys found allows for the efficient destruction of plutonium. The fission must be undertaken in the fast spectrum or the hard neutron spectrum. This is accomplished by the use of lead alloys as coolants. The leading choices for coolant are lithium-7 lead eutectic, lead bismuth eutectic and lead coolants. The metallic thorium-plutonium-hydride fuels function with stability for a deep burn only in hard spectra. The present invention therefore includes, in the first instance, a greatly improved plutonium thorium binary alloy that includes the addition of hydrogen species (protium and deuterium) in computationally engineered amounts, wherein the percentage of thorium in the alloy exceeds 70% and the balance is plutonium, either in weapons grade form or in reactor grade form. Extensive computational studies and analyses directed by the presented inventor revealed that a lightly hydrided/deuterated thorium-plutonium metal fuel worked remarkably well for the disposition of weapons grade plutonium with the ratio of thorium atoms to hydrogen and deuterium atoms being approximately 10:1:1, 10 Thorium, 1 Protium, and 1 Deuterium, and the power setting of the reactor was 400 megawatts thermal. These studies showed that higher hydrogen species doping of the fuel resulted in unsatisfactory performance because the rate of neutron multiplication quickly fell below critical in those cases in which the ratio of thorium to hydrogen was 1:2 and when the ratio of thorium to deuterium and hydrogen was 2:1:1. When no hydrogen was present in the fuel, on the other hand, the rate of neutron multiplication rose above criticality at too high a rate for stable reactor operations. From the analysis conducted to date, it appears that a hydride doping between 8:1:1 (8 Thorium:1 Deuterium:1 Protium and 30:1:1 (30 Thorium:1 Deuterium:1 Protium) is the ideal range for nuclear fuel designed for the disposition of weapons grade plutonium. The optimal ratios of atoms of thorium to atoms of hydrogen species varies as a function of the different output powers for the reactor and thus different sized cores. For reactor grade plutonium the best mix of thorium is between approximately 75% and 85% thorium, ideally between 78% and 82%, with the rest reactor grade plutonium. When neither protium nor deuterium are present in this fuel alloy, the fuel functions well for long periods, i.e., more than 2000 days. When reactor grade plutonium is combined with minor actinides in the same ratio as are produced in light water fuel during operations, and when the ratio of thorium to hydrogen atoms is 5:1 at a power setting of 400 Megawatts, the thorium transuranic fuel lasts for 1000 days. This quantification forms the basis for the invention and allows the transuranics to be consumed as a group without having to separate the neptunium, americium, and curium from plutonium in spent light water fuel. The useful ratio of thorium atoms to hydrogen atom is between 4:1 and 8:1, again depending on the power of the reactor. The present invention therefore provides a new nuclear fuel for use in conventional fuel rods in a novel pool-type rector cooled with depleted lithium lead eutectic, lead bismuth eutectic or lead. The innovation has been computationally modeled, and where hydrogen species doping is modest, the neutron spectrum is energetic enough to fission plutonium-240 and neptunium-237 and to transmute americium and curium from spent light water fuel. The same type of reactor efficiently destroys weapons-grade plutonium-239 in which case the doping of the fuel with deuterium and protium is more modest. Because of its unique capability, the inventive reactor is called a “Special Fast Treatment Reactor (or “SFTR”). The SFTR is a simple modular pool-type reactor in which the metallic thorium-plutonium-hydride fuel is cooled by liquid lead, lead bismuth-eutectic, or depleted lithium-7-lead eutectic. This achieves a spectrum that efficiently fissions surplus plutonium and produces uranium-233. The SFTR is a simple pool type fast reactor design that operates at atmospheric pressure in an enclosed vessel. It is passively safe and has a minimum number of moving parts. The thorium/plutonium/hydrogen species reactor fuel is enclosed in an array of fuel rods clad with HT-9 or EP 823 stainless steel. The fuel rods are surrounded by spectrum shaping rods comprising hydrided or deuterated thorium and by metallic thorium reflector rods that are clad with the same stainless steel used for the fuel rods and the stainless steel jacket retaining the liquid metal pool. The fuel rod array is submerged in liquid lithium-7-lead eutectic, lead or lead-bismuth eutectic, two preferred liquid lead eutectic coolants with natural lead being the third choice. The lead in solid or liquid form metal scatters and conducts neutrons well without moderating them and without capturing them. This permits good neutron communication among the fuel rods, which can therefore have a larger diameter than those in light water reactors because the metal fuel transports heat well and the metal fuel is bonded to the cladding with the same heat conducting lead alloy that used outside of the cladding. Importantly, the SFTR transmutes and fissions reactor grade neptunium, plutonium, americium and curium without needed to separate the transuranic group from each other before the plutonium is alloyed with thorium and hydrogen. Thus, many of the troublesome transuranic isotopes can be permanently destroyed as a group. The invention therefore obviates the need for facilities that separate plutonium from spent nuclear fuel and long term plutonium storage facilities. The invention relieves concerns about geopolitical risks, environmental risks and stewardship costs associated with storing plutonium in weapons grade form or in reactor grade form or as irradiated MOX fuel. Use of the invention fissions away the undesirable materials. The inventive fuel produces uranium-233 during operations in the liquid lead environment. This makes isotopic separation facilities for enrichment of natural uranium to make nuclear fuel obsolete. The SFTR produces uranium-233 that can be used as the fissile component in all nuclear fuels in place of uranium-235. The produced uranium-233 can be denatured with uranium-238 to comply with pertinent global non-proliferation standards. Since uranium-232 is coproduced with uranium-233 in the hard neutron spectrum, the fuel produced in the SFTR will have little attraction for weapons purposes by nation states because the gamma radiation emitted from thallium-208 (in the decay chain starting with uranium-232) gives away the location of the uranium-233 produced in the SFTR. This in addition to denaturing with uranium-238 makes a better more proliferation resistant nuclear fuel for use in the world's civilian fleet. Additionally the presence of the gamma radiation from the decay of thallium-208 makes close working of the uranium metal impractical because exposure to a lethal dose of radiation does not take a long time. These factors could limit interest from sub-national groups interested in using fissile uranium-233 containing small amounts of uranium-232 for illicit radio toxic or explosive purposes. In operation, the SFTR conserves surplus neutrons from fission of plutonium by transmuting thorium into uranium-233. Uranium-233 can be blended with fertile uranium-238 or with fertile thorium-232 (or both) for use as nuclear fuels in existing and future nuclear power stations. Accordingly, the infrastructure needed for enrichment of natural uranium by isotopic separation of uranium-235 is no longer needed to produce fuel grade uranium, since uranium-233 is produced in abundance during SFTR operations with its general production matching the weight of plutonium-239 destroyed. After plutonium and the undesirable transuranics are fissioned in the SFTR, they cease to exist, and, thus, no longer pose an intractable storage problem. Long-term underground storage is not necessary for the plutonium and associated transuranic group that is fissioned away. Long term underground storage is also unnecessary for the recovered and recyclable uranium-238 from light water fuel or other sources, which can be blended with the uranium-233 that is produced in the SFTR and can be used as fuel in present and future reactors. From the foregoing, it should be clear that the inventive system obviates the need for long term storage of weapons grade plutonium, reactor grade plutonium and obviates the need for energy consumptive industrial facilities dedicated to the enrichment of uranium fuels with fissile uranium-235. Each SFTR reactor produces 400 megawatts of thermal energy for sale to electric power generators, and “fresh” nuclear fuel in the form of grade uranium-233 (to be blended with uranium-238 as needed) for use in existing light water reactors or other advanced reactors that breed fuel continuously and use thorium. These benefits are all provided by the stable nuclear fuel disclosed herein. The fuel is used with liquid lead alloys to assure a fast spectrum. The fast spectrum fissions neptunium-237 and plutonium-240. As disclosed above, when reactor grade transuranics are included as a group and combined with thorium, the doping of the alloy with protium is relatively light. For this combination of materials the ratio of thorium atoms to hydrogen atoms in the fuel is 5:1 when reactor power is 400 megawatts. When reactor grade plutonium is separated from spent light water fuel and combined with thorium there is no need to dope with hydrogen for a core having a power of 400 megawatts. When the plutonium is substantially all plutonium-239 as is the case with weapons grade plutonium, the best results reveal that the ratio of thorium to deuterium and protium is 10:1:1 for a power of 400 Megawatts. If the reactor is scaled for more power or less power the optimum ratio will vary so that the fuel should be treated with hydrogen species so that the ratio is richer in hydrogen species when the core volume and power are scaled down and that the ratio is leaner in hydrogen species when core volume and power are scaled higher. The foregoing summary broadly sets out the more important features of the present invention so that the detailed description that follows may be better understood, and so that the present contributions to the art may be better appreciated. There are additional features of the invention that will be described in the detailed description of the preferred embodiments of the invention which will form the subject matter of the claims appended hereto. Those skilled in the art will appreciate that the conception upon which this disclosure is based may readily be used as a basis for designing other structures, methods, and systems for carrying out the several purposes of the present invention. It is important, therefore, that the claims are regarded as including such equivalent constructions as far as they do not depart from the spirit and scope of the present invention. Rather, the fundamental aspects of the invention, along with the various features and structures that characterize the invention, are pointed out with particularity in the claims annexed to and forming a part of this disclosure. For a better understanding of the present invention, its advantages and the specific objects attained by its uses, reference should be made to the accompanying drawings and descriptive matter in which there are illustrated the preferred embodiment. Referring first to FIG. 1, there is shown a chart 100 illustrating the in-fuel and in-reflector spectra of three types of reactor fuel. The coolant for this chart is depleted lithium-7 lead eutectic, with lithium-6 entirely excluded. The hardest spectrum is labeled as the “no H, D” case in which liquid lead was selected as the coolant. The intermediate spectrum is the deuterium case in which the metal fuel was doped with deuterium. The softest spectrum is the hydrogen case. For the two hydrogen species, the ratio of metal atoms to gas atoms in the fuel is 5:19. The hardest spectrum using the lead lithium-7 coolant is superior because it peaks near 0.24 MeV; so that most of the neutrons are energetic enough above the neutron capture resonances of thorium to maximize production of uranium-233. These thorium neutron capture resonances are also near but above the neutron spectrum energy levels at which uranium-233 and plutonium-239 have good fission cross sections. The six spectra are shown with the neutron capture cross sections of thorium, uranium-233 and plutonium-239 and the flux. Three spectra are “in-fuel” and three are “in-reflector” The information from this diagram was developed using Monte Carlo N-Particle Transport Code (“MCNP”). The plot of FIG. 1 shows that protium provides a “super-soft” thermal neutron spectrum in the 3.9% thorium enrichment case with plutonium-239, and an epi-thermal spectrum in the 7.9% thorium enrichment case with plutonium-239 when deuterium is used. The hardest, most energetic spectrum occurs when the fuel is modeled without hydrogen or deuterium. For this case the thorium enrichment with plutonium-239 is 8.3%. Greater mass of fissile and fissable transuranics are present fuels where moderation by hydrogen species is not present. The largest resonance for neutron capture by thorium nuclei ranges from an energy of approximately 50 eV in the epi-thermal range downward to approximately 15 eV. In all cases uranium-233 fuel production takes advantage of the neutron capture resonances of thorium within this range of neutron energies. The graph also includes a line relating to the “super soft” spectrum. This shows that an abundance of thermal neutrons forms when the fuel is moderated primarily with protium (Th:H, 5:19.) This super soft spectrum is produced by the 3.9% plutonium-239 enrichment of the thorium metal fuel. The softness of the spectrum also significantly reduces the amount of uranium-232 co-produced with uranium-233 during operations. This spectrum is useful for producing uranium-233 without co-production of significant amounts of uranium-232 but it tends to be greatly influenced by the presence of fission products, so that the time period that k-eff exceeds one is comparatively brief. This fuel is over hydrided and was computationally found to be unsatisfactory because its useful life is predicted to be much too short. FIG. 2 is a chart 200 showing the results of a computationally simulated ten year run. The power was set at 400 megawatts thermal; the fuel is metallic thorium enriched with fissile reactor grade transuranics. Specifically, neptunium-237 is 1.2% of the mass and plutonium-239 is 8.3% of the fuel mass. The neutron multiplication rate increases for approximately one thousand days and declines thereafter to 1 after two thousand additional days. The beginning of life k-eff is approximately 1.01. It approaches 1.02 at the end of the first thousand day period and declines very gradually thereafter over a two thousand day period. The study documented in FIG. 2 demonstrates the stability of the fissile transuranic-thorium fueled system over a decade of operations. Notable is the gradual change in k-eff between 1.01 and 1.02 and back again to 1 over a long period. This supports the proposition that hydrogen doped plutonium thorium fuel fissioned in the liquid metal environment will allow for a long deep burn of the plutonium fuel when active control measures reduce the variation in the k-eff over time. Traditional methods of neutron capture can be used to actively control the neutron multiplication rate during reactor operations. Referring next to FIG. 3, there is seen a chart 300 showing the fission capture cross sections of uranium-233, plutonium-239 and plutonium-241. These are the leading fissile isotopes in the fuel. These isotopes capture neutrons within complementary ranges easing control issues. The major fissile isotopes for this reactor do not include uranium-235. Plutonium-240 is provided for reference purposes. The fission cross section broadens from ten barns to over one thousand barns over energy range from approximately 6 thousand electron volts to approximately one-tenth of an electron volt. As long as the spectrum stays harder than ten thousand electron volts or one hundredth of a MeV, and leakage is promoted by passive reactor effects the conceptual core seems to exhibit satisfactory stability that can be fine tuned with conventional methods, neutron absorption rods. Stainless steel is used for the cladding that contains the metallic fuel and separates the fuel and the fission products from the liquid metal coolant. HT-9 or EP-823 has good attributes for both cladding and for structural members, the jacket of the liquid metal pool. These similar stainless steel alloys have good resistance to the energetic and high neutron flux of the core region and to heat. These stainless steels are compatible with a variety of the liquid metal coolants: lead, lead bismuth eutectic, lead lithium-7-eutectic or tin-lead eutectic, and so forth. The exterior of the stainless steel plates or rods function to confine the nuclear fuel and the fission gasses and fission products produced during reactor operations. The fuel and reactor are designed to fission plutonium extracted from spent light water reactor fuel or from obsolete weapons. Transuranic reactor grade isotopes of plutonium and neptunium and associated americium and curium or weapons grade plutonium by itself or blended with reactor grade plutonium are alloyed with thorium in computationally optimized proportions and placed in the fuel rods as rolled foil, compacted metal wool or as sintered metal alloy powder. The bonding for the metal fuel under the cladding that assists heat transport is the same lead alloy used as for the coolant on the other side of the cladding. As another option, the thorium-plutonium fuel may be further alloyed with lithium-7, magnesium or aluminum and computationally optimized in a metallic form to enhance fuel expansion effects and homogenous moderation effects when hydrogen species are introduced into the fuel at computationally optimized ratios. Natural lithium and/or lithium-6 could be added to the fuel matrix as beneficial burnable poisons to assist in the control of the reactivity of the reactor over long time periods. The various alloys employed in the preferred embodiments for use in the fuel rods have relatively good heat conduction characteristics that are improved with the addition of lithium-7, aluminum, and/or magnesium. Lithium-7, aluminum, and magnesium all have low neutron capture cross sections in the fast spectrum. However, lithium-7 is preferred as it moderates the hard spectrum modestly and has the lowest neutron capture cross section. Aluminum and magnesium can be added to the thorium transuranic alloy to enhance temperature expansion and heat transfer effects. Lithium-6 can be added to the fuel composition as a burnable poison. The ratio of thorium to reactor grade or weapons grade plutonium in the fuel preferably ranges from 19Th:1Pu, 5% plutonium to 95% thorium, to 3Th:1Pu 30% plutonium 70% thorium. The optimal metallurgical mix for the thorium-plutonium alloy is 3 parts plutonium to 17 parts thorium, 15% plutonium 85% thorium. The percentages are measured by the proportion of atoms, the atomic ratio. The most important additive is hydrogen species. For weapons grade disposition the ideal ratio is approximately between 8-12 Thorium atoms to 1 deuterium atoms and 1 protium atom for a 400 Megawatt power setting. For reactor grade plutonium associated with spent fuel transuranics as they are found in spent fuel the ideal ratio is approximately between 4-6 Thorium atoms to 1 protium atom for a 400 Megawatt power setting, the precise ratio being governed by the intended application. To this alloy the lithium-7, aluminum, magnesium fractions can be added sparingly, not in excess of a total 10% by atomic ratio, so that the final fuel alloy is optimized for long service. The neptunium and the minor actinide isotopes, americium and curium isotopes may also be present in their reactor grade proportions obviating the need to separate these from spent nuclear fuel. In all of the fuel rods, fission gasses migrate to the plenum space to minimize neutron capture in the fuel zone where reactivity is to be maintained at the highest levels. By having the fission gasses migrate to the plenum space at the top of the rod and decay in the plenum above a stainless steel mesh, the fuel and thus power output from the core will be less influenced by xenon transients and by the build up of fission products. Large plenum space in the fuel rods are used to gather disruptive fission gasses and to promote negative reactivity caused by the departure of hydrogen species in fuel hydrides as fuel temperatures exceed pre selected temperature thresholds. For the hydrided fuels the fuel is homogeneous. Moderation although slight takes place in the fuel because hydrogen species are present. If a temperature threshold of 883 degrees C. is exceeded in the fuel, the hydrogen species will dissociate from thorium and enter the plenum space. Because the moderator density is reduced, fewer thermal neutrons will be available to sustain the immediately preexisting rate of neutron multiplication. The rate will decline until temperature in the fuel drops below 883 degrees C. at which point the hydrogen can be reabsorbed by the metallic thorium. As a note, the hydrogen will already have disassociated from plutonium and uranium in the fuel because the dissociation point of these hydrides and deuterides is much lower. The reactor also makes use of spectrum shaping rods. These contain thorium hydride or thorium deuteride and like the fuel rods have large plenum volumes. Neutrons are moderated by the hydrogen species in the spectrum shaping rods until temperature in the liquid metal coolant is elevated to above 883 degrees C. at which time the hydrogen disassociates from the compound. When the spectrum shaping rods are used the loading of the fuel with hydrogen species can be reduced. Practical and simple reactor embodiments use fuel rods arrayed in geometries well known to the art. Referring now to FIG. 4, there is shown in schematic cross-sectional top plan view a preferred embodiment of the rod and coolant geometry used in the reactor core of the present invention. The rods include fuel rods 410, control rods 420, reflector rods 430, and spectrum shaping rods 440, disposed in bundles of 3, 7, or 19, and these are immersed in the selected liquid metal coolant 450. In a preferred embodiment, the rods are arrayed in either in concentric rings or in rows and columns with the fuel rods and shaping rods on the inside and the reflecting rods on the outside. Control is maintained and achieved in the rod system by using traditional neutron absorbing rods and by spectrum shaping rods of thorium hydride to soften the spectrum. In the preferred embodiment, thorium hydride is selected to thermalize some of the fast neutrons to maintain the chain reaction. As liquid metal temperature crosses a threshold, the metal hydride disassociates reducing moderator density. The hydrogen gas is trapped in the spectrum shaping rod plenum above the neutronically active region of the core. This system is expected to be semi-autonomous with the hydride providing an automatic line of defense. Traditional control measures are used, as control rods and safety rods of neutron absorbing materials are deployed and introduced into the neutronically active region to increase or reduce the rate of neutron multiplication and to shut down or start up the reactor. It will be appreciated, then, that control of the inventive spectrum shaping system is provided passively by the dissociation of hydrogen from thorium in the spectrum shaping rods and the thorium in the fuel rods. Generally, when the thorium hydride spectrum shaping rods are removed or are above 883 degrees C. the system is under moderated and the rate of neutron multiplication declines markedly. Depending on the hardness of the spectrum, hafnium, tungsten, tantalum, niobium can be used for the fast and epi-thermal spectra for neutron absorbing safety rods and control rods in embodiments where solid metal performs the safety control functions. Vertical movements of the control rods assist and reinforce the passive and Doppler effects, fuel expansion effects, and moderator density change effects combine to promote safety during operations. In the inventive system, passive control factors are also maximized. Control Features: In the intermediate energy neutron spectrum, lightly hydrided reactor grade or weapons grade plutonium alloyed with fertile thorium fissions predictably for reasonably long time periods as fuel expansion effects, neutron leakage effects and Doppler broadening effects are well managed to maximize passive controls over the neutron multiplication rate. Control over the neutron spectrum is accomplished by the selection of the proportion of protium or deuterium combined with the fuel in the fuel rods and in the spectrum shaping rods, and by the selection of the lead-containing liquid metal coolant alloy. For neptunium-237 to fission, the hardest spectrum must be present. For this purpose the preferred coolant is lead depleted lithium eutectic and the fuel is lightly protiated. This metallic fuel is the preferred embodiment for reactor grade plutonium with associated transuranics because the neutron multiplication rate stays slightly super critical for a simulated computation run time of over 1000 days. The lithium-7 lead eutectic with metal thorium-transuranic fuel has the most stable k-eff discovered so far. For the destruction of weapons grade plutonium the lightly deuterated and hydrided thorium-plutonium metallic fuel alloy is the first choice. During a 2000 day period of operations the reactor's k-eff, the neutron multiplication factor, remains above one increasing at the beginning of life for the first thousand days and declining slowly thereafter. The fast spectrum brings with it the advantages of longer core life and deeper burning because the neutron capture cross section of the fission products is reduced in the higher energy ranges in comparison to the lower energy ranges. Traditionally control and safety have been more difficult to achieve in the fast spectrum than in the thermal spectrum with a large core. However, active features using movable spectrum shaping rods, moveable reflector rods allow the population of thermalized neutrons to be reduced when liquid metal temperature is high. When the reflector rods are retracted and removed more neutrons leak out of the core. When the spectrum shaping rods are retracted and removed fewer neutrons are thermalized. Passive features also assist operations. Neutrons will be absorbed by thorium atoms in the reflector rods, in the shaping rods and in the fuel when fuel and coolant temperature is high because of Doppler effects. Doppler broadening effects are more pronounced in the mid energy ranges because of the resonance regions of thorium's neutron capture spectrum. Fuel expansion effects in metallic fuel are more pronounced in the higher temperature ranges because expansion of the fuel causes significantly fewer fissile atoms to be present in cubic centimeter unit volumes when the fuel is operating at mid-range temperatures. From the foregoing, it will be appreciated by those with skill that there are three aspects promoting negative reactivity in this reactor. The first is the neutron spectrum energy at which thorium is most likely to capture neutrons. The rate of neutron capture and thus the rate of uranium-233 fuel production are influenced by the energy of the neutron population in proximity to the nuclei of thorium. In the higher end of the epi-thermal spectrum, probabilities for capture by thorium are the highest. The spectrum of the active neutron population in this reactor is made to exceed this optimal energy level through the use of lithium-7 in the coolant, the bond in the fuel and the use of spectrum shaping rods containing deuterated thorium. The second aspect promoting negative reactivity involves Doppler broadening effects. The neutron capture cross section diagrams that depict the probabilities of neutron capture by thorium at various neutron energies has specific energies at which the probabilities for neutron capture vary markedly. When temperatures are elevated, the metal atoms move more energetically and this movement “blurs” the peaks and valleys of the capture resonances. As the temperature increases in the thorium metal alloy in the fuel rods, in the reflector rods, Doppler broadening effects cause the resonances to smear together favoring and enhancing neutron capture by the thermally excited nuclei. The Doppler broadening effects compliment active controls by reducing the neutron population when fuel is hot. The third aspect promoting negative reactivity is the fuel expansion effect. Heated atoms of the metal matrix when in full operation occupy a larger volume than atoms of the matrix when at the mid range of operations. This provides fewer fissile nuclei per unit volume of fuel matrix, giving hot fuel fewer fissile nuclei for neutrons to collide with than they would encounter in cold fuel. Fuel expansion effects compliment other control features by decreasing fission captures per unit of volume when fuel is hot. Fuel expansion effects are enhanced by the addition of lithium, aluminum and/or magnesium to the fuel matrix. This is augmented by spectrum hardening effects following reduction of moderator density. This reactor has two fuel configurations. The first configuration uses weapons grade plutonium and thorium, the second uses reactor grade plutonium and the associated transuranics, neptunium, americium and curium and thorium. Destruction of plutonium-239 over a ten year operations period is in the range of a metric ton and the production of uranium-233 over this period is slightly more than the mass of the plutonium consumed with a power of 400 megawatts. The thorium-plutonium fuels must be hydrided to provide advantageous neutronic effects. The addition of the hydrogen species reduces the percentage of fissile material in the fuel needed in order for the core to go critical. Lightly hydriding the fuel so that ratios of metal to gas atoms range from 30 to 1 to 1 5 to 1 provide important benefits. Hydrogen species is kept in the fuel to provide a homogenous moderation effect. When temperature exceeds pre set thresholds, the fuel compound dissociates releasing the hydrogen species as a gas. This reduces moderator density in the fuel and has a direct and prompt impact on the neutron multiplication rate, reducing it as the hydrogen species migrate from the fissile or fissable isotopes in the fuel far enough so that fewer fissions occur because fewer neutrons are moderated to the low energies associated with the highest probabilities for fission capture in plutonium. The above disclosure is sufficient to enable one of ordinary skill in the art to practice the invention, and provides the best mode of practicing the invention presently contemplated by the inventor. While there is provided herein a full and complete disclosure of the preferred embodiments of this invention, it is not desired to limit the invention to the exact construction, dimensional relationships, and operation shown and described. Various modifications, alternative constructions, changes and equivalents will readily occur to those skilled in the art and may be employed, as suitable, without departing from the true spirit and scope of the invention. Such changes might involve alternative materials, components, structural arrangements, sizes, shapes, forms, functions, operational features or the like. Therefore, the above description and illustrations should not be construed as limiting the scope of the invention, which is defined by the appended claims. |
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description | This application is a continuation-in-part of U.S. application Ser. No. 12/047,297, filed Mar. 12, 2008, which is a Continuation-in-Part of U.S. application Ser. No. 11/257,607, filed Oct. 24, 2005, which claims the benefit of U.S. Provisional Application No. 60/621,105, filed Oct. 22, 2004, each of which are incorporated herein by reference. The United States Government has rights in this invention pursuant to Contract No. DE-AC52-07NA27344 between the United States Department of Energy and Lawrence Livermore National Security, LLC. The present invention relates neutron multiplicity counting systems and methods, and more specifically to a multi-mode, real-time neutron multiplicity counter. Neutrons are a fundamental part of any process involving nuclear fission, and thus detection of neutrons is important for radiation protection purposes. Neutron radiation is an ever-present hazard in nuclear reactors. Neutron detectors used for radiation safety must take into account the way damage caused by neutrons varies with energy, and neutron detection techniques may differ depending upon the actual application. Effective neutron detection systems are required to overcome various challenges, such as background noise, high detection rates, neutron neutrality, and low neutron energies. The main components of background noise in neutron detection are high-energy photons (which are not easily shielded), and alpha and beta particles (some of which can be prevented by shielding). Photons are the major source of interference in neutron detection. Unfortunately, both photons and neutrons register similar energies after scattering into a detector from the target, and are thus hard to distinguish from one another. Another challenge is that since the detector typically lies in a region of high beam activity, it is continuously hit by neutrons and background noise at overwhelmingly high rates. This can obfuscate the collected data, since there is extreme overlap in measurement, and separate events are not easily distinguished from each other. It is thus necessary to keep detection rates as low as possible and use a detector that can keep up with high detection rates to yield coherent data. Neutrons are generated through spontaneous fission, induced fission, or alpha particle induced (α,n) reactions. Because neutrons have mass but no electrical charge, they cannot produce ionization in a detector and, therefore, cannot be detected directly. Detecting neutrons requires an interaction of an incident neutron with a nucleus to produce a secondary charged particle that can itself be detected. The presence of emitted neutrons is thus deduced from the presence of neutrons of such secondary charged particles. The energy distribution of fission (spontaneous or induced) neutrons is very different to that of (α,n) neutrons, and can thus be used to help determine the source of neutrons. However, it is generally not possible to use simple energy discrimination to distinguish neutrons from different sources because a measurement consists of both cosmic induced neutrons that cover all energies, as well as those from any unknown source of interest. To improve the analysis process, a characteristic time distribution difference between (α,n) neutrons and fission neutrons is analyzed. Fission neutrons typically produce multiple neutrons (e.g., two or three neutrons) simultaneously, whereas (α,n) neutrons are produced individually and randomly. Coincidence counting techniques can thus be used to distinguish fission neutrons from random (α,n) neutrons. Neutron detection and counting techniques are used to perform non-destructive assays (NDA) of pure samples of plutonium and uranium. Neutron coincidence counting is used to separate the time-correlated fission neutrons from the random, uncorrelated neutrons to determine the fissile mass of the sample. Multiplicity counting is required to analyze impure samples, such as mixed-oxide scrap. Plutonium in bulk form and in waste generates neutrons from spontaneous fission, (α,n) reactions, and induced fission events caused by primary neutrons. Neutron-pair correlation provides the necessary information to determine the spontaneous fission rate and hence the mass of Pu present in a sample, if the isotopic composition is known. The ratio of the (α,n) reaction rate to the spontaneous fission neutron emission rate may be calculated. Coincidence counting requires the effective number of neutron singlets and the effective number of neutron doublets to solve for two unknowns. Multiplicity counting involves the counting of correlated triplets also. With the three quantities (singlets, doublets, and triplets), it is possible to determine three unknowns, such as the spontaneous fission rate, the (α,n) reaction rate, and the detection probability; or the spontaneous fission rate, detection probability, and the neutron multiplication factor. Higher order multiplicity counting is also possible assuming the data is collected in a way to contain the needed information. A current standard approach to neutron multiplicity counting is through the use of a shift-register sliding word that is gated and counted repeatedly. This usually gives data for one gate width, which is set to correspond to the neutron lifetime. A shift-register is a single-input device where pulses can pile up and be lost. This data loss presents a significant disadvantage for current shift-register based detection systems. Another approach to multiplicity counting is a list mode data acquisition system in which every pulse event is stored in memory. In this system, every pulse is assigned a time-tagged value and stored as a word. The volume of data that accumulates can be on the order of many gigabytes if the objective is a non-destructive assay. The disadvantage of this type of system is that a large quantity of data is required to minimize statistical errors, thus requiring massive amounts of system memory. It is therefore desirable to provide a neutron multiplicity counter utilizing multiple gates, with different definitions of the gate and counting approaches, and with a parallel architecture that reduces pulse pile up dead time. In general, multiplicity counters are readily used in conjunction with various types of neutron detectors, and detection hardware refers to the type of neutron detector used and the electronics used in the detector. For example, the most common present detector type is the scintillation detector. The detector hardware defines key experimental parameters, such as source-detector distance, solid angle and detector shielding. Detection software consists of analysis tools that perform tasks such as graphical analysis to measure the number and energies of neutrons striking the detector. Detectors that rely on neutron absorption are generally more sensitive to low-energy thermal neutrons, and are orders of magnitude less sensitive to high-energy neutrons. Scintillation detectors, on the other hand, have trouble registering the impacts of low-energy neutrons. Although it is sometimes facilitated by higher incoming neutron energies, neutron detection is generally a difficult task, and improving scintillator design has been an ongoing process in the industry. Original scintillation detectors were improved with the advent of the PMT (photomultiplier tube), which gives a reliable and efficient method of detection since it multiplies the detection signal tenfold. Even so, scintillation design has room for improvement as do other methods of neutron detection, other than scintillation. For example, gaseous ionization detectors can be adapted to detect neutrons. While neutrons do not typically cause ionization, the addition of a nuclide with high neutron cross-section allows the detector to respond to neutrons. Nuclides commonly used for this purpose are boron-10, uranium-235 and helium-3. Further refinements are usually necessary to isolate the neutron signal from the effects of other types of radiation. As elemental boron is not gaseous, neutron detectors containing boron use boron trifluoride (BF3) enriched to 96% boron-10 (natural boron is 20% B-10, 80% B-11). It is further desirable, therefore, to provide a detection system that effectively detects neutrons by adequately compensating for background noise, high detection rates, neutron neutrality, and low neutron energies. Also desired is a system that preprocesses neutron data into small data files in real time, and reduces processing time required for gigabytes of list mode data. It is yet further desirable to provide a digital data acquisition unit that collects data (e.g., neutron multiplicity data) at high rate and in real-time preprocesses large volumes of data into directly useable forms. Each publication and/or patent mentioned in this specification is herein incorporated by reference in its entirety to the same extent as if each individual publication or patent was specifically and individually indicated to be incorporated by reference. Embodiments are directed to a digital data acquisition method and apparatus that collects data regarding nuclear fission at high rates and performs real-time preprocessing of large volumes of data into directly useable forms. Pulses from a multi-detector array are fed in parallel to individual inputs that are tied to individual bits in a digital word. Data is collected by loading a word at the individual bit level in parallel, so that there is no latency such as in a technique that uses a shift-register. The word is read at regular intervals, with all bits read simultaneously with no manipulation, in order to minimize latency. The electronics then pass the word to a number of storage locations for subsequent processing, thereby removing the front-end problem of pulse pileup. Latency is therefore limited to the latch time in the counter. The word is used simultaneously in several internal processing schemes that assemble the data in a number of more directly useable forms. The technique is useful generally for high-speed processing of digital data, and specifically for non-destructive assaying of nuclear material and assemblies for, typically, mass and multiplication of special nuclear material (SNM). Under an embodiment, a neutron detector system makes measurements on samples that contain fissile material. The neutron detector system provides identification of fission neutron sources through multiplicity analysis. The system is configured to identify fissionable uranium (U) and plutonium (Pu) by evaluating the distribution in time of neutrons that are emitted spontaneously by these materials. It can be used to perform non-destructive assays of samples and/or segregate threat from non-threat nuclear sources. A fundamental characteristic of uranium or plutonium sources (special nuclear materials, SNM) is that the radioactive decay of each nucleus produces multiple neutrons that are released as the nucleus flies apart after spontaneous fission. Detection of these neutrons can aid or confirm identification of SNM sources, but requires a detector that can distinguish cosmic-ray induced background neutrons. SNM fission reactions produce a large number of neutrons, thus making them feasible for use as weapons. The number of neutrons associated with a single nuclear fission is a statistical quantity, referred to as “multiplicity”. The number of emitted neutrons can be in the range of zero to seven, and is typically on the order of two to four. The simultaneously emitted neutrons are produced by a single decay and occur in a short time window, and are referred to as “correlated” neutrons. Correlated neutrons are indicative of both spontaneous fission and neutron multiplication, both of which are present in a weapon or potentially dangerous sample. Neutron Detector System In general, the probability for detecting random neutrons from a sample is constant with time. To determine whether neutron events which are detected are time-correlated, various time periods (measurement gates) are sampled by the coincidence logic for each neutron that is detected. The detector logic identifies those neutron counts that occur within a short time of each other, such as from fission neutrons, closely spaced (α,n) neutrons, and counts due to accidental coincidences. Once one neutron has been detected, the probability of detecting another neutron from the same fission decreases approximately exponentially with time according to the following equation:P(t)=exp(−t/td) where P(t) is the probability of detecting coincidence neutrons in time t, and Td is the die-away time of the moderated detector assembly. FIG. 1A illustrates a neutron detector system for use with a real-time multiplicity counter, under an embodiment. In one embodiment, a neutron detector system 100 for implementing a real-time multiplicity neutron counter comprises a plurality of helium (He-3) neutron detectors 106 and associated hardware, including preamplifier/discriminator stage 104 to detect neutron events and generate count data. An electronics sub-system 102 processes the count data from the detectors 106 and preamp stage 104. This sub-system measures the relative time intervals between neutrons arriving at a detection system and builds a statistical distribution of the multiplicity. The electronic coincidence system takes each neutron detected and looks in a number of time interval gates (e.g., up to 512 gates) to record the time interval between each neutron and others in the data stream from the detectors 106. A software application 110 can be used to analyze the output from the electronic subsystem 102 to determine if the source is consistent with a benign neutron source or an SNM source. The detector system implements a multi-mode counter 120 to monitor how much correlation occurs while sub-sampling each burst of created correlations. The rate that the correlations are created provides an indication of the multiplying system. The counter exploits the way that fission chains are created and provides a way to count the neutrons through a phased observation window. For example, the amount of correlation is observed for a 1 microsecond window, a 2 microsecond window, a 3 microsecond, and so on up to 512 windows. The amount of observable correlations increases as the number of windows are increased up to the point that all correlations are captured. Data related to the created neutrons is stored as a digital word for further processing. In an embodiment, the multi-mode counter 120 has up to five distinct counting modes that vary depending upon trigger condition and measurement parameters. Thus, the various modes of the multi-mode counter have distinctive attributes and provide different ways of counting or observing how and when counts occur, and can be used to fully analyze the correlations that may be present. In a typical setup of neutron detection unit 100, the incoming particles, comprising neutrons and photons, strike the neutron detector portion. This is typically a scintillation detector stage consisting of scintillating material, a waveguide, and a photomultiplier tube (PMT) that are contained in section 106, and that are connected to a data acquisition (DAQ) system to register detection details. The detection signal from the neutron detector 106 is connected to a scaler unit, gated delay unit, trigger unit and an oscilloscope, all of which may be contained within sections 102 and/or 104 of system 100. The scaler unit is used to count the number of incoming particles or events. It does so by incrementing a tally of particles every time it detects a surge in the detector signal from the zero-point. There is very little dead time in this unit, implying that no matter how fast particles are coming in, it is very unlikely for this unit to fail to count an event (e.g., incoming particle). The low dead time is due to sophisticated electronics in the unit, which take little time to recover from the relatively easy task of registering a logical high every time an event occurs. The trigger unit coordinates all the electronics of the system and gives a logical high to these units when the whole setup is ready to record an event run. The oscilloscope registers a current pulse with every event. The pulse is merely the ionization current in the detector caused by this event plotted against time. The total energy of the incident particle can be found by integrating this current pulse with respect to time to yield the total charge deposited at the end of the PMT. This integration is carried out in an analog-digital converter (ADC). The total deposited charge is a direct measure of the energy of the ionizing particle (neutron or photon) entering the neutron detector. This signal integration technique is an established method for measuring ionization in the detector in nuclear physics. The ADC has a higher dead time than the oscilloscope, which has limited memory and needs to transfer events quickly to the ADC. Thus, the ADC samples out approximately one in every 30 events from the oscilloscope for analysis. Since the typical event rate is around 106 neutrons every second, this sampling will accumulate thousands of events every second. The ADC sends its data to a DAQ unit that sorts the data in presentable form for analysis. The key to further analysis lies in the difference between the shape of the photon ionization-current pulse and that of the neutron. The photon pulse is longer at the ends (or “tails”) whereas the neutron pulse is well-centered. This fact can be used to identify incoming neutrons and to count the total rate of incoming neutrons. The steps leading to this separation are gated pulse extraction and plotting-the-difference. Ionization current signals are all pulses with a local peak in between. Using a logical AND gate in continuous time (having a stream of “1” and “0” pulses as one input and the current signal as the other), the tail portion of every current pulse signal is extracted. This gated discrimination method is used on a regular basis on liquid scintillators. The gated delay unit is precisely to this end, and makes a delayed copy of the original signal in such a way that its tail section is seen alongside its main section on the oscilloscope screen. After extracting the tail, the usual current integration is carried out on both the tail section and the complete signal. This yields two ionization values for each event, which are stored in the event table in the DAQ system. In this step, the extracted ionization values are plotted. Specifically, the graph plots energy deposition in the tail against energy deposition in the entire signal for a range of neutron energies. Typically, for a given energy, there are many events with the same tail-energy value. In this case, plotted points are simply made denser with more overlapping dots on the two-dimensional plot, and can thus be used to estimate the number of events corresponding to each energy-deposition. A considerable random fraction ( 1/30) of all events is plotted on the graph. If the tail size extracted is a fixed proportion of the total pulse, then there will be two lines on the plot, having different slopes. The line with the greater slope will correspond to photon events and the line with the lesser slope to neutron events. This is precisely because the photon energy deposition current, plotted against time, leaves a longer “tail” than does the neutron deposition plot, giving the photon tail more proportion of the total energy than neutron tails. The effectiveness of any detection analysis can be seen by its ability to accurately count and separate the number of neutrons and photons striking the detector. The effectiveness of the second and third steps reveals whether event rates in the experiment are manageable. If clear plots can be obtained in the above steps, allowing for easy neutron-photon separation, the detection can be termed effective and the rates manageable. On the other hand, smudging and indistinguishability of data points will not allow for easy separation of events. Detection rates can be kept low in many ways. Sampling of events can be used to choose only a few events for analysis. If the rates are so high that one event cannot be distinguished from another, physical experimental parameters (e.g., shielding, detector-target distance, solid-angle, and so on) can be manipulated to give the lowest rates possible and thus distinguishable events. It is important to observe precisely those variables that matter, since there may be false indicators along the way. For example, ionization currents might get periodic high surges, which do not imply high rates but, just high energy depositions for stray events. These surges will be tabulated and viewed with cynicism if unjustifiable, especially if there is very much background noise in the setup. Every current pulse in the oscilloscope corresponds to exactly one event due to the fact that the pulse lasts about 50 ns, allowing for a maximum of 2×107 events every second. This number is much higher than the actual typical rate, which is usually an order of magnitude less, as mentioned above. This means that is it highly unlikely for there to be two particles generating one current pulse. The current pulses last 50 ns each, and start to register the next event after a gap from the previous event. Detector system 100 may be embodied within a portable fission meter or neutron source identifier unit that is packaged in an appropriate format for portability and use in the field. The He-3 tube assembly and associated hardware 104 and 102 can be integrated within a single housing or modular configuration with internal (e.g., battery) and/or externally provided power supply. The tube assembly 106 may be provided in a hinged housing made of two or more separate components to facilitate packaging within a portable hand-held case. The case enclosing the detector system may be ruggedized for use in different environmental conditions. In an embodiment, the tube assembly 106 includes thirty 1-inch diameter by 19-inch active length He tubes in association with one or more high density polyethylene (HDPE) moderators, with adjacent pairs of detectors sharing a common preamplifier 104. Depending upon application requirements, other tube arrangements are also possible. The system 100 also includes a front panel control and display unit 101 that includes an LCD display to show neutron counts per second, battery capacity, and cycle count, among other information that may be relevant. Controls can be provided to allow user setting of certain operational parameters, such as on/off, start/stop/clear, and so on. The detector system 100 may include or be coupled to a processor based computing systems. This computing system serves as a platform for the fission meter software application 110 and includes a microprocessor, memory 112, input/output ports, and an optional graphical user interface (GUI) 114 process for use with an external display device 116. Other possible components may also be provided, such as input means (e.g., keyboard, mouse, trackball, etc . . . ), operating systems, input/output processes, application interfaces, and the like. The detector may be set to operate in one or more operational configurations, such as mobile search, which is carried out in real-time to locate an SNM source by monitoring total count rate above background (in standard deviations). Another configuration is a static search setting in which the system is put into position and data is collected over a relatively long period (e.g., 15 to 20 minutes). In this configuration, up to 30,000 counts must be obtained from the source, which implies a count time of at least 1000 seconds. A characterization detect configuration can use the determinations of the static search mode to allow the collection of a great deal of data (e.g., up to 1 million counts) for expert analysis off-site. In general, neutron coincidence counting provides two measured values (reals and totals) while in some cases there are three or more unknown variables which need to be determined, that is, the mass of 240Pu-effective, (α,n)-to-(SF,n) ratio, and multiplication factor. In multiplicity counting, a third measured parameter, the distribution of multiple counts, is derived, and thus the three unknowns may be calculated. The multi-mode counter performs the necessary data collection to generate a histogram of multiplicity events recorded to singles, doubles, and triples event rates. These can be used in conjunction with interpretational models to extract the unknown variables for product material meff, α and ML. The output of the system may be provided in graphical form. For this embodiment, the detector system also includes a graphical generation and output process that displays the plots of various characteristics on a display device 116. FIG. 18 illustrates an example of a multiplicity (fission meter) plot that can be generated by detector system 100, under an embodiment, and FIG. 19 illustrates an example of a neutron lifetime plot that can be generated by detector system 100, under an embodiment. The multiplicity plot 1800 of FIG. 18 shows a random (Poisson) distribution, matched in mean count to the data. It is the deviation of the data which may include non-cosmic effects from a random distribution. The neutron lifetime plot 1900 of FIG. 19 can help the detection of moderation material (e.g., polyethylene shielding). As shown in FIG. 1, the neutron detector system 100 includes a multi-mode neutron counter process 120 as part of the electronic subsystem 102. The multi-mode neutron counter 120 is a highly parallelized circuit that loads the inputs from the detector system 106 in parallel, performs the count operation and creates sums, which are then written to memory. The parallel architecture allows for fast read and clear processes compared to present counters that use shift registers. A shift-register based system typically combines a number of detector input channels, such as 4, 8, or 16 inputs into a single shift register, which then counts neutrons in real-time, and stores the data as words for each measurement time interval. This type of system suffers from pulse pile-up and creates an excessively large amount of data (e.g., on the order of hundreds of gigabytes), and also introduces errors due to dead time, and other latencies. FIG. 1B is a block diagram illustrating the components of a multi-mode neutron counter, under an embodiment. The detector system 140 includes a plurality of He-3 tubes, numbered 1 to n, where n is typically 32, but can be any number depending upon system configuration. Each detector provides a separate input to a word generator circuit 134. This parallel architecture provides a much finer division of the input stream than present shift-register systems. The input data is used to produce sum data form each clock cycle, which are stored as a digital word in one or more registers 132. The sum data is then stored in a specific address location of a memory 130, which may be partitioned in a matrix arrangement, as shown in FIG. 1B. For the circuit of FIG. 1B, pulses from a multi-detector array 140 are fed in parallel to individual inputs that are tied to individual bits in a digital word through word generator 134. Data is collected by loading a word at the individual bit level in parallel, so that there is no latency such as in a technique that uses a shift register. The word is read at regular intervals, all bits simultaneously, with no manipulation, to minimize latency. The electronics then pass the word to a number of storage locations in memory 130 for subsequent processing. This scheme removes the front-end problem of pulse pileup and any latency is limited to the latch time in the counter. The word is used simultaneously in several internal processing schemes that assemble the data in a number of more directly useable forms. FIG. 2 is a flow chart illustrating a method of event counting using a multi-mode neutron counter, under an embodiment. The method includes an act of inputting edge triggered input signals into parallel input circuits observing each event to be counted, act 202. In act 204, the system creates a clock to control a minimum summing interval wherein neutron count data is collected (counted). The clock signal is used by a parallel set of adders, wherein each input circuit is operatively connected to multiple private (independent) adders of the parallel set. In act 206, a sum in each adder is read during the minimum summing interval to produce a sum read. Each adder is then zeroed at the end of the minimum summing interval, act 208. The sum is then read into multiple arrays, act 210. In act 212, the system constructs summed sections from the array to build data structures comprising several characteristics. These characteristics include multiple superset interval sizes, interval sizing after an external trigger, event totals in a fixed interval, event totals in a fixed interval after an external trigger, time intervals between events, time intervals between events after an external trigger, and arrival time of certain clump sizes after an external trigger. In a second embodiment of the method of FIG. 2, the method comprises: providing the input signals into parallel input circuits observing each event to be counted; controlling a minimum summing interval in which data is counted for use by a parallel set of adders; producing a sum read; zeroing adder; storing the sum read; and building data structures. In this second method, the input signals may be edge triggered. The minimum summing interval is controlled with a clock. Each input circuit is operatively connected to multiple independent adders of the parallel set. The sum read is produced by reading a sum in each adder during the minimum summing interval. The adders are zeroed at the end of a minimum summing interval. The sum read may be stored into multiple arrays. The data structures are built by constructing summed sections from the array and may comprise data selected from the group consisting of multiple superset interval sizes, interval sizing after an external trigger, event totals in a fixed interval, event totals in a fixed interval after an external trigger, time intervals between events, time intervals between events after an external trigger, and arrival time of certain clump sizes after an external trigger. The data structures may comprise multiple superset interval sizes, interval sizing after an external trigger, event totals in a fixed interval, event totals in a fixed interval after an external trigger, time intervals between events, time intervals between events after an external trigger, and arrival time of certain clump sizes after an external trigger. An apparatus for event counting according to an embodiment may be embodied within one or more hardware circuits that implement the process steps of FIG. 2, and may comprise: means for inputting input signals into parallel input circuits observing each event to be counted; means for controlling a minimum summing interval in which data is counted for use by a parallel set of means for adding; means for producing a sum read; means for zeroing each said means for adding; means for storing said sum read; and means for building data structures. Multi-Mode Implementation The multi-mode counter module 102 and multi-tube detector 106 allow for a multiply pipelined system that allows for massive parallel inputs to separate inputs of a gate array circuit and the independent tallying of data. The parallel architecture reduces deadtime with, no pulse pileup or latency, since all inputs are accepted. The multi-mode counter 120 analyzes different attributes of the count data. The gate array comprising the counter operates off of a single clock so that all data is on the same clock cycle. The real-time parallel processing creates a record of absolute time when the pulses came in. A pre-processor conditions the data to reduce file size as compared to standard time-tagging systems. For example, file sizes on the order or ½ megabyte are realizable as compared to the multi-gigabyte files of current systems. The detector system counts single neutrons with reference to a background threshold to enable the location of neutron sources. In one embodiment, the neutron detector comprises a multi-mode neutron counter circuit or process 120 that is used in conjunction with, or is incorporated within the detector. In general, the neutron counter measures for a defined gate width, such as 1 μs after a fission chain starts. The counter then counts for number of gates or measurement windows. The gates can be sequential such that each subsequent gate (measurement window) immediately follows the previous gate. Alternatively, gates can overlap so that some of the count data is re-used. The counter thus generates a time history of when neutrons arrive by opening measurement windows to observe when correlations occur. In an embodiment, the multi-mode neutron counter accommodates up to five different counting modes depending on different trigger conditions and measurement classes. The counter gathers the neutron data and analyzes for coincidences; singles, doublets, triplets and quads up to a very high order. Neutron multiplicities in various time subgates during each Data Acquisition Cycle (DAC) are recorded. The acquisition cycle may be defined as 512 time bins. In an embodiment, the recorded data is analyzed by the system software, according to the Feynman Variance technique. Multiplicity counting, thus counts separately the number of neutrons detected with a time gate (e.g., none, 1, 2, 3, . . . ), in contrast to neutron coincidence counting, which only looks for pairs of neutrons with a window, and gross counting, which counts all emitted neutrons. FIG. 3A illustrates a matrix delineating the plurality of modes in which the detector system performs neutron counting operations, under an embodiment. As shown in the neutron counting requirements matrix of FIG. 3A, the system operates in two different trigger modes and performs up to three different classes of measurements per trigger mode. Trigger mode 1 comprises Data Acquisition Cycles (DAC) initiated by periodic internal triggers, and is normally used for passive measurement of naturally radioactive samples; and trigger mode 2 comprises DACs initiated by internal or external periodic triggers and can be used for passive measurements or measurements with periodic neutron generators. The three measurement classes for each trigger mode are denoted class A, class B, and class C. The class A measurement records neutron multiplicities using Feynman Variance analysis techniques. The Feynman variance technique consists of detecting the neutron count deviations in the time domain of the counting data from what are expected to be random events. For a random source, the ratio of the variance to the mean is equal to one, and for a multiplying source, the ratio is greater than one. The class B measurement records neutron-pair time intervals by measuring the time space between counts. The class C measurement is a time dependence measurement that measures the intensity of the count occurrence through a multiplet die-away analysis. The die-away time is the characteristic time a neutron will survive before it is absorbed in the He-3 detector tubes or escapes the counter. The neutron die-away time normally ranges from 10 to 128 μs depending upon the counter geometry. Details of the operation and characteristic of each mode within the matrix of FIG. 3A are provided in greater detail in the description below. The modes are denoted 1.A and 1.B (or 1A and 1B) for trigger mode 1, and 2A, 2.B, and 2.C (or 2A, 2B, and 2C) for trigger mode 2. The multi-mode counter simultaneously collects different items of data pertaining to neutron creation, thus allowing for the observation of the physical nuances of a neutron source. This data includes: neutron lifetime information through the use of multiple time gates, energy effects through the use of a number of gates to capture neutron mode effects, multiplicity in sums that best relate to multiplication (modes 1A and 2A), timing intervals that best relate to the count rate detected (mode 1B versus 2A), and the time structure of time-varying sources (modes 2B and 2C triggered). Modes 1A and 2A are most efficient for observing multiplication (i.e., fission meter plot) mode effects (neutron energy effects). Modes 1B and 2B are suitable for low count rate situations where the time between sparse counts best illustrates correlations. Modes 2B and 2C are suitable for cases where an external event, such as a pulsed neutron source, defines the best time to collect counts Mode 2C is optimized for cases where the particular multiplet structure of a stream of counts needs to be observed over time, which is similar to the idea of an oscilloscope for correlated events. The parallel gate counting architecture is well suited to capture any number of counting sums that would group selected information, including simply recording the arrival times of every count, because the data streams are multiply pipelined. In an embodiment, all five modes of the multi-mode counter 120 run simultaneously in parallel and independently of one another. This provides many different ways of counting or observing how and when counts (from neutrons or gammas) occur, and analyze the correlations that may be present. An operator or data analyzer can decide which individual count mode or combination of count modes to look at. The multi-mode counter opens up a gate, which is an observation window randomly in time, and looks at the counts for a number of gate periods. The trigger conditions can be internal or external, and can start with an actual count or a random time tick. In this manner, different triggers and counting parameters can be used to generate count data that provides comprehensive insight into the nature of the correlations present in the neutrons detected by the detector system. The parallel execution of the subroutines for the different count modes of the multi-mode counter increases the parallelism of the detection system 100. In effect this creates parallel pipeline for all of the inputs of the detector stage. FIG. 3B is a functional block diagram of the multi-mode illustrating the parallelism of the different modes, under an embodiment. As shown in FIG. 3B, the n (e.g., 1-32) inputs from the detector are input to an edge detector/register stage 312. Each detector tube of the detector unit 314 produces an input to the edge detector. This may be a gate-array based edge-triggered sample detector is producing with n inputs pins. As described with reference to FIG. 1B, the detector inputs can generate a word which is produced by either the edge detector at front of the gate array or a register stage after the edge detector, depending on the counting mode. The output from the edge detector 312 is divided out and input individually into each subroutine module of the multi-mode counter 310. For the embodiment of FIG. 3B, the multi-mode counter 310 contains five different subroutines 301-305 for the modes 1A, 1B, 2A, 2B, and 2C illustrated in FIG. 3A. Other modes are also possible. The multiple subroutines execute simultaneously and each subroutine 301-305 processes the stream of pulses as they come in from the detector stages 314 and 312. The results then individually summed as they come in and each subroutine processes each input to produce count data that is conditioned for a particular mode 1A, 1B, 2A, 2B, or 2C. This mode data can then be provided to an analysis process 330 that processes the count data to determine the presence or absence of correlations or other characteristics of the count data. For the neutron counting requirements matrix of FIG. 3A, the two trigger modes (trigger mode 1 and trigger mode 2) may actually be described as three modes: self-triggered mode 1, self-triggered mode 2 and externally-triggered mode 2. The self-triggered mode 2 counting is essentially an internally triggered mode, like trigger mode 1. The externally-triggered mode is typically called the neutron generator triggered counting. Trigger mode 1 is typically used for making measurements of neutrons generated by the natural radioactivity of the sample material. In this mode the detector system will employ internally generated, periodic triggers to detect neutrons in data acquisition gates (DAGs). In this mode, the DAGs are uncorrelated with the neutron emission times. FIG. 4 illustrates Data Acquisition Gate (DAG) and Data Acquisition Cycle (DAC) characteristics for trigger mode 1, under an embodiment. As shown in FIG. 2, a DAC for data mode 1.A (trigger mode 1, class A) comprises 512 counting bins of a defined width. The bins are denoted b-1 to b-512 and are of a user selectable width τ0, where the minimum width for τ0 is 1 microsecond. The length of the DAG is identical to that of the DAC (LG=LC) for mode 1.A. In mode 1.A the gates are overlapped to provide a correlated count history which reuses a large amount of the count data, resulting in a relatively smooth data distribution. For mode 1.B, in which neutron-pair time intervals are measured, the DAG comprises two DACs for a total of 1024 counting bins during a data acquisition gate (LG=2LC). Trigger mode 2 is used for measurements of samples with very low natural neutron activity; it may also be useful for measurements on some samples with higher natural activity. Most of the neutrons detected in this mode will be generated by interactions (mainly induced fission) initiated by pulses of 14-MeV neutrons injected into the sample material by an ion-tube (D,T) neutron generator. The periodic triggers for the detector in this mode are provided by the neutron generator at a fixed time relative to the 14-MeV neutron pulses. The DAGs and the induced-fission neutrons emitted by the sample are thus highly correlated in time. FIG. 5 illustrates Data Acquisition Gate (DAG) and Data Acquisition Cycle (DAC) characteristics for trigger mode 2, including the time correlation of the DAGs and the induced-fission neutrons emitted by the sample, under an embodiment. For trigger mode 2, the length of the DAC, Lc, is determined by the PRF of the internal or external trigger. The DAG count starts after a delay, Δ, that is user selectable. A number of DAG cycles may be performed each having a number of subgates (1-512) of width τ0, where the minimum width for τ0 is 1 microsecond. The initial delay Δi, may be the same or different than the delay Δc, between the DAG cycles. There may also be a subgate delay Δs, associated with one or more measurement classes, as well as a beam delay Δo, that is associated with the TTL (transistor-transistor-logic) trigger pulse from the neutron generator. In both mode 1 and mode 2, two classes of measurements (Class A and Class B) are required, and a third class (Class C) can provide valuable information in mode 2, but is not applicable to mode 1. For each class of measurement the neutrons detected within the DAGs must be sorted in different ways. In order to minimize overall data collection time, it is necessary to carry out the various classes of measurements (i.e., implement the different data sorting algorithms) simultaneously. It should be noted that there may be cases in mode 2 in which different beam delays are required for different measurement classes, which would require separate measurements. With reference to FIG. 3A, the class A measurement data will be sorted to record statistics on neutron multiplicities detected within temporal sub-gates with different widths. A Feynman Variance type of analysis can be carried out with this data. Although the same data sorting algorithm (the “Inefficient Implementation”) can be used for both mode 1 and mode 2 measurements, other sorting algorithms (“efficient implementation”) can greatly improve data collection efficiency in mode 1. FIG. 6 illustrates a data acquisition cycle for count performed using mode 1.A and showing the subgate detail, under an embodiment. FIG. 6 shows a data acquisition cycle (DAC) comprising 512 subgates. Mode 1.A utilizes an overlap scheme in which different levels of subgates (denoted gi,j) are defined so that data from each bin is used in more than one subgate. This reuse of the data helps smooth the data distribution. The subgate levels are denoted level-1 to level-n, where n can be any appropriate number depending upon system requirements and constraints. As shown in FIG. 6, each level-1 subgate contains the total counts in one bin (b-n), and there are 512 subgates, where τ1=τ0; each level-2 subgate contains the total counts in two adjacent bins, and there are 511 subgates, where τ2=2τ1=2τ0; each level-3 subgate contains the total counts in four adjacent bins, and there are 127 subgates, where τ3=2τ2=22τ0; each level-4 subgate contains the total counts in eight adjacent bins, and there are 63 subgates, where τ4=2τ3=23τ0; and so on up to the single level-12 subgate that contains the total counts in all 512 counting bins. FIG. 7 shows an example of a count using mode 1.A counting, according to an exemplary embodiment. For the example of FIG. 7, the bin and subgate tally for mode 1.A is shown for bins 1-18, and show the multiplicities occurring within subgates of various widths for a particular DAC. The resulting matrix provides example counts of zeros, single, doubles, and triples for the DAC. Each element in the cumulative data matrix is incremented after each DAC for the duration of the measurement. FIG. 12 illustrates a data acquisition cycle for count performed using mode 2.A and showing the subgate detail, under an embodiment. FIG. 12 shows a data acquisition cycle (DAC) comprising 512 subgates. Mode 2.A utilizes a partial overlap scheme in which different subgate intervals (denoted gi) are defined so that data from each bin may be used in more than one subgate. This reuse of the data helps smooth the data distribution. In certain cases, mode 2.A may be implemented so that the subgates are sequential, in which case no count data is reused. As shown in FIG. 12, subgate g1 consists of a single bin, where τ1=τ0; subgate g2 consists of two adjacent bins, so that τ1=2τ0; subgate g3 consists of three adjacent bins, so that τ1=3τ0; subgate g4 consists of four adjacent bins, so that τ1=4τ0; and so on up to subgate g512, which consists of all 512 bins, so that τ1=512τ0. In mode 2, the DACs are initiated by an external trigger. As shown in FIG. 12, an initial delay, Δi, precedes the first DAG, and each subsequent DAG is preceded by a cycle gap, Δc. The entire DAC is measured from the beginning of the delay preceding the DAG to the end of the delay preceding the next subsequent DAG. FIG. 13 shows an example of a count using mode 2.A counting, according to an exemplary embodiment. For the example of FIG. 13, the bin and subgate tally for mode 2.A is shown for bins 1-18, and show the multiplicities occurring within subgates of various widths for a particular DAC. The resulting matrix provides example counts of zeros, single, doubles, and triples for the DAC. Each element in the cumulative data matrix is incremented after each DAC for the duration of the measurement. For measurement class B, the measurement data is sorted to record statistics on the time intervals between successive neutrons detected within the DAGs. A Rossi-Alpha type of analysis can be carried out with this data. In general, the same data sorting algorithm applies for both mode 1 and mode 2. FIG. 8 illustrates a data acquisition cycle for count performed using mode 1.B and showing example interval measurements for one count in the first bin, under an embodiment. FIG. 8 shows a data acquisition cycle (DAC) comprising two DAGs of 512 subgates each, so that LC=2LG. This mode measures neutron pair intervals for a naturally radioactive source using an internal periodic trigger. As shown in FIG. 8, intervals between any two counts are denoted δ(i,j) where i is the first bin containing a count, and j is the next bin containing a count. In an embodiment, all intervals are measured from the very first bin that contains a count, and is referred to as Type-II analysis in that only one neutron (in the lowest-numbered occupied bin the dual DAG) is considered the virtual trigger neutron. Thus for the example of FIG. 8, there are seven different bins containing detected neutrons (denoted by an *). Two bins, b-17 and b-518 contain two neutrons. This yields eight intervals all measured from the first bin that has a neutron, b-2. FIG. 9 shows an example of a count using mode 1.B counting, according to the exemplary embodiment of FIG. 8. For the example of FIG. 9, the bin and subgate tally for mode 1.B is shown for bins 1 to 1024, and show the interval lengths, as a function of τ0 between bins that contain neutrons. The resulting count data lists the number of neutrons in each time interval based on τ0, that is 0-τ0, 1-τ0, 2-τ0, . . . 1024-τ0, for a 1024 bin DAC. Each element in a similar cumulative data matrix is incremented after each DAC for the duration of the measurement. FIG. 10 illustrates a data acquisition cycle for count performed using mode 1.B and showing example interval measurements for two counts in the first bin, under an embodiment. FIG. 10 shows a data acquisition cycle (DAC) comprising two DAGs of 512 subgates each, so that LC=2LG. This mode measures neutron pair intervals for a naturally radioactive source using an internal periodic trigger. As shown in FIG. 10, intervals between any two counts are denoted δ(i,j) where i is the first bin containing a count, and j is the next bin containing a count. Thus for the example of FIG. 10, there are seven different bins containing detected neutrons (denoted by an *). Three bins, b-2, b-17 and b-518 contain two neutrons. When the first bin that contains any counts contains two counts (e.g., b-2), then this first bin is set to 0, and the intervals are all measured from this bin (Type II). For the example of FIG. 10, this yields ten intervals all measured from the first bin that has two counts, b-2. FIG. 11 shows an example of a count using mode 1.B counting, according to the exemplary embodiment of FIG. 10. For the example of FIG. 11, the bin and subgate tally for mode 1.B is shown for bins 1-1024, and show the interval lengths, as a function of τ0 between bins that contain neutrons. The resulting count data lists the number of neutrons in each time interval based on τ0, that is 0-τ0, 1-τ0, 2-τ0, . . . 1024-τ0, for a 1024 bin DAC. Each element in a similar cumulative data matrix is incremented after each DAC for the duration of the measurement. FIG. 14 illustrates a data acquisition cycle for count performed using mode 2.B and showing example interval measurements, under an embodiment. FIG. 14 shows a data acquisition cycle (DAC) comprising a single DAG of 512 subgates. This mode measures neutron pair intervals for neutron induced activity using an external trigger, such as a pulsed neutron source. The DAC includes the DAG plus the initial gap ΔI and the cycle gap, ΔC. As shown in FIG. 14, intervals between any two counts are denoted δ(k+i,k+j) where i is the first bin containing a count, and j is the next bin containing a count, and k is the first bin following a subgate delay ΔS. In an embodiment, all intervals are measured from the very first bin after the subgate delay that contains a count (Type-II analysis). Thus for the example of FIG. 14, there are six different bins containing detected neutrons (denoted by an *). Two bins contain two neutrons. This yields seven intervals all measured from the first bin after the subgate delay that has a neutron, b−k+3. FIG. 15 shows an example of a count using mode 2.B counting, according to the exemplary embodiment of FIG. 14. For the example of FIG. 15, the bin and subgate tally for mode 2.B is shown for bins 1-512, and show the interval lengths, as a function of τ0 between bins that contain neutrons. The resulting count data lists the number of neutrons in each time interval based on τ0, that is, 0-τ0, 1-τ0, 2-τ0, . . . 512-τ0, for a single DAC. Each element in a similar cumulative data matrix is incremented after each DAC for the duration of the measurement. The examples of FIGS. 14 and 15 illustrate a case in which there are two counts in the first populated bin in the sorting range. For measurement class C, the measurement data is sorted according to the number of multiplets in each time bin within the data acquisition gate. This data allows the measurement of neutron die-away following the injection of the neutron pulse (e.g., 14-MeV) into the sample. FIG. 16 illustrates a data acquisition cycle for count performed using mode 2.C and showing example multiplet count data, under an embodiment. FIG. 16 shows a data acquisition cycle (DAC) comprising 512 subgates. In mode 2.C the time measured from the start of the DAG for a multiplet in bin b-j is defined as τj=(j−1)τo, and the multiplicity for the bin is mj. In mode 2.C, the DACs are initiated by an external trigger (e.g., a pulsed neutron generator). As shown in FIG. 16, an initial delay, Δi, precedes the first DAG, and each subsequent DAG is preceded by a cycle gap, Δc. The entire DAC is measured from the beginning of the delay preceding the DAG to the end of the delay preceding the next subsequent DAG. FIG. 17 shows an example of a multiplicity count using mode 2.C counting, according to the exemplary embodiment of FIG. 16. For the example of FIG. 17, the bin and subgate tally for mode 2.C is shown for bins 1-18 (of 512 total bins), and show the multiplicities, m, occurring within subgates of various widths for a particular DAC. The resulting matrix provides counts of multiplicity in an array specifying bins versus multiplicity count (e.g., 0 to 7). Each element in the cumulative data matrix is incremented after each DAC for the duration of the measurement. In summary, four different data sorting algorithms must be implemented in order to carry out all of the classes of analysis that are necessary for both trigger modes 1 and 2 measurements, although only two are applicable in mode 1 and three are applicable in mode 2. It is desirable to implement simultaneous sorting of data by all four algorithms for all measurements, in order to simplify field operation of the detector system. During operation, a user must select the actual triggering mode (mode 1 or mode 2). Once the trigger mode is selected, the detector system will automatically perform all possible classes of measurement within that mode, thus if trigger mode 1 is selected, the detector will perform both the 1A and 1B measurements; and if trigger mode 2 is selected, the detector will perform the 2A, 2B, and 2C measurements. As shown in FIG. 1, the overall detector system incorporating the multi-mode counter consists of a detector stage 106 and various electronic subsystems. The detector stage 106 comprises He-3 proportional-counter tubes embedded in a polyethylene moderator. The tubes may be in a single pod or in a pair of pods. The output pulses from the tubes are fed to an electronic module containing amplifiers and pulse-sorting circuitry. With respect to the five possible modes, the electronic subsystem 102 has four principal functions: first, it supplies the high-voltage to the He-3 tubes and power for the electronic counting circuitry from a self-contained battery pack; second, it permits user selection of a) one of the two triggering modes, internal (mode 1) or external (Mode II), b) a “Start Delay,” Δ1, for mode 2 (set to the minimum value, 1-μs, for mode 1), c) the width, τo, of the fundamental data-sorting time bins (minimum value currently restricted to 1 μs), and d) the number of Data Acquisition Cycles (DACs) for the measurement (typically 105-108); third, it amplifies and shapes the analog output signals from each tube (e.g., through a separate amplifier and discriminator 104 for each tube) and feeds the signals to a data collection and sorting system; and fourth, it sorts the data collected on each DAC into the four data matrices required for the different modes and analysis types, and appropriately increments the cumulative data matrices at the end of each DAC, and outputs the cumulative data matrices at the end of each measurement. The electronics module also displays and/or prints the average total counting rate in units of neutrons/DAG to allow the operator to adjust the length of the DAG and/or the sample-to-detector distance to achieve good data collection efficiency. It may also print a reminder to the operator that the number of neutrons/DAG needs to be large. (Since the number of counting bins will be fixed at 256, the length of the DAG is determined by the value of τo that is set). The schematic representations of the neutron beam and the beam delay (Δo) shown in FIG. 5 apply only to mode 2. When gathering data from mode 1, the 14-MeV neutron generator (i.e., external trigger input) is not used. The start pulse for the DAC is generated internally. The delay, Δ2, is essentially zero, and Δ1 is kept at the minimum value consistent with the triggering and data sorting requirements for the cycle (approximately 1 μs). The user-selected value, τo, of the fundamental counting bin width, therefore, determines LG (the number of bins is fixed at 256), and (together with the fixed value of Δ1) the length of the DAC (LC) and, of course, its inverse, the pulse repetition frequency (PRF). In trigger mode 2, the user selects the values of τo, Δ1, and the PRF of the neutron generator (within the operational limits of approximately 500-5000 Hz). The neutron generator control module provides a TTL output pulse that serves as the DAC start pulse. The neutron output from the generator occurs at a delay, Δo, approximately 20-40 is after the start of the TTL pulse. The duration of the neutron beam pulse is determined by the selected PRF and the neutron generator duty factor, which is nominally fixed by the manufacturer at some value in the 5% to 10% range, but, in practice, is somewhat PRF dependent. FIGS. 5, 12, 14 and 16 show timing marks. The number of time bins in the DAG will be fixed at 256. Each bin has the same width, τo, which can be selected by the user to adjust the length of the DAG as required by the measurement to be made. The minimum value of τo is fixed at one μs by the current electronics in the system. The sum of neutron counts from all of the He-3 tubes in the detector is recorded in each time bin, as shown in FIGS. 4 and 5. In an embodiment, Δ1 is kept to its minimum value and Δ2 is set to zero in mode 1, in order to maximize data acquisition efficiency. In mode 2, LG, Δ1, and LC can all be set by the user. If these choices are not made judiciously [i.e., if LC<(LG+Δ1)], one could get a negative value of Δ2. In mode 2, the measurement requirements may require the neutron beam to be positioned entirely prior to the start of the DAG, more or less coincident with the DAG, or overlapping part of the DAG. Variability of the PRF, Δ1, and τo allows such flexibility in beam position. Note that the beginning and end of the neutron beam is not well defined in time. Also, the term “beam” is used loosely, here; the 14-MeV neutrons are emitted isotropically by generator, and do not form a spatial beam in the usual sense of the word. FIGS. 6 and 12 show examples of subgate detail. FIG. 14 illustrates another type of subgate counting. The level-1 subgates shown are equivalent to the fundamental time bins. In principle, each level-1 subgate could comprise two or more bins. If longer level-1 subgates are required, this can be achieved by increasing the size of τo. It is possible, in principle, to implement a data sorting algorithm that contains more subgates of level-2 and higher. There are possible modifications of the current implementation (containing the same numbers of subgates of each level) in which some of the longer subgates could comprise different groupings of time bins than the ones indicated in the figure. On any given DAC, the neutron multiplicities in some of those subgates would generally differ from the multiplicities in the illustrated set of subgates. The total multiplicity count in all subgates of a given length would, over a measurement of many DACs, be statistically equivalent for all such variations of the implementation shown. With reference to FIGS. 8, 10, 14 and 16, several constraints apply. For example, the average number of neutrons per DAG needs to be large. Any data acquisition cycles on which only zero or one neutron is detected provide no useful data for the Rossi-Alpha analysis. In order to collect data efficiently, it is necessary that an average of several (say≧10) neutrons be detected on each cycle. If two neutrons are counted in a single bin, the earlier of the two is considered to be the second member of a neutron pair with the nearest preceding neutron; the later neutron is the first member of a pair with the next succeeding neutron; and the two neutrons, themselves, constitute a pair separated by a time interval smaller than τo. This time is arbitrarily defined to be a time interval of “zero” width. If three neutrons occur in a single bin, there are two intervals of zero width, and so on. Embodiments described herein include a method for counting neutrons, comprising: collecting neutrons emitted from a source in a multi-detector array; inputting pulses from the multi-detector array in parallel to a plurality of separate inputs, wherein each input of the plurality of inputs is tied to a respective individual bit of a digital word; reading each digitized word at regular intervals to produce a plurality of read and digitized words, wherein all bits are read simultaneously to minimize latency; and storing each read and digitized word of the plurality of read and digitized words, wherein each read and digitized word is stored in a number of storage locations for subsequent processing. In this method, the multi-detector array may comprise a plurality of individual He-3 detector tubes. The digitized word may encode a count of neutrons emitted from the source. In an embodiment, the count may be obtained by a count process comprising: collecting neutron data in parallel input circuits; controlling a minimum summing interval for counting the data; summing the data in said summing interval, to produce a data sum; storing the data sum in multiple arrays; and building data structures by constructing summed sections from each array of the multiple arrays. The input signals may be edge triggered, and the minimum summing interval may be controlled with a clock. In an embodiment, each input circuit is operatively connected to multiple independent means for adding of the parallel set. The data structures may comprise data selected from the group consisting of multiple superset interval sizes, interval sizing after an external trigger, event totals in a fixed interval, event totals in a fixed interval after an external trigger, time intervals between events, time intervals between events after an external trigger, and arrival time of certain clump sizes after an external trigger. The data structures may also comprise multiple superset interval sizes, interval sizing after an external trigger, event totals in a fixed interval, event totals in a fixed interval after an external trigger, time intervals between events, time intervals between events after an external trigger, and arrival time of certain clump sizes after an external trigger. Embodiments are also directed to a method of neutron event counting, comprising: inputting edge triggered input signals into parallel input circuits observing each neutron event to be counted; controlling a minimum summing interval wherein data is counted, for use by a parallel set of means for adding, wherein each input circuit of said input circuits is operatively connected to multiple independent means for adding of said parallel set; reading a sum in each said means for adding during said minimum summing interval to produce a sum read; zeroing each said means for adding at the end of the minimum summing interval; storing said sum read into multiple arrays; and constructing summed sections from said array to build data structures comprising multiple superset interval sizes, interval sizing after an external trigger, event totals in a fixed interval, event totals in a fixed interval after an external trigger, time intervals between events, time intervals between events after an external trigger, and arrival time of certain clump sizes after an external trigger. Embodiments are further directed to a neutron detector system comprising: a multi-detector array collecting neutrons emitted from a source; an edge-detector having separate inputs, each receiving input pulses from the multi-detector array in parallel, wherein each input of the plurality of inputs is tied to a respective individual bit of a digital word; and a multi-mode counter module coupled to the edge-detector and comprising a plurality of simultaneously executing processing units, each processing unit utilizing a specific combination of trigger condition and count parameter for the input pulses. In this system, the trigger condition may be selected from one of an internal trigger and an external trigger. In an embodiment, the count parameter comprises the number of neutron multiplicities present in one or more time subgate periods during a data acquisition cycle of the detector. Alternatively, the count parameter comprises the distribution of neutron-pair time intervals during one or more data acquisition cycles of the detector. The data acquisition cycle may begin after a defined delay following the trigger condition. The system under an embodiment further comprises a word generator reading each digitized word at regular intervals to produce a plurality of read and digitized words, wherein all bits are read simultaneously to minimize latency. The system can further comprise a memory coupled to the word generator and storing each read and digitized word of the plurality of read and digitized words, wherein each read and digitized word is stored in a number of storage locations in the memory for subsequent processing. Aspects of the circuitry and methodology may be implemented as functionality programmed into any of a variety of circuitry, including programmable logic devices (“PLDs”), such as field programmable gate arrays (“FPGAs”), programmable array logic (“PAL”) devices, electrically programmable logic and memory devices and standard cell-based devices, as well as application specific integrated circuits. Some other possibilities for implementing aspects include: microcontrollers with memory (such as EEPROM), embedded microprocessors, firmware, software, etc. Furthermore, aspects of the memory test process may be embodied in microprocessors having software-based circuit emulation, discrete logic (sequential and combinatorial), custom devices, fuzzy (neural) logic, quantum devices, and hybrids of any of the above device types. The processor or processors mentioned or illustrated herein may be implemented as hardware circuitry embodied in one or more separate integrated circuit devices. It should also be noted that the various functions disclosed herein may be described using any number of combinations of hardware, firmware, and/or as data and/or instructions embodied in various machine-readable or computer-readable media, in terms of their behavioral, register transfer, logic component, and/or other characteristics. Computer-readable media in which such formatted data and/or instructions may be embodied include, but are not limited to, non-volatile storage media in various forms (e.g., optical, magnetic or semiconductor storage media) and carrier waves that may be used to transfer such formatted data and/or instructions through wireless, optical, or wired signaling media or any combination thereof. Examples of transfers of such formatted data and/or instructions by carrier waves include, but are not limited to, transfers (uploads, downloads, e-mail, etc.) over the Internet and/or other computer networks via one or more data transfer protocols (e.g., HTTP, FTP, SMTP, and so on). Unless the context clearly requires otherwise, throughout the description and the claims, the words “comprise,” “comprising,” and the like are to be construed in an inclusive sense as opposed to an exclusive or exhaustive sense; that is to say, in a sense of “including, but not limited to.” Words using the singular or plural number also include the plural or singular number respectively. Additionally, the words “herein,” “hereunder,” “above,” “below,” and words of similar import refer to this application as a whole and not to any particular portions of this application. When the word “or” is used in reference to a list of two or more items, that word covers all of the following interpretations of the word: any of the items in the list, all of the items in the list and any combination of the items in the list. The foregoing description is provided for purposes of illustration and is not intended to be exhaustive or to limit the invention to the precise form disclosed. Many modifications and variations are possible in light of the above teaching. The embodiments disclosed explain the principles of the invention and its practical application to thereby enable others skilled in the art to best use the invention in various embodiments and with various modifications suited to the particular use contemplated. The scope of the invention is to be defined by the following claims. |
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claims | 1. Composite material for neutron shielding and maintenance of sub-criticality comprising:(a) a matrix based on vinylester resin comprising at least one compound chosen from the group consisting of epoxyacrylate vinylester resins, epoxymethacrylate vinylester resins, bisphenol A type vinylester resins, novolac type vinylester resins, halogenated vinylester resins based on bisphenol A, and vinylester resins obtained from isophthalic polyester and urethane; and(b) an inorganic filler capable of slowing and absorbing neutrons, the inorganic filler comprising at least one inorganic compound of boron and at least one hydrogenated inorganic compound. 2. Material according to claim 1 in which the vinylester resin is an epoxy(meth)acrylate bisphenol A type resin complying with the following formula:in which R represents H or CH3. 3. Material according to claim 1 in which the vinylester resin is a novolac resin of formula:in which R represents H or CH3. 4. Material according to claim 1 in which the inorganic compound of boron is chosen from the group consisting of boric acid H3BO3, zinc borates Zn2O14.5H7B6, Zn4O8B2H2 and Zn2O14B6, colemanite Ca2O14B6H10, boron carbide B4C, boron nitride BN and boron oxide B2O3. 5. Material according to claim 1 comprising at least one boron compound chosen among the group consisting of zinc borate Zn2O14.5H7B6, and born carbide B4C. 6. Material according to claim 1 in which the hydrogenated inorganic compound is chosen from the group consisting of alumina hydrates and magnesium hydroxide. 7. Material according to claim 1 in which the quantities of inorganic hydrogenated compound and inorganic compound of boron are such that the boron concentration in the material is equal to 8×1022 to 15×1021 of boron atoms per cm3 and that the hydrogen concentration is 4×1022 to 6×1022 atoms per cm3. 8. Material according to claim 1, comprising 25 to 40% by weight of vinylester resin. 9. Material according to claim 1, with a density equal to or greater than 1.6, preferably 1.65 to 1.9. 10. Material according to claim 1, which can resist a minimum usage temperature of 160° C. 11. Process for preparation of a composite material according to claim 1, including the following steps:prepare a mix of vinylester resin in solution in a vinyl thinner with the inorganic filler,add a catalyst and an accelerator for hardening to the mix,degas the mix under a vacuum,pour the resulting mix in a mould, andallow the resulting mix to set in the mould. 12. Process according to claim 11, in which the vinyl thinner is styrene. 13. Process according to claim 11, in which the mould is a transport and/or storage packaging for radioactive products. 14. Transport and/or storage packaging for radioactive products comprising a shield formed from a composite material according to claim 1. |
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055925200 | summary | TECHNICAL FIELD The present invention relates to a latch handle for use in a control rod in a nuclear reactor and particularly relates to a latch handle which may be readily and easily installed in the control rod and which does not require any welding, machining or additional parts for retention by the control rod and service to release the control rod drive. BACKGROUND As well known, control rods in nuclear reactors form dual functions of power distribution shaping and reactivity control. The rods enter from the bottom of the reactor and are typically connected to bottom mounted, hydraulically actuated drive mechanisms which allow either axial positioning for reactivity regulation or rapid scram insertion. The control rod to control rod drive connection permits each control rod to be attached and detached from its drive during an outage, for example, for refueling, without disturbing the remainder of the control system for the control rod. The control rods are generally cruciform in cross-sectional shape with each blade of the rod containing tubes filled with boron carbide. The bottom of the control rod tube includes rollers for guidance of the rod during insertion and withdrawal as well as a velocity limiter. Each control rod when inserted into the core of the nuclear reactor has a fuel bundle in each of its quadrants. Each control rod typically includes at its lower end below the velocity limiter a coupling socket and a lock plug. The lock plug is mounted on an actuating shaft which passes upwardly coaxially along the control rod to a window in which is mounted a latch handle connected to the shaft. The coupling socket and lock plug are releasably attached to a coupling spud on the control rod drive by operation of the latch handle. Thus, the lock plug and socket receive the coupling spud and lock the control rod to the control rod drive. During an outage, when it is desirable to remove the control rod while leaving the control rod drive intact, and after removal of the fuel bundles, the latch handle is displaced upwardly displacing the lock plug from its locked condition with respect to the socket and coupling spud whereby the control rod is released from the control rod drive and may be removed. Conventional latch handles for control rods have laterally projecting tongues which engage in slots formed along side edges of the window of the control rod. Typically, these slots require either four strips which are welded in place to the control rod or two slots machined into the edges of a thin plate with an overlying strip to capture the latch handle. Welding of strips to the control rod is necessary in these prior art handles to capture the handle in the control rod. Substantial labor is involved in welding the necessary strips to form the slots and, in general, to locate the latch in the control rod window. DISCLOSURE OF THE INVENTION According to the present invention, there is provided a latch handle which performs the same function as latch handles of the prior art, i.e., to releasably lock the control rod and the control rod drive to one another yet which does not require welding or substantial labor to locate the handle in the window of the control rod. To accomplish this, the window in the control rod is formed with flanges projecting from its opposite sides. The latch handle is generally rectilinear in configuration with opposite sides defining slots for receiving the flanges along the opposite sides of the window. The latch handle, however, is sized and configured so that, in a first rotational orientation, the latch handle lies within the peripheral confines of the window and, upon rotation of the latch handle into a second angular orientation, the slots of the latch handle receive the flanges of the window sides. The latch handle in its second orientation is attached to the shaft mounting the lock plug. Hence, by sliding the latch handle in a vertical direction, the lock plug cooperates with the socket and spud to detach or attach the control rod and control rod drive relative to one another. It will be appreciated that the slots on the latch handle and the flanges on the window can be reversed with the flanges lying on the latch handle and the slots along the sides of the window. In a preferred embodiment according to the present invention, there is provided a control rod for a nuclear reactor comprising a control rod body having a plurality of blades projecting generally at right angles to one another, the body having a window defined by a plurality of generally linearly extending sides, a latch handle for connection to the control rod body and location in the window, the latch handle having a plurality of linearly extending sides, at least a pair of sides of one of the window and the latch handle having retaining slots and at least a pair of sides of another of the window and the latch handle having flanges for engaging in the slots, the sides of the latch handle and the window being configured so that the latch handle, in a first rotational orientation relative to the window, is receivable within the peripheral confines of the window and, upon rotation thereof into a second rotational orientation relative to the window, engages the periphery of the window with slots and flanges of the pairs of sides of the window and the latch handle engaging one another, respectively, to retain the latch handle in the window, the pairs of sides engaging one another in the second orientation of the latch handle relative to the window to enable sliding movement of the latch handle in at least one linear direction relative to the window. In a further preferred embodiment according to the present invention, there is provided a control rod for a nuclear reactor comprising a control rod body having an elongated axis and a plurality of laterally projecting blades, the body having a generally rectangular window defined by upper and lower edges and opposite sides, the window lying along the axis, a generally rectangular latch handle for connection to the control rod body and location in the window, the latch handle having a pair of opposite sides, the opposite sides of one of the window and the latch handle having retaining slots and the opposite sides of another of the window and the latch handle having flanges for engaging in the slots, the latch handle and the window being configured so that the latch handle, in a first rotational orientation relative to the window, is receivable within the window with the latch handle sides lying generally in opposition to the upper and lower edges of the window and, upon rotation thereof into a second rotational orientation relative to the window, has the sides thereof engaging with the opposite sides of the window, respectively, with the slots and flanges engaging one another to retain the latch in the window, the opposite sides of the latch handle and the opposite sides of the window engaging one another in the second orientation of the latch handle to enable sliding movement of the latch handle in at least one linear direction relative to the window and along the axis. In a still further preferred embodiment according to the present invention, there is provided a control rod for a nuclear reactor comprising a control rod body having a plurality of laterally projecting blades angularly related to one another, the body having a window with spaced edges, a latch handle for connection to the control rod body and location in the window, the latch handle having spaced edges, at least a pair of edges of one of the window and the latch handle having retaining slots and at least a pair of edges of another of the window and the latch handle having flanges for engaging in the slots, the edges of the latch handle and the window being configured so that the latch handle, in a first rotational orientation relative to the window, is receivable within the peripheral confines of the window and, upon rotation thereof into a second rotational orientation relative to the window, engages the periphery of the window with slots and flanges of the edges engaging one another, respectively, to retain the latch handle in the window, the slots and the flanges engaging one another in the second orientation of the latch handle relative to the window to enable sliding movement of the latch handle in at least one linear direction relative to the window. Accordingly, it is the primary object of the present invention to provide a novel and improved latch handle for a control rod which is readily and easily mounted within the window of the control rod without welding and reduced labor. |
description | This invention relates to a method for packing irradiated nuclear fuel into a cooling pond and to a device for sealing canisters containing irradiated nuclear fuel in a cooling pond, for temporary storage or retreating, or even storage in a deep site. Part of the management of irradiated fuels, after they have been used in a reactor, includes them undergoing a step where they are cooled in a cooling pond in a building, called the fuel building, which is generally located next to the reactor building. When this cooling step is complete, the irradiated fuels are removed from the cooling pond, and then taken to a temporary storage location until they are definitively disposed of either by retreating, or by geological storage. The cooling step in a cooling pond is of limited duration due to the reduced capacity of the cooling ponds. Within this context, it may be envisaged to pack the irradiated fuels in canisters forming the first confinement barrier, wherein each canister is itself placed inside a container forming a second confinement barrier and providing the mechanical resistance of the assembly. This container is designed for the transport of the canister to its temporary storage location. At present, irradiated nuclear fuels are packed in hot cells, which provide the confinement and radiation protection by means of shielded walls. These hot cells have the disadvantage of using heavy and costly constructions. The document FR 2 806 828 discloses a device for sealing under water a canister of irradiated fuel comprising a bell jar designed to seal tightly an open end of the canister, wherein a robot is taken inside the bell jar via a tube, connecting the bell jar to the outside environment, wherein this robot is designed to attach the cap onto the canister. The device permits a tube to be inserted to suck out the water inside the canister prior to the cap being fitted. This device and its associated method do not provide safe packing of the nuclear fuel, as no verification of the sealed confinement of the canister is carried out. Furthermore, when the irradiated fuel is removed from the cooling pond, a single biological protection formed by the canister separates it from the outside environment. Furthermore, the bell jar is connected to the outside environment by a duct permitting the passage, especially by means of welding. Consequently the sealed confinement of the radiation is difficult to implement. The document DE 8 906 938 describes an installation for confining nuclear fuel in a canister, wherein this installation is located in a pool. This two part installation comprises a lower cylindrical housing inside which the canister is directly placed before being filled with the fuel and a bell jar covers the top of the container. At the bottom of the lower housing is an evacuation for the water contained in the canister. This storage only provides a single biological barrier formed by the canister. Furthermore, this installation is very large and difficult to handle, and especially to transport to be placed in other cooling ponds. Furthermore, no verification of the sealed confinement of the container is described. Consequently one purpose of this invention is to propose a method for packing irradiated nuclear fuel which permits storage in open air without the risk of radiation. Another purpose of this invention is to propose a device which permits irradiated nuclear fuel to be packed in a cooling pond safely in packing offering double biological protection. Another purpose of this invention is to propose a device for sealing canisters in cooling ponds which is relatively cheap compared to the known devices. The purposes mentioned above are achieved by a method for packing nuclear fuel in a cooling pond, comprising the steps of: immersion in the cooling pond of a canister inside a container, fitting the sealing device on top of the container, lowering the level of the water in the sealing device to a level below that of the open end of the container, fitting a cap onto the canister, withdrawing the water from the canister, verifying the sealed confinement. A device according to this invention comprises a bell jar that can cover the container which holds the canister, wherein the bell jar is equipped with means of attaching the cap onto the canister, emptying and drying the canister, verifying the absence of water in the canister and the sealed confinement obtained by the cap. The bell jar is stored in the cooling pond and is positioned above a container to cover it when the nuclear fuel is packed. The device according to this invention has the advantage of being easy to transport. It allows the fuels to be packed directly in their temporary storage cooling pond, as well as rendering heavy means, such as hot cells, no longer necessary. The operating cost of such a device is therefore reduced. Furthermore, packing directly in a cooling pond offers the benefit of the biological protection provided by the water and avoids the transport of irradiated fuel without totally safe biological and mechanical protection. The bell jar has equipment which permits the fuels elements to be dried, a sealed cover to be welded onto the canister and the sealed confinement to be checked. All of the operations are carried out remotely and underwater. The subject-matter of the invention is therefore a method for packing nuclear fuel in a cooling pond inside a canister contained in a container, comprising the steps of: immersion of the container and the canister in the cooling pond, loading of the canister with nuclear fuel, fitting a canister device for sealing the top of the container, wherein said device at least comprises a first canister cap, maintaining a pneumatic fluid pressure in the device so as to maintain the water level at a level below that of an open end of the packing, attachment of a cap onto an open end of the canister by means contained inside the device, emptying the canister by emptying means inside the device, verification of the sealed confinement provided by the cap, withdrawal of the device. When emptying the canister, the water is withdrawn by means of a pump, and then the canister is swept with gas to dry the interior of the canister. Advantageously, after emptying, a check is made in the absence of water in the canister, for example by measuring the rise in pressure. The verification of the sealed confinement may include: sweeping with helium of the interior of the canister and pressurising the interior of the canister with helium, use of a helium spectrometer to detect any helium leaks. Preferably, the first cap is attached by welding. A second cap contained in the sealing device, is advantageously attached to the first cap prior to the withdrawal of the sealing device, wherein the latter may comprise gripping means. It is also possible to provide a cap on the container after the withdrawal of the sealing device. Very advantageously, the packing and sealing device are located in a determined position relative to one another, especially permitting the confinement method to be automated. The subject-matter of the invention is also a device for sealing a canister located inside a container in a cooling pond, wherein said canister is loaded with irradiated fuel, comprising a bell jar, at least one first cap equipped with connectors, means of fitting said cap onto the canister, means for attaching the cap onto the canister, means of withdrawing the water, means of verifying the complete withdrawal of the water and means of controlling the sealed confinement of the canister on the cap. The sealing device may comprise an arm forming the means of fitting the cap onto the canister. The emptying means may comprise a pump and means of sweeping the canister with gas so as to dry it, wherein said pump and said sweeping means may be connected by at least one connector. Preferably, the means of verifying the seal comprise means of injecting helium into the canister and a helium spectrometer; the means of injecting helium may be connected to at least one connector. The attachment means are advantageously welding means, wherein these means are brought close to the canister by an arm located inside the bell jar. Advantageously, the sealing device may contain a device for storing several first caps, permitting several canisters to be confined without having to raise the sealing device. The sealing device may also comprise a device for storing several second caps designed to be attached to the first caps. FIGS. 1A and 1B show a cooling pond 2 filled with water in which irradiated nuclear fuels 4 are stored to be cooled. These fuels may be in the form of an assembly of several fuel pencils or fuel plates or any other form. This cooling pond may accommodate a sealing device D according to this invention which may be kept immersed in the cooling pond even when it is not used. The cooling pond 2 comprises a bottom 6 equipped with a base 8 to hold a container 10 inside which a canister 12 is located. This canister 12 is designed to contain the irradiated fuel and forms a first confinement barrier, wherein the container 10 itself forms a second confinement barrier and also provides the mechanical resistance of the assembly. The container 10 is designed to provide the biological confinement and its mechanical protection during transport to a temporary storage location. In the rest of the description, the container 10 and canister 12 assembly will be called packaging and indicated by the reference 11. The method according to the invention is shown in block diagram form in FIG. 2 and especially provides that: in step 100, the container 10 into which the canister 12 has already been placed is immersed, in step 200, the canister 12 is filled with the irradiated fuel, in step 300, the device D is placed above the packaging so as to cover the packaging, wherein a lower end 28 of the device rests on the base 8 in the bottom of the cooling pond (FIG. 3), in step 400, the water level is maintained at a level that is below an open end 30 of the packaging 11, in step 500, a cap is fitted onto the canister and attached to it tightly, wherein the sealing device has means permitting the water contained in the canister to be removed, of drying the inside of the canister and of checking the confinement seal it makes for leaks, in step 600, the water contained in the canister is withdrawn, in step 700, the canister is checked to ensure that there is no water present, in step 800, the sealed confinement of the canister is checked. The device D, shown in the FIG. 3, comprises a bell jar 14 with substantially a cross section that is in the form of an inverted U, that may be moved in the cooling pond so that it may be positioned above the packaging. The bell jar may be moved by means of slings 15 attached to an upper part of the bell jar and connected to an overhead crane. The bell jar 14 has means of fitting the caps, means of attaching them for example by welding, means of emptying the interior of the canister, means of checking for the absence of water in the canister and means of verifying the sealed confinement of the canister (not shown) operating with the cap connectors. The bell jar 14 comprises at least one arm, two arms 16, 16′ in the example shown, to implement the means mentioned above. In the example shown in FIG. 3, the device according to the invention comprises an arm 16′ for fitting the caps and an arm 16 for welding the caps and verifying the confinement. The arms 16, 16′ are , for example formed by two elongated members 16.1, 16.1′, 16.2, 16.2′, articulated with respect to one another, wherein one of the members is equipped for example with gripping means (not shown), such as a clamp. The arms 16, 16′ may be controlled externally by operators or be moved by a computer programme without external intervention. The arm 16′ fits the caps, and advantageously fits the air or helium inlet tubes and attaches them to the cap connectors. The bell jar 14 advantageously comprises a device for storing 22 several canister caps. In the example shown in FIG. 3, this device 22 is placed on an inside wall of the bell jar 14. The storage in the bell jar 14 avoids having to raise the bell jar again between each confinement of a canister. Advantageously, this storage device permits two types of cap to be stored, a first type designed to be in contact with the nuclear fuel and a second type comprising means of gripping the canister and designed to be attached to an outside face of the first cap. Advantageously, the two types of cap are stored in distinct storage devices. The base 8 comprises means of positioning the bell jar 14 and the packaging 11 in a relative position that is identical to each loading of a canister 12. The base 8 comprises, for example a housing 24 with an internal diameter substantially equal to the external diameter of the packaging 11, and a disc 26 with an external diameter substantially equal to the internal diameter of a lower open end 28 of the bell jar 14, wherein the disc 26 and the housing 24 are coaxial in the example shown. Consequently, when the bell jar 14 is fitted onto the base 8, it is positioned in a unique position, as does the canister 12. It is therefore possible to automate the confinement method, as the positions of the packaging, especially the canister and the bell jar are always the same; the arms may then be moved in line with a pre-established command, without the intervention of an operator during the confinement steps. We will now describe in detail the specific steps of the method according to this invention. In step 100, the packaging 11 is immersed and fitted to the housing 24 of the base 8. In step 300, the device 300 is positioned above the packaging 11, the lower end 28 of the bell jar 14 moves into position around the disc 26 of the base 8. Each packaging element 11 and the bell jar 14 are always in the same relative position. In step 400, the water level is maintained below the open end 30 of the packaging 11 by a pressurising system which injects gas, for example pressurised air into the bell jar. Consequently, the upper end 30 of the packaging 11 is in the air even though it is positioned in the cooling pond, and the canister 12 may be confined in a dry environment while still benefiting from the biological protection provided by the water. Advantageously, the inside of the bell jar 14 is kept dry permanently by the pressurising system, to limit any potential damage to the equipment it contains. Furthermore, when it is fitted to the packaging, the water level is maintained by the pressurising system. In step 500, a first cap is removed from its storage device 22 by the arm 16 and is placed on the open end 13 of the canister 12 so that it seals it completely. The first caps comprise connectors which permit the drying to be controlled and Helium to be injected into the canister. These connectors are of the self-sealing type, well known to a person skilled in the art. After the helium has been injected into the canister, the end of the connectors is welded. As will be described below, these connectors permit to check for the presence of water in the canister and the sealed confinement provided by the cap once the cap has been fitted. The connectors also permit the canister to be emptied according to the step 600. This cap is then attached to the canister 12, advantageously by welding. It may also be provided that the cap is screwed on. The welding is for example tungsten welding (TIG). In step 600, the water contained in the canister is removed by means of a pump connected to the end of a connector of the cap. In addition to the extraction by means of a pump, it is provided to dry the inside of the canister by sweeping, for example with air. It may be envisaged to withdraw the water before fitting the cap. The canister 12 is centred in the container 10, which is itself centred in the part 24. In step 700, a verification is made of the absence of water in the canister, for example by measuring the pressure build up. This measurement consists of placing the canister in a depression and verifying the stability of the depression. If the measurement shows that there is still water in the canister, step 600 is repeated, especially as concerns the drying by sweeping with air. Then a new verification is carried out according to the step 700. If the measurement shows that the inside of the canister is dry, then step 800 is carried out, to verify the confinement seal provided by the cap and the weld. For example, a connector of the cap is used to inject a gas, advantageously helium, then using a helium spectrometer a check is made for helium leaks on the weld. If a leak is detected, which is to say that the weld has a fault, then step 1300 is carried out. In this step 1300, it may be provided that this weld is repaired if it is only a small fault. Otherwise the operations are stopped, the bell jar is removed, the canister is also removed from the packaging, then opened to recover the fuel. If no leaks are detected, then step 900 may be carried out, to remove a second cap from the storage device, and to attach it, for example by welding, onto the first cap. This second cap permits the canister to be gripped. For the welding, the means used to weld the first cap onto the canister may be used. If a second cap is not attached to the first cap, then step 1000 is carried out directly (arrow in dotted lines). In a later step 1000, the bell jar is removed. Then in step 1100, a cap is fitted to the container. It may also be envisaged to fit the cover onto the container when the bell jar is present. In step 1200, the container equipped with the loaded canister is removed from the cooling pond. The device for sealing according to this invention comprises a bell jar and all means required to implement the method as previously described. The sealing device therefore comprises: a bell jar, at least one first cap equipped with connectors, an arm to fit said cap onto the canister, means of attaching the cap onto the canister, means of withdrawing the water, formed for example by a pump and gas sweeping means, means of verifying the total withdrawal of the water by measuring a pressure rise, and, means of checking the sealed confinement of the canister on the cap, by injecting helium and detecting a helium leak using a helium spectrometer. Consequently, this device is completely autonomous and isolated from the outside environment by its immersion in the cooling pond. No radiation leaks may occur, as no transfer of equipment between the bell jar and the environment exposed to the air takes place. It may be further provided that the sealing device operates according to a determined sequence without human intervention. On the contrary, it may be provided that each step is controlled by an operator. The bell jar measures for example 2 meters in diameter and 3 meters in height. The sealing device according to this invention advantageously permits heavy means such as hot cells to be avoided, for the packing of the irradiated fuel. The sealing device according to this invention offers the advantage of being able to be transported to different places; therefore it may be used for several cooling ponds, which permits costs to be reduced. Furthermore, the packing in a cooling pond permits the biological protection provided by the water to be conserved. Furthermore, the use of a canister inside a container makes possible the use of a first cover without internal biological protection. The first cover then has a small thickness. Consequently the height of the canister is reduced, which results in better filling of the temporary storage devices. |
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056065892 | claims | 1. An air cross grid shaped as a flat panel to be interposed between the image receptor and the patient's breast spanning the X-ray beam traveling from an X-ray source to the image receptor in an X-ray mammography machine, absorbing scattered secondary radiation while transmitting primary radiation traveling from the X-ray source toward the image receptor, characterized by a first large plurality of closely spaced partition walls generally parallel to each other and arrayed across the width of the flat panel, a second large plurality of closely-spaced partition walls generally parallel to each other and arrayed across the length of the flat panel in a direction substantially transverse to the first partition walls, said transverse pluralities of partition walls intersecting each other at a large plurality of cross junctions and defining between themselves a large plurality of independent open air passages extending completely through the thickness of said flat panel, said panel comprising a stacked plurality of metallic foil sheets bonded together forming a laminated integral stack, each foil sheet having a large plurality of air passage openings extending therethrough defined by intersecting partition segments positioned to comprise the portions of said partition walls formed by that foil sheet, the partition segments in each foil sheet being substantially aligned with the partition segments of the adjacent foil sheet in said stack at a focal point centrally located near one edge of said panel, and slightly offset in progressively increasing misalignment at progressively greater distances from said focal point, said open air passages having central axes generally perpendicular to said panel and all converging at an X-ray source focus positioned on a line perpendicular to said panel passing through said focal point thereon, 2. The air cross grid defined in claim 1, wherein each of the partition walls has a wall thickness between about five percent (5%) and about ten percent (10%) of the minimum dimension of each of the open air passages, measured in the plane of the grid. 3. The air cross grid defined in claim 1 wherein each of the partition walls has a wall thickness of at least seven percent (7%) of the minimum dimension of each of the open air passages measured in the plane of the grid. 4. The air cross grid defined in claim 1 wherein the minimum dimension of each of the open air passages measured in the plane of the grid is between about fifteen percent (15%) and about thirty-five percent (35%) of the overall thickness of the air cross grid panel. 5. The air cross grid defined in claim 1 wherein the minimum dimension of each of the open air passages measured in the plane of the grid is no greater than twenty-five percent (25%) of the overall thickness of the air cross grid panel. 6. The air cross grid defined in claim 1 wherein the partition segments in each foil sheet are slantingly aligned by progressively greater angular amounts at progressively greater distances from said focal point, said partition segments thereby lying in planes substantially parallel to the converging central axes of the adjacent open air passages flanking each said segment, whereby said slight offset misalignments between partition segments of adjacent foil sheets are minimized. 7. The air cross grid defined in claim 1 wherein the metallic foil sheets are formed of a beryllium-copper alloy having a thickness of about 0.004 inches. 8. The air cross grid defined in claim 7 wherein the overall grid thickness is between about 0.050 inches and about 0.100 inches. 9. The air cross grid defined in claim 7 wherein the overall grid thickness does not exceed 0.080 inches. 10. The air cross grid defined in claim 1, wherein the grid panel has a rectangular peripheral shape, with said two pluralities of partition walls intersecting at substantially right angles, one plurality of partition walls all extending in a direction (45) angularly displaced by an acute angle (A), measured from a direction (44) parallel to said focal point edge of said panel, whose trigonometric tangent can be expressed as a ratio of two integers, each integer being less than 20. 11. The air cross grid defined in claim 10, wherein the trigonometric tangent ratio is 11:16, or 0.688, corresponding to an acute angle of 34.6.degree.. 12. The air cross grid defined in claim 1, wherein each partition segment of each foil sheet is encased in a thin metal plated coating. 13. The air cross grid defined in claim 12, wherein the substantially aligned partition segments of the adjacent metallic foil sheets in said stack have their metal plated coatings heat-fused together to form said laminated integral stack. 14. The air cross grid defined in claim 12, wherein each foil sheet is formed of a beryllium-copper alloy, and the thin metal plated coating on each partition segment incorporates a first plated coating of copper, covered by an overlying plated coating of lead. |
summary | ||
abstract | A neutron generator is provided with a flat, rectilinear geometry and surface mounted metallizations. This construction provides scalability and ease of fabrication, and permits multiple ion source functionalities. |
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043404995 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a method for the treatment of concentrated radioactive solutions which contain organic compounds as sequestering agents or complex formers. 2. Description of the Prior Art Such liquid wastes contain radioactive substances and complex formers as for example, citric, oxalic, formic acid or ethylenediaminetetraacetic acid (EDTA) and must never be discharged. These wastes must rather be counted among the radioactive wastes and must be disposed of safely as such, possibly after being concentrated, through appropriate storage, for instance, in salt mines. For this purpose it is necessary to solidify them first. One possibility of this kind is to process them with cement and additives into concrete blocks. Radioactive solutions with organic complex formers, however, are difficult to bind in concrete and in addition, the radionuclide complexes present therein are soluble and can therefore be leached out readily. Heretofore, such liquid wastes with organic complex formers always had to be evaporated, since precipitation reactions for removing the radioactive substances are prevented or at least interfered with by these complex compounds. Since such a method, however, is very complicated and requires considerable energy, the problem arose to find a substantially simpler method for compacting the radioactive substances contained in these liquid wastes. |
052079759 | abstract | According to the present invention, there is provided a hydraulic control rod driving system in which radial clearance is formed between a piston portion and a cylinder, one or two or more step portions each composed of a cylindrical upper part having a large diameter and a cylindrical lower part having a small diameter arranged adjacent to the lower side of the upper part are provided on the piston portion along an axial direction thereof, circumferential grooves for partitioning off respective step portions are provided, and a ratio of the length of each step portion to the maximum outside diameter of the piston portion is set at 0.5 to 1.0. |
claims | 1. A beam control circuit for an ion implantation system comprising:a high voltage switch connected in series with a power supply and an electrode associated with an ion source portion of the ion implantation system, operable to interrupt or reestablish a connection between the power supply and the electrode; anda switch controller operable to control the duty factor of a beam produced within the ion implantation system by controlling the high voltage switch to close before ion implantation, to remain closed during ion implantation, and to open after ion implantation, thereby minimizing particle contamination. 2. The system of claim 1, further comprising one or more protection circuits associated with the high voltage switch, operable to absorb energy from reactive elements external to the high voltage switch, and to limit an over-voltage across the switch. 3. The system of claim 2, wherein the power supply and the electrode associated with the ion source portion of the ion implantation system comprises one or more of a suppression voltage power supply and an extraction voltage power supply and electrode associated with the ion source. 4. The system of claim 2, wherein one of the protection circuits is connected in series with the HV switch it protects. 5. The system of claim 2, wherein one of the protection circuits is connected in parallel with the HV switch it protects. 6. The system of claim 1, further comprising a synchronization circuit operable to synchronize and time two or more switch controllers of two or more beam control circuits for the opening and closing of two or more high voltage switches for the ion implantation system. 7. The system of claim 1, wherein the power supply and the electrode associated with the ion source portion of the ion implantation system comprises a cathode voltage power supply and electrode associated with the ion source. 8. The system of claim 1, wherein the power supply and the electrode associated with the ion source portion of the ion implantation system comprises an arc voltage power supply and electrode associated with the ion source. 9. The system of claim 1, wherein the duty factor of the beam comprises a ratio of a predetermined on-time to the predetermined on-time plus an off-time, wherein the predetermined on-time generally corresponds to the ion implantation time and an ion beam settling time, and the predetermined off-time generally corresponds to an idle time of the beam after ion implantation. 10. The system of claim 1, wherein the switch controller is further operable to control the high-voltage switch to close for a first time interval before a start of the ion implantation and to remain closed for a second time interval after a conclusion of the ion implantation. 11. The system of claim 10, wherein one of the first and second time intervals is between about one millisecond and about two minutes. 12. The system of claim 10, wherein one of the first and second time intervals is about one millisecond or less. 13. The system of claim 1, wherein the switch controller is operable to detect a current or voltage change associated with the electrode and to control one or more high voltage switches to open or close based on the detection to quench an arc associated with the ion source. 14. The system of claim 1, wherein the switch controller is further operable to terminate the ion beam during one of, arrival at a wafer load or unload position, a manual beam OFF switch operation, an arc detection, and a wafer exchange, and to initiate the ion beam during one of, a manual beam ON switch operation, subsequent to a wafer exchange, following a load operation, and upon a command to implant another wafer. 15. The system of claim 1, wherein the switch controller is further operable to receive a beam duty factor command from the ion implantation system or a motion control system to disable the high voltage switch during one of, arrival at a wafer exchange position, receipt of a manual beam OFF switch command, waiting for a next wafer or an implant auto recovery, and prior to a wafer exchange. 16. The system of claim 1, wherein the switch controller is commanded by an external control input. 17. The system of claim 1, further comprising:a trigger control circuit operable to detect a current or voltage change associated with the electrode and to control one or more high voltage switches to open or close based on the detection; andone or more protection circuits, each protection circuit associated with one of the high voltage switches, operable to absorb energy from reactive elements external to the respective high-voltage switch, and to limit an over-voltage across the switch. 18. A beam control circuit for minimizing particle contamination in an ion implantation system comprising:an ion source for producing a quantity of ions which can be extracted in the form of an ion beam;a high voltage switch connected in series between a power supply and an electrode associated with the ion source portion of the ion implantation system, operable to interrupt or reestablish a connection between the power supply and the electrode; anda switch controller operable to initiate the ion beam before ion implantation by closing the high voltage switch, to maintain the ion beam during implantation by having the high voltage switch remain closed, and to terminate the ion beam after ion implantation by opening the high voltage switch, thereby minimizing particle contamination. 19. The system of claim 18, further comprising one or more protection circuits associated with the high voltage switch, operable to absorb energy from reactive elements external to the high voltage switch, and to limit an over-voltage across the switch. 20. The system of claim 19, wherein the power supply and the electrode associated with the ion source portion of the ion implantation system comprises one or more of a suppression voltage power supply and an extraction voltage power supply and electrode associated with the ion source. 21. The system of claim 19, wherein one of the protection circuits is connected in series with the HV switch it protects. 22. The system of claim 19, wherein one of the protection circuits is connected in parallel with the HV switch it protects. 23. The system of claim 18, further comprising a synchronization circuit operable to synchronize and time two or more switch controllers of two or more beam control circuits for the opening and closing of two or more high voltage switches for the ion implantation system. 24. The system of claim 18, wherein the power supply and the electrode associated with the ion source portion of the ion implantation system comprises one or more of a cathode voltage power supply and electrode associated with the ion source. 25. The system of claim 18, wherein the power supply and the electrode associated with the ion source portion of the ion implantation system comprises an arc voltage power supply and electrode associated with the ion source. 26. The system of claim 18, wherein the duty factor of the beam comprises a ratio of a predetermined on-time to the predetermined on-time plus an off-time, wherein the predetermined on-time generally corresponds to the ion implantation time and an ion beam settling time, and the predetermined off-time generally corresponds to an idle time of the beam after ion implantation. 27. The system of claim 18, wherein the switch controller is further operable to control the high-voltage switch to close for a first time interval before a start of the ion implantation and to remain closed for a second time interval after a conclusion of the ion implantation. 28. The system of claim 27, wherein one of the first and second time intervals is between about one millisecond and about two minutes. 29. The system of claim 27, wherein one of the first and second time intervals is about one millisecond or less. 30. The system of claim 19, wherein the switch controller is further operable to terminate the ion beam during one of, arrival at a wafer load or unload position, a manual beam OFF switch operation, an arc detection, and a wafer exchange, and to initiate the ion beam during one of, a manual beam ON switch operation, subsequent to a wafer exchange, following a load operation, and upon a command to implant another wafer. 31. The system of claim 19, wherein the switch controller is further operable to receive a beam duty factor command from the ion implantation system or a motion control system to disable the high voltage switch during one of, arrival at a wafer exchange position, receipt of a manual beam OFF switch command, waiting for a next wafer or an implant auto recovery, and prior to a wafer exchange. 32. The system of claim 19, further comprising:a trigger control circuit operable to detect a current or voltage change associated with the electrode and to control one or more high voltage switches to open or close based on the detection; andone or more protection circuits, each protection circuit associated with one of the high voltage switches, operable to absorb energy from reactive elements external to the respective high-voltage switch, and to limit an over-voltage across the switch. 33. A method of reducing the duty factor of an ion beam to minimize particle contamination in an ion implantation system utilizing a beam control circuit comprising a switch controller and a high-voltage switch connected between a voltage supply and an electrode associated with an ion source of the ion implantation system, the method comprising:receiving one of a beam ON command or a beam OFF command;initiating the ion beam before a start of ion implantation by closing the high-voltage switch connected between the voltage supply and the electrode associated with the ion source, in response to the beam ON command;implanting ions; andterminating the ion beam after a conclusion of ion implantation by opening the high voltage switch in response to the beam OFF command, thereby minimizing the duty factor of the beam and reducing particle contamination. 34. The method of claim 33, further comprising synchronizing two or more switch controllers having two or more high voltage switches used to initiate or terminate the ion beam of the ion implanter. 35. The method of claim 33, wherein the switch controller is further operable to control the high-voltage switch to close for a first time interval before the start of ion implantation and to remain closed for a second time interval after the conclusion of ion implantation. 36. The method of claim 33, wherein the ion beam is forced on or off by the high-voltage switch. 37. The method of claim 33, wherein the beam control circuit is further operable to control the high-voltage switch to open for a predetermined time period following the detection of an arc. 38. The method of claim 33, wherein the switch controller is further operable to terminate the ion beam during one of, arrival at a wafer load or unload position, a manual beam OFF switch operation, an arc detection, and a wafer exchange, and to initiate the ion beam during one of, a manual beam ON switch operation, subsequent to a wafer exchange, following a load operation, and upon a command to implant another wafer. |
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summary | ||
052182096 | abstract | An ion implanter for implanting ions into a batch of semiconductor wafers comprises a wafer holding disk of the centrifugal holding type, and a plurality of wafer rests in the wafer holding disk having a wafer holding surface which is conically curved. When the wafer holding disk is rotated, the wafer is pushed onto the wafer holding surface so that the surface of the wafer is curved nearly in the same manner as the conically-curved inner surface of the peripheral portion of the s wafer holding disk As a result, an ion beam being irradiated upon the surface of the wafer is always perpendicular to the surface of the wafer. |
summary | ||
claims | 1. A collimator for X-ray imaging apparatus, said collimator comprising:a collimator housing including a tube flange;a tube;a locating ring configured to be mounted at an outlet of the tube flange; andat least one tongue set fixed on the locating ring, wherein an outstretching direction of the tongue is towards a center of the locating ring. 2. The collimator according to claim 1, wherein the at least one tongue set further comprises:a tongue set housing including a tongue outlet on a first end face thereof;a tongue in the tongue set housing and including a head configured to move out from the tongue and a tail having a pullback section; anda trigger pin configured to control the movement of the tongue head. 3. The collimator according to claim 2, further comprising a block plate connected to a second end face of the tongue set housing opposite to the first end face, the block plate configured to cover the pullback section of the tongue tail. 4. The collimator according to claim 1, further comprising at least three tongue sets uniformly disposed about the locating ring. 5. The collimator according to claim 2, further comprising at least three tongue sets uniformly disposed about the locating ring. 6. The collimator according to claim 3, further comprising at least three tongue sets uniformly disposed about the locating ring. 7. The collimator according to claim 1, further comprising a rotating lock knob on the locating ring, the rotating lock knob configured to control a rotation of the collimator. 8. The collimator according to claim 2, further comprising a rotating lock knob on the locating ring, the rotating lock knob configured to control a rotation of the collimator. 9. The collimator according to claim 3, further comprising a rotating lock knob on the locating ring, the rotating lock knob configured to control a rotation of the collimator. 10. A method for assembling and disassembling the collimator for X-ray imaging apparatus according to claim 1, the method comprising:an assembling method comprising:rotating an operation platform system to an assembling position to place the tube and the tube flange in position;placing the collimator in the assembling position;pressing a trigger pin; andmounting a block plate; anda disassembling method comprising:rotating the operation platform system to a disassembling position;removing the block plate;pulling back the tongue; andmoving the collimator away from the operation platform system. 11. An X-ray imaging apparatus comprising the collimator according to claim 1. 12. A method for assembling and disassembling a collimator for an X-ray imaging apparatus, the collimator including a collimator housing including a tube flange, a tube, a locating ring configured to be mounted at an outlet of the tube flange, and at least one tongue set fixed on the locating ring, wherein an outstretching direction of the tongue is towards a center of the locating ring, said method comprising:rotating an operation platform system to an assembling position to place the tube and the tube flange in position;placing the collimator in the assembling position;pressing a trigger pin; andmounting a block plate. 13. The method according to claim 12, further comprising:rotating the operation platform system to a disassembling position;removing the block plate;pulling back the tongue; andmoving the collimator away from the operation platform system. 14. An X-ray imaging apparatus comprising a collimator comprising:a collimator housing including a tube flange;a tube;a locating ring configured to be mounted at an outlet of the tube flange; andat least one tongue set fixed on the locating ring, wherein an outstretching direction of the tongue is towards a center of the locating ring. 15. The X-ray imaging apparatus according to claim 14, wherein the at least one tongue set comprises:a tongue set housing including a tongue outlet on a first end face thereof;a tongue in the tongue set housing and including a head configured to move out from the tongue and a tail having a pullback section; anda trigger pin configured to control the movement of the tongue head. 16. The X-ray imaging apparatus according to claim 15, wherein the at least one tongue set further comprises further comprising a block plate connected to a second end face of the tongue set housing opposite to the first end face, the block plate configured to cover the pullback section of the tongue tail. 17. The X-ray imaging apparatus according to claim 16, wherein the at tongue comprises a locating hole and a slotted hole connected to the locating hole, wherein the locating hole is wider than the slotted hole. 18. The X-ray imaging apparatus according to claim 17, wherein the head of the tongue is wider than a remainder of the tongue such that the tail of the tongue passes through an opening on the block plate. 19. The X-ray imaging apparatus according to claim 14, wherein the collimator comprises at least three tongue sets uniformly disposed about the locating ring. 20. The X-ray imaging apparatus according to claim 14, wherein the locating ring comprises rotating lock knob configured to control a rotation of the collimator. |
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claims | 1. A reactor-internal equipment handling method of loading/unloading equipment into/from a reactor vessel with a reactor-internal equipment handling apparatus comprising: control rod holding means for releasably holding a control rod which is loaded in a reactor vessel, fuel support/control rod guide tube holding means for releasably holding both a fuel support, which supports a bottom end of a fuel assembly, and a control rod guide tube, on which the fuel support is placed at a top end, a main body frame to which both the control rod holding means and the fuel support/control rod guide tube holding means are fitted and adapted to be hung down inside the reactor vessel, a holding state detecting mechanism for detecting both a holding state of the control rod holding means about the control rod, and a holding state of the fuel support/control rod guide tube holding means about the fuel support and the control rod guide tube, and a position state detecting mechanism for detecting a positioning state of the main body frame in the reactor vessel, the method comprising: detecting a positioning state of the main body frame in the reactor vessel with the positioning state detecting mechanism in order to confirm that the main body frame is properly positioned; holding the control rod by the control rod holding means and also holding both the fuel support and the control rod guide tube with the fuel support/control rod guide tube holding means; detecting, with the holding state detecting mechanism, both a holding state of the control rod holding means about the control rod and a holding state of the fuel support/control rod guide tube holding means about the fuel support and the control rod guide tube, to confirm that the control rod holding means holds the control rod and the fuel support/control rod guide tube holding means holds the fuel support and the control rod guide tube; and hoisting the main body frame by a fuel exchanger and simultaneously loading or unloading all of the control rod, the fuel support, and the control rod guide tube into or from the reactor vessel. 2. A reactor-internal equipment handling method according to claim 1 , wherein the fuel support/control rod guide tube holding means includes an orifice engaging member which is adapted to engage edge portions of orifices formed in the fuel support and the control rod guide tube, an orifice engaging member linking mechanism for manipulating the orifice engaging member, and orifice engaging member driving means for driving the orifice engaging member linking mechanism, and wherein the orifice engaging member is simultaneously brought into contact with the edge portions of the orifices formed in the fuel support and the control rod guide tube. claim 1 |
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056053614 | claims | 1. A method for replacing a nozzle which is attached to a pressure vessel of a nuclear power facility passing through a bore in the vessel with a replacement nozzle, comprising: removing the entire nozzle to be replaced from the vessel; providing a mechanical attachment of the full replacement nozzle to the vessel with the replacement nozzle passing through the bore in the vessel from which the nozzle to be replaced was removed; and providing a mechanical seal of the full replacement nozzle with the vessel implemented either as part of the step of providing a mechanical attachment of the nozzle to the vessel or separately therefrom; said mechanical attachment and said mechanical seal not including a weld of the nozzle to the vessel. providing a temperature gradient between the vessel adjacent to the bore and the nozzle sufficient to insert the nozzle into the bore and sufficient when the temperature gradient is substantially reduced to provide a mechanical attachment of the nozzle to the vessel in the bore and such that a mechanical seal is produced between the nozzle and the wall of the bore, said mechanical attachment and said mechanical seal not including a weld of the nozzle to the vessel; inserting the nozzle into the bore; and reducing said temperature gradient effective to produce said mechanical attachment and said mechanical seal. providing a nozzle having a first portion with a diameter therealong smaller than the diameter of the bore, a second portion with a diameter therealong larger than the diameter of the bore, and a flange therebetween; inserting the nozzle portion with the smaller diameter into the bore with the flange bearing against the exterior circumference of the pipe; attaching the flange to the pipe with a suitable clamping device; and mechanically sealing the exterior of the nozzle with respect to the pipe without a weld of the nozzle to the pipe. a full nozzle extending from the interior of said vessel, through said bore and exiting said bore to the exterior of said vessel, threads on a portion of the nozzle projecting from said vessel, means associated at least with said nozzle which engages said vessel and prevents said nozzle from being withdrawn through said bore to the exterior of said vessel, and a nut threaded and tightened on said nozzle which bears against the exterior of said vessel and causes said means to firmly engage said vessel without a weld of said nozzle to said vessel; and means for mechanically sealing said nozzle to said vessel not including a weld of said nozzle to said vessel. a full nozzle extending from the interior of said vessel, through said bore and exiting said bore to the exterior of said vessel, a flange attached to or engaging said full nozzle outside of said bore, bolts attaching said flange to the exterior of said vessel; and said full nozzle and said bore having interfering portions which engage when said flange is bolted to said vessel to provide mechanical attachment of said full nozzle to said vessel and also to produce a mechanical seal of said full nozzle and said bore at said interfering portions, said mechanical attachment and said mechanical seal not including a weld of said full nozzle to said vessel. a full nozzle extending from the interior of said vessel, through said bore and exiting said bore to the exterior of said vessel, a flange attached to or engaging said full nozzle outside of said bore, bolts attaching said flange to the exterior of said vessel adjacent the exterior of said vessel to provide a mechanical attachment of said nozzle to said vessel which does not include a weld of said nozzle to said vessel, and gasket material positioned between said flange and the exterior of said vessel which produces a mechanical seal of said nozzle with the exterior of said vessel which does not include a weld of said nozzle to said vessel. a full nozzle extending from the interior of said vessel, through said bore and exiting said bore to the exterior of said vessel, threads on said nozzle and threads in said bore by which said nozzle is threaded to said bore, said nozzle and said bore including interfering portions which engage when said nozzle is tightened into said bore to provide a mechanical attachment of said nozzle to said vessel and a mechanical seal of said nozzle within said bore of the engaging interfering portions of said nozzle and said bore, said mechanical attachment and said mechanical seal not including a weld of said nozzle to said vessel. providing a mechanical attachment of the full nozzle to the vessel passing through the bore in the vessel; and mechanically sealing the full nozzle with the vessel implemented either as part of the step of providing a mechanical attachment of the nozzle to the vessel or separately therefrom; said mechanical attachment and said mechanical seal not including a weld of the nozzle to the vessel. said full nozzle having a flange attached to or engaged with said full nozzle outside of said bore, bolts engaging said flange and threaded to the exterior of said vessel to attach said flange to said vessel. a flange attached to or engaging said full nozzle outside of said bore, bolts attaching said flange to the exterior of said vessel adjacent the exterior of said vessel, and gasket material positioned between said flange and the exterior of said vessel which provide a mechanical attachment of said nozzle to said vessel which does not include a weld of said nozzle to said vessel, a mechanical sealing of said nozzle to said vessel which does not include a weld of said nozzle to said vessel, or both. 2. The method of claim 1 wherein the nozzle to be replaced is welded to the vessel. 3. The method of claim 1 wherein providing the mechanical attachment of the replacement nozzle to the vessel comprises clamping the replacement nozzle to the vessel. 4. The method of claim 1 wherein providing the mechanical attachment of the replacement nozzle to the vessel comprises bolting the replacement nozzle to the vessel. 5. The method of claim 4 wherein bolting the replacement nozzle to the vessel comprises threading the replacement nozzle in the bore. 6. The method of claim 4 wherein bolting the replacement nozzle to the vessel comprises bolting to the vessel a flange of the replacement nozzle which is attached to or engaged with the replacement nozzle. 7. The method of claim 1 wherein providing the mechanical attachment of the replacement nozzle to the vessel comprises attaching the nozzle in the bore with an interference fit. 8. The method of claim 1 wherein providing a mechanical seal of the nozzle to the vessel comprises positioning gasket material between the replacement nozzle and the vessel and pressing that material against the vessel in or adjacent the bore sufficiently to create a seal between the replacement nozzle and the vessel. 9. The method of claim 1 wherein providing a mechanical seal of the nozzle to the vessel comprises configuring the replacement nozzle and the bore so that a mechanical seal is formed therebetween when the replacement nozzle is forced into the bore and forcing at least a part of the replacement nozzle into a sealing engagement with the bore. 10. A method of mechanically attaching a nozzle in a bore of a pressure vessel of a nuclear power facility, the nozzle having a larger outer diameter than the inner diameter of the bore at a same given temperature of the nozzle and the vessel adjacent to the bore, the method comprising: 11. The combination of a pressure vessel of a nuclear power facility which has a bore therein and a full nozzle extending completely through said bore, means for mechanically attaching said nozzle to said vessel which does not include a weld of said nozzle to said vessel and means for mechanically sealing said nozzle to said vessel which does not include a weld of said nozzle to said vessel. 12. The combination of claim 11 wherein said mechanically attaching means comprises threads on a portion of said nozzle projecting from said vessel, means associated at least with said nozzle which engages said vessel and prevents said nozzle from being withdrawn through said bore to the exterior of said vessel, and a nut threaded and tightened on said nozzle which bears against the exterior of said vessel and causes said means to firmly engage said vessel. 13. The combination of claim 12 comprising a spacer between said nut and the exterior of said vessel, said nut bearing against the exterior of said vessel through said spacer. 14. The combination of claim 12 wherein said engaging means comprises a flange attached to said nozzle surrounding said bore on the interior of said vessel. 15. The combination of claim 12 wherein said engaging means comprises interfering portions of said nozzle and said bore. 16. The combination of claim 15 wherein said nozzle and said bore have circular cross sections, and a portion of said nozzle within said bore has a larger diameter than the largest diameter of said bore thereby providing said interfering portions. 17. The combination of claim 16 wherein at least a portion of said bore and at least a portion of said nozzle are tapered. 18. The combination of claim 14 wherein said mechanically sealing means comprises gasket material positioned between said flange and the interior of said vessel. 19. The combination of claim 14 wherein said mechanically sealing means comprises a mechanical engagement of said flange and the interior of said vessel sufficient to produce a seal therebetween. 20. The combination of claim 15 wherein said mechanically sealing means comprises gasket material between said interfering portions of said nozzle and said bore. 21. The combination of claim 15 wherein said mechanically sealing means comprises polished surfaces of said interfering portions of said nozzle and said bore and a mechanical engagement at least between said polished surfaces of said interfering portions of said nozzle and said bore sufficient to produce a seal therebetween. 22. The combination of claim 11 wherein said nozzle comprises a sleeve and a separate nozzle body, and wherein said mechanically attaching means comprises threads on said sleeve and threads in said bore, a flange on said nozzle body engaged by said sleeve threaded in said bore and interfering portions of said nozzle body and said bore which engage when said sleeve is tightened into said bore and mechanically attach said nozzle to said vessel and mechanically seal said nozzle in said bore. 23. The combination of claim 22 wherein said mechanically sealing means comprises polished surfaces of said interfering portions of said nozzle and said bore and a mechanical engagement at least between said polished surfaces of said interfering portions of said nozzle and said bore sufficient to produce a seal therebetween. 24. The combination of claim 22 wherein said mechanically sealing means comprises gasket material between said interfering portions of said nozzle and said bore. 25. The combination of claim 11 wherein said mechanically attaching means comprises threads on said nozzle and threads in said bore, and said nozzle and said bore include interfering portions which engage when said nozzle is tightened into said bore and mechanically attach said nozzle to said vessel and mechanically seal said nozzle in said bore. 26. The combination of claim 25 wherein said nozzle and said bore have circular cross sections, and a portion of said nozzle within said bore has a larger diameter than the largest diameter of said bore, thereby providing said interfering portions. 27. The combination of claim 26 wherein at least a portion of said bore and at least a portion of said nozzle are tapered. 28. The combination of claim 25 wherein said mechanically sealing means comprises gasket material between said interfering portions of said nozzle and said bore. 29. The combination of claim 25 wherein said mechanically sealing means comprises polished surfaces of said interfering portions of said nozzle and said bore and a mechanical engagement between said polished surfaces of said interfering portions of said nozzle and said bore sufficient to produce a seal therebetween. 30. The combination of claim 11 wherein said mechanically attaching means comprises a flange attached to or engaging said nozzle on the exterior of said vessel, and a plurality of bolts engaging said flange and said vessel for attaching said flange and said nozzle to said vessel. 31. The combination of claim 30 comprising means associated at least with said nozzle for engaging said vessel and preventing said nozzle from being drawn into the interior of said vessel. 32. The combination of claim 31 wherein said engaging means comprises interfering portions of said nozzle and said bore. 33. The combination of claim 32 wherein said nozzle and said bore have circular cross sections, and a portion of said nozzle within said bore has a larger diameter than the largest diameter of said bore thereby providing said interfering portions. 34. The combination of claim 33 wherein at least a portion of said bore and at least a portion of said nozzle are tapered. 35. The combination of claim 30 wherein said nozzle and said bore are cylindrical, and wherein said flange is attached to said nozzle, and wherein said mechanically sealing means comprises gasket material at least between said flange and the exterior of said vessel, said flange bearing tightly against said vessel through said gasket material. 36. The combination of claim 32 wherein said mechanically sealing means comprises at least one O-ring type gasket material between said nozzle and said bore. 37. The combination of claim 11 wherein said nozzle and said bore have circular cross sections of approximately the same diameter at corresponding locations thereof, and said mechanically attaching means comprises an interference fit between said nozzle and the interior of said bore along substantial portions thereof. 38. The combination of claim 37 wherein said nozzle and said bore are cylindrical. 39. The apparatus of claim 37 wherein said mechanically sealing means comprises said interference fit being sufficient to produce a seal between said nozzle and said bore. 40. The combination of claim 37 wherein said mechanically attaching means further comprises at least one flanged portion of said nozzle in an end of said nozzle within said vessel in engagement with the interior of said vessel adjacent said bore so as to prevent movement of said nozzle out of said bore to the exterior of said vessel. 41. The combination of claim 40 comprising a plurality of said flanged portions spaced radially apart relative to said bore in engagement with the interior of said vessel adjacent said bore. 42. A method for attaching a nozzle to a pipe in a nuclear power facility which has a bore therein through the circumference thereof comprising: 43. The combination of a pipe in a nuclear power facility which has a bore therein through the circumference thereof and a nozzle, said nozzle having a first portion with a diameter therealong smaller than the diameter of said bore, a second portion with a diameter therealong larger than the diameter of said bore, and a flange therebetween, said smaller diameter nozzle portion passing through said bore, said larger diameter portion projecting from said bore and said flange bearing against the outer circumference of said pipe, a clamp device having a hole therein through which said nozzle larger diameter portion passes attaching said flange and with it said nozzle to said pipe, and means mechanically sealing the exterior of said nozzle with respect to said pipe without a weld of the nozzle to the pipe. 44. The combination of claim 43 wherein said sealing means comprises gasket material positioned between said flange and the outer circumference of the pipe. 45. The combination of a nozzle assembly and a pressure vessel of a nuclear power facility, said vessel having a bore, comprising: 46. The combination of a nozzle assembly and a pressure vessel of a nuclear power facility which has a bore, comprising: 47. The combination of claim 46 comprising gasket material positioned between said engaging interfering portions of said nozzle and said bore. 48. The combination of a nozzle assembly and a pressure vessel of a nuclear power facility which has a bore, comprising: 49. The combination of a nozzle assembly and a pressure vessel of a nuclear power facility which has a bore, comprising: 50. The combination of claim 49 wherein said nozzle comprises a sleeve having said threads and a nozzle body which includes the portion interfering with a portion of said bore, a portion which passes through said sleeve and a shoulder portion which is engaged by said sleeve when said sleeve is tightened to said bore to cause said interfering portions to engage. 51. A method for installing a full nozzle to a pressure vessel of a nuclear power facility which has a bore, comprising: 52. The combination of claim 46 wherein said nozzle is tapered for at least a portion of its length and said bore is correspondingly tapered for at least a portion of its length to provide said interfering portions which engage when said flange is bolted to said vessel. 53. In the combination of a full nozzle and a pressure vessel of a nuclear power facility which has a bore, said full nozzle extending from the interior of said vessel, through said bore and exiting said bore to the exterior of said vessel and being structurally attached to said vessel, the improvement comprising: 54. The combination of claim 53 wherein said bolts engage said flange to compressively load said nozzle to said vessel, and provide a mechanical attachment of said nozzle to said vessel which does not include a weld of said nozzle to said vessel, a mechanical sealing of said nozzle to said vessel which does not include a weld of said nozzle to said vessel, or both. 55. In the combination of a full nozzle and a pressure vessel of a nuclear power facility which has a bore, extending from the interior of said vessel, through said bore and exiting said bore to the exterior of said vessel and being structurally attached to said vessel, the improvement comprising: 56. The method of claim 9 herein configuring the replacement nozzle and the bore comprises providing polished surfaces thereof which when pressed together form said mechanical seal. |
abstract | The present invention provides a method and apparatus for dynamic adjustment of an active sensor list. The method includes providing an active sensor list indicative of at least one sensor associated with at least one processing tool, the at least one sensor being communicatively coupled to a network having an associated bandwidth, receiving information indicative of a state of at least one of the processing tools, and modifying the active sensor list based on the information indicative of the state of the at least one of the processing tools and the associated network bandwidth. |
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abstract | A first apparatus includes an imaging component configured to capture an earthbound image of the sky from a terrestrial location, a chronometric component, a communication component configured to transmit data representative of a captured earthbound image of the sky, and a controller. The controller is configured to cause an earthbound image of the sky to be captured using the imaging component at a time identified by the chronometric component and data representative of the captured earthbound image to be transmitted. A second apparatus includes a computer and a computer readable medium accessible by the computer and including data representative of a master mapping of the sky relative to the Earth and computer-executable instructions for determining a terrestrial location based on data representative of a captured earthbound image of the sky and an identified time at which the earthbound image was captured. A system includes the first and second apparatus. |
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047818832 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS Turning first to FIG. 9, the cask 38 of the present invention includes generally cylindrical container 40 having a grid basket assembly 42 disposed therein. Grid basket assembly 42 provides an array of storage slots 44, configured much like a matrix of vertically disposed pigeonholes, each accommodating a cell 46. The fuel assemblies 20 are placed in the cells 46 for storage. With continuing reference to FIG. 9, container 40 includes a base element 48 having an upper portion configured to receive lid element 50. The cavity 51 provided within base element 48 has substantially cylindrical walls 53 rising from floor 55, which is substantially horizontal during long-term storage. Elements 48 and 50 include carbon steel portions 52 to which inter cladding layers 54 of stainless steel are affixed. Carbon steel portions 52 are approximately 25 cm thick and serve to protect the environment from gamma rays. Layer 54 can be applied to base element 48, for example, by placing it on a turntable and rotating it while welding a continuous spiral path around the interior using stainless steel welding rods, thereby applying a stainless steel surface while completely covers the floor 55 and side walls 53 of element 48. Outside of portions 52, elements 48 and 50 are provided with a layer about 7.0 cm thick of neutron absorbing material 56, which may be a resin. A suitable resin for use as material 56 is commercially available from Bisco Products, Inc., 1420 Renaissance Drive, Park Ridge, Ill. 60068, under Stock No. N.S.- 3. Surrounding material 56 is outer layer 58 of stainless steel to protect cask 38 from the environment. Cask 38 also includes cooling fins 60 of carbon steel, preferably treated to protect the carbon steel from chemical attack by the environment. Fins 60 are welded to portion 52 and extend through material 56 and layer 58. In this embodiment, cask 38 is approximately 4.8 meters high and has an outer diameter of about 2.5 meters, excluding fins 60. When loaded with spent fuel, cask 38 has a mass of over a hundred thousand kilograms. Although not illustrated, it is advantageous to affix a pair of trunnions at the top and bottom of base element 48 to facilitate handling. Turning now to FIG. 3, grid basket assembly 42 includes major plates 62, 64, 66, 68, 70, 72, 74, 76 78 and 80. These plates may be fabricated of aluminum sheets approximately 3.7 meters high and 2.0 cm thick. With reference to FIG. 4, the bottom portion of plate 66, for example, includes five downwardly oriented slots 82 dividing the bottom portion of plate 66 into six segments, each of these six segments forming a portion of one of the four walls 83 (FIG. 3) of a storage slot 44. In a similar manner, the upper portion of plate 76, for example, includes five upwardly extending slots 84. The slots 82 in plate 66 and the slots 84 in plate 76 are about 26.3 cm apart, and it will be apparent that such dimensions provide slots 44 which are positioned about 26.3 cm apart, center-to-center. During assembly, plates 62, 64, 66, 68 and 70 are held at right angles to plates 72, 74, 76, 78 and 80, and the slots 82 are then inserted into slots 84 to provide intersections which run the entire length of the plates. The plates are joined by full length fillet welds along opposite sides of the intersections, so that there are two welds running along each intersection. This is best illustrated in FIG. 5 where, for example, welds 86 are oppositely disposed along the intersection formed by plates 68 and 78. Returning to FIG. 3, minor plates 88 are welded at one end of plate 66 and minor plates 90 are welded at the other end. Similarly, minor plates 92 are welded at one end of plate 76 and minor plates 94 are welded at the other end. Minor plate 96 is welded at the intersection of plates 62 and 72, minor plate 98 is welded at the intersection of plates 62 and 80, minor plate 100 is welded at the intersection of plates 80 and 70, and minor plate 102 is welded at the intersection of plates 70 and 72. Minor plates 103 are welded to plates 64, 68, 74 and 78. Turning next to FIGS. 5 and 6, cell 46 will now be described. To support the cell panels, tabs 104 are welded to the plates adjacent the intersections. Panel portions 106, of stainless steel about 0.25 cm thick, are bent to provide cell walls 107 and flanges 108, which slidably extend beween tabs 104. During storage, walls 107 contact the sides of fuel assembly 20 to keep it positioned at a predetermined location in slot 44 in order to maintain the spacing between fuel assemblies 20. In the preferred embodiment, fuel assemblies are positioned centrally within slots 44 and, for the sake of convenience, it will be said that they are "centered" therein by walls 107, although it will be understood that this does not exclude other uniform positioning patterns. It will be apparent that the distance by which walls 107 project into slot 44 in order to center the fuel assembly is determined by the dimensions of the particular fuel assembly which the cell 46 will enclose. Stainless steel wrapper portions 110, about 0.076 cm thick, are welded to portions 106 in order to support sheets 112 of "neutron poison" such as boron carbide. Sheets 112, which are about 19 cm wide and 0.19 cm thick, are present to moderate neutrons emitted from fuel assemblies 20 when cask 38 is in pool 30 while the fuel assemblies are being loaded into it. Sheets 112 help to maintain the critically factor K.sub.eff at less than 1 in order to obviate the possibility of a chain reaction. That is, the sheets 112 and the spacing limit the nuclear interaction between fuel assemblies 20, which are of course designed to promote chain reactions when they are present in a reactor. It is worth noting that sheets 112 have served their purpose after the borated water has been drained from cask 38 during the loading process, since thereafter the potential for a chain reaction is reduced. The foregoing details concerning the thickness of the stainless steel elements and the neutron poison are characteristic of the remaining cell embodiments disclosed herein and need not be repeated. If desired, holes 114 can be provided in portions 110 in order to permit visual confirmation that sheets 112 are present and to facilitate the escape of borated water when cask 38 is drained. During drainage the temperature within cask 38 rises to the boiling temperature of the borated water, so that a prolonged drying time is not required. With continuing reference to FIG. 6, panel portions 106 include projections 116 which extend beyond grid basket assembly 42 when portions 106 are installed. The projections 116 are bent to provide flanges 118. Turning now to FIG. 7, brace elements 120 are welded to the projections 116 in order to support the portion of cell 46 that extend beyond grid basket assembly 42. It will be apparent from FIG. 7 that flanges 118 provide a funnel to guide fuel assemblies 20 into the cell 46 as cask 38 is being loaded. FIG. 8 illustrates cells 46 installed in grid basket assembly 42. Cells 46 are mechanically attached to grid basket assembly 42 only at lower portion 122 of assembly 42, so that the upper portions of assembly 42 and cells 46 may move with respect to each other in order to accommodate differences in the thermal coefficients of expansion between cells 46 and assembly 42. Differential expansion between grid 42 and cells 46 may amount two or more centimeters within the range of temperatures encountered during operation of cask 38. It will be noted from FIG. 8 that panel portions 106, without poison sheets 112 and wrapper portions 110, provide one wall of each of the two cells 46 between plates 74, 78 and 62. The reason for this is that the cell walls lie at the periphery of assembly 42 and consequently neutrons that are generated within these cells and that travel directly to container 40 need not be moderated in order to avoid a critical mass. In a similar manner, the outer wall of the cells 46 bounded by plates 64, 68 and 72, the outer walls of the cells 46 bounded leg plates 74, 78 and 70, and the outer wall of the cells 46 bounded by plates 64, 68 and 80, can be provided by panel portions 106 without wrapper portions 110 and sheets 112 of neutron poison. FIG. 10 illustrates grid basket assembly 42, having cells 46 installed therein in order to provide storage slots 44 for receiving fuel assemblies 20, mounted within cask 38. In accordance with the present invention, cask 38 and grid basket assembly 42 are fabricated separately. Cells 46 are installed in grid basket assembly 42 (except for the eight cell walls at the periphery of assembly 42 which are formed by panel portions 106 alone, without portions 110 or sheets 112) and, before assembly 42 is inserted within cask 38, stainless steel channel sections (which will be described below) are fitted to the projecting ends of plates 64, 66, 68, 74, 76, 78 and 96, 98, 100 and 102. The portions of stainless steel cladding layer 54 which will receive the channel sections are machined to provide smooth surfaces, and then assembly 42 with the channel sections attached but not permanently joined thereto is inserted into cask 38. Full length fillet wells (preferably applied by a welding robot) are then used to permanently attach the channel sections to cask 38, following which the eight peripheral cell walls formed by panel portions 106 alone are installed. Installation of these eight peripheral cell walls is delayed in this manner in order to permit space for welding the channel sections. It should be noted that welding the channel sections after grid basket assembly 42 is installed, rather than before, avoids the possibility that assembly 42 might "hang up" while partially installed due to the close tolerances involved. Turning now to FIGS. 11, 12 and 13, the coupling of grid basket assembly 42 to cask 38 by way of channel sections will now be described in more detail. These Figures illustrate only three such couplings, but it will be understood that a total of 16 connections are actually present, each of these couplings being accomplished in accordance with one of FIGS. 11-13. In FIG. 11, an elongated channel section 124 is attached via full length fillet welds 126 to the wall 53 of cask 38. Channel section 124 has a generally U-shaped channel 128 which is slightly wider than the thickness of plate 66, thereby providing narrow gaps 130, which are perhaps 0.025-0.05 cm wide if the gaps 130 are equal on each side. However, gaps 130 need not be symmetrical, and no adverse consequences arise if plate 66 contacts one side of channel 128. Before cask 38 is sealed, it is preferably flooded with an inert gas, such as helium, which transmits heat well through narrow gaps. However, other insert gases such as nitrogen can be used and, moreover, caks performance is acceptable with air. There is also a space 132 at the end of channel 128, thereby permitting section 124 to accommodate variations in the dimensions of plate 66 due to differential expansion of cask 38 and assembly 42 at differential temperatures. FIG. 12 illustrates an elongated channel section 134 welded to the inner wall 53 of cask 38 in order to accommodate an edge of plate 64, and FIG. 13 illustrates an elongated channel section 136 welded to the inner wall 53 of cask 38 in order to accommodate plate 96. As was the case in FIG. 11, there are narrow gaps 130 between the plates and the channel sections, these narrow gaps producing little hindrance to heat transfer, particularly in a helium atmosphere, and spaces 132 to accommodate differential expansion of the elements. It is desirable to elevate grid assembly 42 above the floor 55 of cask 38 in order to facilitate the removal of borated water when spent fuel is being prepared for storage in the cask. In order to achieve this elevation, FIG. 14 illustrates the lower portion 138 of channel section 124 as being solid in order to elevate plate 66. The lower portions of the remaining channel sections are, of course, configured in a similar manner. Alternatively, the plates of grid basket assembly 42 may be provided with drainage cut-outs. FIG. 15 illustrates plates 140 and 142, corresponding to plates 66 and 76 of FIG. 4, respectively, with drainage cut-outs 143. A further cell embodiment is illustrated in FIGS. 16 and 17. Unlike the previous embodiment, the embodiment of FIGS. 16 and 17 does not require tabs 104. This is advantageous because the usable space in the storage slot 44 is effectively enlarged, so that the same grid basket assembly 42 which is used for storage of fuel assemblies 20 can, alternatively, be employed to store fuel in consolidated form. Use of cask 38 for consolidated storage will be discussed subsequently. With reference to FIGS. 16 and 17, cell 144 includes a hollow stainless steel shell element 146 having wall portions 148 joined by corner portions 150. As in other cell embodiments, the particular fuel assembly to be stored in cell 144 determines the cross-sectional area defined by wall portions 148. Corner portions 150 contact grid assembly 42 adjacent the intersections of the plates thereof. Four stainless wrapper portions 152 are welded to shell element 146 between corner portions 150 in order to support neutron poison sheets 154. If desired, wrapper portions 152 may having openings to facilitate drainage of borated water and to permit verification that sheets 154 are present. In the embodiment of FIGS. 16 and 17, brace elements 120 are unnecessary in order to support the portion of cell 144 that extends above grid assembly 42, since support in this region is provided by corner portions 150. However, the upper ends of wall portions 148 are bent outward to provide flanges 156 to act as funnels for guiding a fuel assembly 26 into the cell. As was the case with the previous embodiment, the cells 144 can be attached to grid assembly 42 at bottom end 122 thereof in order to permit differential expansion of the cells 144 with respect to assembly 42 at different temperatures. However, differential expansion is still permitted if, instead of affixing cells 144 to grid assembly 42, the cell bottoms are allowed to rest directly on floor 55 (in which case drainage cut-outs, similar to cut-outs 143 in FIG. 15, should be provided). This alternative is, of course, advantageous, since the cells 144 would not have to be installed in assembly 42 before assembly 42 is inserted in container 40, so that it would be unnecessary to dedicate cask 38 at the time of manufacture to the storage of fuel assemblies 20 having particular dimensions. Cells 144 having interior dimensions appropriate for particular fuel assemblies 20 can be installed after cask 38 is manufactured, or cells can be omitted entirely if fuel is to be stored in consolidated form, as will be discussed subsequently. Thus, cask 38 need not be tailored to a particular storage application, a factor which contributes significantly to the versatility of the cask. For the sake of convenience, cells which can readily be installed in cask 38 after it is fabricated, or removed from cask 38 after storage of fuel assemblies 20 so that cask 38 can be prepared to receive fuel in consolidated form, will be deeded "removable" cells. It will be apparent that cells 144 are removable cells, unless they are permanently installed in assembly 42 at the time of manufacture, since they can readily be slid into or out of storage slots 44. Turning next to FIGS. 18 and 19, another removable cell embodiment, one having spacer elements on the walls of the cell in order to center the cell within a storage slot 44, will now be discussed. In these Figures cell 158 includes a four-sided stainless steel shell element 160 whose upper end is provided with flanges 161 to funnel the fuel assembly 20 into the cell. Stainless steel wrapper portions 162 are welded to element 160 in order to support neutron poison sheets 166 between element 160 and portion 162. The spacer elements in this embodiment are provided by dimpled portions 168 in shell element 160 adjacent the upper end thereof and dimpled portions 170 positioned at various heights along wrapper portions 162. The bottom end of cell 158 can be bolted or welded to the lower end 122 of grid basket assembly 42 in the manner previously described or it may simply rest upon a support element as illustrated in FIGS. 22 and 23. In FIGS. 22 and 23, a stainless steel support element 172 is provided with intersecting channels 173 to accommodate the bottoms of the plates of grid basket assembly 42. In FIG. 22, plates 66 and 76 are supported at their intersection, and it will be apparent that the other plate intersections are also provided with support. Elements 72 are welded to grid assembly 42 but not to the floor 55 of cask 38, thereby supporting grid 42 above the floor of cask 38 in order to facilitate the drainage of borated water from cask 38 when spent fuel is being loaded while nevertheless permitting differential expansion of cask 38 with respect to grid basket assembly 42. Support elements 172 are provided with flanges 174 against which the lower ends of cells 158 rest after they are inserted into storage slots 44 and with sloping walls 175 to center the cells and guide the lower ends towards flanges 174 as they are being inserted. In this way, dimpled portions 168, dimpled portions 170, and flanges 174 properly position the cells 158 within grid basket assembly 42 without bolting, welding, or clamping. The embodiment of FIG. 20 is similar to that of FIGS. 18 and 19 except that dimpled portions 168 are replaced by spacer members 176 that are welded to shell element 178. Although not illustrated in FIG. 20, spacer members are also welded to stainless steel wrapper portions 179. The thickness of members 176 is, of course, greater than the thickness of the members welded to wrapper portions 178 so as to facilitate assembly of the cells into the grid basket assembly 42 while also providing lateral support of the cells. Unlike the embodiments of FIGS. 18-20, the removable cell embodiment of FIG. 21 does not rest upon support elements 172 or upon floor 55. Instead, stainless steel spacer member 180 having a hooked portion 182 is welded to stainless steel shell element 184. Member 180 not only spaces the cell away from the slot walls 83, in the manner of member 176 in the embodiment of FIG. 20, it also supports the cell adjacent the top thereof so that the cell hangs from the top rather than being supported from the bottom. It will be apparent that this arrangement, like the previous embodiments, accommodates differential expansion of the cell with respect to grid basket assembly 42. Although not illustrated in FIG. 21, spacing members are also welded to stainless steel wrapper element 186, which supports sheet 188 of a neutron poison such as boron carbide. When cask 38 is loaded entirely with fuel assemblies 20, the cell embodiments previously described prevent a self-sustaining reaction before the borated water is drained both by supporting neutron poison and by positioning the fuel assemblies 20 in storage slots 44 provided by grid basket assembly 42 in order to ensure that fuel assemblies 20 are adequately spaced apart. These factors are sufficient to prevent criticality even under anomalous conditions. FIG. 24 is a graph of the criticality or neutron multiplication K.sub.eff, assuming an anomalous condition, within a cask 38 that has been fully loaded with 24 fuel assemblies 20 following storage in pool 30 of borated water for approximately ten years, each of the fuel assemblies 20 having an array of 15.times.15 fuel rods 26. The anomalous condition which is assumed in FIG. 24 is that fresh, or non-borated, water is introduced into cask 38. In FIG. 24 the abscissa indicates fresh water density D in grams/cm.sup.3, that is, the mass of the liquid water and water vapor in cask 38 divided by the total volume of liquid which can be contained by the cask. It will be apparent that a density of 1.0 gm/cc indicates that the cask is fully flooded with fresh water, and K.sub.eff is highest under this condition both for the cells disclosed herein (solid curve) and for cells which lack neutron poison (dotted curve). It will be noted that the solid curve lies below K.sub.eff =0.95 even at maximum fresh water density. Moreover, as indicated by the dotted curve, K.sub.eff would remain less than 1.0, until cask 38 is over half full of fresh water, even if neutron poison were omitted from all the cells. Particularly if the wrapper portions have openings through which the presence of neutron poison can be visually confirmed, this possibility is extremely remote and does not present a creditable condition for evaluating the performance of cask 38. Moreover, if all of the cells were to lack poison sheets but the cask were to be flooded with borated water rather than fresh water, the maximum K.sub.eff would be approximately 0.83. Thus, K.sub.eff remains below 0.95 under one anomalous condition (cask flooded with fresh water) or another (all poison sheets absent), and only the theoretical possibility that both anomalous conditions might, somehow, occur simultaneously, would result in criticality. Of course, the only water which should ever find its way into cask 38 is borated and the opportunity for visual inspection would ensure that the cells are adequately provided with poison sheets. With poison sheets, K.sub.eff reaches approximately 0.74 when cask 38 is flooded with borated water. It is generally recognized in the art that K.sub.eff is higher when fuel assemblies are stored than when fuel is stored in consolidated form under conditions that are otherwise identical. The reason for this is that water slows or "thermalizes" neutrons, thereby increasing the probability of a fissionable reaction, and that fuel which is packed tightly effectively removes space which would otherwise be available for water. Accordingly K.sub.eff within cask 38 remains below 0.95 if some or all of the storage slots 44 are used for consolidated storage. As in known in the art, in consolidated storage the fuel rods 26 are removed from the fuel assemblies 20 and placed in a consolidated canister which holds a greater number of fuel rods than the number in a single fuel assembly. The nozzles 22 and 24, along with other non-fuel portions of the fuel assemblies 20, such as grid members 28, can then be stored separately in a low-level storage facility. FIG. 25 illustrates a stainless steel consolidation canister which is filled with spent fuel. Canister 190 includes a base portion 192, a hollow body portion 194 having a collar 196 at the upper end thereof, and a lid portion 198 which is affixed by screws 200 to collar 196 in order to close canister 190. Bars 202 are welded to lid portion 198 to support rod 204. Rod 204 is present in order to accept a hook (not illustrated) for hoisting canister 190 into cask 38 after canister 190 is loaded with spent fuel. Both the loading and hoisting, of course, are accomplished by remote control. With continuing reference to FIG. 25, the exterior of body portion 194 is provided with projections 206 which contact the walls of a storage slot 44 when canister 190 is inserted in cask 38 in order to provide lateral support for canister 190 within cask 38. Body portion 194 is also provided with openings 208 to permit passage of borated water. In FIG. 25, fuel rods 26 can be seen through openings 208. FIG. 27 illustrates the upper end of body portion 194 after it has been filled with the spent fuel rods 26 that have been removed from two fuel assemblies 20. Divider 210 is provided as part of body portion 194 and includes a stainless steel plate 212 having a flange 214 that is welded to one side of portion 194 and another plate 216 having a flange 218 that is welded to the other side of portion 194. On the side of plate 212 opposite flange 214 is a leg 220 terminating in a flange 222 that is welded to plate 216. In a similar manner, leg 224 of plate 216 has a flange 226 which is welded to plate 212. With continuing reference to FIG. 27, the rods 26 from one fuel assembly 20 are inserted on one side of divider 210 and the rods 26 from another fuel assembly are inserted on the other side. Different fuel assembly designs, however, have different numbers of rods, and divider 210 permits the canister 190 to be customized when it is manufactured in accordance with the design of the particular fuel assembly 20 whose rods 26 it will store. That is to say, it will be apparent that the space bounded by plate 212, leg 220, plate 216 and leg 224 can readily be adjusted during fabrication of canister 190 in order to provide a snug fit for the fuel rods 26 that canister 190 is to receive. With continuing reference to FIG. 27, threaded bores 228 through collar 196 of portion 194 are provided for receiving screws 200 (FIG. 25) when lid portion 198 (FIG. 25) is attached. Turning next to FIG. 26, base portion 192 includes a foot plate 230 of stainless steel. The sides of plate 230 have hollowed regions 232 communicating with a passageway (not illustrated) through portion 192 in order to facilitate drainage of borated water from body portion 194. Plate 230 also has bevelled regions 234 at the bottom thereof. In the event that a canister 190 is used with a cask 38 having a grid basket assembly 42 that is supported by elements 172 (see FIGS. 22 and 23), the bevelled regions 234 of plate 230 overlap flanges 174 of elements 172 and permit plate 230 to rest securely on floor 55 of cask 38, with edges 236 of plate 230 being nestled against sides 238 of elements 172. With continuing reference to FIG. 26, stainless steel support 240 is welded to plate 230. Support 240 is hollow and has walls which crumple in the event that canister 190 is accidentally dropped along its axis after it has been loaded with spent fuel; this crumpling absorbs energy which might otherwise rupture canister 190 during the accident. Stainless steel pedestal 242 is welded to support 240 and provides an outwardly extending rim 224 abutting the lower end of body portion 194. Platform portion 246 of pedestal 242 extends slightly into the interior of body portion 192 and supports the lower ends of fuel rods 26. Portion 246 provides another way to customize canisters 190 to the fuel rods 26 of particular fuel assembly designs. That is, platform portion 246 is manufactured to telescope further into body portion 194 if relatively short rods 26 are to be stored than if relatively long rods 26 are to be stored. Accordingly, it will be apparent to those skilled in the art that divider 210 and pedestal 242 permit canister 190 to be tailored during manufacture to the fuel rods of particular fuel assemblies 120. Thus, during manufacture, cask 38 need not be dedicated to either storage mode (that is, storage of intact fuel assemblies or storage of consolidated fuel) or to storage of spent fuel from any particular source (that is, spent fuel, either intact or consolidated, from any particular fuel assembly design). All that is necessary is that appropriately configured cells and/or consolidation canisters be present during the loading operation. Since both the cost and manufacturing lead time for cells and canisters are significantly less than for a storage cask, the versatility of cask 38 is expected to have considerable commercial significance. A cask 38 can simultaneously store both consolidated fuel and intact fuel assemblies. Moreover, the storage mode may be changed after storage has begun. For example, if a cask 38 originally stores 24 fuel assemblies 26, it can be returned to pool 30 and opened, the fuel rods 26 can be removed from the fuel assemblies 20 and placed in consolidation canisters 190, and the cask 38 can then be re-loaded with more spent fuel than it originally stored. FIGS. 28 and 29 indicate the effectiveness of cask 38 in isolating spent fuel from the environment. In these Figures, it is assumed that all 24 storage slots 44 in cask 38 have been loaded with fuel in consolidated form, each canister 190 having 450 fuel rods 26 following storage thereof in borated water for approximately ten years. FIG. 28 illustrates the dose rate, in milliroentgen/hour, along the exterior surface of the sides of cask 38 between adjacent fins 60. As can be seen in FIG. 28, the dose rate D varies with the height H, in centimeters, above the bottom of the cask. Between heights of approximately 100 and 350 cm, the maximum neutron dose rate (solid curve) is less than 20 mrem/hr; the maximum primary gamma dose rate (that is, gamma rays from the interior of cask 38 which have passed through the cask walls) is less than 30 mrem/hr (dotted curve); and the maximum secondary gamma dose rate (gamma rays generated by neutron absorbing material 56 when neutrons are absorbed) is less than three mrem/hr (dot-dash curve). Accordingly, the maximum total dose rate at the side surface of cask 38 is less than 60 mrem/hr, and this dose rate falls rapidly with distance from the side surface. FIG. 29 illustrates the dose rate D, in milliroentgen/hour, at the external surface of the bottom of cask 38. The bottom might be exposed, for example, during transportation of cask 38 to a remote storage location. In FIG. 29, "R" indicates the distance in centimeters along the bottom surface from the axis of cask 38 to the side walls, and it will be seen that D drops rapidly as the side walls are approached. At the bottom surface, the maximum neutron dose rate (solid curve) is less than 15 mrem/hr; the maximum primary gamma dose rate (dotted curve) is less than 10 mrem/hr; and the maximum secondary gamma dose rate (dot-dash curve) is less than 2 mrem/hr. From the foregoing description it will be apparent that the present invention provides a versatile and mechanically rugged spent fuel storage cask which reliably shields the environment from radiation produced by spend nuclear fuel stored therein despite temperature variations which arise during the loading of the fuel into the cask and long-term storage. Heat from fuel assemblies stored in the cells and/or fuel rods stored in consolidation canisters is transmitted by the grid assembly to channel sections affixed to the inner cask walls via narrow gaps between the U-shaped channels and the ends of the plates. The channel sections also provide space between the bottoms of the channels and the ends of the plates in order to accommodate radial movement of the grid basket assembly with respect to the cask as temperature changes. The various cell embodiments ensure tha a sustained nuclear reaction cannot occur as fuel assemblies are loaded into the cask and, moreover, several cell embodiments are disclosed which permit ready conversion of the cask and grid basket assembly for the storage of consolidated fuel. Although the foregoing discussion has described the preferred embodiments of the invention with reference to pressurized water reactors, in which case the water in pool 30 would be borated, it will be apparent to those skilled in the art that the present invention could be used with spent fuel from a boiling water reactor. It will be understood that the above description of the present invention is susceptible to various modifications, changes and adaptations, and the same are intended to be comprehended within the meaning and range of equivalents of the appended claims. |
claims | 1. An X-ray multi-layer mirror comprising an Mo/Si alternate layer with a non-uniform, non-periodic film thickness structure, which is produced by conducting optimization processing for widening an X-ray reflection characteristic on a constant film thickness fundamental structure of an Mo/Si alternate layer having the X-ray reflection characteristic, wherein each one of Mo layers and Si layers for forming the non-uniform film thickness structure is designed to have a film thickness of 1.5 nm or more. 2. An X-ray exposure apparatus comprising a projection optical system comprising an X-ray multi-layer mirror according to claim 1, wherein a pattern of an original plate is transferred to a substrate by the projection optical system. |
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claims | 1. A grating with a large aspect ratio, comprising:a multiplicity of recurring alternating grating webs and grating gaps, each of the respective grating webs including a height; anda multiplicity of filler beams, respectively arranged in the grating gaps with a spacing from one another in a direction of the gaps, each of the respective filler beams connecting respectively adjacent grating webs over the respective height of the grating webs, wherein the grating webs and the grating gaps run from a first to a second side of the grating, and the filler beams each include a width in a direction of the gaps, the width being at most 10% of the spacing between two adjacent filler beams, and wherein the spacings between respective adjacent ones of the filler beams in each of the grating gaps do not vary by more than 10% in the entire grating. 2. The grating as claimed in claim 1, wherein the spacings between the respective adjacent filler beams in each of the grating gaps are the same in the entire grating within the scope of procedural accuracy. 3. The grating as claimed in claim 1, wherein the spacing (1) between two of the filler beams in one of the grating gaps satisfies the following geometric condition: l = k h · ( p 2 ) 3 ,where k is a constant, h is a height and p is the recurrence of the grating. 4. The grating as claimed in claim 3, wherein the constant k is less than 5000 μm−1. 5. The grating as claimed in claim 3, wherein the constant k lies in the range between 750 μm−1 and 2500 μm−1. 6. The grating as claimed in claim 1, wherein the filler beams are positioned randomly in one of the grating gaps whilst maintaining the separation criterion. 7. The grating as claimed in claim 1, wherein, in each grating gap, there is a first filler beam adjacent to the first side of the grating, wherein the first filler beam includes a spacing (x) from the first side of the grating, the spacing (x) varying within the following geometric conditions:(l−5 μm)≧x≧5 μm. 8. The grating as claimed in claim 1, wherein a width of the filler beams is not more than 3 μm in a direction of the gaps. 9. The grating as claimed in claim 1, wherein edge-side filler beams in a grating gap are elongated in a direction of the gaps if the spacing of the edge-side beams from the edge of the grating is smaller than a set value. 10. The grating as claimed in claim 1, wherein, during production of the grating, a primary structure is firstly generated by a lithography method and the grating is produced as a negative impression of the primary structure. 11. The grating as claimed in claim 10, wherein the lithography method is an X-ray lithography method. 12. The grating as claimed in claim 10, wherein the grating is produced by electrodeposition in the primary structure. 13. The grating as claimed in claim 10, wherein the primary structure is generated by a negative photoresist, in particular SU-8. 14. The grating as claimed in claim 1, wherein the grating is arranged on a grating base. 15. The grating as claimed in claim 1, wherein the grating is arranged on a foil. 16. A CT system, comprising at least one grating as claimed in claim 1. 17. The grating as claimed in claim 1, wherein the grating is an X-ray optical grating. 18. The grating as claimed in claim 1, wherein the grating is produced by a lithography method. 19. A CT system, comprising at least one grating as claimed in claim 18. |
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description | 1. Field of the Invention Embodiments of the invention relate generally to a method of forming a protection layer on a TEM specimen and a related method of preparing a specimen for transmission electron microscope (TEM) inspection. This application claims priority to Korean Patent Application No. 2004-72456 filed on Sep. 10, 2004, the subject matter of which is hereby incorporated by reference in its entirety. 2. Description of the Related Art In general, a semiconductor device is formed by a complex sequence of processes, including one or more patterning process(es). Conventional patterning processes are adapted to form, for example, a circuit pattern on a semiconductor substrate. The patterning process may include an etching process, a diffusion process, and a metallization process. Recent trends in the use of semiconductor devices are characterized by demands for multi-functionality and high speed performance. In order to meet these demands, semiconductor devices are ever more densely integrated. Increasing integration density demands tighter tolerances on circuit patterns. In order to meet these tighter design tolerances, a greater emphasis has recently been placed on inspection and analysis (e.g., structural and/or chemical) processes and equipment implementing these processes. The transmission electron microscope (TEM) is one such piece of equipment increasingly adapted to the inspection and analysis of circuit patterns formed on semiconductor substrates. Indeed, use of the TEM is wide spread due to its high resolution and general applicability to inspection and analysis processes. However, while the TEM provides very accurate information about an object under inspection, the inspection and analysis results provided by the TEM are heavily dependent on the quality of the exemplary specimen associated with the object. That is, TEM inspection of an entire object or structure is rarely possible under commercial fabrication circumstances, so an exemplary specimen must be prepared. Accordingly, special emphasis is laid upon a related method of preparing the specimen for TEM inspection. Conventional TEM specimens formed to inspect a predetermined layer on a semiconductor substrate have usually been prepared using an ion milling process or a focusing ion beam (FIB) process. In the conventional FIB process, an etching process is initiated from a neighboring point proximate the inspection point on the specimen and then moved towards the inspection point during a period in which the inspection point is being viewed as an electronic image. Using this technique, the thickness of the specimen may be easily controlled. Examples of conventional methods used to form TEM specimens include those disclosed in U.S. Pat. No. 6,194,720 to Li et al., and U.S. Pat. No. 6,080,991 to Tsai. Li et al. disclose a method of forming a TEM specimen including first and second electron transparent segments. The first electron transparent segment is formed using a FIB technique while the second electron transparent segment is formed using a wedge forming technique. Tsai discloses a method of forming the TEM specimen comprising a thin layer taken from a portion of the object or structure being inspected. According to yet another conventional method of preparing a TEM specimen; an inspection point—which contains a defect—is first identified (or selected) using an electronic image. Then, the semiconductor wafer is cut into a wafer slice of about 2 mm×3 mm using an ultrasonic wave cutter such that the inspection point is positioned in a middle of the specimen, thereby forming a first specimen. The first specimen is then ground to a thickness of about 40 μm, thereby forming a second specimen. The second specimen is adhered to a nickel grid, and a protection layer is coated over the inspection point using a deposition process, thereby forming a third specimen. Both side surfaces of the third specimen are then milled by sequentially varying the current applied to a FIB process, thereby forming the TEM specimen having a vertical side surface suitable for inspection. FIG. 1 is a cross sectional view illustrating a conventional method of forming a protection layer on a TEM specimen. More particularly, FIG. 1 illustrates one step in the foregoing conventional method adapted to the preparation of the first specimen. Referring to FIG. 1, a first layer of protection material is deposited on a stage 12 and a wafer 10 (e.g., a preliminary specimen cut from a semiconductor substrate) is placed on the first layer of the protection material. Then, a second layer of the protection material is deposited on wafer 10. Thus, the first and second applications of the protection material form a protection layer 14 on wafer 10. A covering member 16 is then positioned on protection layer 14. Thereafter, wafer 10 is cut into a wafer slice using an ultrasonic wave cutter. As shown in FIG. 1, although top and bottom surfaces of wafer 10 are covered by protection layer 14, the side surface(s) of wafer 10, including a beveled side surface, are barely, if at all, covered by protection layer 14. As a result, when wafer 10 is cut into a wafer slice by the ultrasonic wave cutter, the beveled side surface of wafer 10 may be broken off or otherwise removed. Accordingly, where an inspection point P is located on a beveled side surface of wafer 10, the ultrasonic wave may cause damage to inspection point P (i.e., the point of interest). When inspection point P is damaged by the ultrasonic wave, the specimen is no longer useful and a new specimen must be prepared, since TEM inspection of a broken or damaged specimen if of little value. This outcome is particularly harmful where the inspection point P is associated with a unique defect or point of interest. FIG. 2 is a cross sectional view illustrating one conventional method of grinding a specimen being prepared for TEM inspection. More particularly, FIG. 2 illustrates one step in the foregoing conventional method adapted to the preparation of the second specimen. Referring to FIG. 2, an inspection point P is again located on a beveled side surface of a first specimen 20. The beveled side surface is secured to a stage 22 using an adhesive and a covering member 26 is fixed to the vertical lateral side of the secured first specimen 20. During the grinding process, the specimen may be laterally visualized through covering member 26. Unfortunately, inspection point P cannot be viewed through covering member 26 during the grinding process, and the grinding process must be guided solely by an operator's intuition. Accordingly, the personal skill of an operator has a large effect on the grinding quality, and thus the reliability of the grinding process may be relatively low. In one embodiment, the invention provides a method of forming a protection layer on a wafer slice to form a specimen for a transmission electron microscope (TEM) inspection, the method comprising coating the wafer slice with a protection material on top, bottom and side surfaces of the wafer slice, wherein one of the side surfaces comprises a beveled side surface comprising an inspection point, and compressing the protection material to the wafer slice. In another embodiment, the invention provides a method of forming a specimen for a transmission electron microscope (TEM) inspection, the method comprising cutting a wafer slice from a wafer, the wafer slice comprising a beveled side surface comprising an inspection point, forming a protection layer on the wafer slice protecting the inspection point, forming a first preliminary specimen by cutting the wafer slice, wherein the first preliminary specimen comprises the inspection point, forming a second preliminary specimen by grinding top and bottom surfaces of the first preliminary specimen, wherein the second preliminary specimen comprises the inspection point, and forming a specimen by etching portions of top and bottom surfaces of the second preliminary specimen around the inspection point to respective predetermined depths. Spatially relative terms, such as “beneath”, “below”, “lower”, “above”, “upper” and the like, may be used herein for ease of description to describe one element or feature's relationship to another element(s) or feature(s) as illustrated in the figures. It will be understood that the spatially relative terms are intended to encompass different orientations of the device in use or operation in addition to the orientation depicted in the figures. For example, if the device in the figures is turned over, elements described as “below” or “beneath” other elements or features would then be oriented “above” the other elements or features. Thus, the exemplary term “below” can encompass both an orientation of above and below. The device may be otherwise oriented (rotated 90 degrees or at other orientations) and the spatially relative descriptors used herein interpreted accordingly. FIG. 3 is a flow chart illustrating an exemplary method of forming a protection layer on a TEM specimen. FIGS. 4 to 6 are presented to further illustrate the method of FIG. 3. Referring to FIGS. 3 to 6, a wafer slice 110 for a TEM inspection process is provided. Wafer slice 110 will be coated with a protection layer in a subsequent process. Particularly, a wafer W is prepared for the TEM inspection process in which a layer defect or point of interest is to be inspected. The layer defect may be formed in one or more of the layers formed at a beveled side surface of wafer W. This layer defect or point of interest on the beveled side surface of wafer W will be referred to as an inspection point P. The position of inspection point P on wafer W may be determined using an electron microscope. Once inspection point P has been located, wafer slice 110 is cut from wafer W using a diamond cutter. As shown in FIG. 5, wafer slice 110 comprises the beveled side surface surrounding inspection point P. In one embodiment, wafer slice 110 is cut to a size of about 2 cm×3 cm. Referring to the flow chart of FIG. 3 and the illustration of FIG. 6, a stage 120 is first heated using (e.g.) a hot plate or a heater to a temperature sufficient to reflow a selected protection material (S40). In one embodiment, an epoxy resin compound is used as the protection material and the stage is heated for about 5 minutes to about 10 minutes. Then, the top surface of stage 120 is coated with the protection material using one or more conventional processes, such as a spin coating. Wafer slice 110 is then positioned on stage 120 once it has been coated with the protection material. For example, in the illustrated example, wafer slice 110 is positioned on stage 120 lengthwise and in parallel with stage 120. Thereafter, the top surface and side surfaces of wafer slice 110, including the beveled side surface, are coated with the protection material (S50). In the illustrated embodiment, the protection material is applied such that it covers inspection point P on the beveled side surface of wafer slice 110. In a related aspect of this embodiment, the fact that wafer slice 110 is coated with the protection material only after wafer slice 110 including the beveled side surface is sufficiently heated to reflow the protection material, allows the entire surface of wafer slice 110 to be evenly coated with the protective material. Continuing with the illustrated example, a bottom surface of wafer slice 110 is coated with the protection material when the wafer slice 110 is positioned on stage 120. Whereas, the top and side surfaces of wafer slice 110 are coated with the protection material after wafer slice 110 is positioned on stage 120. Once wafer slice 110 is coated with the protection material, a covering member 140 is placed over it, such that wafer slice 110 is mechanically compressed (S60). In the illustrated example, the combination of the upper surface of stage 120 and covering member 140 form an enclosure of sorts for wafer slice 110 and its coating of protection material. Covering member 140 may take any reasonable form, but will generally be shaped in accordance with the shape of wafer slice 110, such that the top and side surfaces of wafer slice 110 are enclosed within covering member 140. In one embodiment, covering member 140 takes the form of a hollow hexahedron having an opening in its bottom portion. A side portion of this exemplary covering member 140 has a curvature corresponding to that of the beveled side surface of wafer slice 110. Covering member 140 will typically be designed to fit snuggly around wafer slice 110. In one embodiment, covering member 140 is formed from glass having a thickness of about 0.1 cm. As stage 120 is cooled to around room temperature (S70), the protection material is hardened around wafer slice 110 to form a protection layer 130 covering at least inspection point P. Protection layer 130 covers the beveled side surface of wafer slice 110, such that inspection point P is not damaged during subsequent preparation steps, such as a step in which wafer slice 110 is cut using an ultrasonic wave cutter. In the foregoing exemplary embodiment, the protection layer is entirely formed on wafer slice 110 after the wafer slice is cut from wafer W. However, the protection layer may be, at least partially, formed on the wafer before wafer slice 110 is cut. FIG. 7 is a flow chart illustrating an exemplary method of forming a TEM specimen. FIG. 8 is a flow chart illustrating an exemplary method of forming a protection layer on the TEM specimen. FIG. 9 is a flow chart illustrating an exemplary method of forming a second preliminary specimen by grinding a first preliminary specimen. FIGS. 10 to 16 are presented to further illustrate the methods of FIGS. 7-9. Referring to FIGS. 7 and 10, a wafer slice 310 for a TEM inspection process is provided. Wafer slice 310 will be coated with a protection layer in a subsequent process. Particularly, a wafer W is prepared for the TEM inspection process in which a layer defect or point of interest is to be inspected. The layer defect may be formed in one or more of the layers formed at a beveled side surface of wafer W. This layer defect or point of interest on the beveled side surface of wafer W will be referred to as an inspection point P. The position of inspection point P on wafer W may be determined using an electron microscope. Once inspection point P has been located, wafer slice 310 is cut from wafer W using a diamond cutter. As shown in FIG. 11, wafer slice 310 comprises the beveled side surface surrounding inspection point P. In one embodiment, wafer slice 310 is cut to a size of about 2 cm×3 cm. Referring again to FIGS. 7 and 12, a protection layer 330 is formed on wafer slice 310 (S210). An exemplary method of forming protection layer 330 will be described in detail hereinafter with reference to FIGS. 8 and 12. Referring to the flow chart of FIG. 8 and the illustration of FIG. 12, a stage 320 is first heated using (e.g.) a hot plate or a heater to a temperature sufficient to reflow a selected protection material (S211). In one embodiment, an epoxy resin compound is used as the protection material and the stage is heated for about 5 minutes to about 10 minutes. Then, the top surface of stage 320 is coated with the protection material using one or more conventional processes, such as a spin coating. Wafer slice 310 is then positioned on stage 320 once it has been coated with the protection material. For example, in the illustrated example, wafer slice 310 is positioned on stage 320 lengthwise and parallel with stage 320. Thereafter, the top surface and side surfaces of wafer slice 310, including the beveled side surface, are coated with the protection material (S212). In the illustrated embodiment, the protection material is applied such that it covers inspection point P on the beveled side surface of wafer slice 310. In a related aspect of this embodiment, the fact that wafer slice 310 is coated with the protection material only after wafer slice 310 including the beveled side surface is sufficiently heated to reflow the protection material, allows the entire surface of wafer slice 310 to be evenly coated with the protective material. Continuing with the illustrated example, a bottom surface of wafer slice 310 is coated with the protection material when wafer slice 310 is positioned on stage 320. Whereas, the top and side surfaces of wafer slice 310 are coated with the protection material after wafer slice 310 is positioned on stage 320. Referring to FIG. 12, once wafer slice 310 is coated with the protection material, a first covering member 340 is placed over it, such that wafer slice 310 is mechanically compressed (S213). In the illustrated example, the combination of the upper surface of stage 320 and first covering member 340 form an enclosure of sorts for wafer slice 310 and its coating of protection material. First covering member 340 may take any reasonable form, but will generally be shaped in accordance with the shape of wafer slice 310, such that the top and side surfaces of wafer slice 310 are enclosed within first covering member 340. In one embodiment, first covering member 340 takes the form of a hollow hexahedron having an opening at its bottom portion. A side portion of this exemplary first covering member 340 has a curvature corresponding to that of the beveled side surface of wafer slice 310. Covering member 340 will typically be designed to fit snuggly around wafer slice 310. In one embodiment, first covering member 340 is formed from glass having a thickness of about 0.1 cm. As stage 320 is cooled to around room temperature (S214), the protection material is hardened around wafer slice 310 to form a protection layer 330 covering at least inspection point P. Referring to FIGS. 7 and 13, a first preliminary specimen 314 is cut from wafer slice 310 by an ultrasonic wave cutter (S220), and inspection point P is included in first preliminary specimen 314. In one embodiment, first preliminary specimen 314 is cut to a size of about 2 mm×3 mm. Protection layer 330 covers the beveled side surface of wafer slice 310, such that inspection point P is not damaged and first preliminary specimen 314 is not broken during subsequent preparation steps, such as a step in which first preliminary specimen 314 is cut from wafer slice 310 by an ultrasonic wave cutter. In the foregoing exemplary embodiment, the protection layer is entirely formed on wafer slice 310 after the wafer slice is cut from wafer W. However, the protection layer may be, at least partially, formed on the wafer before wafer slice 310 is cut. Referring to FIGS. 7 and 15, both a top surface and a bottom surface of first preliminary specimen 314 are ground to form a second preliminary specimen 316 with a predetermined thickness w (S230). The grinding process with respect to first preliminary specimen 314 will be described in detail hereinafter with reference to FIGS. 9 and 14. Referring to FIGS. 9 and 14, first preliminary specimen 314, having a size of about 2 mm×3 mm, comprises the beveled side surface surrounding inspection point P. First preliminary specimen 314 is secured to stage 320 by an adhesive 350 in such an arrangement that inspection point P faces in a direction parallel with the surface of stage 320 (S231). A second covering member 360 is formed at a side portion of first preliminary specimen 314 and is adapted to support first preliminary specimen 314 (S232). In one embodiment, second covering member 360 is positioned perpendicularly with respect to stage 320, and is secured to stage 320 by adhesive 350. Second covering member 360 is also secured to first preliminary specimen 314 by adhesive 350. Second covering member 360 comprises a transparent material so that inspection point P may be viewed through an electron microscope while first preliminary specimen 314 is ground. Second covering member 360 may comprise, for example, transparent glass. Next, top and bottom surfaces of first preliminary specimen 314 are ground simultaneously (S233), which reduces the width of first preliminary specimen 314. Since the top and bottom surfaces are symmetrical with respect to inspection point P, the width of first preliminary specimen 314 is reduced when the top and bottom surfaces of first preliminary specimen 314 are ground. In one embodiment, first preliminary specimen 314 is reduced to a width within a range of about 40 μm to about 60 μm. Second preliminary specimen 316, shown in FIG. 15, is formed when first preliminary specimen 314 is ground while inspection point P faces in a direction parallel with the surface of stage 320. As a result, inspection point P of the beveled side surface of first preliminary specimen 314 may be viewed through an electron microscope while performing the grinding process on first preliminary specimen 314. FIG. 15 shows a second preliminary specimen 316 created by performing a grinding process on first preliminary specimen 314. Referring to FIGS. 7 and 16, a portion of second preliminary specimen 316 around inspection point P is removed from the top and bottom surfaces thereof to respective predetermined depths by an FIB process, thereby forming a TEM specimen 318 (S240). As shown in FIG. 16, TEM specimen 318 has recessed portions that are symmetrical with respect to inspection point P created by the FIB etching process. A reduced width w1 of TEM specimen 318 around inspection point P is much smaller than width w of an end portion of TEM specimen 318. In one embodiment TEM specimen 318 has the reduced width w1 of about 40 nm around inspection point P, so that electrons ejected from the TEM may sufficiently penetrate the TEM specimen around inspection point P. The FIB etching process may be performed by sequentially varying a current in a range from about 2700 picoamperes (pa) to about 70 pa in order to reduce the width of TEM specimen 318 around inspection point P from width w to reduced width w1. First in the FIB etching process, a coarse etching is performed at a portion of second preliminary specimen 316 neighboring inspection point P using a high current FIB process. Then, as the depth of the FIB etching portion increases, and consequently the surface of the etching portion approaches inspection point P, the current used in the FIB etching process is reduced, and as a result the FIB etching process is finer as the FIB etching portion approaches inspection point P. Accordingly, the width of second preliminary specimen 316 around inspection point P is quickly and accurately reduced to reduced width w1, thereby forming TEM specimen 318. Also, an additional protection layer (not shown) may be formed on inspection point P of second preliminary specimen 316 prior to the FIB process to protect the TEM specimen from being damaged during either the FIB process for viewing inspection point P of TEM specimen 318 or a micro-etching process. The additional protection layer may include a tungsten (W) layer, a platinum (Pt) layer, a carbon (C) layer, an aluminum (Al) layer, etc. These can be used alone or in combinations thereof. Thereafter, TEM specimen 318 is adhered to a grid, thereby completing the formation of the TEM specimen. In accordance with embodiments of the present invention, a protection layer is formed on a beveled side section of a specimen on which an inspection point is formed, so that the specimen is not broken and the inspection point not damaged when the specimen is cut by an ultrasonic wave cutter. Accordingly, less time is lost during the formation of the specimen. Furthermore, the inspection point is viewed through an electron microscope during a specimen grinding process, thereby improving the specimen quality. Although exemplary embodiments of the present invention have been described, the present invention should not be limited to these exemplary embodiments, but rather, one of ordinary skill in the art will understand that various changes and modifications can be made while remaining within the scope of the present invention as claimed hereinafter. |
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claims | 1. A charged particle beam application apparatus having a stage device used to move a sample in at least three axial directions, a charged particle beam optical system having an optical axis inclined from a surface of said sample to irradiate the sample with a charged particle beam, and a display device that displays an image formed by the charged particle beam optical system, the apparatus comprising:a correction table indicating a relationship between focal distance and a position of said sample and a relationship between optical conditions for said charged particle beam optical system and a position of said sample, and an arithmetic section that calculates the position of said sample, the arithmetic section calculating the amount of three-dimensional correction for the position of said sample using said correction table, so that when the focal distance of said charged particle beam changes, a position of a target on said sample is placed in the center of a visual field of a screen of said display device, wherein said correction table is created by assuming that a position [Z] of said sample in a Z direction is expressed using a linear expression with a position [X] of said sample in an X direction, a position [Y] of said sample in a Y direction, and optical conditions [P] for the charged beam optical system. 2. The charged particle beam application apparatus according to claim 1, wherein said charged particle beam optical system includes a plurality of charged particle beam optical systems, and even when said charged particle beam optical system is switched, the position of the target on said sample remains in the center of the visual field of the screen of said display device. 3. The charged particle beam application apparatus according to claim 2, wherein if the same target on said sample is observed without changing the height of said sample even when said charged particle beam optical system is switched, then the focal distance of said charged particle beam optical system is changed and said sample is moved in the X and Y directions by an amount equal to the following differential (dX, dY),dX=SX(1)−FX(1)dY=SY(1)−FY(1)where FX(1) and FY(1) denote the position of the sample in the charged particle beam optical system before the switching and SX(1) and SY(1) denote the position of the sample in the charged particle beam optical system after the switching. |
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claims | 1. A method for making a radiopharmaceutical pig comprising:forming a shielding element of nuclear shielding material to have an open top, a closed bottom, a wall extending between the top and bottom, an inner surface, an outer surface, and an interior cavity;placing the shielding element into an injection molding machine;injecting a polymer material into the molding machine so that an entirety of the shielding element including, but not limited to, the inner surface and outer surface, thereof, is completely covered by the polymer material; andallowing the polymer material to harden. 2. The method set forth in claim 1 wherein the polymer material comprises polycarbonate resin. 3. The method set forth in claim 1 wherein forming a shielding element of nuclear shielding material comprises pouring molten nuclear shielding material into a mold and allowing it to solidify. 4. The method set forth in claim 3 wherein pouring molten nuclear shielding material into a mold comprises pouring molten lead into the mold. 5. The method set forth in claim 1 wherein forming a shielding element of nuclear shielding material comprises forming a base shielding element. 6. The method set forth in claim 1 wherein forming a shielding element of nuclear shielding material comprises forming a cap shielding element. 7. A method for making a radiopharmaceutical pig comprising:forming a base shielding element of nuclear shielding material to have an open top, a closed bottom, a wall extending between the top and bottom, an inner surface, an outer surface, and an interior cavity;placing the base shielding element into an injection molding machine;injecting a polymer material into the molding machine so that an entirety of the base shielding element including, but not limited to, the inner surface and the outer surface thereof is completely covered by the polymer material;allowing the polymer material covering the entirety of the base shielding element to harden;forming a cap shielding element of nuclear shielding material to have an open top, a closed bottom, a wall extending between the top and bottom, an inner surface, an outer surface, and an interior cavity;placing the cap shielding element into an injection molding machine;injecting a polymer material into the molding machine so that an entirety of the cap shielding element including, but not limited to, the inner surface and outer surface thereof is completely covered by the polymer material;allowing the polymer material covering an entirety of the cap shielding element to harden; andcoupling the cap shielding element to the base shielding element. 8. The method set forth in claim 7 wherein the polymer material that covers an entirety of the cap shielding element comprises polycarbonate resin. 9. The method set forth in claim 8 wherein the polymer material that covers an entirety of the base shielding element comprises polycarbonate resin. 10. The method set forth in claim 9 wherein the nuclear shielding material of the base shielding element comprises lead. 11. The method set forth in claim 10 wherein the nuclear shielding material of the cap shielding element comprises lead. 12. The method set forth in claim 7 wherein forming a base shielding element of a nuclear shielding material comprises pouring molten nuclear shielding material into a mold and allowing it to solidify. 13. The method set forth in claim 12 wherein pouring molten nuclear shielding material into a mold comprises pouring molten lead into the mold. 14. The method set forth in claim 7 wherein forming a cap shielding element of nuclear shielding material comprises pouring molten material into a mold and allowing it to solidify. 15. The method set forth in claim 14 wherein pouring molten material into a mold comprises pouring molten lead into the mold. |
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description | Not Applicable. Not Applicable. The present invention relates generally to diagnosis of problems with measurements from a measurement device. More specifically, the invention relates to a method for the diagnosis of measurements from measurement devices that have a relatively high data input rate and a relatively low data output rate. Quite often, a physical property, characteristic, or phenomenon requires measurement. Various meters and measurement devices have been developed in a wide variety of industries to measure a characteristic-of-interest. For example, a person may wish to measure characteristics of the atmosphere, a fluid flow, or a moving object. Measurement devices to monitor fluid flow include ultrasonic meters, coreolis meters, magnetic flow meters, turbine meters, and orifice plates. Measurement devices are not perfect, however. They are known to make errors in measurements, there being many reasons why measurement devices may not measure a characteristic-of-interest accurately. Furthermore, diagnosing problems with measurement devices in the field can be a difficult and troublesome experience. This is particularly true if a large amount of data (i.e. a high input data rate) is being processed into a small amount of data (i.e. a low output data rate but with a high value content). The problem is exacerbated when input data varies considerably from one field location to another, resulting in no single or small data set being representative for typical conditions in the field. An ultrasonic meter provides a good example of a measurement device with a high input data rate and a low output data rate, where measurement data varies from one location to another. An ultrasonic meter, such as disclosed in U.S. Pat. No. 4,646,575, hereby incorporated by reference for all purposes, can have a 100-fold or more reduction in quantity between input data and output data. For example, an ultrasonic meter typically has a spoolpiece through which there is a fluid flow. Along the perimeter of the spoolpiece are one or more sets of transducers that act as transceivers, each transducer generating an ultrasonic signal and then receiving an ultrasonic signal from the respective transducer in the transducer pair. This may happen thousands of times per minute. Transit times are thus measured along each chord (i.e. ultrasonic wave path) for the upstream and downstream ultrasonic signals. The difference in travel times between upstream and downstream ultrasonic signals indicates the velocity of the fluid flow within the pipeline. The ultrasonic meter also includes electronics that sample and record pertinent ultrasonic signal information. Each ultrasonic signal generated by a transducer (either upstream or downstream) is identified by numerous pieces of information when sampled and recorded. These include a wavenumber, path identifier (Aup, Adwn, Bup, Bdwn, etc.), gain (AGC), hold number (delay from generation of the ultrasonic signal until the time at which the data is recorded, indicating a window during which the ultrasonic signal is expected to arrive), and a value for each of, e.g., 256 samples in the received waveform. The sample rate must also be known. To determine transit time accurately, a batch of e.g., 20 ultrasonic signals along the same chord and same direction are taken and then processed to provide velocity and speed of sound for each chord. Thus, in this example, over ten thousand pieces of information for each chord are transformed into two: velocity and speed of sound for each chord. This information may then be averaged to compute average velocity and speed of sound for the fluid passing through the spoolpiece. This is an example of a high data input and low data output. Unfortunately, measurement errors occur. In the case of an ultrasonic meter, one or more of the measurements may deviate so significantly from a benchmark that it indicates a problem either with the fluid flow or with the meter itself. Typically, it will be difficult to diagnose the problem based on only the velocity and speed of sound measurements. Similarly, it may be difficult to determine based only on velocity and speed of sound measurements whether there exists a measurement error or whether an unusual or notable event is occurring in the medium being measured by the measurement device. A technician may be dispatched to the site of the measurement device to analyze the apparent problem based on the greater amount of data available at the meter location. One approach to investigating the apparent problem includes recording the raw data, partially calculated values, or final answers at predetermined moments in time. This may be referred to as inserting “measurement points” into the data sampling. Examination of the recorded data is then made and an attempt made to identify the problem. One problem with such an approach, however, is its failure to collect data in between the temporal measurement points. As a result, a substantial amount of data is not collected. If a cause or condition of a meter problem takes place while data is not being recorded, detection of the cause or condition can be missed. Alternately, an in-circuit emulator at the meter location may be used to try and identify the apparent problem. An in-circuit emulator is a device separate from the measurement device or a programmed feature in the measurement device electronics that waits for a trigger condition (such as an unusually high maximum transit time). Upon occurrence of the trigger condition, the in-circuit emulator triggers a secondary effect—it either records data corresponding to that instant in time or stops program execution. This recorded data or program memory is then examined in an effort to identify the problem. A problem with this approach, however, is that it does not collect data prior to the trigger event. This is a problem when the trigger event is only a culmination of a trend or ongoing problem in existence before the trigger event. In such a case, it may be difficult to identify the problem by use of an in-circuit simulator. A drawback with both of these approaches is that they rely greatly on the knowledge and training of the technical persons who are sent to investigate the problem at the meter location. There is little opportunity to carefully diagnose the problem elsewhere because much of the analysis and effort to rectify a problem takes place at the meter location. Both of these approaches record only limited data, either at the time of an event or at spaced intervals. This is a problem because a more complete set of data is important when attempting to identify a problem with the measurement device. A problem with the measurement of fluid flow may correlate to any one of a number of different issues. A peak selection error may be due to noise, turbulence in the fluid flow, or incorrect default values in the meter's software. However, because of the high amount of data that is processed or refined down in a high input-low output processing scheme, crucial information may go unreported or underreported, leading to an inability to adequately describe fleeting anomalous conditions. If these fleeting conditions are not captured by the method used to identify a problem with the measurement device, they go undetected. One embodiment of the invention is a method for analyzing data measurement, including programming the measurement device with a predetermined trigger value corresponding to a characteristic of interest, recording a first set of data by the measurement device that occurs prior to an occurrence for the trigger value, recording a second set of data by the measurement device that occurs at the time of the occurrence for the trigger value, and playing the first and second sets of data on a simulator at a time after the occurrence of the predetermined trigger value. The data includes temporal information and the method may also include recording and playing a third set of data collected after occurrence of the third trigger value in the characteristic of interest. A second embodiment of the invention is a method for analyzing measurement device performance, including recording an uninterrupted stream of data from a measurement device to a first location, the duration of the uninterrupted stream of data being of a substantially longer duration than fluctuations in the characteristic of interest, retrieving the stream of data from the first location, and playing the uninterrupted stream of the data on a replay device. A third embodiment of the invention is a measurement device diagnostic system including a measurement device for taking measurements of a characteristic of interest, a processor not located in the measurement device, means to record data corresponding to the measurements to a memory device, and means to transmit the data from the memory device to the processor. The measurement device may be an ultrasonic meter. A fourth embodiment of the invention is a method for analyzing measurement data from a measurement device including producing measurement data at a data acquisition rate, producing output data from the measurement data at an output data rate (where the output data rate is lower than the measurement data acquisition rate), reordering the measurement data, transmitting temporal data corresponding to the measurement data along with the measurement device to a location outside the measurement device, and playing the measurement data on a reply device. Thus, the present invention comprises a combination of features and advantages which enable it to overcome various problems of prior devices. The various characteristics described above, as sell as other features, will be readily apparent to those skilled in the art upon reading the following detailed description of the preferred embodiments of the invention, and by referring to the accompanying drawings. The novel solution of the invention is a system and method to take a “snapshot” of meter performance. This includes: 1) recording a continuous stream of raw data into a format suitable for transport to a site better suited for debug and diagnosis; and 2) replaying the data on a replay device such that time, as seen by the system, may be “frozen” at will and any instant of the stream examined. FIG. 1A shows one type of ultrasonic meter suitable for measuring gas flow. Spoolpiece 100, suitable for placement between sections of gas pipeline, has a predetermined size and thus defines a measurement section. Alternately, a meter may be designed to attach to a pipeline section by, for example, hot tapping. As used herein, the term “pipeline” when used in reference to an ultrasonic meter may be referring also to the spoolpiece or other appropriate housing across which ultrasonic signals are being sent. A pair of transducers 120 and 130, and their respective housings 125 and 135, are located along the length of spoolpiece 100. A path 110, sometimes referred to as a “chord” exists between transducers 120 and 130 at an angle θ to a centerline 105. The position of transducers 120 and 130 may be defined by this angle, or may be defined by a first length L measured between transducers 120 and 130, a second length X corresponding to the axial distance between points 140 and 145, and a third length D corresponding to the pipe diameter. Distances D, X and L are precisely determined during meter fabrication. Points 140 and 145 define the locations where acoustic signals generated by transducers 120 and 130 enter and leave gas flowing through the spoolpiece 100 (i.e. the entrance to the spoolpiece bore). In most instances, meter transducers such as 120 and 130 are placed a specific distance from points 140 and 145, respectively, regardless of meter size (i.e. spoolpiece size). A fluid, typically natural gas, flows in a direction 150 with a velocity profile 152. Velocity vectors 153–158 indicate that the gas velocity through spool piece 100 increases as centerline 105 of spoolpiece 100 is approached. Transducers 120 and 130 are ultrasonic transceivers, meaning that they both generate and receive ultrasonic signals. “Ultrasonic” in this context refers to frequencies above about 20 kilohertz as required by the application. Typically, these signals are generated and received by a piezoelectric element in each transducer. To generate an ultrasonic signal, the piezoelectric element is stimulated electrically, and it responds by vibrating. This vibration of the piezoelectric element generates an ultrasonic signal that travels across the spoolpiece to a corresponding transducer of the transducer pair. Similarly, upon being struck by an ultrasonic signal, the receiving piezoelectric element vibrates and generates an electrical signal that is detected, digitized, and analyzed by electronics associated with the meter. It is these electronics (and software) that process the sampled data to yield output data. Initially, D (“downstream”) transducer 120 generates an ultrasonic signal that is then received at, and detected by, U (“upstream”) transducer 130. Some time later, U transducer 130 generates a return ultrasonic signal that is subsequently received at and detected by D transducer 120. Thus, U and D transducers 130 and 120 play “pitch and catch” with ultrasonic signals 115 along chordal path 110. During operation, this sequence may occur thousands of times per minute. The transit time of the ultrasonic wave 115 between transducers U 130 and D 120 depends in part upon whether the ultrasonic signal 115 is traveling upstream or downstream with respect to the flowing gas. The transit time for an ultrasonic signal traveling downstream (i.e. in the same direction as the flow) is less than its transit time when traveling upstream (i.e. against the flow). In particular, the transit time t1, of an ultrasonic signal traveling against the fluid flow and the transit time t2 of an ultrasonic signal travelling with the fluid flow is generally accepted as being defined as: t 1 = L c - V x L ( 1 ) t 2 = L c + V x L ( 2 ) where, c=speed of sound in the fluid flow; V=average axial velocity of the fluid flow over the chordal path in the axial direction; L=acoustic path length; x=axial component of L within the meter bore; t1=transmit time of the ultrasonic signal against the fluid flow; and t2=transit time of the ultrasonic signal with the fluid flow. The upstream and downstream transit times are typically calculated separately on an average of a batch of measurements, such as 20. These upstream and downstream transit times may then be used to calculate the average velocity along the signal path by the equation: V = L 2 2 x t 1 - t 2 t 1 t 2 ( 3 ) with the variables being defined as above. The upstream and downstream travel times may also be used to calculate the speed of sound in the fluid flow according to the equation: c = L 2 t 1 + t 2 t 1 t 2 ( 4 ) To a close approximation, equation (3) may be restated as: V = c 2 Δ t 2 x ( 5 ) where,Δt=t1−t2 (6)So to a close approximation at low velocities, the velocity v is directly proportional to Δt. Given the cross-section measurements of the meter carrying the gas, the average velocity over the area of the meter bore may be used to find the volume of gas flowing through the meter or pipeline 100. In addition, ultrasonic gas flow meters can have one or more paths. Single-path meters typically include a pair of transducers that projects ultrasonic waves over a single path across the axis (i.e. center) of spoolpiece 100. In addition to the advantages provided by single-path ultrasonic meters, ultrasonic meters having more than one path have other advantages. These advantages make multi-path ultrasonic meters desirable for custody transfer applications where accuracy, repeatability, and reliability are crucial. Referring now to FIG. 1B, a multi-path ultrasonic meter is shown. Spool piece 100 includes four chordal paths A, B, C, and D at varying levels through the gas flow. Each chordal path A–D corresponds to two transceivers behaving alternately as a transmitter and receiver. Also shown is an electronics module 160, which acquires and processes the data from the four chordal paths A–D. This arrangement is described in U.S. Pat. No. 4,646,575, the teachings of which are hereby incorporated by reference. Hidden from view in FIG. 1B are the four pairs of transducers that correspond to chordal paths A–D. The precise arrangement of the four pairs of transducers may be more easily understood by reference to FIG. 1C. Four pairs of transducer ports are mounted on spool piece 100. Each of these pairs of transducer ports corresponds to a single chordal path of FIG. 1B. A first pair of transducer ports 125 and 135 includes transducers 120 and 130 recessed slightly from the spool piece 100. The transducers are mounted at a non-perpendicular angle θ to centerline 105 of spool piece 100. Another pair of transducer ports 165 and 175 including associated transducers is mounted so that its chordal path loosely forms an “X” with respect to the chordal path of transducer ports 125 and 135. Similarly, transducer ports 185 and 195 are placed parallel to transducer ports 165 and 175 but at a different “level” (i.e. a different radial position in the pipe or meter spoolpiece). Not explicitly shown in FIG. 1C is a fourth pair of transducers and transducer ports. Taking FIGS. 1B and 1C together, the pairs of transducers are arranged such that the upper two pairs of transducers corresponding to chords A and B form an X and the lower two pairs of transducers corresponding to chords C and D also form an X. Referring now to FIG. 1B, the flow velocity of the gas may be determined at each chord A–D to obtain chordal flow velocities. To obtain an average flow velocity over the entire pipe, the chordal flow velocities are multiplied by a set of predetermined constants. Such constants are well known and were determined theoretically. Thus, transit time ultrasonic flow meters measure the times it takes ultrasonic signals to travel in upstream and downstream directions between two transducers. This information, along with elements of the geometry of the meter, allows the calculation of both the average fluid velocity and the speed of sound of the fluid for that path. In multipath meters the results of each path are combined to give an average velocity and an average speed of sound for the fluid in the meter. The average velocity is multiplied by the cross sectional area of the meter to calculate the actual volume flow rate. Referring to FIG. 2, a method according to the invention is shown. A measurement device measures raw data corresponding to a characteristic of interest at step 200. It is to be understood that the term “raw data” includes data prior to final processing that undergoes additional processing. “Raw data” refers most broadly to any data prior to final processing. It is to be understood that a high data input-low data output measurement device includes a routine to process the raw data into output data. Such a routine includes a number of steps, and may be programmed in the device as a single processing software chain. This routine would normally be stored in firmware in the ROM or PROM in electronics associated with the measurement device. At step 210, the measurement device exports the raw data. Exporting the data means to send it from one program or computer system to another program or computer system. In this case, raw data should be exported, as it is sampled, from an embedded (computer) system to a portable computer so that the data may be stored (such as on a hard drive) for further analysis. To export input data that corresponds to each waveform, a feature or routine in the original signal processing software chain should be included to export the input data as it is arriving at the start of the chain. The data may be exported at all times into a circular buffer (either on the hard disk of an external computer or within the measurement device itself). Any number of trigger conditions may be programmed or pre-programmed into the software that would cause the recording device to “save off” the older data and continue to record new data. For example, it may be done manually or automatically. Either way, the data recorded earlier in time than the detected trigger condition would not be written over. Data available for later analysis would include an uninterrupted stream including data recorded before the trigger condition, the data recorded at the time of the trigger condition, and the data recorded after the trigger condition. Also included with the exported data may be any other information such as gain or other physical parameters that were used to collect the data Temporal data indicating when each sample of the ultrasonic waveform was taken should also be included. Adequate temporal information could constitute timestamps corresponding to the moment at which each group of data was received or collected. The inclusion of temporal data with the exported data is a particularly notable aspect of the invention. Identification of meter or measurement device behavior is most effective when a detailed recording of the measurement data has been made so that even small perturbations can be detected and identified. To achieve the desired level of detail, it is important that the relative timing for all the data is precise. Temporal data such as timestamps provides this precision among the many ultrasonic signals that are sampled and recorded. Temporal data combined with the input data results in a stream of data that is in a form suitable to save to a hard drive or other appropriate device. The stream of data may alternately be transmitted to a remote location, either along a cable or wirelessly. If desired, prior to save or transmission the data may be compressed by well known data-compression techniques but the data should be fully reproducible at the site of diagnosis and debug. Enough information should be transmitted to allow the characteristic-of-interest to be reproduced deterministically. This allows a full analysis of the data to be run. At a minimum, a large percentage of the data should be retrievable, so that to the extent possible the problem may be identified. A second aspect to the invention is replaying the exported data at a remote location or later in time, occurring at step 220. A replay device according to the invention allows recorded data to be fed to signal processing software in the same manner as was being done by signal processing software at the field location. The replay device may then replay (i.e., process) raw data in the same manner as it was originally received by the signal processing chain at the field location. For example, the data measured by the measurement device (e.g. an ultrasonic meter) may be played on a simulator that yields a set of output data that substantially reproduces the characteristic of interest recorded at the field location. The set of output data from the simulator is substantially the same as that from the measurement device for the characteristic of interest. The term “substantially” in this context means that the output data is the same with respect to the characteristic or characteristics of interest that is or are being measured to such an extent the characteristic of interest can be reproduced. Ultimately, this means that the data must be similar enough to detect the same problem or interesting phenomenon that caused the measurement data to be of interest initially. A laptop or portable computer may be used as a replay device. In the case of a laptop computer, client software installed on the laptop would receive the digitized data over a communication link. The laptop computer would save this data to, e.g., a file on the hard drive. Software installed on the laptop would operate as a software simulator, capable of reading the saved file and playing it back on demand at a later date or a different place. The playback parameters would depend on any number of desired and important criteria that would be specific to the system. The most important feature of the playback system would be to present the data intended for play back to the signal processing chain in such a manner that the signal processing chain in the simulator would behave in a substantially identical manner to the signal processing chain in the field (where the data was originally exported). Additionally, the data would be played back in a manner (utilizing the stored parameters that were stored along with the raw data) so as to reproduce the detected conditions and characteristic(s)-of-interest, and hence the issue, that was desired to be “snapshot” at the field site. The replay portion of the invention allows time, as seen by the system, to be frozen at will and any part of the raw and any other pertinent data examined. The input data may then be replayed back one sample at a time or at any desired rate to enable adequate debugging of the original problem. Because the data is already recorded, and the corresponding output has already been calculated, new (including modified) software for a data processing chain can be tested. Calculations are executed on the data and by the results of those calculations are compare to known correct results to determine if there is any error in the new software. This will also help detect viruses that corrupt software. The snapshot-and-playback feature will be immensely useful to detect anomalous conditions that cause a measurement error. Additionally, this feature can be used extensively to further the capabilities of the software processing chain to handle new flow conditions. These tests can be made in the laboratory. For example, signal processing software implements a set of algorithms. Each algorithm has been derived/determined to work on a certain set of data. There are inherent assumptions to the data that reflects a set of flow conditions. These assumptions are or were based on the original set(s) of data that were examined to determine the algorithms originally (when the product was designed). Any flow conditions that adhere to these assumptions will be measured properly. Any physical changes in the flow conditions that invalidate the original assumptions will lead to an erroneous measurement in the field. A new set of data (corresponding to these new flow conditions) can be snapshot and then examined in the lab at a later date to see how the signal processing software may be modified to suitable process this new flow condition(s). An example is noise introduced by flow valves. Current software for an ultrasonic meter of the assignee may assume that the signal-to-noise ratio in a measured signal is below a certain value (such as 30 (signal strength) to 1 (noise level)). A field condition may be encountered where noise is particularly bad and there is only a 10 to 1 signal-to-noise ratio. The raw data could be snap shot (exported and recorded), brought back to a lab, analyzed (e.g. using FFT (Fast Fourier Transforms) spectrum analysis) and a new digital filter developed that reduces noise levels. The flow could then be accurately measured even in such adverse conditions. This way new flow conditions can be handled by utilizing the snapshot and playback feature of the system. An additional benefit of this feature is that since the data is being “played back” from a deterministic source, it is extremely repeatable. Thus the data can be used for “what if” scenarios where software processing chain changes are compared with the same data set run through both processors and the results compared. This also makes the replay aspect of the invention a good training tool. Another advantage of the invention is an ability to control the snapshot feature remotely; there is no need to dispatch a technician to the meter location. Another aspect of the invention is that the snapshot feature may be triggered either manually (such as by a switch on the measurement device or a switch remote from the measurement device) or automatically (such as when an event occurs that exceeds a predetermined threshold). While preferred embodiments of this invention have been shown and described, modifications thereof can be made by one skilled in the art without departing from the spirit or teaching of this invention. The embodiments described herein are exemplary only and are not limiting. Many variations and modifications of the system and apparatus are possible and are within the scope of the invention. Accordingly, the scope of protection is not limited to the embodiments described herein, but is only limited by the claims which follow, the scope of which shall include all equivalents of the subject matter of the claims. |
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description | The present application is a U.S. national stage application under 35 U.S.C. §371 of International Patent Application No. PCT/US2014/019042 filed Feb. 27, 2014, which claims the benefit of U.S. Provisional Patent Application Ser. No. 61/770,213 filed Feb. 27, 2013, which are incorporated herein by reference in their entireties. The present invention relates to nuclear reactor vessels, and more particularly to a nuclear reactor shroud surrounding the fuel core. Many nuclear reactor designs are of circulatory type wherein the water heated in the reactor fuel core region must be separated from the cooler water outside of it. Such a nuclear reactor may be typically equipped with a cylindrical shroud around the fuel core. The shroud serves to separate the internal space in the reactor vessel between an “up-flow” (e.g. riser) region in which primal) coolant heated by the core flows inside the shroud and the “downcomer” region in which colder primary coolant returned to the reactor vessel from the Rankine cycle steam generating system flows outside the shroud. It is desirable to minimize heat transfer from the heated hot reactor water inside the riser region of the shroud to the colder downcomer water outside the shroud which is deleterious to the thermodynamic performance of the reactor. The standard practice in shroud design has typically consisted of hermetically enclosing a fibrous or ceramic insulation in a stainless steel (or another corrosion resistant alloy) enclosure. Such a shroud works well until a leak in the enclosure develops, usually caused by the thermal stresses and strains that are inherent to any structure operating under a temperature differential. Concerns regarding failure of the shroud and subsequent dismembering of the insulation have been a source of significant and expensive ameliorative modification efforts in many operating reactors. The present disclosure provides a reactor shroud which minimizes heat transfer between the hot reactor riser water and cold downcomer water in a manner which eliminates drawbacks of the foregoing insulated enclosure designs. In an embodiment of the present invention, the shroud may be comprised of a series of concentric cylindrical shells separated by a small radial clearance. The top and bottom extremities of the shells are each welded to common top and bottom annular plates (“Closure plates”) to create an essentially isolated set of narrow & tall annular cavities. Each cavity is connected to its neighbor by one or more small drain holes such that submerging the multi-shell body in water (e.g. demineralized primary coolant in a reactor vessel) would fill all of the internal cavities with water and expel virtually all entrapped air, thereby creating water-filled annular cavities. In one non-limiting embodiment, the thin walled concentric shells may be buttressed against each other with a prescribed gap by small fusion welds made by a suitable process such as spot, plug, or TIG welding. In such a welding process, a small piece of metal (e.g. spacer) equal in thickness to the radial gap or clearance in the cavity serves to enable a fusion nugget to be created between the two shell walls. The number of such nuggets is variable, but preferably is sufficient to prevent flow induced vibration of the shroud weldment during reactor operation. One principal advantage of the multi-shell closed cavity embodiment described herein is that it is entirely made of materials native to the reactor's internal space, namely demineralized water (e.g. primary coolant) disposed within the radial gaps between the concentric shells and metal such as stainless steel. No special insulation material of any kind is used in the reactor shroud (which may degrade and fail over time). Advantageously, the present shroud design provides the desired heat transfer minimization between the hot reactor water inside the riser region of the shroud to the colder downcomer water outside the shroud without insulation, thereby preserving the thermodynamic performance of the reactor. According to one exemplary embodiment, a nuclear reactor vessel includes an elongated cylindrical body defining an internal cavity containing primary coolant water, a nuclear fuel core disposed in the internal cavity; an elongated shroud disposed in the internal cavity, the shroud comprising an inner shell, an outer shell, and a plurality of intermediate shells disposed between the inner and outer shells; and a plurality of annular cavities formed between the inner and outer shells, the annular cavities being filled with the primary coolant water. In one embodiment, the annular cavities are fluidly interconnected by a plurality of drain holes allowing the primary coolant to flow into and fill the cavities from the reactor vessel. According to another embodiment, a shroud segment for a nuclear reactor vessel includes an elongated inner shell; an elongated outer shell; a plurality of elongated intermediate shells disposed between the inner and outer shells; the inner shell, outer shell, and intermediate shells being radially spaced apart forming a plurality of annular cavities for holding water; a top closure plate attached to the top of the shroud segment; and a bottom closure plate attached to the bottom of the shroud segment, wherein the top and bottom closure plates are configured for coupling to adjoining shroud segments to form a stacked array of shroud segments. A method for assembling a shroud for a nuclear reactor vessel is provided. The method includes: providing a first shroud segment and a second shroud segment, each shroud segment including a top closure plate and a bottom closure plate; abutting the top closure plate of the second shroud segment against the bottom closure plate of the first shroud segment; axially aligning a first mounting lug on the first shroud segment with a second mounting lug on the second shroud; and locking the first mounting lug to the second mounting lug to couple the first and second shroud segments together. In one embodiment, the locking step is preceded by pivoting a mounting clamp attached to the first shroud segment from an unlocked open position to a locked closed position. All drawings are schematic and not necessarily to scale. Parts given a reference numerical designation in one figure may be considered to be the same parts where they appear in other figures without a numerical designation for brevity unless specifically labeled with a different part number and described herein. The features and benefits of the invention are illustrated and described herein by reference to exemplary embodiments. This description of exemplary embodiments is intended to be read in connection with the accompanying drawings, which are to be considered part of the entire written description. Accordingly, the disclosure expressly should not be limited to such exemplary embodiments illustrating some possible non-limiting combination of features that may exist alone or in other combinations of features. In the description of embodiments disclosed herein, any reference to direction or orientation is merely intended for convenience of description and is not intended in any way to limit the scope of the present invention. Relative terms such as “lower,” “upper,” “horizontal,” “vertical”, “above,” “below,” “up,” “down,” “top” and “bottom” as well as derivative thereof (e.g., “horizontally,” “downwardly.” “upwardly,” etc.) should be construed to refer to the orientation as then described or as shown in the drawing under discussion. These relative terms are for convenience of description only and do not require that the apparatus be constructed or operated in a particular orientation. Terms such as “attached,” “affixed,” “connected,” “coupled,” “interconnected” and similar refer to a relationship wherein structures are secured or attached to one another either directly or indirectly through intervening structures, as well as both movable or rigid attachments or relationships, unless expressly described otherwise. Referring to FIG. 1, a reactor vessel 20 includes a vertically elongated cylindrical both defining a longitudinal axis LA and having a top 21, closed bottom 22, and a circumferentially extending sidewall 24 extending between the top and bottom. Sidewall 24 defines an internal cavity 25 configured for holding a nuclear fuel core 26. Internal cavity extends axially along the longitudinal axis from the top 21 to the bottom 22 of the reactor vessel 20 in one embodiment. The bottom 22 may be closed by a lower head 23, which may be without limitation dished or hemispherical in configuration. In one embodiment, the internal cavity 25 may be filled with a liquid such as primary coolant which may be demineralized water. The reactor vessel 20 may be made of any suitable metal, including without limitation coated steel or stainless steel for corrosion resistance. Referring to FIGS. 1-3 and 7, a vertically elongated shroud 30 is provided which is disposed in the internal cavity 25 of the reactor vessel 20. Shroud 30 may be cylindrical in shape with a circular annular cross-section; however, other suitable shapes may be used. Shroud 30 is coaxially aligned with the reactor vessel 20 along the longitudinal axis LA. The fuel core 26 may be located inside the shroud 30, and in one non-limiting embodiment nearer to the bottom 22 of the reactor vessel 20. Shroud 30 includes a top 34 and bottom 35 which may be spaced vertically apart from the bottom 22 of reactor vessel 20 to provide a flow passage into the shroud 30 at the bottom of the reactor vessel 20 (see. e.g. directional flow arrows FIGS. 1 and 8). In one embodiment as best shown in FIG. 8, the bottom 35 of the shroud 30 may be spaced apart from bottom 22 of reactor vessel 20 and supported by a plurality of radially oriented and circumferentially spaced apart support plates 42. Support plates 42 are configured to engage the reactor vessel bottom 22 at one extremity and bottom 35 of shroud 30 at another extremity. In one embodiment, support plates 42 may include one or more flow holes 41 to allow primary coolant to flow and circulate through the plates at the bottom of the reactor vessel. In other embodiments, the holes may be omitted. The shroud 30 divides the internal cavity 25 of reactor vessel 20 into an outer annular space which defines a vertical downcomer region 28 (i.e. down-flow region) and an inner space which defines a vertical riser region 27 (up-flow region). Primary coolant flows downwards in reactor vessel 20 through the annular downcomer region 28, reverses direction and enters the bottom 35 of the shroud 30, and flows upwards through riser region 27 though the fuel core 26 where the primary coolant is heated for generating steam in an external steam generator. In one embodiment, the shroud 30 may comprise an elongated outer shell 31, an inner shell 32, and a plurality of intermediate shells 33 disposed between the outer and inner shells. Shells 31-33 are cylindrically shaped in one embodiment. Shells 31-33 are concentrically aligned with respect to each other and spaced radially apart forming an array comprised of a plurality of relatively thin concentric annular cavities 40 between the outer and inner shell 31, 32. In one embodiment, the cavities 40 are fluid-filled with primary coolant, as further described herein. Annular cavities 40 extend longitudinally from the top 34 to bottom 35 of shroud 30. Accordingly, the annular cavities 40 have a length or height substantially coextensive with the length of the shells 31-33. The shells 31-33 may be conned of a suitable corrosion resistant metal, such as coated or stainless steel for example. In one exemplary embodiment, the number of intermediate shells 33 may be at least two to provide at least three annular cavities 40. In non-limiting preferred embodiments, at least six or more intermediate shells 33 (divider shells) may be provided to divide the space between the inner and outer shells 32 and 31 into at least seven annular cavities 40. In one representative embodiment, without limitation, eight intermediate shells 33 are provided to create nine intermediate shells 33. The number of water-filled annular cavities 40 selected correlates to the insulating effect and heat transfer reduction from the inner shell 32 through the shroud to the outer shell 31. The number of intermediate shells 33 will be one less than the number of water-filled annular cavities 40 to be created. In order to provide inter-shell connectivity and maintain the radial gap of annular cavities 40 between intermediate shells 33 and between the innermost and outermost intermediate shells and inner shell 32 and outer shell 31 respectively, spacers 80 may be provided as shown in FIG. 2. Spacers 80 are disposed in annular cavities 40 between the shells 31-33 and have a radial thickness sufficient to provide the desired radial width of each annular cavity. Each annular cavity 40 preferably includes spacers 80 in an exemplary embodiment. To retain the spacers 80 in their desired vertical position, the spacers may be rigidly attached to a shell 31-33 by any suitable means such as fusion welding in an exemplary embodiment. In one embodiment, a spot weld 81 may be used to attach spacer 80 to a shell 31-33 as shown. The spot welds 81 may have any suitable diameter, such as without limitation about 1 inch as a representative example. The number of spot welds 81 (spot nuggets) needed for joining neighboring shells 31-33 together may be estimated by the following empirical formula: Number=(shroud diameter times height (in inches)/100). Preferably, the spot welds 81 and spacers 80 should be spaced as uniformly as possible. In one embodiment, the spacers 80 may be radially staggered such that the spacers between adjacent shells 31-33 do not lie on the same radial axis (see, e.g. FIG. 2 showing a set of spacers aligned radially only in every other annular cavity 40). Other suitable arrangements of spacers 80 may be used. Spacers 80 may have any suitable shape, including circular or polygonal configurations. Preferably, spacers 80 may be formed of metal such as steel or other. Referring to FIGS. 2, 3, and 7, each annular cavity 40 may be connected to its adjoining cavities by one or more small fluid drain holes 90. Drain holes 90 are configured and arranged to hydraulically or fluidly interconnect all of the annular cavities 40. The outer shell. 31 includes drain holes 90 which fluidly connect the outermost annular cavity 40 in shroud. 30 to the annular downcomer region 28 in reactor vessel 20. This allows the primary coolant water to enter the outermost cavity 40 and then flow inwards successively through the plurality of drains holes in intermediate shells 33 for filling all the annular cavities with the fluid. Submerging the multi-shell shroud 30 body in the water-filled reactor vessel (e.g. demineralized primary coolant) will fill all of the internal annular cavities 40 with water and expel virtually all entrapped air, thereby creating water-filled annular cavities. In one arrangement, the drains holes 90 may be radially staggered as best shown in FIG. 7 so that the holes in one shell 31 or 33 do not radially align with holes hi its neighboring shells. This forms a staggered flow path through the shroud 30. The inner shell 32 may not have drain holes 90 and is solid in one embodiment. Preferably, a plurality of drain holes 90 are spaced both circumferentially and longitudinally apart along the entire height or length of the shroud 30 in each shroud segment 30A-C Referring to FIG. 2, the inner and outer shells 32 and 31 may have thicknesses T2 and T1 respectively which are larger than the intermediate shells 33 in one embodiment to stiffen and strengthen the shroud 30. For example, in one representative example without limitation inner and outer shells 32 and 31 may have a plate thickness (T1 and T2) of about ¼ inch and intermediate shells 33 may have a thickness T3 of about ⅛ inch. Each annular cavity 40 has a depth D2 (measured in the radial direction transverse to longitudinal axis LA) which is less than the total depth D1 between the inner and outer shells 32 and 31. In one embodiment, the water-filled annular cavities 40 may have a depth D2 that is less than the thickness T1-T3 of the shells 31-33. In one representative example without limitation, the depth of cavity 40 may be about 3/16 inch. This arrangement provides a plurality of thin water films or chambers comprised of primary coolant sandwiched between the inner and outer shells 32 and 31 in the multi-shell weldment (MSW) shroud wall construction. The thin water films have an insulating effect for shroud 30 which minimizes heat transfer between the hot riser region 27 and colder downcomer region 28 (see FIG. 1). Advantageously, the water films eliminate the need for traditional insulation materials in the shroud which may be wetted or otherwise damaged. In one embodiment, inner shell 32, outer shell 31, and intermediate shells 33 may have vertical heights or lengths which are substantially coextensive. According to one aspect of the invention, the shroud 30 may comprise a plurality of vertically stacked and coupled shroud sections or segments 30A, 30B, and 30C. Referring to FIGS. 1 and 3, each shroud segment 30A-C includes an upper end 48, lower end 49, an annular top closure plate 36 attached to upper end 48, and an annular bottom closure plate 37 attached to lower end 49. The top closure plate 36 and bottom closure plate 37 may be formed of a suitable metal such as steel. Corrosion resistant closure plates 36, 37 formed of coated or stainless steel may be used. Within each shroud segment 30A-C the annular cavities 40 and shells 31-33 extend longitudinally between the top and bottom closure plates 36 and 37, and may have coextensive lengths or heights. The outer shell 31, inner shell 32, and intermediate shells 33 in each segment 30A-C may be rigidly attached to the top and bottom closure plates, such as via a rigid connection formed by welding for structural strength. In one embodiment, the shells 31-33 may be hermetically seal joined to the top and bottom closure plates such as with full circumferential seal welds. This forms a water-tight joint between the shells 31-33 and the top and bottom closure plates 36 and 37, respectively. Each shroud segment 30A-C is a self-supporting structure which may be transported, raised, and lowered individually for ease of maneuvering and assembly to adjoining segments during fabrication of the shroud 30. To facilitate handling the shroud segments 30A-C individually, the top closure plates 36 may include radially extending lifting lugs 38 which include a rigging hole 39 for attachment of lifting slings or hoists. A suitable number of lifting lugs 38 circumferentially spaced apart at appropriate intervals are provided to properly and safely hoist the shroud segments 30A-C. The weight of each shroud segment 30A-C may be vertically supported by the shroud segment immediately below with the weight being transferred through the top and bottom closure plates 36 and 37, respectively. Accordingly, in some embodiments, the entire weight of the shroud segments 30A-C may be supported by support plates 42 (see, e.g. FIGS. 1 and 8). In one embodiment, adjoining shroud segments 30A-C may be coupled together at joints 43 between segments via a plurality connectors 76 such as of clamps 50. Referring to FIGS. 1 and 3-5, clamps 50 are configured to detachably join and engage the bottom closure plate 37 of one shroud segment (e.g. 30B) to top closure plate 36 of the adjoining lower shroud segment (e.g. 30C). Clamps 50 each include a U-shaped body 51 defining a recess 52 configured to receive a mounting lug 55 formed on bottom closure plate 37 and a mating mounting lug 56 formed on top closure plate 36 as shown. Mounting lugs 55 and 56 are radially extending and circumferentially spaced apart on bottom and top closure plates 37 and 36, respectively. Each mounting lug 55 is arranged in a pair and coaxially aligned along the longitudinal axis LA with a corresponding mounting lug 56. In one embodiment, the mounting lugs 55 and 56 are integrally formed with and a unitary structural part of the bottom and top closure plates 37, 36. Accordingly, the mounting lugs 55, 56 may preferably be formed of metal similarly to bottom and top closure plates 37, 36 for structural strength. In one arrangement, clamps 50 may each be pivotably connected to a mounting lug 55 on the bottom closure plate 37 by a pivot pin 54 which defines a pivot axis. Pivot pins 54 are oriented parallel to longitudinal axis LA so that the clamp 50 may be pivotably swung or moved transversely to the longitudinal axis LA between a closed locked position (see, e.g. FIG. 4 and open unlocked position (see. e.g. FIG. 5). In one embodiment, pivot pin 54 is disposed proximate to one end 58 of the clamp body 51 and the opposing end 57 is open to receive mounting lug 56 of a top closure plate 36 into recess 52. Pivot pin 54 extends axially through the mounting lug 55 and the bottom and top flanges 59, 60 of clamp 50. To secure the clamp 50 in the closed locked position shown in FIG. 4, a locking fastener such as set screw 53 may be provided which is configured and arranged to engage a top surface of mounting flange 55. Set screw 53 may be threadably engaged in threaded bore 61 formed in top flange 60 of clamp 50. The bore 61 extends completely through top flange 60 to allow the bottom end of the set screw shaft to be projected into or withdrawn from clamp recess 51 for engaging or disengaging mounting flange 55. Raising or lowering the set screw 53 alternatingly disengages or engages the set screw with the mounting flange 55. Set screw 53 is preferably withdrawn from the clamp recess 52 when the mounting flange 55 is inserted therein. A method for assembling shroud 30 comprised of segments 30A-C using clamps 50 will now be described. For brevity, assembly of shroud segment 30B onto segment 30C will be described; however, additional shroud segments may be mounted in a similar manner. Referring to FIG. 3, a pair of shroud segments 30B and 30C are provided each configured as shown. Clamps 50 are in the open unlocked position (see, e.g. FIG. 5). Shroud segment 30B is first axially aligned, along longitudinal axis LA with segment 30C. Segment 30B may then be rotated as needed to axially align mounting flanges 55 on bottom closure plate 37 with mounting flanges 56 on top closure plate 36 of segment 30C. Each pair of mounting flanges 55 and 56 may be brought into abutting relationship. In the process, bottom closure plate 37 is brought into abutting contact with top closure plate 36 forming the joint 43 between segments 30B and 30C. Clamp 50 is then pivoted about pivot pin 54. Mounting flanges 55 and 56 are inserted into recess 51 of clamp 50 between flanges 59 and 60 (see, e.g. FIG. 5). The set screw 53 is then tightened to secure the clamp 50 in the closed locked position shown in FIG. 5. It will be appreciated that the order of performing the steps of the fore steps may be varied. In addition, numerous variations of the foregoing assembly process are possible. Referring to FIG. 2, a sealing gasket 44 may be provided in between each pairing of a top closure plate 36 and bottom closure plate 37 to seal the interface at joint 43 therebetween. In one embodiment, the gasket 44 may be metallic formed of steel, aluminum, or another seal material suitable for the environment within a reactor vessel 20. The gasket 44 may be situated in an annular groove 45 formed in the bottom closure plate 37 as shown, or alternatively in the top closure plate 36 (not shown), to seal water seepage at the interface of joint 43 and also provide a certain level of verticality alignment capability during installation and joining of shroud segments 30A-C. In one embodiment, gasket 43 may be circular in transverse cross-section prior to the joint 43 being closed which will compress and deform the gasket. According to another aspect of the invention, a plurality of lateral seismic restraints such as restraint springs 70 may be provided to horizontally support and protect the structural integrity of the shroud 30 inside reactor vessel 20 during a seismic event. In one embodiment as shown in FIGS. 4 and 5, a dual purpose connector 76 (fastener or coupler for joints 43 between shroud segments 30A-C and lateral restraint) may be provided which combine the clamps 50 and seismic springs 70 into a single assembly. Referring to FIGS. 1 and 3-5, seismic springs 70 are disposed between and engage shroud 30 and the interior surface 74 of the reactor vessel 20. A plurality of seismic springs 70 are provided which are circumferentially spaced apart on the outer shell 28 of the shroud 30. In one embodiment, the seismic springs 70 may be spaced apart at equal intervals. Seismic springs 70 are elastically deformable to absorb lateral movement of the shroud 30. In one embodiment, each spring. 70 may be in the form of an arcuate leaf spring comprised of a plurality of individual leaves 75 joined together to function as a unit. The leaves 75 may be made of suitable metal such as spring steel having an elastic memory. Other appropriate materials however may be used. The thickness and number of leaves 75 may be varied to adjust the desired spring force K of the spring 70. Seismic springs are arranged with the concave side facing outwards away from shroud 30 and towards reactor vessel 20 when in the fully mounted and active operating position. Opposing ends 72 and 73 of each seismic spring 70 are arranged to engage the interior surface 74 of reactor vessel 20. In one embodiment, seismic springs 70 may be rigidly attached to shroud 30 to provide a stable mounting for proper operation and deflection of the spring to absorb energy during a seismic event. In one possible arrangement, seismic springs 70 may be rigidly attached to clamps 50 via a fastener 71 or another suitable mounting mechanism. Spring 70 may be fastened to clamp 50 at the midpoint between ends 72 and 73 in one embodiment. Accordingly, seismic springs 70 may be pivotably movable with clamps 50 in the manner already described herein. In FIG. 1, for example, the seismic spring 70 and clamp 50 shown between shroud segments 30A and 30B is in the open unlocked position. In this same figure, seismic springs 70 shown between shroud segments 30B and 30C are in the pivoted closed locked position in which the seismic springs 70 are in the active operating position with ends 72 and 73 engaged with the reactor vessel 20. During a seismic event when the shroud 30 may shift laterally/horizontally in one or more directions, the seismic springs 70 will deform and deflect assuming a more flattened configuration until the seismic load is removed, thereby returning the spring elastically to its original more arcuately-shaped configuration shown. In one embodiment, each joint 43 between shroud segments 30A, 30B, and 30C may include seismic springs 70 to horizontal support the shroud 30 intermittently along its entire height. Underlying Operating Principle of the Shroud The multi-shell weldment (MSW) design for shroud 30 described herein is based on the principle in applied heat transfer which holds that an infinitely tall and infinitesimally thin closed end cavity filled with water would approximate the thru-wall thermal resistance equal to that of the metal walls and the water layer conductances. The governing dimensionless quantity that provides the measure of departure from the ideal (conduction only is Rayleigh number defined as the product of the Prandtl number (Pr) and the Grashof number (Gr). Heat transfer in a differentially heated vertical channel, of height H and gap L is characterized by Nusselt number correlation as a function of Rayleigh number as follows:Nu=0.039Ra1/3 Where:Nu is Nusselt Number (=hL/k)h is heat transfer coefficientk is conductivity of waterRa is Rayleigh number (=gβΔTL3ρ2/μ2)*Prg is gravitational accelerationβ is coefficient of thermal expansion of waterΔT is hot-to-cold face temperature differenceρ is density of waterμ is water viscosityAs Rayleigh number defined above exhibits an L3 scaling it follows that gap reduction substantially affects Ra number. For example a factor of 2 gap reduction cuts down Ra number by a factor of 8 (almost by an order of magnitude). Thus engineering the shroud with small gaps has the desired effect of minimizing heat transfer. To further restrict heat transfer a multiple array of gaps are engineered in the shroud lateral space to have the effect of resistances in series. An example case is defined and described below to illustrate the concept. A Small Modular Reactor (SMR), such as the SMR-160 available from SMR. LLC of Jupiter, Fla., may have a particularly long shroud (e.g. over 70 feet). In such a case, the principal design concerns are: ease of installation, removal, verticality of the installed structure, mitigation of thermal expansion effects and protection from flow induced vibration of the multi-wall shell. The design features, described below to address the above concerns for such an SMR, can be applied to any shroud design. A Narrow cavity geometry: The height of each shroud (e.g. shroud segments 30A-C) is approximately three times its nominal diameter. The innermost and outer most shells (e.g. shells 32 and 31) are relatively thick compared to the intermediate (inner) shells (e.g. shells 33). The water cavity is less than 0.1% of the Shrouds height. The table below provides representative dimensions for demonstrating the concept: Dimensions of a typical shroud in SMR-160: Inner diameter 71⅛ inch Height 71 ft. (Shroud built in four stacked sections (segments), 3×20 ft, (lower) and 1×11 ft. (top)) Number of water annuli (cavities) 9 Thickness of inner most shell ¼ inch Thickness of outermost shell ¼ inch Thickness of interior shell walls ⅛ inch Thickness of water cavities 3/16 inch B. Inter-shell connectivity: The number of spot nuggets (approximately 1 inch diameter) joining, neighboring shells should be estimated by the following empirical formula: Number=(shroud diameter times height tin inches)/100). The spot welds should be spaced as uniformly as possible. C. Handling: The top plate 36 of each shroud segment 30A-C is equipped with lift lugs 38 for handling and installation. Typically six lift lug locations, evenly spaced in the circumferential direction, will suffice. D. Stacked construction: The multi-shell weldments (MSW) of shroud segments 30A-C are stacked on top of each other as shown in FIGS. 1 and 2. One or more round metallic gaskets 44 as described above are provided at the interface between the annular top and bottom closure plates 36, 37 of successive stacks of shroud segments 30A-C. The gaskets 44 situated in the annular grooves 45 in the bottom closure plate 37 serve to seal water seepage at the interface of joint 43 and also provide a certain level of verticality alignment capability. E. Thermal expansion: The axial thermal expansion of a tall stack of shroud segments 30A-C will cause severe stresses in adjoining structures such as the return piping that delivers the reactor coolant from the steam generator to the reactor's outer annulus (downcomer). To mitigate the thermal stresses, the upper region of the shroud may be equipped with a multiply bellows type expansion joint. F. Seismic restraints. The junctions or joints 43 of the MSW shroud segments 30A-C provide the “hard” locations to join them and to secure them against lateral movement during earthquakes. The dual purpose connector 76 (fastener and lateral restraint) design concept shown in FIGS. 3-5 comprising the clamps 50 and seismic springs 70 as described herein provide the joining and lateral restraint functionality. This dual purpose connector 76 has the following capabilities: (i) The two interfacing closure plates 36 and 37 are prevented from significant rotation or separation from each other during earthquakes. (ii) The connector 76 is amenable to remote installation and removal. (iii) The connector 76 is equipped with the seismic springs 70 (e.g. leaf springs) to enable it to establish a soft contact or a small clearance with the reactor's inside wall under operating condition (hot). A set of three connectors 76, equipment-spaced in the circumferential direction at each closure plate 36, 36 elevation, is deemed to be adequate for the SMR described above. Additional connectors may be employed in other reactor applications at the designer's option. Performance assessment: The efficacy of the MSW design is demonstrated by the case of the SMR-160 described above. Calculations show that the decrease in the hot leg temperature (primary coolant inside shroud 30) using water-filled annular cavities 40 due to heat loss across the shroud is merely 0.355 deg. F. As a point of reference, the idealized temperature loss would be 0.092 deg. F. if the water layers were instead omitted and “solid,” heat transferred only by conduction through the shroud. It can be seen that the Rayleigh effect, responsible for the movement of water in closed cavities, has been largely suppressed by the MSW design of shroud 30. Extension to vessels and conduits: The concept of establishing a thin water layer inside pipes (hereafter called “water lining”) carrying heated water is proposed to be employed at the various locations in the power plant where minimizing heat loss from the pipe is desired. For example, the lines carrying hot and cooled reactor coolant are water lined to limit heat loss. Water lining is achieved by the following generic construction: (i) An inner thin walled (liner) pipe that is nominally concentric with the main pipe. The liner pipe has a few small holes to make the narrow annulus communicate with the main flow space. (ii) The small gap between the main and liner pipes is held in place by small spacer nuggets attached to the outside surface of the liner pipe. (iii) In piping runs subject to in-service inspection of pressure boundary welds, the liner pipe is discontinued at the location of such welds. The foregoing, water lining approach is also proposed to be used to reduce thermal shock to pressure retaining vessel/nozzle junctions (locations of gross structural discontinuity) where large secondary stresses from pressure exist. This is true of penetrations in the reactor vessel, steam generator as well as the superheater. Water lined pressure boundaries will experience significantly reduced fatigue inducing, cyclic stresses which will help extend the service life of the owner plant. While the foregoing description and drawings represent exemplary embodiments of the present disclosure, it will be understood that various additions, modifications and substitutions may be made therein without departing from the spirit and scope and range of equivalents of the accompanying claims. In particular, it will be clear to those skilled in the art that the present invention may be embodied in other forms, structures, arrangements, proportions, sizes, and with other elements, materials, and components, without departing from the spirit or essential characteristics thereof. In addition, numerous variations in the methods/processes described herein may be made within the scope of the present disclosure. One skilled in the art will further appreciate that the embodiments may be used with many modifications of structure, arrangement, proportions, sizes, materials, and components and otherwise, used in the practice of the disclosure, which are particularly adapted to specific environments and operative requirements without departing from the principles described herein. The presently disclosed embodiments are therefore to be considered in all respects as illustrative and not restrictive. The appended claims should be construed broadly, to include other variants and embodiments of the disclosure, which may be made by those skilled in the art without departing from the scope and range of equivalents. |
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description | The present invention relates to semiconductor manufacturing technology, and more particularly, to a method and system of non-uniformity pattern identification. A conventional semiconductor factory typically includes the requisite fabrication tools necessary to process semiconductor wafers for a particular purpose, such as photolithography, chemical-mechanical polishing, or chemical vapor deposition. During manufacture, the semiconductor wafer passes through a series of process steps, which are performed by various fabrication tools. For example, in the production of an integrated semiconductor product, the semiconductor wafer passes through up to 600 process steps. The costs for such automated production are influenced to a great extent by the question as to how well and efficiently the manufacturing process can be monitored or controlled, so that the ratio of defect-free products to the overall number of products manufactured (i.e., yield ratio) achieves as great a value as possible. The individual process steps, however, are subject to fluctuations and irregularities, which in the worst case may mean, for example, the defect of a number of chips or the entire wafer. Therefore, each individual process step must be carried out as stably as possible in order to ensure an acceptable yield after the completed processing of a wafer. The fluctuations, irregularities and instability of a process step will cause so-called non-uniformity patterns, reducing yield. There may be various types of with-in-wafer (WIW) non-uniformity patterns of particular data, e.g., in-line process manufacturing parameters, wafer acceptance test (WAT) parameters, circuit probing (CP) test parameters and the like, subject to various fabrication issues. In the past, simple calculation algorithms, such as range value, and standard deviation, with predetermined thresholds have been used to determine whether a wafer suffers from WIW non-uniformity. Identification of WIW non-uniformity patterns, however, is done by human effort. The labor-intensive nature of WIW non-uniformity pattern identification using conventional means severely hinders efficiency. Therefore, a need exists for a system and method of non-uniformity pattern identification, to not only improve efficiency, but also provide a more effective and reliable result. An embodiment of a system for non-uniformity pattern identification comprises a storage device and a processing unit. The storage device stores multiple theoretical patterns and measurements. Each measurement corresponds to a region on a wafer. The processing unit acquires the theoretical patterns and the measurements on at least two wafers, calculates pattern scores for the respective theoretical patterns of each wafer according to the measurements, and groups at least two of the theoretical patterns into at least one factor according to the pattern scores to identify non-uniformity patterns for the wafers. Each pattern score represents the extent of similarity between one of the theoretical patterns and the measurements in one of the wafers. The processing unit may further output a graph corresponding to the factor to an output device. The graph may comprise a contour, a box plot chart or a histogram. An embodiment of methods for non-uniformity pattern identification comprises acquiring multiple theoretical patterns, acquiring multiple measurements on at least two wafers, calculating pattern scores for the respective theoretical patterns of each wafer according to the measurements, and grouping at least two of the theoretical patterns into at least one factor according to the pattern scores to identify non-uniformity patterns for the wafers. Each measurement corresponds to a region on one of the wafers. Each pattern score represents the extent of similarity between one of the theoretical patterns and the measurements on one of the wafers. Preferably, the method additionally comprises outputting a graph corresponding to the factor to an output device, in which the graph may be a contour, a box plot chart or a histogram. An embodiment of a machine-readable storage medium stores a computer program which when executed performs the method of non-uniformity pattern identification. Preferably, the theoretical patterns may comprise a uniformity pattern and a plurality of non-uniformity patterns. The theoretical patterns may be implemented in a matrix, a two-dimensional array, a linked list or a tree. The region may cover one or more dies on a wafer, or cover a portion of one die. The measurements may be electrical measurements or physical measurements, acquired during wafer acceptance test (WAT) or in-line processing measurement. In pattern score calculation, in one example, the pattern scores for the respective theoretical patterns of each wafer may be calculated by a correlation analysis algorithm or a data classification method according to the measurements. In another example, the pattern scores may be calculated by the following equation: MT m × m × [ W1 ⋮ ⋮ ⋮ Wm ] × 1 L = [ P1 P2 ⋮ ⋮ Pm ] ,where MTm×m represents the m-by-m matrix for m theoretical patterns, W1 to Wm represent measurements individually occurring in the respective regions, L represents an individual standardization factor, which is the square root of the sum of the square of the cell values for each rows 1 to m, and P1 to Pm represent the pattern scores. In factor generation, the theoretical patterns are grouped into factors using a principal component analysis (PCA) or a data clustering algorithm. It is understood, however, that the following disclosure provides many different embodiments, for examples, for implementing different features of the invention. Specific examples of components and arrangements are described below to simplify the present disclosure. These are, of course, merely examples and are not intended to be limiting. In addition, the present disclosure may repeat reference numerals and/or letters in the various examples. This repetition is for the purpose of simplicity and clarity and does not in itself dictate a relationship between the various embodiments and/or configurations discussed. FIG. 1 is a diagram of an embodiment of a hardware environment. The description of FIG. 1 is provides a brief, general description of suitable computer hardware and a suitable computing environment in conjunction with which at least some embodiments may be implemented. The hardware environment of FIG. 1 includes a processing unit 11, a memory 12, a storage device 13, an input device 14, an output device 15 and a communication device 16. The processing unit 11 is connected by buses 17 to the memory 12, storage device 13, input device 14, output device 15 and communication device 16 based on Von Neumann architecture. There may be one or more processing units 11, such that the processor of the computer comprises a single central processing unit (CPU), a micro processing unit (MPU) or multiple processing units, commonly referred to as a parallel processing environment. The memory 12 is preferably a random access memory (RAM), but may also include read-only memory (ROM) or flash ROM. The memory 12 preferably stores program modules executed by the processing unit 11 to perform experiment management functions. Generally, program modules include routines, programs, objects, components, or others, that perform particular tasks or implement particular abstract data types. Moreover, those skilled in the art should understand that at least some embodiments may be practiced with other computer system configurations, including hand-held devices, multiprocessor-based, microprocessor-based or programmable consumer electronics, network PCs, minicomputers, mainframe computers, and the like. Some embodiments may also be practiced in distributed computing environments where tasks are performed by remote processing devices linked through a communication network. In a distributed computing environment, program modules may be located in both local and remote memory storage devices based on various remote access architecture such as DCOM, CORBA, Web object, Web Services or other similar architectures. The storage device 13 may be a hard drive, magnetic drive, optical drive, a portable drive, or nonvolatile memory drive. The drives and their associated computer-readable media (if required) provide nonvolatile storage of computer-readable instructions, data structures, program modules and experiment lot processing records. The processing unit 11, controlled by program modules received from the memory 12 and from an operator through the input device, directs experiment management functions. The storage device 13 may comprise a database management system, an object base management system, a file management system, or others, to store multiple experiment plan records, merge constraints and scheduling rules. This embodiment described in the following discloses methods for non-uniformity pattern identification implemented in program modules and executed by the processing unit 11. FIG. 2 is a flowchart showing an embodiment of methods for non-uniformity pattern identification. The process of FIG. 2 begins in step S211 to acquire multiple theoretical patterns from the storage device 13. Theoretical patterns including uniformity pattern and non-uniformity patterns may preferably be simulated by a matrix (also called a feature space) or others. In one example, equation (1) shows the 9-by-9 matrix MT9×9 for theoretical patterns. Equation ( 1 ) : MT 9 × 9 = [ 1 1 1 1 1 1 1 1 1 0 1 - 1 - 1 1 2 0 - 2 0 0 - 1 - 1 1 1 0 - 2 0 2 0 - 2 0 2 0 1 - 1 - 1 1 0 0 2 0 - 2 1 - 1 - 1 1 0 0 0 0 0 - 1 - 1 1 1 0 1 - 1 1 - 1 0 0 0 0 4 1 1 1 1 - 2 - 2 - 2 - 2 4 - 2 - 2 - 2 - 2 1 1 1 1 ] , where each column indicates a particular measurement region, and each row represents a theoretical pattern. Preferably, any two of the theoretical patterns are orthogonal. Those skilled in the art will recognize that less or more rows can be used to simulate less or more theoretical patterns, and less or more columns can be used to indicate less or more measurement regions. Those skilled in the art will also appreciate that various data structures, such as two-dimensional arrays, linked lists, trees, and the like, may be used to represent theoretical patterns. The implementation of theoretical patterns is not limited to vectors, but may be implemented as equations and the like. FIG. 3 is a schematic diagram of an exemplary wafer. In this example, actual patterns for a wafer may be examined by acquiring data from nine measurement locations 31 to 39. Referring to equation (1), columns 1 to 9 in this matrix respectively associate with measurement locations 31 to 39. FIGS. 4a to 4i are 3D diagrams showing theoretical patterns. For example, row 1 simulates the uniformity pattern of FIG. 4a. Row 2 simulates the uniformity pattern of FIG. 4b wherein the distribution of measurement results on a wafer descends gradually from the lower-right region to the upper-left region. Row 3 simulates the uniformity pattern of FIG. 4c wherein the distribution of measurement results on a wafer descends gradually from the upper-right region to the lower-left region. Row 9 simulates the uniformity pattern of FIG. 4i wherein the distribution of measurement results on a wafer descends circularly from the center to the edge. Next, a loop (steps S221 to S241) is used to calculate pattern scores of theoretical patterns representing the extent of similarity between theoretical patterns and measurements wafer by wafer. In step 221, measurements (e.g., electrical or physical measurements) in different regions on a wafer are received from the storage device 13. Each region may cover one or more dies on a wafer, or cover portions of one die. Electrical or physical parameter measurement may be acquired during wafer acceptance tests (WAT), in-line processing measurements and the like. Each measurement may represent an electrical value, such as voltage level, resistance, power level and the like, or a physical value, such as line width, overlay, thickness and the like, for one or more semiconductor devices. For example, wafer acceptance test (WAT) data is generated by electrical measurements of these test structures after completion of the entire fabrication process. Several sites located on the fixed locations on each wafer are selected, from which over 100 WAT parameters are measured. In step S231, patterns scores with theoretical patterns for a wafer are calculated, representing the extent of similarity between theoretical patterns and actual measurements on wafers. In one example, equation (2) shows the formula for calculating pattern scores. Equation ( 2 ) : MT 9 × 9 × [ w1 w2 w3 w4 w5 w6 w7 w8 w9 ] × 1 L = [ P1 P2 P3 P4 P5 P6 P7 P8 P9 ] , where MT9×9 represents the 9-by-9 matrix for theoretical patterns as shown in equation (1), w1 to w9 represent measurements in different regions, L represents an individual standardization factor, which is the square root of the sum of the square of the cell values for each rows 1 to 9, and P1 to P9 represent pattern scores. For example, standardization factor in row 9 is L9=sqrt(4^2+(−2)^2+(−2)^2+(−2)^2+(−2)^2+1^2+1^2+1^2+1^2)=6. Those skilled in the art will also appreciate that pattern scores may also be calculated by various techniques, such as a variety of correlation analysis algorithms, data classification methods, or others, with relevant implementation of theoretical patterns. In step S241, the process determines whether a wafer that has not been analyzed is present, if so, the process proceeds to step S221, and otherwise, to step S251. Thus, suspicious non-uniformity patterns of analyzed wafers may be the theoretical patterns with the highest pattern score. Although the suspicious non-uniformity pattern of analyzed wafers can be determined by pattern scores of theoretical patterns, in most situations, a real non-uniformity pattern is a combination of two or more theoretical patterns. In step S251, non-uniformity theoretical patterns are grouped into factors according to pattern scores using various factor analysis techniques, such as principal component analysis (PCA), a variety of data clustering algorithms. FIG. 5 is a diagram of exemplary results of factor generation. Rows 51 to 54 illustrate factor analysis results for four WAT parameters “Isat_N4”, “Isat_P4”, “Isat_N6” and “Isat_P6”. The factor composition column shows theoretical pattern combinations for factors, for example, in WAT parameters “Isat_N4”, 3 factors are generated by PCA. Three factors individually contain pattern combination (P6, P9, P4, P3, P2), (P8, P5) and (P8, P7). The first factor has the highest explanability (40.81%) over the other two factors, thus, the first factor may be the most likely suspicious non-uniformity pattern. In step S261, graphs, such as contours, box plot charts, histograms and the like, corresponding to a factor are sent to the output device 15. FIG. 6 is an exemplary contour of a non-uniformity pattern depicting an identified pattern combination. The output device 15 may be a display device, such as a monitor screen, a projector and the like, or a printing device, such as a printer, a plotter and the like. Embodiments of the invention provide additionally a storage medium as shown in FIG. 7 storing a computer program 720 for executing the disclosed methods of non-uniformity pattern identification. The computer program product includes a storage medium 70 having computer readable program code embodied in the medium for use in a computer system, the computer readable program code comprising at least computer readable program code 721 receiving multiple theoretical patterns, computer readable program code 722 calculating pattern scores of theoretical patterns representing the extent of similarity between theoretical patterns and measurements wafer by wafer, computer readable program code 723 grouping non-uniformity theoretical patterns into factors according to pattern scores and computer readable program code 724 outputting graphs based on a factor. The methods and systems of the embodiments, or certain aspects or portions thereof, may take the form of program code (i.e., instructions) embodied in tangible media, such as floppy diskettes, CD-ROMS, hard drives, or any other machine-readable storage medium, wherein, when the program code is loaded into and executed by a machine, such as a computer, the machine becomes an apparatus for practicing the invention. The methods and apparatus of the present invention may also be embodied in the form of program code transmitted over some transmission medium, such as electrical wiring or cabling, through fiber optics, or via any other form of transmission, wherein, when the program code is received and loaded into and executed by a machine, such as a computer, the machine becomes an apparatus for practicing the invention. When implemented on a general-purpose processor, the program code combines with the processor to provide a unique apparatus that operates analogously to specific logic circuits. While the invention has been described by way of example and in terms of preferred embodiment, it is to be understood that the invention is not limited thereto. Those who are skilled in this technology can still make various alterations and modifications without departing from the scope and spirit of this invention. Therefore, the scope of the present invention shall be defined and protected by the following claims and their equivalents. |
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063109298 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to an in-core fixed nuclear instrumentation system and a power distribution monitoring system of a reactor such as a boiling water type reactor. 2. Description of the Prior Art A reactor, for example, a boiling water type reactor (hereinafter, referred simply to as BWR) is provided with a power distribution monitoring system which monitors a reactor operating mode and a reactor power distribution (hereinafter, in this specification, the reactor power distribution is described as "in-core power distribution", "core power distribution", or the like), as generally shown in FIG. 23 and FIG. 24. In the BWR, as shown in FIG. 23, a reactor pressure vessel 2 is housed in a primary containment vessel 1, and a reactor core 3 is accommodated in the reactor pressure vessel 2. As shown in FIG. 24, the reactor core 3 is constructed in a manner that many fuel assemblies 4 and control rods 5 are mounted therein. In the reactor core 3, an in-core nuclear instrumentation assembly 6 is located on a position in the reactor core 3, which is surrounded by fuel assemblies 4. As shown in FIG. 24, the in-core nuclear instrumentation assembly 6 is arranged in a corner water gap G formed by four fuel assemblies 4, and neutron detectors 8 are discretely arranged on several positions along a core axial direction in a nuclear instrumentation tube (pipe) 7. The neutron detector 8 is a so-called fixed type, and in a boiling water type reactor (BWR), usually, four neutron detectors are discretely arranged on a fuel effective portion at equal intervals. Further, in the nuclear instrumentation tube 7, a TIP (Traversing In-Core Probe) conduit pipe 9 is arranged, and in the TIP conduit pipe (tube) 9, one traversing neutron detector (TIP) 10 is located so as to be movable in an axial direction of the TIP conduit pipe 9. Moreover, as shown in FIG. 23, there is located a movable type neutron flux measuring system for axially continuously measuring neutron flux by means of a matrix device 11, a TIP driving device 12, a TIP control and neutron flux signal processing system 13 or the like. A reference numeral 14 denotes a penetration section, a reference number 15 denotes a valve mechanism and a reference number 16 denotes a shielding container. These neutron detectors 8 and 10, control systems such as signal processing systems 13 and 17 (will be described later) for neutron detectors 8 and 10 constitute a reactor nuclear instrumentation system 18. On the other hand, the in-core fixed neutron detectors 8 {LPRM (local power range monitor) detector} arranged in the reactor core are divided into some groups, and then, an average signal {APRM (average power range monitor) signal} for each group is generated so that an power level of a power range of the reactor core 3 is monitored on the basis of these APRM signals. More specifically, when an abnormally transient phenomenon or accident such that neutron flux rapidly rises up happens, the LPRM detectors 8 detect the transient phenomenon and the occurrence of accident so that, according to the APRM signals generated by the detected signals of the LPRM detectors 8, a reactor safety protection system (not shown) rapidly makes a scram operation of a reactor scram system (not shown) such as a control rod driving mechanism or the like in order to prevent fuel assemblies or the reactor core from being break down. That is, the LPRM detector 8 is constituted as a part of the reactor safety protection system. By the way, in individual in-core fixed neutron detectors 8, a sensitivity change takes place by neutron irradiation or the like. In order to calibrate a sensitivity of each neutron detector 8 for each predetermined period during operation, the TIP (traversing neutron detector) 10 is operated so as to obtain a continuous power distribution in a core axial direction, and the sensitivity change of each neutron detector 8 is corrected by means of a gain adjusting function of a neutron detector (LPRM) signal processing system 17. A detection signal S2 detected by the neutron detector 8 is processed by means of the signal processing system 17, and thereafter, is transmitted to a process computer 20 which will be described later. In general, the BWR is provided with a process control computer 20 for monitoring an operating mode and power distribution of a nuclear (atomic) power plant. The process control computer 20 is provided with a nuclear instrumentation control system 21 for monitoring and controlling the reactor nuclear instrumentation system 18, a power distribution simulating system 22 including a physics model having three-dimensional thermal-hydraulics simulation code, and an input-output system 23. The reactor power distribution simulating system 22 is incorporated in one or plural process control computers 20 as a program. Further, the reactor power distribution simulating system 22 includes a power distribution simulating module 24 and a power distribution learning (adaptive) module 25. Neutron flux signal obtained by the TIP 10 of the reactor nuclear instrumentation system 18 is processed as a nuclear instrumentation signal corresponding to a core axial direction position by means of the TIP neutron flux signal processing system 13 of the reactor nuclear instrumentation system 18. Then, the nuclear instrumentation signal is read via the nuclear instrumentation control system 21 of the process control computer 20 into the power distribution simulating system 22 as a reference power distribution in a three-dimensional nuclear thermal-hydraulics simulation. On the other hand, core state data S3 (process quantity) including a control rod pattern, a core coolant flow rate, an internal pressure of the reactor pressure vessel, flow of feed water, a temperature of feed water (a core inlet coolant temperature) and so on, which are used as various operating parameters indicative of a reactor operating mode (state) and obtained from a core state data measuring apparatus 26 as a reactor core state data measuring means, is read into a core state data processing system 27, and then, is processed so that a reactor thermal output or the like is calculated. Then, the reactor core state data S3 including the calculated reactor thermal output is transmitted to the reactor power distribution simulating system 22 via the nuclear instrumentation control system 21 of the process control computer 20. In fact, the reactor core state data measuring apparatus 26 is composed of a plurality of monitoring devices. In addition, the reactor core state data measuring apparatus 26 is a general name of an apparatus for collecting process data of various operating parameters of the reactor, and is shown as one measuring apparatus in FIG. 23 for simplification. Moreover, the core state data processing system 27 may be used as one function of the process control computer 20. The detection signals S2 and the core state data S3 transmitted in the aforesaid manner are transmitted to the power distribution simulating system 22 of the process control computer 20. In the power distribution simulating system 22, a core power distribution is simulated on the basis of the transmitted core state data S3 and the three-dimensional nuclear thermal-hydraulics simulation code of the power distribution simulating module 24. Further, the power distribution simulating system 22 learns a reference power distribution of the core nuclear instrumentation data by a learning function (adapting function) of the power distribution learning (adapting) module 25, and then, corrects the simulation result (core power distribution) while referring to the reference power distribution. As a result, in a power distribution predictive simulation after that, it is possible to accurately simulate a reactor power distribution. In the conventional in-core nuclear instrumentation assembly 6, in place of the traversing neutron detector 10, as shown in a perspective view partly in section of FIG. 25, there may be provided a traversing .gamma.-ray detector 10A which is moved in a core axial direction so as to continuously measure a .gamma.-ray flux in the core axial direction. The .gamma.-ray is generated in proportional to a fission quantity in the reactor core; for this reason, the .gamma.-ray flux is measured, and whereby, it is possible to measure a fission quantity profile in the reactor core. By using the traversing neutron detector 10 and the .gamma.-ray detector 10A, it is possible to calibrate a dispersion of each detection accuracy of each plural neutron detector 8 arranged in the core axial direction. As described above, in the conventional reactor nuclear instrumentation system 18, in order to continuously measure a power distribution in the axial direction of the reactor core 3, the traversing neutron detector 10 and the traversing .gamma.-ray detector 10A constituting a movable type in-core instrumentation system have been used. However, in the traversing (movable type) neutron detector 10 and the .gamma.-ray detector 10A, measurement is made in a manner of vertically moving at least one neutron detector 10 or .gamma.-ray detector 10A over the entire length (core axial length) of the core 3 in the TIP conduit tube 9 from the outside of the reactor pressure vessel 2. For this reason, there are the problems that a mechanical driving and operating mechanism for moving and operating the traversing neutron detector 10 and the .gamma.-ray detector 10A is made large and its structure is complicate so that a moving operation and maintenance of the mechanical driving and operating mechanism are troublesome. In particular, maintenance and management are required for the mechanical driving and operating mechanisms such as the detector driving device 12 for moving and operating the traversing neutron detector 10 and the traversing .gamma.-ray detector 10A, the matrix device 11 for selecting the TIP conduit tube 9, the valve mechanism 15, the shielding container 16 or the like. In addition, the traversing detectors 10 and 10A are activated; for this reason, there is the possibility that an operator (worker) which is carried out the above maintenance and management works of the traversing neutron detector 10 and the traversing .gamma.-ray detector 10A is exposed. In view of the above problem, in the reactor nuclear instrumentation system, there is made a demand for a method of monitoring a reactor operating mode and a power distribution in a core axial direction without using a traversing type measuring (nuclear instrumentation) system. The in-core nuclear instrumentation assembly 6 used in the conventional reactor nuclear instrumentation system is usually provided with four fixed neutron detectors 8, one traversing neutron detector (TIP) 10 or traversing .gamma.-ray detector 10A, and a hollow conduit (TIP conduit tube 9) for housing movably the traversing neutron detector (traversing .gamma.-ray detector). In place of the TIP 10, a study has been made such that a fixed .gamma.-ray heating detector is arranged in the same manner as the fixed neutron detector 8. However, in the case where a plurality of, for example, four fixed .gamma.-ray heating detectors, are located in a core axial direction, it is impossible to measure a power on an upper and lower portions of the reactor core 3. Further, in the case of extrapolating measurement data on the upper and lower portions of the reactor core 3 from four measurement data or interpolating these four measurement data, a behavior of power distribution change differs between individual portions in the core axial direction; for this reason, a great measurement error is caused so that an accuracy becomes worse. If only fixed measuring (nuclear instrumentation) devices are located at several positions in an axial direction in the reactor nuclear instrumentation system, a measurement error is great in a power distribution of the core axial direction. For this reason, there is a need of previously setting a great margin on a limiting condition (operational thermal limit) for a reactor operation. As a result, a degree of the margin of the reactor operation is reduced, so that there is the possibility of giving a bad influence to an availability factor of the reactor. Moreover, in order to improve a measurement accuracy of the power distribution in the core axial direction, it is considered that a plurality of fixed .gamma.-ray heating detectors are arranged in the core axial direction. However, in this case, a number of detector signal cables is increased, and there is a limit to arrange many .gamma.-ray heating detectors in view of restriction of the number of detector connecting cables capable of passing through the nuclear instrumentation tube 7 of the in-core nuclear instrumentation assembly 6. As disclosed in Japanese Unexamined Patent Publication No. 6-289182, there has been proposed a reactor nuclear instrumentation system in which many .gamma.-ray heating detectors (called as GT or GT detector) are arranged. However, in the reactor nuclear instrumentation system, a .gamma.-ray heating contributing range analysis and a knowledge of .gamma.-ray heating are insufficient. Since at least one of .gamma.-ray heating detectors located on upper and lower ends of the core is arranged at a position within 15 cm from the upper and lower ends of the fuel effective portion of the core axial direction, it is difficult to accurately detect each .gamma.-ray heating value on the upper and lower ends of the fuel effective portion. In the case of measuring a power distribution of reactor core with the use of many fixed .gamma.-ray heating detectors (GT detectors), a part of these many GT detectors is arranged in the vicinity of an LPRM detector, and thereby, the following technique has been proposed; more specifically, since the GT detector has characteristic of less variation of bias, sensitivity or gain of the LPRM detector is adjusted by means of the GT detectors, or a GT assembly, which has a plurality of GT detectors arranged in an axial direction, is used as core axial direction power distributing measuring means, in place of the traversing neutron detector or the traversing gamma-ray detector. In the .gamma.-ray heating detector (GT detector) used in the conventional reactor nuclear instrumentation system, a differential thermocouple is used for detecting a .gamma.-ray heating temperature. For this reason, the following report has been made; more specifically, almost no aged deterioration; however, a voltage output of the thermocouple with respect to a gamma heat value lowers depending upon an elapse time for unit of week or month, and a saturated phenomenon of the voltage output happens after a stay time of the GT detector in the core (mounted time of the GT detector in the core) to some degree. Therefore, with the use of a heater incorporated into a gamma-ray thermometer (GT) assembly comprising a plurality of GT detectors, a sensitivity {sensitivity coefficient (constant); a value for determining a relationship between a thermocouple output voltage of each GT detector and a .gamma.-ray heating value (unit: W/g) per unit weight} is periodically measured. Then, the measured sensitivity value is checked, and when the measured sensitivity value is changed over a constant level, it is necessary to calculate a .gamma.-ray heating value of the GT detector from a thermocouple output voltage signal with the use of a new sensitivity coefficient corresponding to the changed sensitivity value. In this specification, the aforesaid processing, that is, a processing of measuring a sensitivity of each GT detector with the use of the heater, and in the case where a sensitivity change with respect to the measurement result exceeds a constant level, setting a new sensitivity coefficient corresponding to the changed sensitivity, is described as "sensitivity calibration processing". Moreover, when carrying out the aforesaid sensitivity calibration processing, a GT signal outputted from the GT detector is bypassed so as not to be used for power distribution measurement processing. In addition, the GT detector or GT assembly, which is not used for the aforesaid power distribution measurement processing because the sensitivity of the GT detector or GT assembly is being calibrated or the sensitivity of that shows a defective value so that the GT detector or GT assembly is out of order, is called as bypassed GT detector or bypassed GT assembly. By the way, the GT assembly is incorporated in the same in-core nuclear instrumentation tube integrally with the LPRM detector assembly which are thermal neutron detectors. A sensitivity with respect to a thermal neutron of the LPRM detector is determined depending upon a change by an in-core irradiation quantity of U235 and U234 coated onto an inner surface of a fission detector. If the sensitivity with respect to a thermal neutron of the LPRM detector gets to be a constant value or less, the in-core nuclear instrumentation assembly, that is, the in-core nuclear instrumentation tube including the LPRM detectors and the GT assembly having a sensitivity lowering to a constant value or less, is replaced together. Therefore, in actual use, an in-core mounted elapse time of the gamma-ray thermometer (GT) assembly differs for each in-core nuclear instrumentation assembly. An actual sensitivity of the GT detector in the output voltage sensitivity calibration processing by a heater incorporated into the GT assembly, is measured by the following equation (2) which will be described later, on the basis of an increase of the thermocouple voltage signal by a additional heating value of the GT detector. For this reason, in a thermal equilibrium state that additional heating by a built-in heater is sufficiently completed, an average value must be obtained from large number of time series data of one GT detector signal. Therefore, a time of approximately 30 to 60 seconds per GT assembly is required to collect the GT signal (output voltage signal). In an ABWR (Advanced Boiling Water Reactor) in a 135 ten-thousand kWe range, the reactor core is provided with 52 in-core nuclear instrumentation tubes each including the aforesaid GT assembly. Therefore, in the case of carrying out a calibration by the built-in heater of the GT assembly, if three circuits are prepared for each tube, about 9 to 20 minutes are required depending upon the number of power supply circuits of the heater and heating value measuring circuits of the heater. Moreover, in a core mounted lifetime (approximately 7 years) of the GT assembly, if a heater calibration of the GT assembly is carried out according to heating by the heater, there is the possibility that the heater is latently break down. Thus, by avoiding unnecessary calibration by heater heating, and shortening a time spent for the calibration by heater heating, it is desirable to reduce a time inoperable of measuring a power distribution of the whole core by the GT signal as much as possible. On the other hand, during a calibration of the GT assembly by heater heating and in-core power distribution measurement by the calibrated GT assembly, it is necessary that the core or in-core power distribution is a steady state for a predetermined time or more (approximately one hour when a gamma decay chain becomes a substantially equilibrium state) on the basis of the principle of measuring a gamma-ray heating value. In the BWR, the process control computer has a built-in three-dimensional simulator, and a core power distribution simulation is periodically or always carried out with the use of the parameter of core state data such as a reactor pressure, a core heat output, a core coolant flow rate, a control rod pattern or the like, and thus, it is confirmed that a fuel assembly satisfies a core operational thermal condition (limit). Before a time (within about one hour) relatively shorter than the point of time of a periodically core power distribution simulation, for example, in the case where the core power distribution varies by a change of the control rod pattern or a great change of the core coolant flow rate, the LPRM detector can instantaneously output neutron flux signal corresponding to a power distribution change. However, a signal (GT signal) of the GT detector becomes a precise signal level after a predetermined time, for example, one hour or more elapses because a delayed gamma-ray source slowly varies. Therefore, it is impossible to carry out the in-core power distribution adaptive correction processing or LPRM detector sensitivity and gain adjustment processing until the GT signal becomes the precise signal level. For this reason, the in-core power distribution adaptive correction processing or LPRM detector sensitivity and gain adjustment processing can not be periodically or always carried out. SUMMARY OF THE INVENTION The present invention is directed to overcome the foregoing problems. Accordingly, it is a first object of the present invention to provide an in-core fixed nuclear instrumentation system and a power distribution monitoring system, which accurately carry out a heater heating calibration control of a gamma-ray thermometer assembly (GT assembly) so as to improve a sensitivity calibration or gain calibration of a GT detector or an LPRM detector, making it possible to carry out a high accurately power distribution simulation. A second object of the present invention is to provide an in-core fixed nuclear instrumentation system and a power distribution monitoring system, which decrease a heater heating calibration time of a gamma-ray thermometer assembly so as to reduce a heat damage probability so as to be able to accurately adjust a sensitivity of an LPRM detector or gain thereof. Further, a third object of the present invention is to provide an in-core fixed nuclear instrumentation system and a power distribution monitoring system, which accurately perform a heater heating calibration control of gamma-ray thermometer assemblies on the basis of the difference between the respected in-core mounted times of the respected gamma-ray thermometer assemblies or the difference between the respected in-core irradiation burn-ups of the respected gamma-ray thermometer assemblies so as to prevent a wasteful sensitivity calibration of a part of the gamma-ray thermometer assemblies which is in a sensitivity stabilizing state, thereby reducing a load of an operator. Moreover, a fourth object of the present invention is to provide an in-core fixed nuclear instrumentation system and a power distribution monitoring system, which accurately carry out a heater heating calibration control of a gamma-ray thermometer assembly so as to improve, when a GT signal level of a GT detector of the gamma-ray thermometer assembly is in an equilibrium state, a sensitivity adjustment accuracy or gain adjustment accuracy of an LPRM detector and an in-core power distribution simulation accuracy. Still furthermore, a fifth object of the present invention is to provide an in-core fixed nuclear instrumentation system and a power distribution monitoring system, which accurately carry out a heater heating calibration control of a gamma-ray thermometer assembly whereby, even in a case where GT signal level of a GT detector of the gamma-ray thermometer is in a non-equilibrium state, to periodically or always carry out a sensitivity adjustment or gain adjustment of an LPRM detector, and further, accurately to carry out an in-core power distribution simulation with the use of a calibration value of an LPRM signal or an equilibrium predictive value of the GT detector. A sixth object of the present invention is to provide a power distribution monitoring system which captures a high accurate GT signal or LPRM signal in a power distribution simulating system so as to accurately carry out a power distribution simulation after a core state (operating mode) varies, thereby accurately monitoring the power distribution in the core according to the simulated power distribution. In order to achieve such objects, according to one aspect of the present invention, there is provided an in-core fixed nuclear instrumentation system for a reactor, comprising: a plurality of in-core nuclear instrumentation assemblies each having a nuclear instrumentation tube, a plurality of fixed neutron detectors housed in the nuclear instrumentation tube and adapted to detect neutron flux of a local power distribution of a power range in a core of the reactor and a gamma-ray thermometer assembly housed in the nuclear instrumentation tube, the gamma-ray thermometer assembly having a plurality of fixed .gamma.-ray heating detectors for detecting .gamma.-ray heating values and a heater built therein and adapted to calibrate the fixed .gamma.-ray heating detectors, the fixed .gamma.-ray heating detectors being arranged at least close to the fixed neutron detectors; means for processing a neutron flux detection signal based on the detected neutron flux by each of the fixed neutron detectors; means for processing a gamma-ray thermometer signal based on the detected .gamma.-ray heating value by each of the fixed .gamma.-ray heating detectors of each of the gamma-ray thermometer assemblies; means for electrically energizing the heater in each of the gamma-ray thermometer assemblies; means for storing a plurality of predetermined time intervals therein; and means for selecting one of the predetermined time intervals for specified .gamma.-ray thermometer assemblies respectively, wherein the energizing means is adapted to control an electrical energy supplied to the heater according to the selected one of the predetermined time intervals so as to heat the heater, thereby executing a heater calibration of output voltage sensitivities of the fixed .gamma.-ray heating detectors of the gamma-ray thermometer assembly. In preferred embodiment of this aspect, the energizing means has means for measuring increase sensitivities of the output voltage of the fixed .gamma.-ray heating detectors of the gamma-ray thermometer assemblies and current and voltage corresponding to the supplied electrical energy so as to execute the heater calibration according to the measured increase sensitivities and the measured currents and voltages, the increase sensitivities being caused to the heating of the heaters, and the selection means has means for storing at least one of an in-core mounted time of each of the gamma-ray thermometer assemblies and an in-core irradiation quantity of each of the fixed .gamma.-ray heating detectors thereof, which are calculated while the reactor operates, the in-core mounted time of each of the gamma-ray thermometer assemblies representing an operating time of the reactor after each of the gamma-ray thermometer assemblies is mounted in the core, and means for selecting one of the predetermined time intervals according to at least one of the in-core mounted time and the in-core irradiation quantity corresponding to the gamma-ray thermometer assembly. This aspect of the present invention has an arrangement that the energizing means is adapted to store time series data of the output voltage sensitivity of the fixed .gamma.-ray heating detector of each of the gamma-ray thermometer assemblies, estimate a change curve of each output voltage sensitivity thereof by sampling latest two and over time series data points from a present point of time by using the time series data, in a case where the change curve of the output voltage sensitivity exceeds a predetermined judgement value preset with respect to a predetermined future time, set another one of the predetermined time intervals prior to the selected time interval, the another one of the time intervals being shorter than the selected time interval or being a maximum time interval of the predetermined time intervals satisfying the predetermined judgement value, and execute the heater calibration of the output voltage sensitivity by heating of the heater at a resettled future time in accordance with the set another one of the time intervals. In preferred embodiment, this aspect further comprises: means for detecting a core state data representing in a state of the core, wherein the energizing means is adapted to detect a change of the core state according to the detected core state data, judge whether or not a predetermined time after detecting the change of the core state elapses and execute the heater calibration in a case where the predetermined time after detecting the change of the core state elapses. For achieving such objects, according to another aspect of the present invention, there is provided a power distribution monitoring system for a reactor, comprising: a plurality of in-core nuclear instrumentation assemblies each having a nuclear instrumentation tube, a plurality of fixed neutron detectors housed in the nuclear instrumentation tube and adapted to detect neutron flux of a local power distribution of a power range in a core of the reactor and a gamma-ray thermometer assembly housed in the nuclear instrumentation tube, the gamma-ray thermometer assembly having a plurality of fixed .gamma.-ray heating detectors for detecting .gamma.-ray heating values and a heater built therein and adapted to calibrate the fixed .gamma.-ray heating detectors, the fixed .gamma.-ray heating detectors being arranged at least close to the fixed neutron detectors; means for processing a neutron flux detection signal based on the detected neutron flux by each of the fixed neutron detectors; means for processing gamma-ray thermometer signals based on the detected .gamma.-ray heating values by the fixed .gamma.-ray heating detectors of each of the gamma-ray thermometer assemblies; means for electrically energizing the heater in each of the gamma-ray thermometer assemblies; means for detecting a core state data representing in a state of the core; means for detecting a change of the core state according to the detected core state data so as to judge whether or not a predetermined time after detecting the change of the core state elapses; and means for gathering at least one of the gamma-ray thermometer signal processed by the gamma-ray thermometer process means from a predetermined fixed .gamma.-ray heating detector and the gamma-ray heating value calculated with the gamma-ray thermometer signal therefrom. In preferred embodiment, the another aspect further comprises means for adjusting at least one of a sensitivity and a gain of the fixed neutron detector in a case where the predetermined time after detecting the change of the core state elapses by using the gamma-ray heating value simulated by the gamma-ray thermometer signal from the predetermined fixed .gamma.-ray heating detector, the predetermined fixed .gamma.-ray heating detector and the adjusted fixed neutron detector being housed in the identical in-core instrumentation tube, the predetermined fixed .gamma.-ray heating detector being located identically to a core axial direction of the adjusted fixed neutron detector. This another aspect of the present invention has an arrangement comprising means for simulating a power distribution in the core according to at least one of the gamma-ray thermometer signal and the neutron flux signal, wherein, in a case where a predetermined time after detecting the change of the core state does not elapse, the detecting means inputs the neutron flux signal to the simulating means in a place of the gamma-ray thermometer signal, and wherein the simulating means has a memory for storing adaptive correction quantities by simulating power distribution according to the gamma-ray thermometer signals at a latest point of time, executes a power distribution simulation corresponding to the core state at the present point of time based on the above predetermined adaptive correction quantities and the current core state, obtains a pseudo gamma-ray thermometer signal by correcting a difference in response between the change of the neutron flux signal predicted by a simulation from the latest point of time and a change of a gamma-ray thermometer signal predicted by a simulation while accounting a change of a control rod state and void fraction of fuel nodes around the neutron flux detectors until the present point of time from the latest point of time, simulates the power distribution at the present point of time while adapting the power distribution by interpolating and extrapolating correction ratios obtained by making a comparison between the pseudo gamma-ray thermometer signal and the simulated equilibrium value of the gamma-ray thermometer signal in an axial direction of the core so as to obtain an additional correction ratio of all axial nodes, evaluates a power distribution even in a case where a gamma-ray thermometer signal is in a non-equilibrium transient state, and executes a zero-clear process of the additional correction ratio when the gamma-ray thermometer signal is in an equilibrium state by simulating the power distribution so as to obtain the adaptive correction quantity according to the gamma-ray thermometer signal so as to store it therein. To achieve such objects, according to further aspect of the present invention, there is provided a power distribution monitoring system for a reactor, comprising: a plurality of in-core nuclear instrumentation assemblies each having a nuclear instrumentation tube, a plurality of fixed neutron detectors housed in the nuclear instrumentation tube and adapted to detect neutron flux of a local power distribution of a power range in a core of the reactor and a gamma-ray thermometer assembly housed in the nuclear instrumentation tube, the gamma-ray thermometer assembly having a plurality of fixed .gamma.-ray heating detectors for detecting .gamma.-ray heating values and a heater built therein and adapted to calibrate the fixed .gamma.-ray heating detectors, the fixed .gamma.-ray heating detectors being arranged at least close to the fixed neutron detectors; means for processing a neutron flux detection signal based on the detected neutron flux by each of the fixed neutron detectors; means for processing gamma-ray thermometer signals based on the detected .gamma.-ray heating values by the fixed .gamma.-ray heating detectors of each of the gamma-ray thermometer assemblies; means for electrically energizing the heater in each of the gamma-ray thermometer assemblies; means for detecting a core state data representing in a state of the core; means for detecting a change of the core state according to the detected core state data so as to judge whether or not a predetermined time after detecting the change of the core state elapses; in a case where predetermined time after detecting the change of the core state does not elapse, means for estimating equilibrium signal levels of the gamma-ray thermometer signals of the gamma-ray heating detectors after the predetermined time elapses; and means for executing a power distribution simulation corresponding to the core state at a present point of time while adapting the power distribution by interpolating and extrapolating correction ratios obtained by making a comparison between gamma-ray thermometer signal reading values simulated from the simulated power distribution and the evaluated equilibrium signal levels of the gamma-ray thermometer signals in an axial direction of the core, so as to obtain a correction ratio of all axial nodes, thereby evaluating a power distribution even in a case where gamma-ray thermometer signals are in a non-equilibrium transient state. In preferred embodiment of this further aspect, the estimation means is adapted to estimate an equilibrium signal level of the gamma-ray thermometer signal after a required time elapses by using time series gamma-ray thermometer signal readings, read a gamma-ray thermometer signal each new time while a prediction function execution instruction is inputted to the estimation means, cancel an oldest data of reading value of the gamma-ray thermometer signal so as to update estimated equilibrium reading value thereof by using a least square approximation, and transmit the updated estimated equilibrium value of the gamma-ray thermometer signal to the execution means, whereby the execution means executes the power distribution simulation by using the updated estimated equilibrium reading values of the gamma-ray thermometer signals. In preferred embodiment of this further aspect, the estimation means is adapted to gather a number of gamma-ray thermometer signals of time series, estimate an equilibrium signal level of the gamma-ray thermometer signal after a required time elapses, read the gamma-ray thermometer signal each new time, and cancel an oldest data of the reading value of the gamma-ray thermometer signal so as to update estimated equilibrium reading value thereof by using a least square approximation, and wherein the execution means corrects the simulated power distribution according to the estimated equilibrium gamma-ray thermometer signal reading values. |
abstract | In the case of monitoring a resolution of a scanning electron microscope, it is required to prepare a sample and to use a measuring algorithm so as to reduce the pattern dependency of an index value of resolution to be measured in order to measure a variation in the size of an electron beam with a high degree of accuracy. According to the present invention, there is used a sample having a sectional shape which is appropriate for monitoring the resolution, that is, the sample has a pattern with such a sectional shape that a side wall of the pattern is inclined so as to prevent an electron beam irradiated on the sample from impinging upon the side wall of the pattern. With this configuration, it is possible carry out such resolution monitor that does not depend upon a sectional shape of a pattern. |
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abstract | A method and apparatus for inspecting equipment using focal plane array imaging sensor data and dynamic sensor data. Methods involve capturing focal plane array imaging sensor data using a focal plane array imaging sensor such as an infrared camera or a visible camera, or acquiring imaging sensor data from an electronic data storage source, and involve capturing dynamic sensor data, such as vibration or ultrasonic data using a dynamic sensor such as an accelerometer or ultrasound system. Methods also provide for analyzing imaging and dynamic sensor data using such techniques as thermography and fast fourier transformation. Apparatuses include a portable instrument with sensor interfaces for collecting imaging sensor data and dynamic sensor data. A sensor suite is provided that includes vibration sensor, sonic sensors, ultrasonic sensors, oil sensors, flux sensors and current sensors. A base station is included to collect and analyze data from one or more portable instruments. |
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claims | 1. A method for producing a polymeric article containing radiopaque nano-materials comprising the steps of:mixing a radiopaque nano-material with a polymer to create a polymeric mixture, wherein said radiopaque nano-material is surface modified to enhance its dispersion within said polymeric mixture;heating said polymeric mixture until it assumes a liquid form; and,forming said polymeric mixture into an article. 2. The method of claim 1 wherein said polymeric mixture is mixed and heated in an extruder and then deposited on an endless conveyor. 3. The method of claim 1 wherein said liquid polymeric mixture is applied to a sheet of fabric or other pliable material before it solidifies into a film. 4. The method of claim 1 wherein said radiopaque nano-material is combined in said polymeric mixture with larger sized radiopaque materials. 5. The method of claim 1 wherein said radiopaque nano-material is selected from the group of nano-sized lead, tin, tungsten, barium, boron, tantalum, bismuth, silver, gold, platinum, aluminum, copper, depleted uranium, barium, cerium oxide (CeO2), yttrium oxide (Y2 O3), lanthanum oxide (La2O3) and neodymium oxide (Nd2 O3). 6. A method for producing a hazard protective article comprising the steps of:mixing a nano-material which provides radiation, chemical, biological, fire and/or projectile protection with a polymer to create a polymeric mixture, wherein said nano-material is surface modified to enhance its dispersion within said polymeric mixture;heating said polymeric mixture until it assumes a liquid form;applying said liquid polymeric mixture to a sheet of fabric or other pliable material; and,constructing an article from said fabric or other pliable material composite. 7. The method of claim 6 wherein said polymeric mixture is mixed and heated in an extruder before being applied to said sheet of fabric or other pliable material. 8. The method claim 6 wherein said polymeric mixture is sprayed in a liquid form onto said sheet of fabric or other pliable material. 9. A method for producing a radiation protective article of clothing comprising the steps of:mixing a radiation protective nano-material with a polymer to create a polymeric mixture, wherein said radiation protective nano-material is surface modified to enhance its dispersion within said polymeric mixture;heating said polymeric mixture until it assumes a liquid form;applying said liquid polymeric mixture to a sheet of fabric or other pliable material; and,constructing an article of clothing from said fabric or other pliable material composite. 10. The method of claim 9 wherein said radiation protective nano-material is selected from the group of nano-sized lead, tin, tungsten, barium, boron, tantalum, bismuth, silver, gold, platinum, aluminum, copper, depleted uranium, barium, cerium oxide (CeO2), yttrium oxide (Y2 O3), lanthanum oxide (La2O3) and neodymium oxide (Nd2 O3). 11. The method of claim 9 wherein said radiation protective nano-material is selected from the group of nano-clays, nano-spheres, nano-hemispheres and nano-parabolas. 12. The method of claim 1 wherein said surface modification is done through compatibilization. 13. The method claim 12 wherein said nano-material is a nanoclay and said compatibilization method is either onium ion modification or ion-dipole interaction. 14. The method of claim 6 wherein said surface modification is done through compatibilization. 15. The method claim 14 wherein said nano-material is a nanoclay and said compatibilization method is either onium ion modification or ion-dipole interaction. 16. The method of claim 9 wherein said surface modification is done through compatibilization. 17. The method claim 16 wherein said nano-material is a nanoclay and said compatibilization method is either onium ion modification or ion-dipole interaction. |
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claims | 1. A radiation image read-out method, comprising:using a stimulable phosphor sheet;projecting stimulating light onto the stimulable phosphor sheet from the protective layer side; anddetecting stimulated emission emitted from the stimulable phosphor layer upon exposure to the stimulating light by imaging the stimulated emission on a line sensor by an imaging lens from the protective layer side while moving the stimulable phosphor sheet relatively to the line sensor in a direction intersecting the direction in which the line sensor extends,wherein the stimulable phosphor sheet comprises a stimulable phosphor layer and a protective layer, the protective layer is formed of a rigid transparent material and the stimulable phosphor layer is in contact with the protective layer, andwherein the stimulable phosphor layer is in contact with the protective layer at a plurality of discontinuous contact areas. 2. A radiation image read-out apparatus, comprising:a stimulable phosphor sheet;a stimulating light projecting means which projects stimulating light onto the stimulable phosphor sheet from the protective layer side;a detecting means which detects stimulated emission emitted from the stimulable phosphor layer upon exposure to the stimulating light by imaging the stimulated emission on a line sensor by an imaging lens from the protective layer side; anda conveyor means which moves the stimulable phosphor sheet relatively to the detecting means in a direction intersecting the direction in which the line sensor extends,wherein the stimulable phosphor sheet comprises a stimulable phosphor layer and a protective layer, the protective layer is formed of a rigid transparent material and the stimulable phosphor layer is in contact with the protective layer, andwherein the stimulable phosphor layer is in contact with the protective layer at a plurality of discontinuous contact areas. 3. A radiation image read-out apparatus as defined in claim 2 in which the conveyor means conveys the stimulable phosphor sheet relatively to the detecting means so that the surface profile of the surface of the stimulable phosphor layer facing the protective layer is positioned within the range of the focal depth of the imaging lens. 4. A stimulable phosphor sheet comprising a stimulable phosphor layer and a protective layer, stimulating light being projected onto the stimulable phosphor layer from the protective layer side and stimulated emission emitted from the stimulable phosphor layer upon exposure to the stimulating light being detected by a line sensor through an imaging lens from the protective layer side, wherein the improvement comprises thatthe protective layer is formed of a rigid transparent material and the stimulable phosphor layer is in contact with the protective layer, andwherein the stimulable phosphor layer is in contact with the protective layer at a plurality of discontinuous contact areas. 5. A stimulable phosphor sheet as defined in claim 4 in which the protective layer is in the range of not smaller than 0.2 mm and not larger than 10 mm in thickness. 6. A stimulable phosphor sheet as defined in claim 4 comprising:an elastic member,wherein the elastic member presses the stimulable phosphor layer toward the protective layer from the side opposite to the protective layer so that the surface of the stimulable phosphor layer facing the protective layer is brought into contact with the protective layer. 7. A stimulable phosphor sheet as defined in claim 4 in which the surface of the stimulable phosphor layer facing the protective layer is bonded to the surface of the protective layer facing the stimulable phosphor layer by contact bonding under heat. 8. The stimulable phosphor sheet of claim 4 wherein the stimulable phosphor layer is bonded to the protective layer. 9. The stimulable phosphor sheet of claim 4, comprising:an adhesive tape,wherein the adhesive tape attaches the stimulable phosphor layer to the protective layer so that their respective surfaces are in contact with the other. 10. A stimulable phosphor sheet comprising a stimulable phosphor layer and a protective layer, stimulating light being projected onto the stimulable phosphor layer from the protective layer side and stimulated emission emitted from the stimulable phosphor layer upon exposure to the stimulating light being detected by a line sensor through an imaging lens from the protective layer side, wherein the improvement comprises thatthe protective layer is formed of a rigid transparent material and the stimulable phosphor layer is in contact with the protective layer, andwherein the surface of the stimulable phosphor layer facing the protective layer is within the range of ±100 μm in surface profile error. 11. A stimulable phosphor sheet comprising a stimulable phosphor layer and a protective layer, stimulating light being projected onto the stimulable phosphor layer from the protective layer side and stimulated emission emitted from the stimulable phosphor layer upon exposure to the stimulating light being detected by a line sensor through an imaging lens from the protective layer side, wherein the improvement comprises thatthe protective layer is formed of a rigid transparent material and the stimulable phosphor layer is in contact with the protective layer, andwherein the surface of the stimulable phosphor layer facing the protective layer be in the range of not smaller than 0.05 μm and not larger than 5 μm in center line surface roughness. 12. A stimulable phosphor sheet comprising a stimulable phosphor layer and a protective layer, stimulating light being projected onto the stimulable phosphor layer from the protective layer side and stimulated emission emitted from the stimulable phosphor layer upon exposure to the stimulating light being detected by a line sensor through an imaging lens from the protective layer side, wherein the improvement comprises thatthe protective layer is formed of a rigid transparent material and the stimulable phosphor layer is in contact with the protective layer,wherein the stimulable phosphor layer is in contact with the protective layer by way of filler with a void formed between the stimulable phosphor layer and the protective layer, andwherein the filler is positioned on a coating layer formed on the stimulable phosphor layer and/or the protective layer. 13. A stimulable phosphor sheet as defined in claim 12 in which the thickness of the coating layer is in the range of not smaller than 0.1 μm and not larger than 20 μm, and the particle diameter of the filler is in the range of not smaller than 0.2 μm and not larger than 50 μm. 14. A radiation image read-out method in which a stimulable phosphor sheet comprising a stimulable phosphor layer in contact with a protective layer which is rigid and transparent is used, stimulating light is projected onto the stimulable phosphor sheet from the protective layer side, stimulated emission emitted from the stimulable phosphor layer upon exposure to the stimulating light is detected by imaging the stimulated emission on a line sensor by an imaging lens from the protective layer side while moving the stimulable phosphor sheet relatively to the line sensor in a direction intersecting the direction in which the line sensor extends, and further comprising:providing the stimulable phosphor layer and the protective layer in direct physical contact with the stimulable phosphor layer at a plurality of locations,wherein the phosphor layer and the protective layer are separated by an air cushion at locations at which both layers are not in direct physical contact with each other. 15. A radiation image read-out apparatus comprising a stimulable phosphor sheet having a stimulable phosphor layer in contact with a protective layer which is rigid and transparent, a stimulating light projecting means which projects stimulating light onto the stimulable phosphor sheet from the protective layer side, a detecting means which detects stimulated emission emitted from the stimulable phosphor layer upon exposure to the stimulating light by imaging the stimulated emission on a line sensor by an imaging lens from the protective layer side, and a conveyor means which moves the stimulable phosphor sheet relatively to the detecting means in a direction intersecting the direction in which the line sensor extends,wherein the stimulable phosphor layer and the protective layer are in direct physical contact with the stimulable phosphor layer at a plurality of locations, andwherein the phosphor layer and the protective layer are separated by an air cushion at locations at which both layers are not in direct physical contact with each other. 16. A stimulable phosphor sheet comprising a stimulable phosphor layer and a protective layer, stimulating light being projected onto the stimulable phosphor layer from the protective layer side and stimulated emission emitted from the stimulable phosphor layer upon exposure to the stimulating light being detected by a line sensor through an imaging lens from the protective layer side, wherein the improvement comprises thatthe protective layer is formed of a rigid transparent material and the stimulable phosphor layer is in contact with the protective layer,wherein the stimulable phosphor layer and the protective layer are in direct physical contact with the stimulable phosphor layer at a plurality of locations, andwherein the phosphor layer and the protective layer are separated by an air cushion at locations at which both layers are not in direct physical contact with each other. |
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description | Under 35 U.S.C. §120, this continuation application claims priority to and benefits of U.S. patent application Ser. No. 13/949,018 (TI-72039), filed on Jul. 23, 2013, the entirety of which is incorporated herein by reference. This disclosure is generally directed to gas cells. More specifically, this disclosure is directed to a multiple-cavity vapor cell structure for micro-fabricated atomic clocks, magnetometers, and other devices. Various types of devices operate using radioactive gas or other gas within a gas cell. For example, micro-fabricated atomic clocks (MFACs) and micro-fabricated atomic magnetometers (MFAMs) often include a cavity containing a metal vapor and a buffer gas. In some devices, the metal vapor and the buffer gas are created by dissociating cesium azide (CsN3) into cesium vapor and nitrogen gas (N2). This disclosure provides a multiple-cavity vapor cell structure for micro-fabricated atomic clocks, magnetometers, and other devices. In a first example, an apparatus includes a vapor cell having multiple cavities fluidly connected by one or more channels. At least one of the cavities is configured to receive a first material able to dissociate into one or more gases that are contained within the vapor cell. At least one of the cavities is configured to receive a second material able to absorb at least a portion of the one or more gases. In a second example, a system includes a vapor cell and an illumination source. The vapor cell includes multiple cavities fluidly connected by one or more channels. At least one of the cavities is configured to receive a first material able to dissociate into one or more gases that are contained within the vapor cell. At least one of the cavities is configured to receive a second material able to absorb at least a portion of the one or more gases. The illumination source is configured to direct radiation through at least one of the cavities. In a third example, a method includes, in a vapor cell having multiple cavities fluidly connected by one or more channels, placing a first material into at least one of the cavities and placing a second material into at least one of the cavities. The method also includes dissociating at least a portion of the first material into one or more gases that are contained within the vapor cell. The method further includes absorbing at least a portion of the one or more gases using the second material. Other technical features may be readily apparent to one skilled in the art from the following figures, descriptions, and claims. FIGS. 1 through 9, discussed below, and the various examples used to describe the principles of the present invention in this patent document are by way of illustration only and should not be construed in any way to limit the scope of the invention. Those skilled in the art will understand that the principles of the present invention may be implemented in any suitable manner and in any type of suitably arranged device or system. FIGS. 1 through 6 illustrate an example multiple-cavity vapor cell structure and fabrication technique in accordance with this disclosure. The multiple-cavity vapor cell structure can be used, for example, to receive an alkali-based material (such as cesium azide) and to allow dissociation of the alkali-based material into a metal vapor and a buffer gas (such as cesium vapor and nitrogen gas). However, this represents one example use of the multiple-cavity vapor cell structure. The vapor cell structure described here could be used in any other suitable manner. As shown in FIGS. 1 and 2, the multiple-cavity vapor cell structure includes a bottom wafer 102 and a middle wafer 104. The bottom wafer 102 generally represents a structure on which other components of the vapor cell structure can be placed. The bottom wafer 102 is also substantially optically transparent to radiation passing through the vapor cell structure during operation of a device, such as a micro-fabricated atomic clock, magnetometer, or other devices. The bottom wafer 102 can be formed from any suitable material(s) and in any suitable manner. The bottom wafer 102 could, for instance, be formed from glass, such as PYREX or BOROFLOAT glass. The middle wafer 104 is secured to the bottom wafer 102, such as through bonding. The middle wafer 104 includes multiple cavities 106-110 through the middle wafer 104. Each cavity 106-110 could serve a different purpose in the vapor cell structure. For example, the cavity 106 can receive a material to be dissociated, such as cesium azide (CsN3) or other alkali-based material. The cavity 106 can be referred to as a “reservoir cavity.” The cavity 108 can receive gas from the cavity 106, such as a metal vapor and a buffer gas. Laser illumination or other illumination could pass through the cavity 108 during operation of a device, such as a micro-fabricated atomic clock, magnetometer, or other device. The cavity 108 can be referred to as an “interrogation cavity.” The cavity 110 can receive at least one getter material, which can be used to absorb gas or other material(s) from the other cavities 106-108. The cavity 110 can be referred to as a “getter cavity.” One or more channels fluidly connect each adjacent pair of cavities in the vapor cell structure. For example, at least one channel 112 connects the reservoir cavity 106 and the interrogation cavity 108, and at least one channel 114 connects the interrogation cavity 108 and the getter cavity 110. Each channel 112-114 represents any suitable passageway through which gas or other material(s) can flow. The middle wafer 104 could be formed from any suitable material(s) and in any suitable manner. For example, the middle wafer 104 could represent a silicon wafer. The cavities 106-110 and the channels 112-114 could be formed in the silicon wafer using one or more wet or dry etches. Each cavity 106-110 and channel 112-114 could have any suitable size, shape, and dimensions. Also, the relative sizes of the cavities 106-110 and channels 112-114 shown in FIGS. 1 through 6 are for illustration only, and each cavity 106-110 or channel 112-114 could have a different size relative to the other cavities or channels. Further, the relative depth of each channel 112-114 compared to the depth(s) of the cavities 106-110 is for illustration only, and each cavity 106-110 and channel 112-114 could have any other suitable depth. In addition, while each cavity 106-110 is shown as being formed completely through the wafer 104, each cavity 106-110 could be formed partially through the wafer 104. As shown in FIGS. 3 and 4, a material 116 is deposited into the reservoir cavity 106. The material 116 could represent any suitable material or combination of materials used to create one or more gases for the vapor cell structure. In some embodiments, the material 116 represents cesium azide, although any other suitable material(s) could be used in the vapor cell structure. The material 116 could be deposited into the reservoir cavity 106 in any suitable manner. As shown in FIGS. 5 and 6, a top wafer 118 having a getter material 120 is secured to the middle wafer 104, such as through bonding. The top wafer 118 generally represents a structure that caps the cavities 106-110 and channels 112-114 of the middle wafer 104, thereby helping to seal material (such as gas) into the vapor cell structure. The top wafer 118 is also substantially optically transparent to radiation passing through the vapor cell structure during operation of a device, such as a micro-fabricated atomic clock, magnetometer, or other device. The top wafer 118 can be formed from any suitable material(s) and in any suitable manner. The top wafer 118 could, for instance, be formed from borosilicate glass, such as PYREX or BOROFLOAT glass. In this example, the getter material 120 is located on the top wafer 118 so that it is positioned within the getter cavity 110 when the wafers 104, 118 are secured together. The getter material 120 can be used to absorb material within the cavities 106-110. For example, the getter material 120 could absorb at least a portion of the buffer gas released when the material 116 is dissociated. The getter material 120 can also be used to absorb any undesirable materials that may be present in the cavities 106-110, such as water vapor. The getter material 120 represents any suitable material for absorbing or otherwise removing gaseous or other material from the vapor cell structure, such as a zirconium-based alloy. In particular embodiments, the getter material 120 is used to absorb nitrogen gas (N2) from the vapor cell structure. The getter material 120 could have any suitable size, shape, and dimensions. The getter material 120 could also be formed in any suitable manner. For example, the getter material 120 could be deposited as a film on the top wafer 118 and then patterned and etched into the appropriate form. However, the getter material 120 could be deposited directly into the getter cavity 110. The getter material 120 can also be activated in any suitable manner. In particular embodiments, the getter material 120 is activated at temperatures associated with bonding or otherwise securing the middle wafer 104 to the top wafer 118 (such as around 300° C. or more). Note, however, that the getter material could be placed on the bottom wafer 102 as well. At this point, various additional processing steps could occur to make the vapor cavity structure ready for use. For example, the vapor cavity structure generally or the reservoir cavity 106 in particular can be exposed to ultraviolet (UV) radiation to dissociate at least part of the material 116 in the reservoir cavity 106. In some embodiments, UV radiation can be used to dissociate cesium azide into cesium vapor and nitrogen gas (N2). Note, however, that other mechanisms could be used to initiate the dissociation, such as thermal dissociation. The dissociation of the material 116 creates gas inside the reservoir cavity 106, which can flow into the interrogation cavity 108 through the channel 112 and into the getter cavity 110 through the channel 114. Some of the gas can be absorbed by the activated getter material 120 in the getter cavity 110. The activated getter material 120 can absorb any suitable amount of gas depending on, for example, the area and thickness of the getter material 120. In conventional devices, material is often dissociated in a single cavity, and the resulting gas is kept in the same cavity. Radiation can be passed through the gas in that single cavity during operation of a device, but residue from the original material may still exist in that single cavity. This residue can interfere with the optical properties of the cavity and lead to device failure. In accordance with this disclosure, the material 116 can be placed in one cavity 106 and dissociated, and the resulting gas can be used in a different cavity 108 during device operation. Even if residue exists in the cavity 106, it may not interfere with the optical properties in the cavity 108. Also, the amount or composition of metal vapor and buffer gas in the interrogation cavity 108 can be precisely controlled. This is because the dissociation of the material 116 into the metal vapor and the buffer gas typically occurs at a known ratio and is controllable, such as based on the UV dosage selected. As a particular example, cesium azide dissociates into a known ratio of cesium vapor and nitrogen gas, and the amount of each can be controlled by controlling the UV exposure. Moreover, the getter material 120 can absorb a known quantity of gas, such as nitrogen gas, based on various factors like its dimensions. By controlling the UV dosage and the dimensions of the getter material 120, a manufacturer can precisely control the gas content created in the interrogation cavity 108. This allows the optical properties of a vapor cell structure to be controlled, which allows accurate, reliable, and reproducible operation to be obtained in various vapor cell structures manufactured in this way. Moreover, different applications (such as micro-fabricated atomic clocks and micro-fabricated atomic magnetometers) often require different amounts of metal vapor and buffer gas in the interrogation cavity 108. This approach allows a vapor cell structure to be easily customized for a specific application. Overall, this approach provides a reliable technique for the manufacture of vapor cells for use in various applications with precise optical properties. This approach also increases the reliability and performance of the vapor cells since, for example, the getter material 120 absorbs unwanted gases and other materials present in the vapor cells. The vapor cell structure could include any number of additional features depending on the implementation. For example, a portion 122 of the top wafer 118 (shown in outline in FIG. 6) could be thinner than the remainder of the top wafer 118. This may help to facilitate easier UV irradiation of the material 116. Note that any wafer 102, 104, 118 in the vapor cell structure could have a non-uniform thickness at any desired area(s) of the wafer(s). Also note that the portion 122 of the top wafer 118 could have any suitable size, shape, and dimensions and could be larger or smaller than the reservoir cavity 106. The portion 122 of the top wafer 118 could be thinned in any suitable manner, such as with a wet isotropic etch. Although FIGS. 1 through 6 illustrate one example of a multiple-cavity vapor cell structure and fabrication technique, various changes may be made to FIGS. 1 through 6. For example, the vapor cell structure need not include three cavities and could include two cavities or more than three cavities. As a particular example, the cavities 106 and 110 could be combined into a single cavity that receives both the material 116 and the getter material 120. Also, the fabrication technique shown here is for illustration only, and other or additional operations could be used to fabricate the vapor cell structure. Further, the cavities 106-110 need not be arranged linearly, and the channels 112-114 need not be straight. Moreover, the interrogation cavity 108 need not be located between the reservoir cavity 106 and the getter cavity 110. Any arrangement of cavities connected by channels could be used, including non-linear and multi-level arrangements. In addition, the vapor cell structure could be used with any other material(s) 116 and getter material(s) 120 and is not limited to alkali-based materials or metal vapors and buffer gases. FIGS. 7 and 8 illustrate example devices containing at least one multiple-cavity vapor cell structure in accordance with this disclosure. As shown in FIG. 7, a device 700 represents a micro-fabricated atomic clock or other atomic devices. The device 700 here includes one or more illumination sources 702 and a vapor cell 704. Each illumination source 702 includes any suitable structure for generating radiation, which is directed through the vapor cell 704. Each illumination source 702 could, for example, include a laser or lamp. The vapor cell 704 represents a multi-cavity vapor cell structure, such as the vapor cell structure described above. The radiation from the illumination source(s) 702 passes through the interrogation cavity 108 of the vapor cell 704 and interacts with the metal vapor. The radiation can also interact with one or more photodetectors that measure the radiation passing through the interrogation cavity 108. For example, photodetectors can measure radiation from one or more lasers or lamps. Signals from the photodetectors are provided to clock generation circuitry 706, which uses the signals to generate a clock signal. When the metal vapor is, for example, rubidium 87 or cesium 133, the signal generated by the clock generation circuitry 706 could represent a highly-accurate clock. The signals from the photodetectors are also provided to a controller 708, which controls operation of the illumination source(s) 702. The controller 708 helps to ensure closed-loop stabilization of the atomic clock. As shown in FIG. 8, a device 800 represents a micro-fabricated atomic magnetometer or other atomic magnetometer. The device 800 here includes one or more illumination sources 802 and a vapor cell 804. Each illumination source 802 includes any suitable structure for generating radiation, which is directed through the vapor cell 804. Each illumination source 802 could, for example, include a laser or lamp. The vapor cell 804 represents a multi-cavity vapor cell structure, such as the vapor cell structure described above. The radiation from the illumination source(s) 802 can pass through the interrogation cavity 108 of the vapor cell 804 and interact with the metal vapor. The radiation can also interact with one or more photodetectors that measure the radiation passing through the interrogation cavity 108. For example, photodetector(s) can measure radiation from one or more lasers or lamps. Signals from the photodetector(s) are provided to a magnetic field calculator 806, which uses the signals to measure a magnetic field passing through the interrogation cavity 108. The magnetic field calculator 806 here is capable of measuring extremely small magnetic fields. The signals from the photodetector(s) can also be provided to a controller 808, which controls operation of the illumination source(s) 802. Although FIGS. 7 and 8 illustrate examples of devices 700 and 800 containing at least one multiple-cavity vapor cell structure, various changes may be made to FIGS. 7 and 8. For example, the devices 700 and 800 shown in FIGS. 7 and 8 have been simplified in order to illustrate example uses of the vapor cell structure described above. Atomic clocks and atomic magnetometers can have various other designs of varying complexity with one or multiple vapor cell structures. FIG. 9 illustrates an example method 900 for forming a multiple-cavity vapor cell structure in accordance with this disclosure. As shown in FIG. 9, multiple cavities are formed in a middle wafer of a vapor cell structure at step 902. This could include, for example, forming cavities 106-110 in a silicon wafer or other middle wafer 104. Any suitable technique could be used to form the cavities, such as a wet or dry etch. Channels are formed in the middle wafer of the vapor cell structure at step 904. This could include, for example, forming channels 112-114 in the silicon wafer or other middle wafer 104. Any suitable technique could be used to form the channels, such as a wet or dry etch. The formation of the cavities and channels could also overlap, such as when the same etch is used to form both the cavities 106-110 and the channels 112-114. The middle wafer is secured to a lower wafer at step 906. This could include, for example, bonding the middle wafer 104 to the bottom wafer 102. If the cavities 106-110 are formed completely through the middle wafer 104, securing the middle wafer 104 to the bottom wafer 102 can seal the lower openings of the cavities 106-110. A first material to be dissociated is deposited in one of the cavities at step 908. This could include, for example, depositing the material 116 into the reservoir cavity 106. Any suitable deposition technique could be used to deposit any suitable material(s) 116, such as an alkali-based material. A second material is formed on a top wafer of the vapor cell structure at step 910. This could include, for example, forming a film of getter material 120 on the top wafer 118 and then patterning and etching the film. The getter material 120 can be positioned on the top wafer 118 in a location suitable for insertion into the getter cavity 110 of the middle wafer 104. The top wafer is secured to the middle wafer at step 912. This could include, for example, bonding the top wafer 118 to the middle wafer 104. Securing the top wafer 118 to the middle wafer 104 can seal the upper openings of the cavities 106-110. At this point, the second material is located within the getter cavity 110. The second material is activated at step 914. This could include, for example, applying heat of at least about 300° C. to the getter material 120. In some embodiments, the heat could be generated during step 912 when the top wafer 118 is secured to the middle wafer 104. The first material is dissociated to create metal vapor and buffer gas at step 916. This could include, for example, applying UV radiation to the material 116 through the bottom wafer 102 or the top wafer 118. This could also include converting at least a portion of the material 116 into the metal vapor and buffer gas. Note, however, that other dissociation techniques could also be used. At least a portion of the buffer gas, undesirable material(s), or other material(s) in the vapor cell structure is/are absorbed using the second material at step 918. This could include, for example, absorbing at least some of the buffer gas, water vapor, or other material(s) using the getter material 120. In this way, the amounts of metal vapor and buffer gas in the interrogation cavity 108 of the vapor cell structure can be precisely controlled. As a result, the vapor cell structure can be manufactured with specific optical properties suitable for a particular application. Moreover, multiple vapor cell structures can be fabricated using this technique, and each vapor cell structure can have optical properties suitable for each particular application. Although FIG. 9 illustrates one example of a method 900 for forming a multiple-cavity vapor cell structure, various changes may be made to FIG. 9. For example, as noted above, various modifications can be made to the fabrication process. Also, while shown as a series of steps, various steps in FIG. 9 could overlap, occur in parallel, or occur in a different order. It may be advantageous to set forth definitions of certain words and phrases used throughout this patent document. The terms “top,” “middle,” and “bottom” refer to structures in relative positions in the figures and do not impart structural limitations on how a device is manufactured or used. The term “secured” and its derivatives mean to be attached, either directly or indirectly via another structure. The terms “include” and “comprise,” as well as derivatives thereof, mean inclusion without limitation. The term “or” is inclusive, meaning and/or. The phrase “associated with,” as well as derivatives thereof, may mean to include, be included within, interconnect with, contain, be contained within, connect to or with, couple to or with, be communicable with, cooperate with, interleave, juxtapose, be proximate to, be bound to or with, have, have a property of, have a relationship to or with, or the like. The phrase “at least one of,” when used with a list of items, means that different combinations of one or more of the listed items may be used, and only one item in the list may be needed. For example, “at least one of: A, B, and C” includes any of the following combinations: A, B, C, A and B, A and C, B and C, and A and B and C. While this disclosure has described certain embodiments and generally associated methods, alterations and permutations of these embodiments and methods will be apparent to those skilled in the art. Accordingly, the above description of example embodiments does not define or constrain this disclosure. Other changes, substitutions, and alterations are also possible without departing from the spirit and scope of this disclosure, as defined by the following claims. |
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abstract | A feeding system for an absorber liquid containing a neutron poison, in particular for a quick shut-down of a nuclear reactor, has a storage container for the absorber liquid and is configured for high operational reliability with simple construction. In particular, a chemical decomposition of the absorber liquid or corrosion of the container wall of the storage container is to be excluded. For this purpose, the storage container is connected to a pressure container via an overflow line, wherein the pressure container is filled with a motive fluid. |
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055090419 | abstract | An x-ray lithography method for irradiating an object (14) to form a pattern thereon uses an x-ray mask (10) having a membrane (18). The membrane (18) has an open membrane surface (26), and x-ray radiation (16) is passed through the open membrane surface (26) to irradiate the object (14). During this irradiation, the open membrane surface (26) is substantially uniformly exposed to the x-ray radiation (16) so that stress-induced distortion of the membrane (18) is reduced. |
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063303019 | claims | 1. An x-ray analysis system comprising: a focusing optic for focusing an x-ray beam to a focal point; a first slit optically coupled to said focusing optic; a second slit optically coupled to said first slit; and an x-ray detector, wherein said focal point is located between said second slit and said x-ray detector. conditioning an x-ray beam with a lens directing said x-ray beam through a first aperture; directing said x-ray beam through a second aperture; and focusing said x-ray beam with said lens at a point after it exits said second aperture. a plate; an opening formed in said plate, said opening having an outline which converges to a vertex; an aperture formed in said plate a known distance from said vertex; an x-ray detector used as feedback in orienting the x-ray beam, positioning an x-ray beam to travel through an opening formed in a plate, said opening having an outline which converges to a vertex; observing said x-ray beam with an x-ray detector providing feedback to determine if said x-ray beam is traveling through said opening; indexing said x-ray beam in relative fashion with respect to said opening until said x-ray beam reaches said vertex of said opening; and displacing said x-ray beam a known distance to the aperture. a focusing optic for focusing an x-ray beam to a focal point; a first slit optically coupled to said focusing optic; a second slit optically coupled to said first slit to form and define said x-ray beam in conjunction with said first slit; a third slit optically coupled to said second slit to block scattering from said second slit; a sample housing for holding a sample to be illuminated by said x-ray beam; and an x-ray detector for detecting a scattering pattern created by illuminating said sample, wherein said focal point is located between said second slit and said x-ray detector. 2. The x-ray analysis system of claim 1 further comprising a third guard slit optically coupled to said second slit to block parasitic scattering from said second slit. 3. The x-ray analysis system of claim 1, wherein said focusing optic is a Bragg reflector. 4. The x-ray analysis system of claim 3, wherein said Bragg reflector is a multilayer. 5. The x-ray analysis system of claim 4, wherein said Bragg reflector is depth graded. 6. The x-ray analysis system of claim 4, wherein said Bragg reflector is laterally graded. 7. The x-ray analysis system of claim 1, wherein said focusing optic is a total reflection mirror. 8. The x-ray analysis system of claim 1, wherein said focusing optic has an elliptical surface. 9. The x-ray analysis system of claim 1, wherein said focusing optic is a Kirkpatrick-Baez side-by-side optic. 10. The x-ray analysis system of claim 2, wherein said first, second and third slits are pinholes. 11. A method for reducing diffraction noise in an x-ray analysis system comprising: 12. The method of claim 11 further comprising the step of directing said x-ray beam through a sample structure. 13. The method of claim 12 further comprising the step of detecting said x-ray beam after it exits said sample structure. 14. An apparatus for shaping an x-ray beam comprising: 15. The apparatus of claim 14, wherein said opening has a triangular shaped end portion which converges to said vertex. 16. The apparatus of claim 14 further comprising a rotating aperture plate having a plurality of apertures. 17. The apparatus of claim 16, wherein each of said apertures is a different size. 18. A method of directing an x-ray beam through an aperture comprising: 19. An x-ray analysis system comprising: 20. The x-ray analysis system of claim 18, wherein said focusing optic is a Bragg reflector. |
summary | ||
abstract | A single-use neutron generator includes a power supply. The single-use neutron generator includes a fuel source configured to provide neutron-producing fuel. The single-use neutron generator includes a plasma confinement device coupled to the power supply and the fuel source and configured to generate a plasma pinch of the neutron-producing fuel. At least one component of the single-use neutron generator is configured for single use. |
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claims | 1. A pressurized water nuclear reactor comprising:a pressure vessel;an upper removable head for sealably engaging an upper opening in the pressure vessel;a core having an axial dimension supported within the pressure vessel;a plurality of nuclear fuel assemblies supported within the core, at least some of the fuel assemblies having at least one instrumentation thimble extending axially therethrough;an upper internals assembly supported above the core and having axially extending instrumentation guide paths supported therethrough with each of the instrumentation thimbles that are configured to receive instrumentation through the upper internals assembly being aligned with one of the instrumentation guide paths, the upper internals assembly including an instrumentation grid assembly plate supported above the instrumentation guide paths and axially movable relative to a lower portion of the upper internals; andat least one in-core instrumentation thimble assembly extending through a corresponding one of the instrumentation guide paths into an instrumentation thimble, the in-core instrumentation thimble assembly having an upper portion connected to the instrumentation grid assembly plate and retractable with raising of the instrumentation grid assembly plate to move a portion of the in-core instrumentation thimble assemblies within the instrumentation thimbles, into the upper internals assembly, the in-core instrumentation thimble assembly including a lower section comprising a sensor region and an upper section through which signal cabling is routed, both the lower section and the upper section being enclosed within an outer sheath that encloses the corresponding instrument guide path, the outer sheath having an upper portion connected to the instrumentation grid assembly plate with the signal cabling extending at least partially through the instrumentation guide path, through and around an outside of the outer sheath, making a plurality of revolutions around the outer sheath prior to extending through a passage from the interior of the reactor vessel to an exterior thereof, with the number of revolutions sufficient to stretch along an extended length of the outer sheath as the outer sheath is extended with a raising of the instrumentation grid assembly plate to remove the in-core instrumentation thimble assembly from the corresponding instrument thimble, without disconnecting the signal cabling that was housed within the pressure vessel and upper removable head. 2. The pressurized water nuclear reactor of claim 1 wherein the signal cabling is coiled around the outside of an upper portion of the outer sheath. 3. The pressurized water nuclear reactor of claim 2 wherein the coil is in the form of a spiral spring. 4. The pressurized water nuclear reactor of claim 1 wherein the passage from the interior to the exterior of the pressure vessel is an outwardly extending flange on the upper internals assembly. 5. The pressurized water nuclear reactor of claim 1 wherein the instrumentation grid assembly plate is configured to move axially from a lower position to an upper position and the instrumentation guide paths are formed from a tubular housing that substantially extends up to the lower position. 6. The pressurized water nuclear reactor of claim 5 wherein the instrumentation grid assembly plate in the upper position is spaced above the tubular housing. 7. The pressurized water nuclear reactor of claim 6 wherein an upper portion of the tubular housing is configured as a telescoping tube with an upper portion of the telescoping tube connected to the instrumentation grid assembly plate. 8. The pressurized water nuclear reactor of claim 1 wherein the signal cabling exits the outer sheath below an upper end of the outer sheath. |
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abstract | The invention refers to an arrangement for positioning substrates, in particular for positioning wafers, within a device that is provided for exposure of the substrates and/or for measurement on the substrates by means of radiation under high-vacuum conditions. The following are provided according to the present invention: a retention system (4), displaceable on a linear guidance system (3), for receiving the substrate, the guidance direction of the linear guidance system (3) being oriented parallel or substantially parallel to the Y coordinate of an X, Y, Z spatial coordinate system; drives for limited modification of the inclination of the guidance direction relative to the Y coordinate; drives for limited rotation of the linear guidance system (3), including the retention system (4), about the guidance direction; and drives for parallel displacement of the linear guidance system (3), including the retention system (4), in the direction of the X coordinate, the Y coordinate, and/or the Z coordinate. |
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053274695 | abstract | An arrangement for the underwater storage of environmentally hazardous waste, particularly radioactive or chemical waste, includes at least one secondary capsule (1) in the form of a cylindrical concrete body. The concrete body has a central, axially extending storage cavity (3). The cavity has the form of a shaft which is open at one end thereof and into which a waste-containing primary capsule (2) can be inserted, whereafter the open end of the shaft or cavity is sealed. Arranged in spaced relationship around the circumference of the concrete body (1) are a number of ballast chambers (4) which can be filled with water to varying degrees and the total, combined volume of the chambers is such as to enable the concrete body to be brought to a buoyant state, by emptying the chambers. A plurality of such secondary capsules (1) enclosing waste-containing primary capsules (2) can be stored on the sea bed in an annular concrete structure (6) which is provided with a large number of circumferentially distributed and vertically extending cylindrical compartments (9) each capable of accommodating a secondary capsule (1). The annular concrete structure (6) resting on the sea bed is also provided with a large number of ballast chambers which can be filled with water to varying degrees and which have a total, combined volume such as to enable the annular concrete structure to be brought to a buoyant state by emptying the ballast chambers. |
054535620 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS Many types of contaminated inert materials, such as soil, sand, sludge and hazardous wastes such as contaminated pond sludges, filter cakes, etc., can successfully be treated in accordance with the present invention. The contaminated materials to be treated may collectively be referred to hereafter as "feed." The term "solids" and "feed" are used interchangeably and refer to any pumpable or non-pumpable contaminated materials comprising at least 30 wt % solids, and more preferably at least 50 wt % solids. The removal of contaminants by thermal evolution of vapor may be referred to as "thermal desorption," "distillation," "vaporization" or "drying." Contaminated feeds are obtained from a variety of sources including soil contaminated by chemical spills and industrial discharges, solids from clarifiers and thickeners, sludges from waste holding tanks and treatment ponds, and solids from filtration processes. Mixed wastes are common at weapon research and production locations. Mixed wastes can be generally characterized as solid materials contaminated with hazardous organic, inorganic and radioactive compounds. Many such mixed wastes are contaminated with radionuclides and volatile and semi-volatile chemically oxidizable pollutants comprising both organic and inorganic compounds. In addition, the mixed wastes may contain non-volatile pollutants. The volatile chemically oxidizable pollutants include compounds, such as alcohols, ketones, esters, aromatic hydrocarbons (benzene, toluene, ethylbenzene, etc.), chlorinated hydrocarbons (TCE, TCA, PCE, carbon tetrachloride, etc.), ammoniacal compounds, cyanide and sulfur-containing compounds, and a variety of other chemicals that are known pollutants. The semi-volatile and non-volatile pollutants typically comprise organics, such as petroleum hydrocarbons, oils, polychlorinated biphenyls (PCBs), and minerals, which are broadly defined as any element, inorganic compound or mixture occurring or originating in the earths'crust and atmosphere, including all metals and nonmetals, their compounds and ores. Included in the solids can be compounds of heavy metals such as barium copper, beryllium, aluminum, nickel, zinc, cadmium, mercury, arsenic and lead. Radioactive contaminants include such elements as uranium, plutonium, cesium, thorium, radium, americium, neptunium, strontium and cobalt. While it would not be possible to list every contaminant to which the presently claimed thermal separation process may be applied, the inventive treatment process should be effective in separating volatile and semi-volatile chemical contaminants whose vapor pressures (at 5.degree. C.) range from 0.000001 to over 300 millimeters Hg. The process is capable of treating feeds with contamination levels ranging from 1 ppm to 20 weight percent of the feed charge with a more preferred contamination level of less than 5 weight percent. The mechanism by which complex materials are dried so that substantially complete removal of contaminants from inert materials occurs is complex and not completely understood. The structure of the solids in the feed, the type of liquid contaminants and other liquids in the feed, the concentration of liquids, and the saturation of the gas phase determine the mechanism by which internal liquid flow and vaporization may occur. Fluid flow mechanisms can include (1) diffusion, (2) capillary flow, (3) flow caused by shrinkage and pressure gradients, (4) flow caused by gravity and (5) flow caused by vaporization-condensation sequence. Drying of feeds wherein the solids are of a complex structure and texture does not occur as a single continuous process but involves a number of distinct phases. A first phase in drying contaminated inert materials involves evaporation of liquids, which may be contaminants, water, or other liquids, from the saturated surface on the solid. This is followed in turn by a period of evaporation from a saturated surface of gradually decreasing area and, finally, when the surface of the solids in the feed is no longer saturated, to a period of evaporation from the interior of the solids. In the present batch process, the drying rate accordingly varies with temperature, vacuum level, time, solids composition, and moisture content. In a plot comparing vapor evolution versus time, distinct phases may be recognized. There is usually a first phase of gradually increasing evolution of vapors as the feed warms up. A second phase, known as the constant-rate phase, corresponds to the period in which a constant amount of vapor is evolved. The constant-rate phase continues until a point at which the rate of drying begins to fall, known as the point at which the "critical-moisture content" point is reached. After reaching the critical-moisture content point, the next phase is called the falling-rate phase. This phase is typified by a continuously changing rate throughout the remainder of the drying cycle, corresponding to the decrease in saturated surface area. A next point in the curve occurs when all of the exposed surfaces become completely unsaturated. This marks the start of the portion of the drying cycle during which the rate of internal moisture movement controls the drying rate. Generally, the drying rate depends on factors affecting the diffusion of moisture away from the evaporating surface and those affecting the rate of internal moisture movement. Moisture which is held in the interstices of solids, or held as liquid on the surface, or is held as free moisture in cell cavities moves by gravity and capillary flow, provided that passageways for continuous flow are present. Moisture may move by vapor diffusion through the feed, provided that a temperature gradient is established by heating, thus creating a vapor-pressure gradient. Vaporization and vapor diffusion may occur in any feed in which heating takes place from one direction, drying from the other, and in which liquid is isolated between or inside granules of solid. In the terminal phase, the drying rate is governed by the rate of internal moisture movement; the influence of external variables diminishes. This period usually predominates in determining the overall drying time to lower moisture content. The presence of a vacuum during the heating and agitation of the contaminated soil and/or sludge material influences the efficiency and rate of evaporation of volatile and semi-volatile chemical compounds. In addition to regulating a vacuum, the presence of a sweep of inert gas, such as nitrogen, aids in removing the volatile chemical compounds from the contaminated soil. In this way, the concentration of evolved vapors in the gas phase around the drying solids is lowered, and it becomes easier for the heated liquids to pass from the liquid phase into the vapor phase. It is also the experience of the inventors that the presence of a small amount of water in the feed improves the effectiveness of the overall decontamination process. It is believed that as water in the interstitial spaces in the inert materials vaporizes and goes into the vapor phase, it carries contaminants along with it or otherwise facilitates the vaporization of the contaminants i.e., by conditioning the gas phase to lower the vapor pressure at which the contaminants will pass into the vapor phase. Even though the largest portion of water present in the feed vaporizes at around the boiling point of water, some water nevertheless goes into the vapor phase together with low boiling organics, and sufficient residual water i remains to be vaporized even in the feed that has been heated to a temperature above the boiling point of water, so that water is believed to play a significant role in increasing effectiveness of decontamination throughout a very broad range of temperatures. The vaporization step of the present invention is carried out using any known vessel capable of being indirectly heated and sealed sufficiently to maintain a vacuum of at least 25 mm Hg. The vessel should also contain a means for agitating and/or mixing the contaminated soil and/sludge material during the heating and cooling step while evolved vapors of volatile and semi,volatile chemical compounds are removed. A preferred vessel is one that is jacketed to allow for circulation of a heated fluid, such as hot oil, and has rotating internal shafts equipped with paddles or blades for agitating and/or mixing the inert solids. The mixing vessel used for vaporization of the feed typically operates under a vacuum. The level of vacuum is dictated by many factors, including quantity and type of feed contaminants. The actual pressure in the mixing vessel changes as a function of time during the process, and is controlled by a microprocessor. The microprocessor monitors the temperature of the system as well as the vapor evolution rate in order to determine the proper pressure level. The pressure in the system typically ranges from a high of just below atmospheric pressure when the system is first being purged to remove the oxygen, to a low of from about 400 mm Hg to about 50 mm Hg at a point when most of the water has been vaporized and it is mostly the contaminants which remain in the feed. By controlling the pressure in this fashion, it is not necessary for the system to operate at higher temperatures where energy consumption is increased. Alternatively, the system may operate at higher temperatures in which case the vacuum greatly reduces the residence time of the feed materials, once again lessening the energy consumption. A secondary advantage of the vacuum is that it ensures that if the system is not positively airtight, any leakage that might occur will draw air into the system, and not the reverse. This will minimize environmental emissions. The temperature of the heating fluid in the jacketed vessel is monitored and controlled to maintain the temperature of inert solids below 600.degree. F. Heating fluid may also be circulated through the rotating internal shafts for providing more efficient heat transfer to the contaminated soils and/or sludges. The strong vacuum and low temperature at which this process operates allows the volatile and semi-volatile components of the contaminated material to vaporize to form a gas phase, leaving behind a solid phase that is contaminated with radioactive pollutants and/or non-volatile chemical components. The gaseous phase, alternatively referred to as the evolved vapors, which may contain water vapor, air, an inert carrier gas, and vaporized contaminants such as volatile and semi-volatile organic compounds and PCBs, is continuously drawn off from the dryer. The water and organics are condensed and collected for further treatment. The specific operating parameters will vary depending on the moisture content of the feed, the concentration and boiling point(s) of contaminant(s) in the feed (which can vary over a wide range), and the percentage of the contaminants to be removed from the feed. Accordingly, the temperature and pressure required to vaporize the volatile chemical compounds, the degree of vacuum needed and the residence time in the vessel may vary widely. However, the maximum average solids temperature should not exceed 600.degree. F. Water or steam may also be positively employed in the inventive process to help strip contaminants from the interstitial spaces. It is believed that as water volatilizes within and around the interstitial spaces it helps volatilize or strip organics, and that the flow of steam entrains and helps carry organics out the dryer in the effluent gas stream. An inert gas other than steam is preferably introduced into the system for introductory purging of oxygen, and as a sweep gas for additional stripping efficiency. The sweep gas velocity is also controlled by the microprocessor and is also dependant upon the temperature and vapor evolution rate of the system. The inert gas is used in the process primarily for safety to eliminate the risk of a fire in the dryer and to reduce the partial pressure of the overall atmosphere to more easily distill or boil off organic contaminants. Nitrogen is preferably used for reasons of convenience and practicality. However, other inert gases such as, but not limited to, carbon dioxide, helium and argon, could also be used subject to price considerations, availability and composition of the feed material being processed. The gas phase, alternatively referred to as the evolved vapors, that is formed as the contaminants are removed from the contaminated soil and/or sludge in the dryer, comprises inert gases, water vapor, and volatile chemical contaminants. The evolved vapors are removed by the vacuum and pulled through a primary filter system to remove entrained particulates. The water and organics are then condensed and discharged for off-site disposal. Condensation can be performed by any method known to the art. One method is to pass the gaseous reaction product through an economizer to utilize its latent heat to supply the heating requirements of the vaporization step. Another method involves using a refrigerated cooling system. After treatment the solids are cooled to a temperature below 300.degree. F. A typical batch run requires a residence time of up to eight hours. As used herein, "residence time" is defined as the time it takes to load the vessel, vaporize the contaminants, and cool and unload the vessel. This system may be operated to remove virtually all organic compounds leaving an inert material that is decontaminated, contaminated only with radionuclides, contaminated only with non-volatile hazardous chemical contaminants, or in some cases, contaminated with both radionuclides and non-volatile hazardous chemical contaminants. If the remaining solids are contaminated with only radionuclides or non-volatile hazardous chemical components, then the solids can be treated or properly disposed. If the remaining solids are contaminated with radionuclides and non-volatile chemical components, then they should be further treated. One preferred method of treatment is to stabilize the hazardous non-volatile chemical components as is known in the art to prevent their leachability. After the stabilization the radioactive waste can be properly disposed of. A more complete understanding of the inventive concept of this invention may be obtained by a review of the accompanying figure, which presents a preferred embodiment of the invention. The presentation of this embodiment is not intended to exclude from the scope of the inventive concept those other embodiments set out herein or other reasonable and normal modifications of the inventive concept. Details, such as miscellaneous pumps, heaters, and coolers, condensers, start-up lines, valving, and similar hardware, have been omitted as being nonessential to a clear understanding of the preferred embodiment of the invention. It will also become apparent that the apparatus and conditions may be varied widely while retaining the basic principles of the present invention. The example is to be considered illustrative, and is not in any way restrictive. The inventive process will now be explained with reference to FIG. 1, although it will be understood that the spirit of the presently claimed invention is in no way limited to that particular embodiment. The process equipment needed to perform the present process and which is depicted in the process flow scheme of FIG. 1 can be easily incorporated into a small mobile system that requires only minimal mobilization and demobilization activities at the site location. In a preferred embodiment, the process equipment can be contained on two semi-tractor trailers, including self-contained generators which supply all needed electrical requirements, thus eliminating the need for access to external conventional energy sources. If mixed wastes are involved, the process equipment must be carefully constructed and monitored to ensure that no radioactive wastes escape. Referring again to FIG. 1, a batch of contaminated soil and/or sludge is loaded via line 1 into a jacketed vacuum dryer 3. Typically up to 15 tons (approximately 300 cubic feet) of contaminated soil and/or sludge material constitute a batch, and this requires about one half to about one hour to load. After the dryer is loaded and sealed, the vacuum pump 20 is started and an inert purge gas, such as nitrogen, is added to the system through line 2. Subsequently, after the system is purged, an internal agitator or other means of mixing the soil and/or sludge is started along with the circulation of a heating fluid, such as oil, through the jacketed vacuum dryer and agitator. As the vacuum pump 20 continues to draw a vacuum, nitrogen gas is added to the dryer as a sweep gas through line 2. The sweep gas helps purge the vacuum dryer 3 of volatilized liquids and aids in the stripping efficiency of the process. Both the pressure level and the sweep gas velocity are controlled by a microprocessor which monitors the temperature and vapor evolution rate of the system. The pressure in the dryer 3 typically drops to a final pressure from about 400 mm Hg to about 50 mm Hg as the water vapor is driven off to allow for operation of the dryer at temperatures below 600.degree. F. (and preferably below 500.degree. F.), to lower the residence time, and to prevent any environmental emissions. Both the heating and agitation rates can be computer controlled to accommodate differing feed material characteristics. As the materials in the dryer are heated and the vacuum is applied, the contaminants will begin to vaporize. The mixture of inert sweep gas, particulates, volatilized organic materials, steam and air are continuously drawn off by the vacuum through line 5. The withdrawn vapors first pass through a primary filter system 6 which removes entrained particulates, such as soil which has been carried out of the dryer by the vapor. The vapor is then carried through line 7 to a first condenser 8. This condenser 8 is preferably a fin fan type condenser in which the vapor stream is cooled to a temperature 10.degree. F. higher than the ambient air temperature in order to condense a portion of the hydrocarbons and water existing in the vapor phase. This stream passes via line 9 to a first knockout pot 10 where the vapor and liquid are separated. The liquid passes via line 11 to a first condensate collection tank 12. The vapor passes through line 13 to a second condenser unit 14, which is preferably a shell and tube condenser with circulating refrigerated liquid from the refrigerated cooling unit 15 on the cooling side. In this condenser, the vapor is cooled so that the temperature of the exiting stream 16 is less than 50.degree. F. Stream 16 passes to a second knockout pot in which the liquid and vapors are separated. The liquid stream 18 feeds into the condensate collection tank 12. The purpose of the first and second condensers 8 and 14 is to reduce the volume of vapor in line 19, which passes to the vacuum pump 20, by condensing a majority of the condensable liquids. Therefore, the size of the vacuum pump 20 can be held to a minimum thus decreasing costs significantly. As the vapor passes through the vacuum pump 20 it is compressed to atmospheric pressure, and a resulting increase in temperature also takes place. Therefore, the vapor leaving the vacuum pump 20 via line 21 is once again passed through a third condenser 22. This condenser is preferably also a shell and tube type condenser in which the circulating refrigerated liquid from the refrigerated cooling unit 15 is on the cooling side. The exiting stream 23 is cooled to about 40.degree. F. in order to condense any remaining condensable vapors. The liquid and vapor components of stream 23 are separated in a third knockout pot 24, and the liquid stream 25 feeds into a second condensate collection tank 26. The vapor stream 27 feeds into a high efficiency filter 28 to remove any remaining particulates, and the filtered stream 29 passes through an activated carbon bed 30 to remove any organic contaminants which were not condensed. After treatment, the vapor stream 31 is in condition for discharge into the atmosphere, or a portion of the stream may be recirculated back to the dryer as the sweep gas. The condensate collection tanks 12 and 26 which contain the condensed contaminants are discharged to off-site disposal. The entire operation typically takes about four to eight hours per batch. After the treatment is complete, the dryer contents are cooled to less than 300.degree. F. and the system is returned to atmospheric pressure. The solid phase is discharged through the dryer discharge port 2. The discharged material is then properly disposed of or sent for further treatment. Variations on the design or operation of the above illustrative embodiments may be readily made to adapt the inventive process to various operational demands, all of which are within the scope and spirit of the present invention. The present invention has been described in terms of certain preferred embodiments. Of course, numerous other embodiments not specifically described may fall within the spirit or scope of the following claims. |
claims | 1. An electron generating panel comprising:an outer sheathing of a low Z metal which forms an outer housing around the electron generating panel;an electron generator comprising a high Z metal disposed within the sheathing and forming an emitter with a positive output pin extending through the sheathing, electrically insulated from the sheathing;an electron charge collector comprising a low Z metal disposed between the emitter and the sheathing on a side of the emitter, with a negative output pin extending through the sheathing and the collector, electrically insulated from the sheathing and, spaced and electrically insulated from the positive output pin;a first layer of insulation material occupying spaces between the emitter and the collector; anda second layer of insulation material occupying spaces between the sheathing and the collector. 2. The electron generating panel of claim 1 wherein the high Z metal comprises Lead or Tungsten. 3. The electron generating panel of claim 1 wherein the low Z metal comprises Inconel or Steel alloy. 4. The electron generating panel of claim 1 wherein the layer of insulation comprises aluminum oxide or boron-10. 5. The electron generating panel of claim 1 wherein the layer of insulation surrounding the emitter is a cylinder approximately 1 mm thick. 6. The electron generating panel of claim 1 wherein the electron generating panel is positioned between an outer perimeter of a nuclear reactor core comprising a plurality of fuel assemblies and a structural member surrounding the plurality of fuel assemblies. 7. The electron generating panel of claim 6 wherein the structural member is a baffle-former plate structure, wherein a baffle-former plate structure comprises a baffle plate attached to a former plate and the electron generating panel is fastened to the baffle plate of the baffle-former plate structure surrounding the core. 8. The electron generating panel of claim 1 wherein the electron generating panel is approximately 3 mm thick. 9. The electron generating panel of claim 1 further comprising a layer of cobalt-59 between the sheathing and the emitter on a side of the emitter opposite that of the collector with the first layer of insulation material extending around the emitter to between the cobalt-59 and the emitter and the second layer of insulation material extending around the collector and the cobalt-59 to between the sheathing and the cobalt-59. 10. The electron generating panel of claim 1 wherein the collector spans opposite sides of the emitter with the sides of the collector on opposite sides of the emitter in electrical communication with each other. 11. The electron generating panel of claim 10 wherein the collector is a cylinder that surrounds the emitter. 12. The electron generating panel of claim 10 is supported on a wall of a fuel assembly cell of a modular fuel rack. |
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abstract | A multi-energy imaging system and method for selectively generating high-energy X-rays and low-energy X-ray beams are described. A pair of optic devices are used, one optic device being formed to emit high X-ray energies and the other optic device being formed to emit low X-ray energies. A selective filtering mechanism is used to filter the high X-ray energies from the low X-ray energies. The optic devices have at least a first solid phase layer having a first index of refraction with a first photon transmission property and a second solid phase layer having a second index of refraction with a second photon transmission property. The first and second layers are conformal to each other. |
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abstract | A therapeutic energy system for performing interstitial laser therapy and brachytherapy includes two categories of components. The first category includes components usable to perform both interstitial laser therapy and brachytherapy. The second category of components includes components usable to perform either interstitial laser therapy or brachytherapy, but not both. The components co-act to apply therapeutic energy to tissue. The components of the first system include components inserted percutaneously into the tissue, such that interstitial laser therapy and brachytherapy can be performed sequentially without removing and re-inserting percutaneous components. Components of the second category include components not requiring additional puncturing of the skin of a patient, such that removing and inserting components of the second category from a patient is done easily and painlessly. An energy probe component does not maintain a cavity around the tumor mass. Surgical excision of tissue can be performed coincident to therapeutic energy treatment as disclosed. |
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description | This application claims priority to U.S. Provisional Patent Application No. 62/712,704 filed on Jul. 31, 2018, the entire content of which is incorporated herein by reference. The semiconductor integrated circuit (IC) industry has experienced exponential growth. Technological advances in IC materials and design have produced generations of ICs where each generation has smaller and more complex circuits than the previous generation. In the course of IC evolution, functional density (i.e., the number of interconnected devices per chip area) has generally increased while geometry size (i.e., the smallest component (or line) that can be created using a fabrication process) has decreased. This scaling down process generally provides benefits by increasing production efficiency and lowering associated costs. Such scaling down has also increased the complexity of processing and manufacturing ICs. For example, the need to perform higher resolution lithography processes grows. One lithography technique is extreme ultraviolet lithography (EUVL). The EUVL employs scanners using light in the extreme ultraviolet (EUV) region, having a wavelength of about 1-100 nm. Some EUV scanners provide 4×reduction projection printing, similar to some optical scanners, except that the EUV scanners use reflective rather than refractive optics, i.e., mirrors instead of lenses. One type of EUV light source is laser-produced plasma (LPP). LPP technology produces EUV light by focusing a high-power laser beam onto small tin droplet targets to form highly ionized plasma that emits EUV radiation with a peak maximum emission at 13.5 nm. The EUV light is then collected by a LPP EUV collector mirror and reflected by optics towards a lithography target, e.g., a wafer. The LPP EUV collector mirror is subjected to damage and degradation due to the impact of particles, ions, radiation, and most seriously, tin deposition. The following disclosure provides many different embodiments, or examples, for implementing different features of the provided subject matter. Specific examples of components and arrangements are described below to simplify the present disclosure. These are, of course, merely examples and are not intended to be limiting. For example, the formation of a first feature over or on a second feature in the description that follows may include embodiments in which the first and second features are formed in direct contact, and may also include embodiments in which additional features may be formed between the first and second features, such that the first and second features may not be in direct contact. In addition, the present disclosure may repeat reference numerals and/or letters in the various examples. This repetition is for the purpose of simplicity and clarity and does not in itself dictate a relationship between the various embodiments and/or configurations discussed. Further, spatially relative terms, such as “beneath,” “below,” “lower,” “above,” “upper” and the like, may be used herein for ease of description to describe one element or feature's relationship to another element(s) or feature(s) as illustrated in the figures. The spatially relative terms are intended to encompass different orientations of the device in use or operation in addition to the orientation depicted in the figures. The apparatus/device may be otherwise oriented (rotated 90 degrees or at other orientations) and the spatially relative descriptors used herein may likewise be interpreted accordingly. In addition, the term “made of” may mean either “comprising” or “consisting of.” The present disclosure is generally related to extreme ultraviolet (EUV) lithography system and methods. More particularly, it is related to apparatus and methods for mitigating contamination on an EUV collector mirror in a laser produced plasma (LPP) EUV radiation source. The EUV collector mirror, also referred to as an LPP EUV collector mirror or an EUV collector mirror, is a component of the LPP EUV radiation source. It collects and reflects EUV radiation and contributes to overall EUV conversion efficiency. However, it is subjected to damage and degradation due to the impact of particles, ions, radiation, and debris deposition. In particular, tin (Sn) debris is one of the contamination sources of the EUV collector mirror. EUV collector mirror life time, the time duration where the reflectivity decays to half of the initial reflectivity, is one of the factors for an EUV scanner. The major reason of reflectivity decay of the EUV collector mirror is residual metal contamination (tin debris) on the EUV collector mirror surface caused, inevitably, by the EUV light generation procedure. FIG. 1 is a schematic and diagrammatic view of an EUV lithography system. The EUV lithography system includes an EUV radiation source apparatus 100 to generate EUV light, an exposure tool 200, such as a scanner, and an excitation laser source apparatus 300. As shown in FIG. 1, in some embodiments, the EUV radiation source apparatus 100 and the exposure tool 200 are installed on a main floor MF of a clean room, while the excitation source apparatus 300 is installed in a base floor BF located under the main floor. Each of the EUV radiation source apparatus 100 and the exposure tool 200 are placed over pedestal plates PP1 and PP2 via dampers DP1 and DP2, respectively. The EUV radiation source apparatus 100 and the exposure tool 200 are coupled to each other by a coupling mechanism, which may include a focusing unit. The lithography system is an extreme ultraviolet (EUV) lithography system designed to expose a resist layer by EUV light (or EUV radiation). The resist layer is a material sensitive to the EUV light. The EUV lithography system employs the EUV radiation source apparatus 100 to generate EUV light, such as EUV light having a wavelength ranging between about 1 nm and about 100 nm. In one particular example, the EUV radiation source 100 generates an EUV light with a wavelength centered at about 13.5 nm. In the present embodiment, the EUV radiation source 100 utilizes a mechanism of laser-produced plasma (LPP) to generate the EUV radiation. The exposure tool 200 includes various reflective optic components, such as convex/concave/flat mirrors, a mask holding mechanism including a mask stage, and wafer holding mechanism. The EUV radiation EUV generated by the EUV radiation source 100 is guided by the reflective optical components onto a mask secured on the mask stage. In some embodiments, the mask stage includes an electrostatic chuck (e-chuck) to secure the mask. Because gas molecules absorb EUV light, the lithography system for the EUV lithography patterning is maintained in a vacuum or a-low pressure environment to avoid EUV intensity loss. In the present disclosure, the terms mask, photomask, and reticle are used interchangeably. In the present embodiment, the mask is a reflective mask. One exemplary structure of the mask includes a substrate with a suitable material, such as a low thermal expansion material or fused quartz. In various examples, the material includes TiO2 doped SiO2, or other suitable materials with low thermal expansion. The mask includes multiple reflective multiple layers (ML) deposited on the substrate. The ML includes a plurality of film pairs, such as molybdenum-silicon (Mo/Si) film pairs (e.g., a layer of molybdenum above or below a layer of silicon in each film pair). Alternatively, the ML may include molybdenum-beryllium (Mo/Be) film pairs, or other suitable materials that are configurable to highly reflect the EUV light. The mask may further include a capping layer, such as ruthenium (Ru), disposed on the ML for protection. The mask further includes an absorption layer, such as a tantalum boron nitride (TaBN) layer, deposited over the ML. The absorption layer is patterned to define a layer of an integrated circuit (IC). Alternatively, another reflective layer may be deposited over the ML and is patterned to define a layer of an integrated circuit, thereby forming an EUV phase shift mask. The exposure tool 200 includes a projection optics module 210 for imaging the pattern of the mask on to a semiconductor substrate with a resist coated thereon secured on a substrate stage of the exposure tool 200. The projection optics module generally includes reflective optics. The EUV radiation (EUV light) directed from the mask, carrying the image of the pattern defined on the mask, is collected by the projection optics module, thereby forming an image onto the resist. In the present embodiments, the semiconductor substrate is a semiconductor wafer, such as a silicon wafer or other type of wafer to be patterned. The semiconductor substrate is coated with a resist layer sensitive to the EUV light in the present embodiment. Various components including those described above are integrated together and are operable to perform lithography exposing processes. The lithography system may further include other modules or be integrated with (or be coupled with) other modules. As shown in FIG. 1, the EUV radiation source 100 includes a target droplet generator 115 and a LPP EUV collector mirror 110, enclosed by a chamber 105. The target droplet generator 115 generates a plurality of target droplets DP. In some embodiments, the target droplets DP are tin (Sn) droplets. In some embodiments, the tin droplets each have a diameter about 30 microns (μm). In some embodiments, the tin droplets DP are generated at a rate about 50 droplets per second and are introduced into a zone of excitation ZE at a speed about 70 meters per second (m/s). Other material can also be used for the target droplets, for example, a tin containing liquid material such as eutectic alloy containing tin or lithium (Li). The excitation laser LR2 generated by the excitation laser source apparatus 300 is a pulse laser. In some embodiments, the excitation layer includes a pre-heat laser and a main laser. The pre-heat laser pulse is used to heat (or pre-heat) the target droplet to create a low-density target plume, which is subsequently heated (or reheated) by the main laser pulse, generating increased emission of EUV light. In various embodiments, the pre-heat laser pulses have a spot size about 100 μm or less, and the main laser pulses have a spot size about 200-300 μm. The excitation laser (laser pulses) LR2 are generated by the excitation laser source 300. The laser source 300 may include a laser generator 310, laser guide optics 320 and a focusing apparatus 330. In some embodiments, the laser source 310 includes a carbon dioxide (CO2) or a neodymium-doped yttrium aluminum garnet (Nd:YAG) laser source. The excitation laser (laser light) LR1 generated by the laser generator 300 is guided by the laser guide optics 320 and focused into the excitation laser LR2 by the focusing apparatus 330, and then introduced into the EUV radiation source 100. The excitation laser (laser light) LR2 is directed through windows (or lenses) into the zone of excitation ZE. The windows adopt a suitable material substantially transparent to the laser beams. The generation of the pulse lasers is synchronized with the generation of the target droplets. As the target droplets move through the excitation zone, the pre-pulses heat the target droplets and transform them into low-density target plumes. A delay between the pre-pulse and the main pulse is controlled to allow the target plume to form and to expand to an optimal size and geometry. When the main pulse heats the target plume, a high-temperature plasma is generated. The plasma emits EUV radiation EUV, which is collected by the EUV collector mirror 110. The EUV collector mirror 110 further reflects and focuses the EUV radiation for the lithography exposing processes. In some embodiments, a droplet catcher 120 is installed opposite the target droplet generator 115. The droplet catcher 120 is used for catching excessive target droplets. For example, some target droplets may be purposely missed by the laser pulses. The EUV collector mirror 110 is designed with a proper coating material and shape to function as a mirror for reflection, and focusing. In some embodiments, the EUV collector mirror 110 is designed to have an ellipsoidal geometry. In some embodiments, the coating material of the EUV collector mirror 110 is similar to the reflective multilayer of the EUV mask. In some examples, the coating material of the EUV collector mirror 110 includes a ML (such as a plurality of Mo/Si film pairs) and may further include a capping layer (such as Ru) coated on the ML to substantially reflect the EUV light. In some embodiments, the EUV collector mirror 110 may further include a grating structure designed to effectively scatter the laser beam directed onto the EUV collector mirror 110. For example, a silicon nitride layer is coated on the EUV collector mirror 110 and is patterned to have a grating pattern. In such an EUV radiation source apparatus, the plasma caused by the laser application creates physical debris, such as ions, gases and atoms of the droplet, as well as the desired EUV radiation. It is necessary to prevent the accumulation of material on the EUV collector mirror 110 and also to prevent physical debris exiting the chamber 105 and entering the exposure tool 200. As shown in FIG. 1, in some embodiments, a buffer gas is supplied from a first buffer gas supply 130 through the aperture in the EUV collector mirror 110 by which the pulse laser is delivered to the tin droplets. In some embodiments, the buffer gas is H2, He, Ar, N or another inert gas. In certain embodiments, H2 is used as H radicals generated by ionization of the buffer gas can be used for cleaning purposes. The buffer gas can also be provided through one or more second buffer gas supplies 135 toward the EUV collector mirror 110 and/or around the edges of the EUV collector mirror 110. Further, the chamber 105 includes one or more gas outlets 140 so that the buffer gas is exhausted outside the chamber 105. Hydrogen gas has low absorption to the EUV radiation. Hydrogen gas reaching to the coating surface of the EUV collector mirror 110 reacts chemically with a metal of the droplet forming a hydride, e.g., metal hydride. When tin (Sn) is used as the droplet, stannane (SnH4), which is a gaseous byproduct of the EUV generation process, is formed. The gaseous SnH4 is then pumped out through the outlet 140. However, it is difficult to exhaust all gaseous SnH4 from the chamber and to prevent the SnH4 from entering the exposure tool 200. To prevent the SnH4 or other debris deposition on the EUV collector mirror 110, one or more debris collection mechanisms 150 are employed in the chamber 105. As shown in FIG. 1, one or more debris collection mechanisms 150 are disposed along optical axis A1 between the zone of excitation ZE and an output port 160 of the EUV radiation source 100. FIG. 2A is a front view of the debris collection mechanism 150 and FIG. 2B is a schematic side view of debris collection mechanism 150. FIGS. 2A to 2C is a partial picture of the debris collection mechanism 150. The debris collection mechanism 150 includes a frustoconical support frame 151, a first end support 153 and a second end support 154 that operably support a plurality of vanes 152 that rotate within the housings. The first end support 153 has a larger diameter than the second end support 154. The debris collection mechanism 150 serves to prevent the surface of the EUV collector mirror 110 and/or other elements/portions of the inside the chamber 105 from being coated by Sn vapor by repelling slow Sn atoms and/or SnH4. The plurality of vanes 152 project radially inwardly from the frustoconical support frame 151. The vanes 152 are thin and elongate plates. In some embodiments, each of the vanes has a triangular or trapezoid or trapezium shape in plan view. The vanes 152 are aligned so that their longitudinal axes are parallel to the optical axis A1 so that they present the smallest possible cross-sectional area to the EUV radiation EUV. The vanes 152 project towards the optical axis A1, but do not extend as far as the optical axis. In some embodiments, a central core of the debris collection mechanisms 150 is empty. The debris collection mechanisms 150 is rotated by a drive unit including one or more motors, one or more belts and/or one or more gears, or any rotating mechanism. One mechanism for preventing gaseous tin and/or gaseous SnH4 from depositing on the EUV collector mirror 110 is applying a magnetic field. In some embodiments, superconducting magnets are placed on either side of the EUV collector mirror 110 to generate a strong magnetic field and confine the plasma in a direction perpendicular to the magnetic field. However, the plasma may not be completely confined in a direction parallel to the magnetic field, resulting in deposition of tin 70 and/or SnH4 near the magnets. FIG. 3A shows the EUV collector mirror 110 after use, on which tin debris are deposited, and FIG. 3B shows the EUV collector mirror 110 after cleaning the surface thereof. As set forth above, the EUV collector mirror 110 contamination by the residual metal from the EUV light generation procedure is the major cause of the EUV scanner exposure power loss and throughput down trend. The EUV collector mirror life time is maintained at about 3 months, for example, and then it is generally necessary for a week or more of down time to swap the EUV collector mirror 110 with a new, clean EUV collector mirror to maintain high exposure power and throughput. While the magnetic field can be helpful to reduce collector mirror contamination, it may not result in sufficient prevention of deposition of tin and/or SnH4 on the EUV collector mirror 110. It is to be noted that because of space constraints caused by positioning of the droplet generator, droplet catcher and other components near the EUV collector mirror 110, it is generally difficult to place additional superconducting magnets around the EUV collector mirror 110, especially given that the superconducting magnets tend to be fairly large in size because of accompanying cooling and electric systems. Thus, alternative methods for prevention of the EUV collector mirror contamination may be beneficial. As shown in FIG. 4, the EUV vessel 1000 includes a plurality of inlet ports 1100 and exhaust ports 1200 for receiving and transmitting cleaning gas. In some embodiments, the plurality of inlet ports 1100 include a first cleaning gas port 1120, a second cleaning gas port 1140, and third and fourth cleaning gas ports 1160, 1180. The EUV vessel 1000 includes one or more exhaust ports 1200 so that the cleaning gas is exhausted outside the EUV vessel 1000. In some embodiments, a width W1 is in the range from about 80 mm to about 100 mm. In some embodiments, a width W2 is in the range from about 710 mm to about 750 mm. In some embodiments, a width W3 is in the range from about 340 mm to about 380 mm. In some embodiments, a height H1 is in the range from about 950 mm to about 1000 mm. In some embodiments, a height H2 is in the range from about 1310 mm to about 1360 mm. While the EUV vessel 1000 shown in FIG. 4 includes the above inlet and outlet ports, alternate embodiments of the EUV vessel 1000 may include a different number and/or arrangement of inlet and/or outlet ports. Hydrogen gas has low absorption to the EUV radiation. Hydrogen gas reaching to the coating surface of the EUV collector mirror 110 reacts chemically with a metal of the droplet forming a hydride, e.g., metal hydride. When tin (Sn) is used as the droplet, Sn ions and/or stannane (SnH4), which is a gaseous byproduct of the EUV generation process, is formed. The gaseous SnH4 is then pumped out through the exhaust ports 1200. However, it is difficult to exhaust all gaseous SnH4 from the chamber and to prevent the SnH4 from contaminating the collector mirror 110. In some embodiments, the EUV vessel 1000 includes a trajectory correcting device 1500 connected to the EUV collector mirror 110. In some embodiments, the trajectory correcting device 1500 includes one or more first charging point 1520 embedded in the EUV collector mirror 110 and a second charging point 1530 attached to a lower cone 190 of the EUV radiation source 100. In some embodiments, the first charging point 1520 embedded in the EUV collector mirror 110 is provided with a higher electrical potential than the electrical potential at the second charging point 1530 attached to the lower cone 190 of the EUV radiation source 100. In alternative embodiments, it is possible to provide a lower electrical potential to the first charging point 1520 than the electrical potential at the second charging point 1530 attached to the lower cone 190 of the EUV radiation source 100. In some embodiments, the first and second charging points 1520, 1530 include Ni—Cr alloy wires and/or Fe—Cr—Al alloy wires. A pulse timing module 1540 provides electric power to the first charging point 1520 via a DC bias circuit 1545 and controls the modulating frequency. In some embodiments, the DC bias circuit 1545 provides a biasing DC voltage to the EUV collector mirror 110. In some embodiments, the DC bias circuit 1545 provides delays the biasing DC voltage. In some embodiments, the DC bias circuit 1545 is connected to a controller 1547 and the biasing DC voltage is controlled by the controller. In other embodiments, the trajectory correcting device 1500 provides an electrical field to the reflective surface and/or the back surface of the EUV collector mirror 110. In some embodiments, the first charging point 1520 is divided into a plurality of sections, which are independently controlled by the pulse timing module 1540. With this feature, it is possible to locally provide an electrical field to a part of the EUV collector mirror 110 where the debris is heavily deposited. In some embodiments, the EUV collector mirror 110 is provided with a voltage in a range from about 1 Vdc to about 50 Vdc. In other embodiments, the EUV collector mirror 110 is provided with an electric potential equal to 5 V, 20 V, 30 V, 40 V, 50 V, or any other potential between any two of these values. Further, by modulating the voltage, it is possible to create an electric field that causes the tin ions (Sn′) to be rapidly redistributed away from the EUV collector mirror 110. In some embodiments, the EUV collector mirror 110 is configured such that a desired direction of the electric field is obtained when the EUV collector mirror 110 is provided with an electric potential so that the electric field causes the tin ions (Sn′) to be rapidly redistributed in a direction away from the EUV collector mirror 110 during EUV radiation source operation. The EUV collector mirror 110 includes one or more pulse timing module 1540 to prevent metal debris (contamination) from depositing on the surface of the EUV collector mirror 110. The trajectory correcting device 1500 provides an electric potential in the EUV collector mirror 110 so that the voltage is adjusted through the pulse timing module 1540. In some embodiments, the pulse timing module includes a configurable pattern of the pulse to adjust the trajectory of the metal debris when necessary. The configurable pattern of the pulse is controlled by control circuitry in some embodiments. As shown in FIG. 5A, in some embodiments, the trajectory correcting device 1500 may further includes first electrodes 1550 having an electrical potential V+ arranged in the EUV collector mirror 110 adjacent to the vanes 152 to provide an electric field. In some embodiments, the trajectory correcting device 1500 may further include second electrodes 1560 having an electrical potential V− arranged in adjacent to the lower cone 190 of the EUV vessel 1000. In some embodiments, the electrical potential V+ at the first electrodes 1550 is higher than the electrical potential V− at the second electrodes 1560 adjacent to the exhaust ports 1200 of the EUV vessel. In the foregoing embodiment, tin ions (Snx+) are caught by the electric field EF, the first electrodes 1550 and the second electrodes 1560 guide the Sn ions along the electric field and discharge the Sn ions to the exhaust ports 1200. In some embodiments as shown in FIG. 5B, the first electrodes form a mesh structure 1555 in a plan view extending along and adjacent the droplet axis A2. In some embodiments, the mesh structure 1555 has an opening in a center region of a shape such as for example, a donut shape, where the excitation laser beam passes through. In certain embodiments as shown in FIG. 5C, the second electrodes form a ring-shaped structure 1565 adjacent to the lower cone 190 and the exhaust ports 1200 along the optical axis A1. In some embodiments as shown in FIG. 5D, the trajectory correcting device 1500 further includes one or more isolated electrode regions 1549 in the EUV collector mirror 110. FIGS. 5E, 5F, and 5G are detailed views showing different arrangements of the isolated electrode regions 1549 on the back side of the EUV collector mirror 110. The isolated electrode regions 1549 separated by electrically insulating regions 1570 connected to the DC biasing circuit 1545 according to an embodiment of this disclosure. In one or more embodiments, as shown in FIG. 5H, the trajectory correcting device 1500 further includes isolated electrodes 1580 arranged in the EUV collector mirror 110. The isolated electrodes 1580 are a plurality of electrically isolated electrodes 1580 connected to the DC biasing circuit 1545 according to an embodiment of this disclosure. While the specific arrangement are provided for the electrodes disclosed above, these are merely exemplary embodiments and that the present system may utilize any appropriate configuration of the trajectory correcting device 1500. Further, depending upon the application of the present system, specific capabilities of the foregoing electrodes may be required. These various configurations may be contemplated by the present disclosure. As shown in FIG. 5A, in some embodiments, the trajectory correcting device 1500 may further include first electrodes 1550 arranged adjacent to the EUV collector mirror 110 to provide an electric field. In particular embodiments, such electrodes 1550 may form a mesh structure 1555 in a plan view arranged alongside the droplet axis A2. FIG. 6 shows an exemplary view of the EUV collector vessel 1000 according to an embodiment of the present disclosure. In FIG. 6 the electric field potential across the length of the EUV collector vessel 1000 is illustrated. As shown, in some embodiments, the electric field potential ranges from 10 V adjacent the collector mirror surface to 0 V adjacent the lower cone. In some embodiments, the tin liquid droplet is irradiated by the pre-pulse laser light and then by the carbon dioxide gas laser light to emit EUV light. When tin ions (Snx+) are caught by the electric field, the trajectory correcting device 1500 of this disclosure guides and discharges the Sn ions along the electric field to the exhaust ports 1200. In some embodiments, the exhaust ports 1200 have a debris exhausting structure 1210 provided on an outlet portion 1080 as shown in FIG. 6. The debris exhausting structure 1210 collects the tin ions (Snx+) redistributed away from the EUV collector mirror 110 and guided by the electric field caused by the trajectory correcting device 1500 to the outside of the EUV vessel 1000. Hydrogen gas introduced into the chamber reacts chemically with the tin ions forming a gas, stannane (SnH4). The gaseous SnH4 is then pumped out through the outlet 140. FIGS. 7A, 7B, 7C, 7D, 7E and 7F show the results of a simulation of the distribution of Sn ions and electrons when no potential is applied to the EUV collector mirror 110. In some embodiments, as shown in FIG. 7A, a laser pulse is arranged with a pre-pulse of 100 ns and a main pulse of 70 ns as shown in FIGS. 7A-7F. An interval between the pre-pulse and the main pulse is in a range of about 20 μs in some embodiments. In some embodiments, the pulse timing module 1540 is connected to the trajectory correcting device 1500 to provide an electrical potential of 0 Vdc. However, the location of the pulse timing module 1540 is not limited to the arrangement of FIG. 4. In some embodiments, the pulse timing module 1540 is located at or near the lowest position of the EUV collector mirror 110 when the EUV collector mirror 110 is installed in an EUV radiation source apparatus. In some embodiments, FIGS. 7B and 7C illustrate electrons that are spreading radially after about 0.5 μs and 1 μs of time has passed by with no electric field generated by the trajectory correcting device 1500. After about 51 μs of time has passed from the main laser pulse, as shown in FIG. 7D, tin ions (Sn9+) are observed. In some embodiments, FIGS. 7E and 7F illustrate tin ions (Sn9+) that are spreading radially after about 501 μs and 891 μs of time has passed by with no electric field generated by the trajectory correcting device 1500. In some embodiments, as shown in FIG. 7F, the tin ions (Sn9+) are located near and adjacent to the EUV collector mirror 110. Therefore, it is likely that gaseous tin and/or gaseous SnH4 will be deposited on the EUV collector mirror 110. FIGS. 8A, 8B, 8C, 8D, 8E and 8F show the results of a simulation of the distribution of Sn ions and electrons when a 10V potential is applied to the EUV collector mirror 110 according to embodiments of the present disclosure. As can be seen in FIG. 8B, the EUV collector mirror 110 immediately attracts all the electrons. FIGS. 8C through 8F illustrate the Sn ions are repelled away from the EUV collector mirror 110 quickly within about 191 μs of time by the electric field with the 10V potential generated by the trajectory correcting device 1500 as illustrated in FIGS. 8C through 8F. In some embodiments, as shown in FIG. 8F, the tin ions (Sn9+) are located near and adjacent to the lower cone 190 of the EUV radiation source 100 shown in FIG. 4. In this embodiment, it is likely the gaseous tin and/or gaseous SnH4 formed by the tin reacting with H2 gas will be exhausted through the debris exhausting structure 1210 shown in FIG. 4. In some embodiments, the application of the potential is synchronized with the main laser pulse so as to prevent the potential from interfering with the plasma reionization process which results in the emission of the EUV radiation. In such embodiments, the electric potential is applied as a pulse after a certain predetermined time period following the arrival of the main laser pulse in the zone of excitation. For example, in an embodiment, the electric potential is applied about 1 μs, about 2 μs, about 5 μs, about 10 μs or any time delay between any two of these values, after the main laser pulse. In some embodiments, a pulse duration of the electric potential is the same as that of the main laser pulse. In some embodiments, the pulse duration of the electric potential is different from that of the main laser pulse. In some embodiments, the application of the electric potential according to embodiments of this disclosure obviates the need for using superconducting magnets, thereby saving the costs associated with the magnets and their operation. In some embodiments, the application of the electric potential renders the superconducting magnets as a fail-safe mechanism, thereby increasing desired redundancy in the system. The application of the electric potential to the EUV collector mirror 110 of the EUV radiation source prevents deposition of tin on the EUV collector mirror surface and increases the working time of the EUV radiation source, thereby increasing the throughput of the fabrication process. Moreover, because the EUV collector mirror 110 remains contamination free, the efficiency of the EUV radiation source is also increased. In some embodiments, the DC voltage applied to the EUV collector mirror 110 does not heat the EUV collector mirror 110. In other words, the EUV collector mirror 110 is not electrically connected to a potential other than the power supply. An EUV collector mirror for an extreme ultra violet (EUV) radiation source apparatus includes an EUV collector mirror body on which a reflective layer as a reflective surface is disposed, a trajectory correcting device attached to or embedded in the EUV collector mirror body and a trajectory correcting device to adjust the trajectory of a metal from the reflective surface of the EUV collector mirror body to an opposite side of the EUV radiation source apparatus away from the EUV collector mirror body. In accordance with one aspect of the present disclosure, an EUV vessel for an extreme ultra violet (EUV) radiation source apparatus includes an EUV collector mirror body, on which a reflective layer as a reflective surface is disposed, and a trajectory correcting device attached to or embedded in the EUV collector mirror body. In some embodiments, the trajectory correcting device is configured adjust the trajectory of metal ion towards an opposite side of the EUV vessel away from the EUV collector mirror body. In some embodiments, the trajectory correcting device includes a pulse timing module coupled to a first charging point at the EUV collector mirror body and a second charging point at a lower cone of the EUV vessel. In some embodiments, a conduit connects the first charging point at the EUV collector mirror body and the pulse timing module. In one or more of the foregoing embodiments, the first charging point is configured to positively bias the collector by applying a voltage in a range from 1 V to 50 V to the collector mirror body. In some embodiments, the pulse timing module is configured to provide an amplitude modulation of the pulse. In other embodiments, the pulse timing module is also configured to provide a frequency modulation of the pulse. In one or more embodiments, the pulse timing module is configured to provide an electric to cause the tin ions to be redistributed in a direction away from the EUV collector mirror body. In accordance with another aspect of the present disclosure, a method of preventing contamination of a collector of an extreme ultraviolet (EUV) radiation source comprises providing an EUV collector mirror body that a reflective layer is disposed on the EUV collector mirror body as a reflective surface. The method of preventing contamination of a collector also includes providing a trajectory correcting device attached to or embedded in the EUV collector mirror body. The method of preventing contamination of a collector further includes applying an electric field to an EUV collector mirror body in order to adjust the trajectory of metal debris towards an opposite side of the EUV radiation source away from the EUV collector mirror body. In some embodiments, the method providing the trajectory correcting device includes a pulse timing module that is coupled to a first charging point at the EUV collector mirror body and a second charging point at a lower cone of the EUV vessel. In some embodiments, the first charging point at the EUV collector mirror body is connected to the pulse timing module by a conduit. In some embodiments, the method further comprises positively biasing the collector by applying a voltage in a range from 1 V to 50 V to the EUV collector mirror body. In some embodiments, the method further comprises positively biasing the collector by applying a voltage in a range from 1 V to 50 V to the EUV collector mirror body. In some embodiments, the method further comprises modulating the amplitude of the voltage using the pulse timing module. In some embodiments, the method further comprises modulating a frequency of voltage pulses using the pulse timing module. In accordance with another aspect of the present disclosure, an extreme ultra violet (EUV) radiation source apparatus comprises a chamber enclosing an EUV vessel. In some embodiments, the EUV vessel comprises a collector mirror configured to reflect EUV radiation, a debris collection mechanism disposed over the collector mirror and a lower cone disposed over the debris collection mechanism. In some embodiments, the EUV vessel also comprises at least one first charging point attached to the collector mirror and a second charging point attached to the lower cone. In some embodiments, a pulse timing module and a DC bias circuit are coupled to the first charging point and the second charging point. In some embodiments, the timing module, DC bias circuit, and the first charging point are configured to positively bias the collector by applying a voltage in a range from 1 V to 50 V to the collector mirror. In some embodiments, the pulse timing module modulates a frequency and amplitude of the modulation of a DC pulse that is applied to the collector mirror. In some embodiments, the at least one charging points are attached to the collector mirror adjacent to vanes of the debris collection mechanism. In some embodiments, the EUV radiation source apparatus further comprises isolated electrode regions on a rear side of an EUV collector mirror. In some embodiments, the isolated electrode regions are separated by electrically insulating regions. The foregoing outlines features of several embodiments or examples so that those skilled in the art may better understand the aspects of the present disclosure. Those skilled in the art should appreciate that they may readily use the present disclosure as a basis for designing or modifying other processes and structures for carrying out the same purposes and/or achieving the same advantages of the embodiments or examples introduced herein. Those skilled in the art should also realize that such equivalent constructions do not depart from the spirit and scope of the present disclosure, and that they may make various changes, substitutions, and alterations herein without departing from the spirit and scope of the present disclosure. |
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description | The disclosed subject matter relates to enhancing power from an extreme ultraviolet light source with feedback from a spatially-extended target distribution. Extreme ultraviolet (EUV) light, for example, electromagnetic radiation having wavelengths of around 50 nm or less (also sometimes referred to as soft x-rays), and including light at a wavelength of about 13 nm, can be used in photolithography processes to produce extremely small features in substrates, for example, silicon wafers. Methods to produce EUV light include, but are not necessarily limited to, converting a material that has an element, for example, xenon, lithium, or tin, with an emission line in the EUV range into a plasma state. In one such method, often termed laser produced plasma (LPP), the plasma can be produced by irradiating a target material, for example, in the form of a droplet, stream, or cluster of material, with an amplified light beam that can be referred to as a drive laser. For this process, the plasma is typically produced in a sealed vessel, for example, a vacuum chamber, and monitored using various types of metrology equipment. In some general aspects, a method includes releasing a stream of target material droplets toward a target region, the droplets in the stream traveling along a trajectory from a target material supply system to the target region; producing a spatially-extended target distribution by directing a first pulse of light along a direction of propagation toward the first target material droplet while the first droplet is between the target material supply apparatus and the target region, the impact of the first pulse of light on the first target material droplet increasing a cross-sectional diameter of the first target material droplet in a plane that faces the direction of propagation and decreasing a thickness of the first target material droplet along a direction that is parallel to the direction of propagation; positioning an optic to establish a beam path that intersects the target location; coupling a gain medium to the beam path; and producing an amplified light beam that interacts with the spatially-extended target distribution to produce plasma that generates extreme ultraviolet (EUV) light by scattering photons emitted from the gain medium off of the spatially-extended target distribution, at least some of the scattered photons placed on the beam path to produce the amplified light beam. Implementations can include one or more of the following features. For example, the EUV light can be generated without providing external photons to the beam path. The stream can include a plurality of target material droplets, each separated from one another along the trajectory, and separate spatially-extended target distributions are produced from more than one of the droplets in the stream. The first pulse of light can have a wavelength of 1.06 μm. A cross-sectional diameter of the spatially-extended target distribution in the plane that is transverse to the direction of propagation can be 3 to 4 times larger than the cross-sectional diameter of the first target material droplet. The spatially-extended target distribution can be produced a time period after the first light pulse impacts the first target material droplet. The first pulse of light can have a duration of 10 ns. The amplified light beam can have a foot-to-foot duration of 400-500 ns. The amplified light beam can have wavelength of 10.6 μm. The amplified light beam can have a wavelength that is about ten times the wavelength of the first pulse of light. The method can include sensing that a first target material droplet in the stream of droplets is between the target material supply system and the target region. The spatially-extended target distribution can be in the form of a disk. The disk can include a disk of molten metal. The amplified light beam can interact with the spatially-extended target distribution to generate extreme ultraviolet (EUV) light without any coherent radiation being produced. The optic can be positioned at a side of the gain medium opposite to the target location to reflect light back on the beam path. In other general aspects, an extreme ultraviolet light source includes an optic positioned to provide light to a beam path; a target supply system that generates a stream of target material droplets along a trajectory from the target supply system to a target location that intersects the beam path; a light source positioned to irradiate a target material droplet in the stream of target material droplets at a location that is between the target supply system and the target location, the light source emitting light of an energy sufficient to physically deform a target material droplet into a spatially-extended target distribution; a gain medium positioned on the beam path between the target location and the optic; and a spatially-extended target distribution positionable to at least partially coincide with the target location to define an optical cavity along the beam path and between the spatially-extended target distribution and the optic. The spatially-extended target distribution and the target material droplets comprise a material that emits EUV light in a plasma state. Implementations can include one or more of the following features. For example, the target material can include tin, and the target material droplets can include droplets of molten tin. The spatially-extended target distribution can have a cross-sectional diameter in a plane that is perpendicular to direction of propagation of an amplified light beam that is produced by the optical cavity, and the cross-sectional diameter of the spatially-extended target distribution can be 3-4 times larger than a cross-sectional diameter of the target material droplet. Implementations of any of the techniques described above may include a method, a process, a target, an assembly or device for generating optical feedback from a spatially-extended target distribution, a kit or pre-assembled system for retrofitting an existing EUV light source, or an apparatus. The details of one or more implementations are set forth in the accompanying drawings and the description below. Other features will be apparent from the description and drawings, and from the claims. Techniques are described that enhance power from an extreme ultraviolet light source with feedback from a target material that has been modified prior to entering a target location into a spatially-extended target distribution or extended target. The feedback from the spatially-extended target distribution provides a nonresonant optical cavity because the geometry of the path over which feedback occurs, such as the round-trip length and direction, can change in time, or the shape of the spatially-extended target distribution may not provide a smooth enough reflectance. However, it may be possible that the feedback from the spatially-extended target distribution provides a resonant and coherent optical cavity if the geometric and physical constraints noted above are overcome. In any case, the feedback can be generated using spontaneously emitted light that is produced from a non-oscillator gain medium. In particular, the shape of a droplet of a target material is modified as it travels toward a target location with a pre-pulse optical beam so that the reflectivity of the modified target material when it reaches the target location is much greater than the reflectivity of the target material droplet. In this way, it is possible to provide feedback in a beam path that includes a gain medium by irradiating the highly-reflective spatially-extended target distribution with the light produced from the optical gain medium if a reflecting optic is positioned to reflect light on a beam path that intersects the target location so that the modified target material and the optic form an oscillating optical cavity. The oscillating optical cavity produced by the reflection off of the spatially-extended target distribution can be considered a random laser with incoherent feedback if the light that reflects from the spatially-extended target distribution provides a scattering surface that reflects light along distinct paths so that the reflected light may not return to its original position (for example, at the reflecting optic) after one round trip. The spatial resonances for the electromagnetic field may be absent in such a cavity and thus, the feedback in such a laser is used to return part of the energy or photons to the gain medium. In this scenario, many modes in the optical cavity interact with the gain medium as a whole, and the statistical properties of the laser emission in this case can be approximated or close to those of the emission of an extremely bright black body in a narrow range of a spectrum. Also, there may be no spatial coherence. The target material droplets are a part of a stream of target material that is released toward the target location. The target location is on the axis of the beam path and the optical gain medium. Prior to reaching the target location, the pre-pulse optical beam irradiates the target material droplet to form the spatially-extended target distribution, which is a modified shape of the target material such as a flattened or disk-shaped target. The modified shape of the target material can be a mist, cloud of fragments, or a hemisphere-like target that can have similar properties to a disk-shaped target. In any case, the modified shape of the target material has a larger extent or a larger surface area that faces the amplified light beam in the target location. Compared to the original target material droplet, the spatially-extended target distribution has a larger diameter and has a greater reflectivity. The spatially-extended target distribution arrives at the target location, which aligns with the beam path, and begins to generate feedback in the gain medium. The oscillating optical cavity can be considered a laser with some coherent feedback if the light that reflects from the spatially-extended target distribution provides a non-scattering surface that reflects light along the beam path so that some of the reflected light returns to its original position (for example, at the reflecting optic) after each round trip. The spatial resonances for the electromagnetic field may be present in such a cavity and thus, the feedback in such a laser is used to return more of the energy or photons to the gain medium. The spatially-extended target distribution can be used in a laser produced plasma (LPP) extreme ultraviolet (EUV) light source. The spatially-extended target distribution includes a target material that emits EUV light when in a plasma state. The target material can be a target mixture that includes a target substance and impurities such as non-target particles. The target substance is the substance that is converted to a plasma state that has an emission line in the EUV range. The target substance can be, for example, a droplet of liquid or molten metal, a portion of a liquid stream, solid particles or clusters, solid particles contained within liquid droplets, a foam of target material, or solid particles contained within a portion of a liquid stream. The target substance, can be, for example, water, tin, lithium, xenon, or any material that, when converted to a plasma state, has an emission line in the EUV range. For example, the target substance can be the element tin, which can be used as pure tin (Sn); as a tin compound, for example, SnBr4, SnBr2, SnH4; as a tin alloy, for example, tin-gallium alloys, tin-indium alloys, tin-indium-gallium alloys, or any combination of these alloys. Moreover, in the situation in which there are no impurities, the target material includes only the target substance. The discussion below provides examples in which the target material is a target material droplet made of molten metal. In these examples, the target material is referred to as the target material droplet. However, the target material can take other forms. With reference to FIG. 1, a general description of an exemplary laser produced plasma (LPP) extreme ultraviolet (EUV) light source 100 in which the techniques are implemented is initially provided as background. The LPP EUV light source 100 is formed by irradiating a target mixture 114 at a target location 105 with the amplified light beam 110 that travels along a beam path toward the target mixture 114. The target location 105, which is also referred to as the irradiation site, is within an interior 107 of a vacuum chamber 130. When the amplified light beam 110 strikes the target mixture 114, a target material within the target mixture 114 is converted into a plasma state that has an element with an emission line in the EUV range. The created plasma has certain characteristics that depend on the composition of the target material within the target mixture 114. These characteristics can include the wavelength of the EUV light produced by the plasma and the type and amount of debris released from the plasma. The light source 100 also includes a target material delivery system 125 that delivers, controls, and directs the target mixture 114 in the form of liquid droplets, a liquid stream, solid particles or clusters, solid particles contained within liquid droplets or solid particles contained within a liquid stream. The target mixture 114 includes the target material such as, for example, water, tin, lithium, xenon, or any material that, when converted to a plasma state, has an emission line in the EUV range. For example, the element tin can be used as pure tin (Sn); as a tin compound, for example, SnBr4, SnBr2, SnH4; as a tin alloy, for example, tin-gallium alloys, tin-indium alloys, tin-indium-gallium alloys, or any combination of these alloys. The target mixture 114 can also include impurities such as non-target particles. Thus, in the situation in which there are no impurities, the target mixture 114 is made up of only the target material. The target mixture 114 is delivered by the target material delivery system 125 into the interior 107 of the chamber 130 and to the target location 105. The light source 100 includes a drive laser system 115 that produces the amplified light beam 110 due to a population inversion within the gain medium or mediums of the laser system 115. The light source 100 includes a beam delivery system between the laser system 115 and the target location 105, the beam delivery system including a beam transport system 120 and a focus assembly 122. The beam transport system 120 receives the amplified light beam 110 from the laser system 115, and steers and modifies the amplified light beam 110 as needed and outputs the amplified light beam 110 to the focus assembly 122. The focus assembly 122 receives the amplified light beam 110 and focuses the beam 110 to the target location 105. In some implementations, the laser system 115 can include one or more optical amplifiers, lasers, and/or lamps for providing one or more main pulses and, in some cases, one or more pre-pulses. Each optical amplifier includes a gain medium capable of optically amplifying the desired wavelength at a high gain, an excitation source, and internal optics. The optical amplifier may or may not have laser mirrors or other feedback devices that form a laser cavity. Thus, the laser system 115 produces an amplified light beam 110 due to the population inversion in the gain media of the laser amplifiers even if there are no permanent feedback devices that form a laser cavity. Moreover, the laser system 115 can produce an amplified light beam 110 that is a coherent laser beam if there is a laser cavity to provide enough feedback to the laser system 115. The term “amplified light beam” encompasses one or more of: light from the laser system 115 that is merely amplified but lacks a permanent optical feedback device and thus, may not necessarily provide coherent laser oscillation, and light from the laser system 115 that is amplified (externally or within a gain medium in the oscillator) and is also a coherent laser oscillation due to a permanent optical feedback device. The optical amplifiers in the laser system 115 can include as a gain medium a filling gas that includes CO2 and can amplify light at a wavelength of between about 9.1 μm and about 11 μm, and in particular, at about 10.6 μm, at a gain greater than or equal to 1000. In some examples, the optical amplifiers amplify light at a wavelength of 10.59 μm. Suitable amplifiers and lasers for use in the laser system 115 can include a pulsed laser device, for example, a pulsed, gas-discharge CO2 laser device producing radiation at about 9.3 μm or about 10.6 μm, for example, with DC or RF excitation, operating at relatively high power, for example, 10 kW or higher and high pulse repetition rate, for example, 50 kHz or more. The optical amplifiers in the laser system 115 can also include a cooling system such as water that can be used when operating the laser system 115 at higher powers. FIG. 2 shows a block diagram of an example drive laser system 180. The drive laser system 180 can be used as the drive laser system 115 in the source 100. The drive laser system 180 includes three power amplifiers 181, 182, and 183. Any or all of the power amplifiers 181, 182, and 183 can include internal optical elements (not shown). The power amplifiers 181, 182, and 183 each include a gain medium in which amplification occurs when pumped with an external electrical or optical source. For example, each of the power amplifiers 181, 182, 183 includes a pair of electrodes on each side of a gain medium to provide an external electrical source. Additionally, a reflective optic 112 is placed along a beam path defined between the amplifiers 181, 182, 183. Spontaneously emitted photons from within the gain media of the amplifiers 181, 182, 183 can be scattered by the spatially-extended target distribution (as discussed below) when the spatially-extended target distribution is within the target location, and at least some of these scattered photons are placed on a beam path in which they travel through each of the amplifiers 181, 182, 183. This beam path is described next. Light 184 travels between the power amplifier 181 and the power amplifier through coupling window 185 of the power amplifier 181 and a coupling window 189 of the amplifier 182 by being reflected off a pair of curved mirrors 186, 186. The light 184 also passes through a spatial filter 187. The light 184 is amplified in the power amplifier 182 and directed out of the power amplifier 182 through another coupling window 190 as light 191. The light 191 travels between the amplifier 183 and the amplifier 182 as it is reflected off fold mirrors 192 and enters and exits the amplifier 183 through a coupling window 193. The amplifier 183 amplifies the light 191 and the light 191 that exits the amplifier 183 toward the beam transport system 120 travels through coupling window 194 as an amplified light beam 195. A fold mirror 196 can be positioned to direct the amplified beam 195 upwards (out of the page) and toward the beam transport system 120. The spatial filter 187 defines an aperture 197, which can be, for example, a circular opening through which the light 184 passes. The curved mirrors 186 and 188 can be, for example, off-axis parabola mirrors with focal lengths of about 1.7 m and 2.3 m, respectively. The spatial filter 187 can be positioned such that the aperture 197 coincides with a focal point of the drive laser system 180. The example of FIG. 2 shows three power amplifiers. However, more or fewer power amplifiers can be used. Referring again to FIG. 1, the light source 100 includes a collector mirror 135 having an aperture 140 to allow the amplified light beam 110 to pass through and reach the target location 105. The collector mirror 135 can be, for example, an ellipsoidal mirror that has a primary focus at the target location 105 and a secondary focus at an intermediate location 145 (also called an intermediate focus) where the EUV light can be output from the light source 100 and can be input to, for example, an integrated circuit beam positioning system tool (not shown). The light source 100 can also include an open-ended, hollow conical shroud 150 (for example, a gas cone) that tapers toward the target location 105 from the collector mirror 135 to reduce the amount of plasma-generated debris that enters the focus assembly 122 and/or the beam transport system 120 while allowing the amplified light beam 110 to reach the target location 105. For this purpose, a gas flow can be provided in the shroud that is directed toward the target location 105. The light source 100 can also include a master controller 155 that is connected to a droplet position detection feedback system 156, a laser control system 157, and a beam control system 158. The light source 100 can include one or more target or droplet imagers 160 that provide an output indicative of the position of a droplet, for example, relative to the target location 105 and provide this output to the droplet position detection feedback system 156, which can, for example, compute a droplet position and trajectory from which a droplet position error can be computed either on a droplet by droplet basis or on average. The droplet position detection feedback system 156 thus provides the droplet position error as an input to the master controller 155. The master controller 155 can therefore provide a laser position, direction, and timing correction signal, for example, to the laser control system 157 that can be used, for example, to control the laser timing circuit and/or to the beam control system 158 to control an amplified light beam position and shaping of the beam transport system 120 to change the location and/or focal power of the beam focal spot within the chamber 130. The target material delivery system 125 includes a target material delivery control system 126 that is operable in response to a signal from the master controller 155, for example, to modify the release point of the droplets as released by a target material supply apparatus 127 to correct for errors in the droplets arriving at the desired target location 105. Additionally, the light source 100 can include a light source detector 165 that measures one or more EUV light parameters, including but not limited to, pulse energy, energy distribution as a function of wavelength, energy within a particular band of wavelengths, energy outside of a particular band of wavelengths, and angular distribution of EUV intensity and/or average power. The light source detector 165 generates a feedback signal for use by the master controller 155. The feedback signal can be, for example, indicative of the errors in parameters such as the timing and focus of the laser pulses to properly intercept the droplets in the right place and time for effective and efficient EUV light production. The light source 100 can also include a guide laser 175 that can be used to align various sections of the light source 100 or to assist in steering the amplified light beam 110 to the target location 105. In connection with the guide laser 175, the light source 100 includes a metrology system 124 that is placed within the focus assembly 122 to sample a portion of light from the guide laser 175 and the amplified light beam 110. In other implementations, the metrology system 124 is placed within the beam transport system 120. The metrology system 124 can include an optical element that samples or re-directs a subset of the light, such optical element being made out of any material that can withstand the powers of the guide laser beam and the amplified light beam 110. A beam analysis system is formed from the metrology system 124 and the master controller 155 since the master controller 155 analyzes the sampled light from the guide laser 175 and uses this information to adjust components within the focus assembly 122 through the beam control system 158. Thus, in summary, the light source 100 produces the amplified light beam 110 that is directed along the beam path when at least some of the spontaneously emitted photons on the beam path from the laser system 115 are reflected from the spatially-extended target distribution and from the reflecting optic 112 to produce more light at wavelengths within the gain band of the gain medium along the beam path to provide laser action in the laser system 115 (there is enough stimulated emission). In this way, enough energy is imparted to the target material within the spatially-extended target distribution to thereby convert the target material into plasma that emits light in the EUV range. The amplified light beam 110 operates at a particular wavelength (that is also referred to as a source wavelength) that is determined based on the design and properties of the laser system 115. At least some of the amplified light beam 110 is reflected back into the beam path off of the spatially-extended target distribution to provide feedback into the laser system 115. Referring to FIG. 3, a top plan view of an exemplary optical imaging system 300 is shown. The optical imaging system 300 includes an LPP EUV light source 305 that provides EUV light 306 to a lithography tool 310. The light source 305 can be similar to, and/or include some or all of the components of, the light source 100 of FIGS. 2A and 2B. The light source 305 includes a drive laser system 315, an optical element 322, a pre-pulse source 324, a focusing assembly 326, a vacuum chamber 340, and an EUV collecting optic 346. The EUV collecting optic 346 directs EUV light emitted from a target location 342 to the lithography tool 310. The EUV collection optic 346 can be the collector mirror 135 (FIG. 1), and the target location 342 can be at a focal point of the collection optic 346. The drive laser system 315 produces an amplified light beam 316. The drive laser system 315 can be, for example, the drive laser system 180 of FIG. 2. The pre-pulse source 324 emits a pulse of radiation 317. The pre-pulse source 324 can be, for example, a Q-switched Nd:YAG laser, and the pulse of radiation 317 can be a pulse from the Nd:YAG laser. The optical element 322 directs the amplified light beam 316 and the pulse of radiation 317 from the pre-pulse source 324 to the chamber 340. The optical element 322 is any element that can direct the amplified light beam 316 and the pulse of radiation 317 along similar paths and deliver the amplified light beam 316 and the pulse of radiation 317 to the chamber 340. The amplified light beam 316 is directed to the target location 342 in the chamber 340. The pulse of radiation 317 is directed to a location 341. The location 341 is displaced from the target location 342 in the “−x” direction. In this manner, the pulse of radiation 317 is a “pre-pulse” that can irradiate a target material droplet at a location that is physically distinct from the target location 342 at a time before it reaches the target location 342. FIG. 4 shows a side view of an exemplary light source 400 that produces EUV light. FIG. 4 shows the light source 400 at a first time, t=t1. FIGS. 5-7 show the light source 400 at later times t=t2, t=t3, and t=t4, with each time being later than the preceding time. FIGS. 4-7 show a target material droplet 405b transforming into a spatially-extended target distribution and subsequently providing more photons along the beam path that includes the gain medium to increase gain in the gain band of the gain medium. As discussed below, the light source 400 produces amplified light at wavelengths within the gain band of the gain medium 420 on a beam path 410 by forming an optical cavity between a reflective optic 412 and a spatially-extended target distribution. To create the spatially-extended target distribution, a target material droplet 405b is irradiated with a pulse of radiation 417 while the target material droplet 405b is between a target material supply apparatus 447 to a target location 442. When the formed spatially-extended target distribution arrives at the target location 442, the optical cavity (which may be non-resonant) is formed between the optic 412 and the spatially-extended target distribution. Referring to FIG. 4, the light source 400 includes the optic 412, an optical gain medium 420, a vacuum chamber 440, an EUV collection optic 446, and a target material supply apparatus 447. The light source 400 also can include one or more droplet imagers 460, and a droplet position detection feedback system 456. The target material supply apparatus 447 can be similar to the target material supply apparatus 127 (FIG. 1). The droplet imagers 460 and the droplet position detection feedback system 456 can be similar to the droplet imagers 160 and the droplet position detection feedback system 156 (FIG. 1). The position detection feedback system 456 can include an electronic processor and a tangible computer-readable medium that stores instructions that, when executed, cause the electronic processor to determine a position of a target material droplet based on information from the droplet imagers 460. At t=t1 (as shown in FIG. 4), the target material supply apparatus 447 has released the target material droplet 405b and a target material droplet 405a. The droplets 405a and 405b travel in the “x” direction toward the target location 442. The target location 442 is a location within the chamber 440 that corresponds to a focal point of the EUV collection optic 446. The target location 442 also intersects the beam path 410, which is a path along which the reflective optic 412 directs light. The beam path 410 is defined by the configuration of the optical gain medium 420 and apertures and spatial filters that may be within the arrangement of the optical gain medium 420. The optic 412 can be, for example, a partially or completely reflective mirror. The source 400 also includes the optical gain medium 420. In the example of FIG. 4, the optical gain medium 400 includes a plurality of optical amplifiers 420a, 420b, and 420c. Each of the optical amplifiers 420a, 420b, 420c includes a pair of electrodes on each side of its respective gain medium to provide an external electrical source. The amplifiers 420a, 420b, and 420c can be similar to the amplifiers 181, 182, and 183 discussed with respect to FIG. 2. The optical gain medium 420 is coupled to and partly defines the beam path 410. That is, light that reflects from the optic 412 enters and can pass through the optical gain medium 420. Spontaneously emitted photons from within the gain media of the amplifiers 420a, 420b, and 420c can exit the gain medium 420 onto and along the beam path 410. The source 400 also includes the one or more droplet imagers 460, which are connected to a droplet position detection feedback system 456. As the target material droplet 405b travels to the target location 442, the imagers 460 measure data that the droplet position detection feedback system 456 uses to determine a position of the target material droplet 405b in the “x” direction. Shortly before the target material droplet 405b reaches a location that is a distance “d” from the beam path 410 in the “−x” direction, a pulse of radiation 417 arrives at the location and irradiates the target material droplet 405b. The distance “d” is large enough to enable the irradiated target material droplet to adequately change its shape before reaching the target location 442. The distance “d” can be, for example, between about 100 μm and 200 μm, or about 120 μm. The pulse of radiation 417 can be generated from a source that is similar to the pre-pulse source 324 (FIG. 3A). In some implementations, the pulse of radiation 417 can have a wavelength of 1 micrometer (μm), a pulse duration (measured as full width at half maximum) of 10 nanoseconds (ns), and an energy of 1 mJ (milliJoule). In other implementations, the pulse of radiation 417 can have a wavelength of 1 μm, a pulse duration of 2 ns (when measured using a full width at half maximum or FWHM metric), and an energy of 0.5 mJ. In yet other implementations, the pulse of radiation 417 can have a wavelength of 1 μm, a FMHM pulse duration of 10 ns, and an energy of 0.5 mJ. The pulse of radiation 417 can have a wavelength of 1-10 μm, a FWHM duration of 10-60 ns, and an energy of 10-50 mJ. Referring to FIG. 5, the source 400 is shown at time t=t2, a time after the pulse of radiation 417 strikes the target material droplet 405b. The impact of the pulse of radiation 417 on the target material droplet 405b physically deforms the target material droplet 405b into a geometric distribution 505 that includes target material. The geometric distribution 505 can be, for example, a region of molten metal with few or no voids. The geometric distribution 505 is elongated in the “x” direction as compared to the target material droplet 405b. The geometric distribution 505 also can be thinner along the “z” direction than the target material droplet 405b. The geometric distribution 505 continues to expand in the “x” direction as it travels toward the target location 442. Referring to FIG. 6, at the time t=t3, the geometric distribution 505 has expanded into a spatially-extended target distribution 605 and is at a location just before the beam path 410 in the “−x” direction. The disk shaped target 605 arrives at the beam path axis 410 without being substantially ionized. That is, the spatially-extended target distribution 605 can be considered to be pre-formed before reaching the beam path axis 410. The spatially-extended target distribution 605 has a longitudinal extent 606 and latitudinal extent 607. The extents 606 and 607 depend on the amount of time elapsed between t=t1 (when the target material droplet 405b is struck by the pulse of radiation 417) and t=t3, as well as the pulse duration and energy of the pulse of radiation 417. The extent 606 generally increases as the amount of elapsed time increases. For an elapsed time of 2000 ns, the extent 606 can be about 80-300 μm. In comparison, a similar dimension of the target material droplet 405a is about 20-40 μm. Referring to FIG. 7, at the time t=t4, the target 605 intersects with the beam path 410 and an optical cavity 702 (represented by the solid double arrowed line) is formed between the target 605 and the optic 412. The spontaneously emitted photons on the beam path are reflected from the spatially-extended target distribution 605 and from the reflecting optic 412 to produce more light in the gain band of the gain medium 420 along the beam path 410, and if enough feedback is provided, the losses in the chain are overcome by the buildup from the feedback and all of the energy stored in the gain medium is converted into stimulated emission (to produce the amplified light beam). While the spatially-extended target distribution 602 is in the target location 442 and thus intersects the beam path 410, the amplified light beam irradiates the spatially-extended target distribution 602. In this way, enough energy is imparted to the target material within the spatially-extended target distribution to thereby convert the spatially-extended target distribution 605 into plasma that emits light in the EUV range. And, this is done without using a separate coherent light source to provide the photons to the target location. Further, because the spatially-extended target distribution 605 has a greater extent 606 than the target material droplet 408b from which the spatially-extended target distribution 605 is formed, the spatially-extended target distribution 605 reflects more light back into the optical amplifiers 420, thereby enhancing the light production within the gain band of the optical amplifiers 420. The light produced using the spatially-extended target distribution 605 to form the optical cavity 702 can generate about 2-10 times more light than would be generated with the use an unmodified target material droplet. Additionally, because the spatially-extended target distribution 605 has a smaller extent 605 in a direction along which the light beam propagates, the spatially-extended target distribution 605 is more easily converted into a plasma that emits EUV light. The relative thinness of the extent 606 means that the spatially-extended target distribution 605 presents more target material to the light beam (the thin shape allows an incident light beam to irradiate more of the target material in the spatially-extended target distribution). Consequently, more of the spatially-extended target distribution is converted to plasma. This results in greater conversion efficiency and less debris. Finally, a smaller initial target material droplet can be used because the technique of using the pulse of radiation 417 to modify the physical shape of the target material droplet 405b increases the extent 606. Using a smaller target material droplet can improve the lifetime of the light source 400. FIG. 8 shows an example of a pulsed radiation beam 802 used to deform a target material droplet and a light beam 804 that is produced using the deformed target material to form an oscillating optical cavity. The pulsed radiation beam 802 has a wavelength of 1 μm, a pulse duration of 10 ns, and an energy of 1 mJ. The light beam 804 has a duration (measured along a baseline, for example foot-to-foot) of 400-500 ns. FIG. 9 is a flow chart of an exemplary process 900 for producing an amplified light beam. The process 900 can be performed on any EUV light source that emits a pulsed radiation beam capable of deforming a target material droplet. The example process 900 is discussed with respect to the EUV light source 400. A stream of target material droplets is released from the target material supply apparatus 447 (910). The stream of target material droplets includes the target material droplets 405a and 405b. The stream of target material droplets is released or emitted toward the target location 442. The droplet position feedback system 456 may be used to determine that the droplet 405b is between the target material supply apparatus 447 and the target location 442 (920). An example of the target material droplet 405b being between the target supply apparatus 447 and the target location 442 is shown in FIG. 4. In some implementations, the target material droplet 405b is displaced about 120 μm in the “−x” direction when it is determined that the target material droplet 405b is between the target supply apparatus 447 and the target location 442. The spatially-extended target distribution 605 is produced (930). Directing the pulse of radiation 417 toward the target material droplet 405b while the droplet 405b is between the target supply apparatus 447 and the target location 442, and allowing the resulting physically deformed target material droplet to expand, produces the spatially-extended target distribution 605. As shown in FIG. 5, the interaction between the pulse of radiation 417 and the target material droplet 405b deforms the droplet into the geometric distribution 505. A finite period of time passes after the interaction, and the geometric distribution 505 elongates while moving toward the target location 442 and forms the spatially-extended target distribution 605. The pulse of radiation 417 is directed toward the target material droplet 405b before it reaches the target location 442. In this manner, the target 605 is pre-formed and not substantially ionized when it reaches the target location 442. As compared to the target material droplet 405b, the spatially-extended target distribution 605 has a greater cross-sectional diameter in a plane that faces an oncoming pulsed radiation beam. A plane that faces the oncoming pulsed radiation beam can be a plane that is transverse to the direction of propagation of the beam. In other examples, the plane can be angled relative to the direction of propagation of the pulsed radiation beam at an angle that is not transverse to the direction of propagation but still allows the spatially-extended target distribution 605 to reflect light back into the amplifier 420. The larger cross-sectional diameter allows the spatially-extended target distribution 605 to reflect more light into the amplifier 420 than the target material droplet 405b. The reflective optic 412 is positioned to reflect some of the light on the beam path 410 (940). The beam path 410 intersects the target location 442. Thus, when the spatially-extended target distribution 605 coincides with the beam path 410 in space, the spatially-extended target distribution 605 and the reflective optic form the optical cavity 702, which may be non-resonant (FIG. 7). An amplified light beam is produced between the spatially-extended target distribution 605 and the reflective optic 412 (950). The process 900 can be repeated with another target material droplet to improve the gain or amplification of the gain medium 420. The second light beam can be formed 20-40 ns after the first. In this way, a train of light pulses can be generated by repeatedly forming an optical cavity between the reflected optic 412 and spatially-extended target distribution that is formed by irradiating a target material droplet with a pulse of radiation. FIG. 10 shows another exemplary EUV light source 1000. The EUV light source 1000 is similar to the EUV light source 400, and the EUV light source 1000 physically transforms the target material droplet 405b into the spatially-extended target distribution 605 by irradiating the target material droplet 405b with the pulse of radiation 417. However, the light source 1000 includes an external laser source 1002. The external laser source 1002 supplies photons to the optical path 410 that are within the gain band of the amplifier 420. There are few ways that light from the source 1002 could be injected, such as at the other end of the chain of gain media 420, for example, through a hole in a turning mirror at the end. This light could reflect off of the spatially-extended target distribution first and then back into the laser. The EUV light source 1000 is shown at a time just before the spatially-extended target distribution 605 reaches the target location 442. When the spatially-extended target distribution 605 reaches the target location 442, additional photons that are supplied to the optical path 410 (for reflection off the distribution 605) add to the photons that are emitted by spontaneous emission from within the amplifiers 420a, 420b, and 420c. The photons from the laser source 1002 can be the same wavelength as the gain band of the amplifiers 420a, 420b, and 420c. The presence of additional photons that are amplified by the amplifiers 420a, 420b, and 420c can assist the generation of a light between the spatially-extended target distribution 605 and the reflective optic 412. For example, as compared to a similar EUV light source that lacks the laser source 1002, the light can be generated with less light reflected from the spatially-extended target distribution 605. Other implementations are within the scope of the claims. For example, the spatially-extended target distribution 605 can have a shape that varies slightly from a disk. The spatially-extended target distribution can have one or more flatted sides and/or an indented surface, for example. The spatially-extended target distribution can have a bowl-like shape. In the example shown in FIG. 3, the drive laser system 315 and the pre-pulse source 324 are shown as separate sources. However, in other implementations, it is possible that both the pulse of radiation 317 (which can be used as the pulse of radiation 417) and the amplified light beam 316 can be generated by the drive laser system 315. In such an implementation, the drive laser system 315 can include two CO2 seed laser subsystems and one amplifier. One of the seed laser subsystems can produce an amplified light beam having a wavelength of 10.26 μm, and the other seed laser subsystem can produce an amplified light beam having a wavelength of 10.59 μm. These two wavelengths can come from different lines of the CO2 laser. Both amplified light beams from the two seed laser subsystems are amplified in the same power amplifier chain and then angularly dispersed to reach different locations within the chamber 340. In one example, the amplified light beam with the wavelength of 10.26 μm is used as the pre-pulse 317, and the amplified light beam with the wavelength of 10.59 μm is used as the amplified light beam 316. In other examples, other lines of the CO2 laser, which can generate different wavelengths, can be used to generate the two amplified light beams (one of which is the pulse of radiation 317 and the other of which is the amplified light beam 316). The optical element 322 (FIG. 3) that directs the amplified light beam 316 and the pulse of radiation 317 to the chamber 340 can be any element that can direct the amplified light beam 316 and the pulse of radiation 317 along similar paths. For example, the optical element 322 can be a dichroic beamsplitter that receives the amplified light beam 316 and reflects it toward the chamber 340. The dichroic beamsplitter receives the pulse of radiation 317 and transmits the pulses toward the chamber 340. The dichroic beamsplitter can be made of, for example, diamond. In other implementations, the optical element 322 is a mirror that defines an aperture. In this implementation, the amplified light beam 316 is reflected from the mirror surface and directed toward the chamber 340, and the pulses of radiation pass through the aperture and propagate toward the chamber 340. In still other implementations, a wedge-shaped optic (for example, a prism) can be used to separate the main pulse 316, the pre-pulse 317, and the pre-pulse 318 into different angles, according to their wavelengths. The wedge-shaped optic can be used in addition to the optical element 322, or it can be used as the optical element 322. The wedge-shaped optic can be positioned just upstream (in the “−z” direction) of the focusing assembly 326. Additionally, the pulse of radiation 317 can be delivered to the chamber 340 in other ways. For example, the pulse 317 can travel through optical fibers that deliver the pulses 317 and 318 to the chamber 340 and/or the focusing assembly 326 without the use of the optical element 322 or other directing elements. In these implementations, the fiber can bring the pulse of radiation 317 directly to an interior of the chamber 340 through an opening formed in a wall of the chamber 340. |
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054426685 | summary | BACKGROUND OF THE INVENTION The present invention relates to passive decay heat removal in a pressure tube light water cooled and moderated reactor with a large power rating on the order of 1,000 MWe. A large amount of decay energy is continually generated by nuclear fuel after the reactor is shut down. Conventionally, actively pumped redundant cooling systems provide the emergency coolant to remove this decay heat under abnormal conditions such as a loss of coolant accident. Conventional reactors have the additional disadvantage that their power density profile has a maximum average power ratio or peaking factor of about, 2, even when measures are employed to minimize this peaking. SUMMARY OF THE INVENTION One object of the present invention is to insure that the fuel remains undamaged and within safety limits even in the total absence of coolant in the primary system of the reactor. Another object of the present invention is to provide a heat path to ambient air which is provided by natural heat transfer mechanisms such as, conduction, radiation and free convection and which eliminates the reliance on the availability of pumped cooling. A further object of the present invention is to considerably flatten the power density profile of the reactor across the core to increase safety margins or increase power rating. These and other features of the present invention are achieved in accordance with the present invention in one embodiment wherein a wet calandria design utilizes a fuel matrix in a pressure tube with light water as the coolant, the calandria is surrounded by a solid reflector having a mixture of a light water moderator and a void or graphite in the calandria space and arranged such that the light water moderator is always in contact with the calandria tube and provides a heat sink, and the optional use of a thermal switch in the gap between the pressure tube and the calandria tube which provides thermal insulation during normal operation to minimize heat loss and enhance heat transport across the gap during accidents. In accordance with another embodiment of the present invention wherein a dry calandria is utilized, the present invention is carried out utilizing a fuel matrix in a pressure tube with light water as the coolant, a dry calandria surrounded by a solid reflector connected by passages to a light water pool kept outside the calandria via a gas lock and self-actuated means which initiates calandria flooding by releasing the gas pressure within the calandria space during accidents which could lead to temperatures exceeding the safety limits. As a result of the above-mentioned features of the present invention, both the wet calandria and the dry calandria embodiments can survive the loss of coolant in an accident without a scram and without replenishing primary coolant, there is a super flat power density profile which considerably increases safety margins or allows operation at higher power ratings, very tight neutronic core coupling is achieved which allows for reactor control to be performed from outside the core and eliminating the possibility of local criticality and giving absolute stability towards xenon spatial oscillations, long prompt neutron lifetime in comparison with conventional light water reactors which reduces potential concerns with prompt reactivity excursions and negative coolant void coefficient is attained. With respect to the dry calandria embodiment, other advantages include the fact that it can operate in post-critical heat flux without exceeding safe fuel and matrix temperature limits and has one additional barrier in the silicon carbide matrix coating to fission product release. While all of the above-mentioned embodiments utilize water as a coolant, the wet calandria design can also be adapted to use gas coolants and both versions can employ organic coolants. These and other features and advantages of the present invention will be seen from the following detailed description of the invention taken with the attached drawings, wherein: |
053575474 | claims | 1. In a nuclear reactor pressure vessel having a plurality of fuel assemblies disposed therein and having liquid coolant flowing through the pressure vessel, a vibration dampener for dampening flow-induced vibration of an instrumentation tube adapted to extend into an associated one of the fuel assemblies and adapted to slidably receive a flux measuring instrument therein for measuring the flux in the fuel assembly, the instrumentation tube defining a longitudinal axis therethrough and sized to be received in a tubular conduit capable of guiding the instrumentation tube into the fuel assembly, the tubular conduit having an inside surface for surrounding the instrumentation tube, comprising: (a) a sleeve coaxially alignable with the instrumentation tube and connectable to the inside surface of the conduit and sized to be interposed between the conduit and the instrumentation tube, said sleeve defining a longitudinal axis therethrough, said sleeve having an interior surface for surrounding the instrumentation tube and an exterior surface for matingly engaging the inside surface of the conduit, the interior surface and the exterior surface of said sleeve defining an annular wall therebetween; (b) a pair of coaxially-aligned inwardly-directed dimples integrally attached to the interior surface of said sleeve for supporting the instrumentation tube; and (c) an elongate inwardly-directed flexible spring member disposed radially opposite said dimples and formed from a pair of parallel slots cut through the wall of said sleeve and interposed between the slots for biasing the instrumentation tube into abutting engagement with said dimples, said spring member being disposed at an angle of approximately 180.degree. with respect to said dimples, said spring member including: whereby as flow-induced vibration causes the instrumentation tube to radially outwardly translate out of coaxial alignment with said sleeve to further engage said spring member and disengage said dimples, said spring member radially outwardly flexes a predetermined distance after which the spring member radially inwardly flexes to bias the instrumentation tube into abutting relationship with said dimples so that the instrumentation tube is brought into coaxial alignment with said sleeve; and whereby as said sleeve and the instrumentation tube are brought into coaxial alignment, the vibration of the instrumentation tube is dampened. (a) a second pair of dimples disposed at an angle of approximately 90.degree. with respect to said first pair of dimples; and (b) a second pair of slots and a second spring member interposed therebetween, the second pair of slots and said second spring member disposed at an angle of approximately 180.degree. with respect to said second pair of dimples. (a) wherein the slots are cut longitudinally through the wall of said sleeve; and (b) wherein said spring member is interposed between the slots extends longitudinally in the wall of said sleeve. 2. The dampener of claim 1, further comprising: 3. The dampener of claim 1, 4. The dampener of claim 1, wherein the slots and said spring member interposed therebetween extend circumferentially in the wall of said sleeve. |
summary | ||
description | This application is a National Stage of International Application No. PCT/US2018/033145, filed May 17, 2018, which claims the benefit of U.S. Patent Application No. 62/508,836, filed May 19, 2017, the contents of which are hereby incorporated by reference. This invention was made with government support under Grant No. FA9550-15-1-0428 awarded by the Air Force Office of Scientific Research. The government has certain rights in the invention. Proposals for separating and enriching isotopes came about almost immediately after isotopes were discovered. In 1919, Lindemann and Aston examined a vast array of possible methods including fractional distillation, chemical separation, gaseous diffusion, and gravitational and centrifugal separation, along with separation of positive ions with electric and magnetic fields (1). Their early analysis concluded that isotopes “must be separable in principle though possibly not in practice.” The Manhattan Project in the 1940s ushered in large scale practical implementation of many of these techniques. Fractional distillation, gaseous diffusion and magnetic sector mass spectrometers (Calutrons) were all used on an industrial scale to enrich 23U (2, 3). Today, isotope separation and enrichment underpin advanced technologies in a wide variety of fields, including isotopic labeling in the life sciences and radioisotopes in medicine. Microelectronics may also begin to utilize isotopic enrichment as isotopically enriched materials have increased thermal conductivity and electron transport properties, as well as improved spin properties for quantum information platforms (4-6). Gaseous diffusion, distillation and gas centrifuges exhibit small isotopic separation effects which are overcome through large scale installations where many separation steps are performed in sequence. Alternatively, laser-based techniques such as atomic vapor laser isotope separation (AVLIS) and magnetically activated and guided isotope separation (MAGIS) can separate isotopes to a much higher degree, but require ionization or excitation of the target isotope (7, 8). Provided are systems and methods for separating isotopes. The systems and methods are based on supersonic beam diffraction. In one aspect, method for separating isotopes are provided. An embodiment of such a method comprises directing a supersonic beam characterized by an average velocity v and velocity distribution Δv/v, the beam comprising a first isotope and a second isotope, at a single-crystalline surface at an angle of incidence θi such that the first isotope elastically scatters from the surface with a peak angle θf1 and the second isotope elastically scatters from the surface with a peak angle θf2; and selectively collecting the scattered first isotope, the scattered second isotope, or both. In another aspect, apparatus for carrying out the methods are provided. An embodiment of such an apparatus comprises a source configured to provide a supersonic beam characterized by an average velocity v and velocity distribution Δv/v, the beam comprising a first isotope and a second isotope, wherein the apparatus is further configured to direct the supersonic beam at a single-crystalline surface at an angle of incidence θi such that the first isotope elastically scatters from the surface with a peak angle θf1 and the second isotope elastically scatters from the surface with a peak angle θf2; and a collector configured to selectively collect the scattered first isotope, the scattered second isotope, or both, as a function of angle θf, or time-of-flight, or both. Other principal features and advantages of the disclosure will become apparent to those skilled in the art upon review of the following drawings, the detailed description, and the appended claims. Provided are systems and methods for separating isotopes. The systems and methods are based on supersonic beam diffraction. The systems and methods may be used to separate isotopes, e.g., 22Ne and 20Ne, by diffracting a monovelocity supersonic beam of an isotopically mixed gas from a crystalline surface (e.g., CH3—Si(111)). The isotopes do not need to be the same element, but only need to differ in atomic mass. The relative abundances in the native mixture can be determined from the relative intensities of their respective diffraction peaks. In an embodiment, a method for separating isotopes comprises directing a supersonic beam characterized by an average velocity v and velocity distribution Δv/v, the beam comprising a first isotope and a second isotope, at a surface of a crystal at an angle of incidence θi such that the first isotope elastically scatters from the surface with a peak angle θf1 and the second isotope elastically scatters from the surface with peak angle θf2; and selectively collecting the scattered first isotope, the scattered second isotope or both. The peak angles θf1 and θf2 differ, thereby providing separation of the isotopes in space. (See FIG. 4.) In addition, at a given θf, kinematic conditions dictate that the isotopes of interest must have different velocities (see FIG. 6), therefore providing separation of the isotopes in time. In other words, the method achieves both angular and temporal separation. Such separation is the basis of the selective collection of the isotopes as further described below. The supersonic beam may be generated by a variety of atomic or molecular beam sources configured to supersonically expand a gas through a variable temperature nozzle. The gas comprises the first and second isotope. By “first and second isotope” it is meant elements which differ in atomic mass. The first and second isotope could be, but need not be, the same element. The gas may comprise additional isotopes, each isotope of which may be separated using the method. The composition of the gas and the type of isotopes are not particularly limited. However, as atomic mass decreases, the mass separation between the target isotopes increases. Thus, the method is particularly suitable for elements or molecules having an atomic mass of 50 AMU or less. This includes elements or molecules having an atomic mass of 40 AMU or less. Illustrative examples include isotopes of hydrogen (1H, 2H, and 3H), helium (3He and 4He), lithium (6Li and 7Li), boron (10B and 11B), carbon (12C, 13C, and 14C), nitrogen (14N and 15N), oxygen (16O, 17O, and 18O), magnesium (24Mg, 25Mg, and 26Mg), silicon (28Si, 29Si, and 30Si), sulfur (32S, 35S, 34S, and 36S), chlorine (35Cl and 37Cl), argon (36Ar, 38Ar, and 40Ar) and potassium (39K, 40K, and 41K). Illustrative examples of low molecular mass molecules include 1H1H, 2H2H, and 1H2H; 12CH4, 13CH4, and 14CH4; 14N14N and 15N15N; 16O16O, 18O18O, and 16O18O; 6Li1H and 7Li1H; 35Cl1H and 37Cl1H; 14N1H3 and 15N1H3; 1H19F and 2H19F; 28Si1H4, 29Si1H4, and 30Si1H4. The velocity distribution of the beam may be selected (e.g., minimized) to increase (e.g., maximize) the separation of the isotopes. In embodiments, the velocity distribution is no more than 1%, no more than 3%, no more than 5%, no more than 7%, no more than 9%, or in the range of from 1% to 10%. Similarly, the average velocity of the beam may be selected (e.g., minimized) to increase (e.g., maximize) the separation of the isotopes. The selected average velocity will depend upon the isotopes to be separated as well as the crystal surface. However, by way of illustration, in embodiments involving the separation of 20Ne and 2Ne, suitable average velocities include those of no more than 360 m/s, no more than 400 m/s, no more than 425 m/s, no more than 450 m/s, no more than 475 m/s, or in the range of from 360 m/s to 510 m/s. Adjustment of the velocity distribution and average velocity may be accomplished by beam seeding and/or in-line velocity selectors. The supersonic beam may be pulsed, e.g., via mechanical chopping. A wide variety of crystals may be used. The crystal is single-crystalline by which it is meant that the extended crystal lattice of the solid is substantially continuous and substantially unbroken with few or substantially no grain boundaries. The crystal may be characterized by its surface atom spacing. The surface atom spacing may be selected (e.g., minimized) in order to increase (e.g., maximize) the separation of the isotopes. In embodiments, the surface atom spacing is no more than 2.5 Å, no more than 3.5 Å, no more than 4.5 Å, no more than 5.5 Å, no more than 7.0 Å, or in the range of 2.5 Å to 7.5 Å. The crystal may be characterized by its Debye temperature, or stiffness. The Debye temperature may be selected (e.g., maximized) in order to increase (e.g., maximize) the flux of separated isotopes. In embodiments, the Debye temperature is at least 100 K, at least 250 K, at least 500 K, at least 750 K, at least 950 K, or in the range of from 100 K to 1000 K. Illustrative crystals include diamond, hydrogen terminated diamond, graphite, graphene, CH3—Si(111), CH3—Ge(111), LiF, NaCl, GaAs, Ni, Pt, and Au as well as, for example, O or H covered crystalline metallic surfaces, and crystallized self-assembled molecular interfaces including self-assembled alkane thiols. The selection of the crystal is determined, at least in part, so that it is inert with respect to the isotopes to be separated. By way of illustration, NaCl or LiF are suitable crystals for separating isotopes of O2, while graphite is not since graphite and atomic oxygen react. The crystals NaCl or LiF will also work in separating isotopes of HCl and HF. Selective collection of separated isotopes may be carried out as follows. In one embodiment, for collecting diffracted beams emerging from the surface of the crystal at θf1 and θ2, two adjacent apertures are used. These two apertures are placed at each diffraction angle θf1, θf2 to capture each of the diffracted beams, respectively. The width of each aperture may be selected to maximize the collection of the desired isotope over the other isotopes. Each aperture then leads to a separate vacuum chamber that is pumped by a high vacuum pump where the exhaust of each high vacuum pump contains each diffracted beam's contents. In an alternative embodiment, a method of collecting the diffracted beams involves two cryogenically cooled surfaces placed at θf1 and θf2, respectively. Isotopes, e.g., neon atoms, striking a surface below 7 Kelvin will condense and remain frozen on the surface. As the collection surfaces are filled with condensed isotopes they may periodically be warmed up to release the trapped isotopes. Thus, selective collection involves some surface which receives a diffracted beam containing the separated isotope of interest, e.g., an aperture coupled to a vacuum chamber or a cryogenically cooled surface. The surface that receives the diffracted beam may also be an active surface of a detector such as the mass spectrometer detector described in the Example, below. Thus, any of these surfaces and similar surfaces may be referred to as “collection surfaces” and the assembly of components including the collection surfaces may be referred to as a “collector,” as further described below. Also, in embodiments, “collection” encompasses “detection.” The phrase “selective collection” is used in reference to the fact that different isotopes in a mixture can be selectively collected (i.e., collected with specificity) from the mixture since the present method achieves separation in both space and time, as described above. In turn, the collection conditions for each isotope in a mixture of isotopes can be separately optimized so as to improve (e.g., maximize) collection of each isotope at its optimized condition. These “conditions” can include collection at a specific θf as described above. These conditions can also refer to collection over an angular width (or aperture width) and/or collection of isotopes at a specific time-of-flight or range of times-of-flight. The ability to achieve selective collection by the present method can be quantified via an enrichment factor, e.g., the factor by which a desired isotope is enriched in the material collected. By way of illustration, the Example below finds that the enrichment factor for detecting 22Ne in a neon beam containing 22Ne and 20Ne at their natural abundances is about 3.5. In other embodiments, the method may be characterized by the ability to achieve an enrichment factor for a selected isotope of at least 1.01, at least 2, at least 3, at least 5, at least 7, at least 9, or at least 10. The method may be characterized by the ability to achieve an angular resolution, i.e., Δ(θf1−θf2), of at least 0.1°, at least 0.5°, at least 1.0°, at least 1.5°, or at least 2°. The methods may be carried out using an apparatus comprising a source of the supersonic beam, components configured to mount the crystal and direct the beam as described above, and a collector. (See FIG. 1B.) The collector comprises any of the collection surfaces described above. Thus, the collector could be an assembly comprising the apertures/vacuum chambers, an assembly comprising the cryogenically cooled surface, or a mass spectrometer as described above. Such an apparatus may further comprise any of the components typically associated with atomic or molecular beam sources. This example uses supersonic beam diffraction as an isotope separation technique. Among isotopic separation methods, supersonic beam diffraction has the unique combination of being a non-ionizing/dissociative process that can achieve high separation effects. This high degree of separation is only achievable via the narrow velocity distribution of a supersonic beam, which translates into a narrow angular distribution that is scattered from a highly periodic surface. While effusive beam sources have been used for atomic and molecular diffraction since pioneering experiments in the 1930s, a very small percentage of the beam flux is within a few percent of the mean beam velocity (9), preventing any meaningful degree of isotopic purification by atomic diffraction. In contrast, the advent of supersonic nozzle sources with high Mach numbers affords considerably smaller velocity distributions—here, as low as Δv/v˜6%. Such narrow velocity distributions, when coupled with a high-quality, high Debye temperature surface, make separation of atomic isotopes via atomic diffraction feasible. The existence of isotopically unique diffraction channels for neon scattering from LiF(001) has been contemplated, but the feature was not resolved (10), nor has there been any consideration of either enrichment or separation, nor of time separation. Here, the separation of the 20Ne and 22Ne isotopes via atomic diffraction is observed for the first time when a neon beam with a natural abundance of each isotope is scattered from a methyl-terminated Si(111) surface as shown schematically in FIG. 1A. When paired with the extreme sensitivity of scattered angle with respect to the mass differences of the incident atoms, diffraction experiments may be used as an isotopic separation technique. Methods The ultra-high vacuum (UHV) scattering apparatus required for this experiment is illustrated in FIG. 1B, and has been described in greater detail elsewhere (11). Briefly, it is comprised of three primary sections: a differentially pumped beam source, a UHV chamber that houses the crystal, and a rotatable mass spectrometer detector. A natural abundance (90.48% 20Ne and 9.25% 22Ne) neon beam with a narrow energy distribution is generated by supersonically expanding ultra-high purity Ne gas through a 15 μm diameter nozzle source which is cooled by a closed-cycle helium refrigerator. The incident energy distribution of this beam is measured with an in-line mass spectrometer and is minimized to Δv/v˜6% by adjusting the backing pressure of Ne. Similarly, the beam energy, which is determined by the nozzle temperature, is optimized to 50 K (EB˜10 meV) in order to limit the incident energy while avoiding the formation of clusters. For diffraction and time-of-flight measurements, a pre-collision chopper is used to modulate the beam with a duty cycle of 50%; the time-of-flight measurements are performed by modulating the beam with a pseudorandom chopping sequence for cross-correlation analysis (12). The spatial profile of the beam is minimized by collimation through a series of apertures, resulting in a 4 mm spot size on the crystal (chopper-to-crystal distance=0.4996 m). After the collision with the surface, which is mounted on a six-axis manipulator in order to control the incidence angle (θi), azimuth (φ), and tilt (χ) of the crystal, the neon atoms travel along a 0.5782 m (crystal-to-ionizer distance) triply differentially pumped rotatable detector arm with an angular resolution of 0.29° FWHM, are ionized by electron bombardment, and then pass through a quadrupole mass spectrometer (QMS) before striking an electron multiplier. The QMS is adjusted to selectively filter either the 20Ne or 22Ne isotope. The angular distributions for diffraction scans are obtained by scanning the detector at 0.1° increments over a range of 35°, all while holding the incident angle at a fixed value. Between scattering experiments, the temperature of the crystal was flashed to 200 K to eliminate unwanted surface adsorbates and maximize elastic scattered intensity. The crystal used for the isotopic separation by diffraction, CH3—Si(111), was created by the Lewis group at the California Institute of Technology (13), and shipped under argon to the University of Chicago for the neon scattering experiments. This crystal was chosen for its relatively small surface atom spacing (3.82 Å), the relatively high surface Debye temperature which limits diffusive scattering, and the high quality and long-range periodicity achieved in the synthesis of the crystal, which is described in greater detail elsewhere (14-16). Results and Discussion When molecules elastically scatter from a surface, they can undergo a discrete exchange of parallel momentum ΔK with the surface, as governed by the equationΔK=ki(sin(θf)−sin(θi)), (1)where ki is the incident wavevector of the beam, and θi and θf are, respectively, the incident and final scattered angles of the molecular beam as measured from the surface normal. This condition for elastic diffraction is met when the change in parallel momentum is equal to a sum of the reciprocal lattice vectors b, according to the equationΔK=h+k. (2) As is evident from equation (1), the angular location of a diffraction peak is determined in part by its incident wavevector (ki), which in turn is dependent on the velocity of the incident beam. A measured time-of-flight and velocity distribution for supersonic neon is shown in FIGS. 2B and 2A, respectively. For an elastic gas-surface interaction, the incident velocity distribution of the molecular beam can be transformed into a theoretical angular distribution of the scattered beam through the implementation of equation (1). FIGS. 3A-3B show the predicted angular distribution of Ne scattered from CH3—Si(111) for both an effusive (FIG. 3B) and a supersonic molecular beam (FIG. 3A). Experimental angle scans of the (11) diffraction peak for 20Ne and 22Ne are shown in FIG. 4 and can be considered a figure of merit for the feasibility of separating isotopes in a supersonic beam via diffraction. These two spectra were recorded under identical incident neon beam and surface conditions and illustrate the angular separation of the isotopes observed for the (11) diffraction peak. The peak intensities observed in FIG. 4 can be used to quantify the enrichment capability of this isotope separation technique; for these experimental conditions, a collector with an angular width of 0.67° positioned at the maximum of the 22Ne diffraction peak yields neon with an abundance of 67.6% 20Ne and 32.4% 22Ne (an enrichment factor of 3.50±0.30 for 22Ne; i.e., (abundance of 22Ne in detected signal)/(natural abundance of 22Ne)=32.4%/9.25%˜3.5)). That is, the detected signal when the collector is at the maximum of the 22Ne diffraction peak (with angular width of 0.67°) is effectively enriched in 22Ne by a factor of about 3.5. While the partial overlap of the isotopes indicates that the angular separation of 20Ne and 22Ne is incomplete, FIG. 5 shows helium and neon scattering under similar incidence conditions and demonstrates that the angular width of the observed (11) diffraction peaks is a result of the velocity spread of the incident beam and not the instrumental resolution or surface quality. Because the final angle and momentum of a diffracted species are directly dependent on its incident momentum, the efficiency of isotopic separation via diffraction is limited by the velocity spread of the atomic beam, which in turn is a function of the source in which the beam is created. While a common method for narrowing the velocity distribution is seeding the beam with a light gas (e.g. He, H2) (15, 17, 18), the increased average velocity of this mixture would bring the angular positions of the diffraction peaks closer together, limiting the degree of separation, as predicted by equation (1). However, a more straightforward solution would be the addition of an in-line pre-collision velocity selector which would directly lead to more complete angular separation of the two isotopes (19). Velocity selection techniques can also be implemented after the atoms collide with the surface. For a given θf at which there is angular overlap between the 20Ne and 22Ne non-zeroth order diffraction peaks, the two isotopes will necessarily have different velocities, as required by equation (1). This is demonstrated in FIG. 6, which shows time-of-flight spectra for both isotopes at the midway point between their (11) diffraction peak maxima. The pronounced difference in arrival time between the two isotopes opens up the possibility for complete isotopic separation mediated by velocity selection techniques. The practical throughput of diffractive isotope separation can be maximized by thoughtful consideration of the incidence parameters and the choice of diffracting surface. As established by equation (1), the angles at which atoms will scatter from a surface depend upon the incident wavevector of the atomic beam (ki) and the spacing between diffraction peaks (ΔK), which is in turn dependent on the real-space distance between atoms at the surface. The incident flux of an atomic beam can be concentrated into a smaller number of accessible diffraction channels by lowering the incident wavevector/beam velocity (e.g. by seeding in a heavier gas such as xenon) or increasing the angular spread between diffraction peaks by choosing a surface with a smaller lattice parameter, such as graphite (lattice constant=2.46 Å). The choice of surface can also affect the relative flux scattered into various diffraction channels. Higher ratios of scattered intensity between non-zeroth order diffraction and specular peaks have been demonstrated to be correlated with increased surface corrugation (10, 20-22). Additionally, the amount of flux that is scattered diffusely from a surface is strongly affected by the surface hardness, which is quantified by the surface Debye temperature (14). When gases diffract from surfaces with high Debye temperatures, less of the incident flux is scattered into diffuse elastic channels due to the Debye-Waller effect than for soft surfaces, resulting in a more directed channeling of the incident beam into diffraction peaks. The angular and temporal separation effects of supersonic molecular beam diffraction provide a promising isotope enrichment method that does not require ionization or laser excitation of the target isotope. The necessity of a supersonic expansion for this technique is demonstrated, and as a proof of concept natural abundance neon has been shown to diffract into separate, isotopically dependent diffraction lab frame angles. The experimental set-up may be adjusted to achieve maximum separation and throughput, with the velocity spread of the incident beam serving as the most determining factor in thorough separation. As atomic diffraction has been observed for species with masses as high as 50 amu(23), this isotope separation technique is applicable to a wide range of co-expanded atoms and molecules. Overall, separation of atoms and molecules into isotopically pure diffraction channels is an interesting and novel application of supersonic molecular beam assemblies. 1. F. A. Lindemann, F. W. Aston, XLVIII. The possibility of separating isotopes. Philos. Mag. Ser. 6. 37, 523-534 (1919). 2. L. O. Love, Electromagnetic Separation of Isotopes at Oak Ridge: An informal account of history, techniques, and accomplishments. Science. 182, 343-52 (1973). 3. A. L. Yergey, A. K. Yergey, Preparative scale mass spectrometry: A brief history of the calutron. J. Am. Soc. Mass Spectrom. 8, 943-953 (1997). 4. T. Ruf et al., Thermal conductivity of isotopically enriched silicon. Solid State Commun. 115, 243-247 (2000). 5. G. Balasubramanian et al., Ultralong spin coherence time in isotopically engineered diamond. Nat. Mater. 8, 383-387 (2009). 6. K. M. Itoh, H. Watanabe, Isotope engineering of silicon and diamond for quantum computing and sensing applications. MRS Commun. 4, 143-157 (2014). 7. T. R. Mazur, B. Klappauf, M. G. Raizen, Demonstration of Magnetically Activated and Guided Isotope Separation. Nat. Phys. 10, 601-605 (2014). 8. P. A. Bokhan et al., Laser Isotope Separation in Atomic Vapor (2006). 9. J. B. Anderson, R. P. Andres, J. B. Fenn, Supersonic nozzle beams. Adv. Chem. Phys. 10, 275-317 (1966). 10. G. Boato, P. Cantini, L. Mattera, A study of the (001)LiF surface at 80 K by means of diffractive scattering of He and Ne atoms at thermal energies. Surf Sci. 55, 141-178 (1976). 11. B. Gans, P. A. Knipp, D. D. Koleske, S. J. Sibener, Surface dynamics of ordered Cu3Au(001) studied by elastic and inelastic helium atom scattering. Surf Sci. 264, 81-94 (1992). 12. D. D. Koleske, S. J. Sibener, Generation of pseudorandom sequences for use in cross-correlation modulation. Rev. Sci. Instrum. 63, 3852 (1992). 13. N. T. Plymale, Y. G. Kim, M. P. Soriaga, B. S. Brunschwig, N. S. Lewis, Synthesis, Characterization, and Reactivity of Ethynyl- and Propynyl-Terminated Si(111) Surfaces. J. Phys. Chem. C. 119, 19847-19862 (2015). 14. J. S. Becker, R. D. Brown, E. Johansson, N. S. Lewis, S. J. Sibener, Helium atom diffraction measurements of the surface structure and vibrational dynamics of CH(3)-Si(111) and CD(3)-Si(111) surfaces. J. Chem. Phys. 133, 104705 (2010). 15. K. J. Nihill et al., Experimental and theoretical study of rotationally inelastic diffraction of H2(D2) from methyl-terminated Si(111). J. Chem. Phys. 145, 84705 (2016). 16. H. Yu et al., Low-temperature STM images of methyl-terminated Si(111) surfaces. J. Phys. Chem. B. 109, 671-674 (2005). 17. N. Isa, K. D. Gibson, T. Yan, W. Hase, S. J. Sibener, Experimental and simulation study of neon collision dynamics with a 1-decanethiol monolayer. J. Chem. Phys. 120, 2417-2433 (2004). 18. G. Scoles, Atomic and Molecular Beam Methods Volume I (Oxford University Press, New York, 1988). 19. C. Szewc, J. D. Collier, H. Ulbricht, Note: A helical velocity selector for continuous molecular beams. Rev. Sci. Instrum. 81, 10-13 (2010). 20. A. Politano et al., Helium reflectivity and Debye temperature of graphene grown epitaxially on Ru(0001). Phys. Rev. B. 84, 35450 (2011). 21. K.-H. Rieder, W. Stocker, Observation of Pronounced Neon Diffraction from Low-Index Metal Surfaces. Phys. Rev. Lett. 52, 352-355 (1984). 22. M. W. Cole, D. R. Frankl, Atomic and molecular beam scattering from crystal surfaces in the quantum regime. Surf. Sci. 70, 585-616 (1978). 23. M. Minniti et al., Helium, neon and argon diffraction from Ru(0001). J. Phys. Condens. Matter. 24, 354002 (2012). 21. M. Minniti, C. Diaz, J. L. Fernandez Cunado, A. Politano, D. Maccariello, F. Martin, D. Farias, and R. Miranda, J. Phys. Condens. Matter 24, (2012). The word “illustrative” is used herein to mean serving as an example, instance, or illustration. Any aspect or design described herein as “illustrative” is not necessarily to be construed as preferred or advantageous over other aspects or designs. Further, for the purposes of this disclosure and unless otherwise specified, “a” or “an” means “one or more”. The foregoing description of illustrative embodiments of the disclosure has been presented for purposes of illustration and of description. It is not intended to be exhaustive or to limit the disclosure to the precise form disclosed, and modifications and variations are possible in light of the above teachings or may be acquired from practice of the disclosure. The embodiments were chosen and described in order to explain the principles of the disclosure and as practical applications of the disclosure to enable one skilled in the art to utilize the disclosure in various embodiments and with various modifications as suited to the particular use contemplated. It is intended that the scope of the disclosure be defined by the claims appended hereto and their equivalents. |
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abstract | An image processing apparatus includes a display configured to display a medical image; an input unit configured to receive n (n being an integer equal to or greater than three) number of input points with respect to the displayed medical image; and a controller configured to set a window in the medical image based on an area in a shape of a polygon, the area being defined by the input points, and to perform image processing of reducing at least one of brightness and definition of the medical image in a remaining area except for an area of the window. |
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048067679 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS Now, the invention will be described in detail in conjunction with an illustrative embodiment by referring to FIGS. 3 and 4. FIG. 3 is a sectional view of an electron lens assembly according to an embodiment of the invention. Parenthetically, it is to be noted that like elements or parts are attached with like reference characters throughout all the drawings, and description is not repeated, unless it is necessary. Referring to FIG. 3, a lower yoke member 2 and an electron beam passage defining pipe 9' are welded together by way of a non-magnetic member (interconnecting member) 20 by Heliarc or the like welding at portions where the yoke member 2 and the pipe 9' are located close to each other, as indicated by 21 and 22, whereby the lower yoke member 2 and the pipe 9' are implemented in an integral structure. The non-magnetic member 20 may be implemented in a sufficient thickness to prevent penetration of gases therethrough (e.g. thickness of about 1 to 1.5 mm). On the other hand, an O-ring 13' which is disposed between the lower yoke member 2 and the spacer 10' for the purpose of realizing a vacuum tight sealing is constituted by a metal O-ring. In this connection, it should be mentioned that when a metal gasket such as of copper, aluminum or the like is employed, a clamping force of very great magnitude is required and besides an additional clamping must be performed later on in view of poor resiliency. In contrast, the metal O-ring exhibits a high resiliency and requires no additional clamping, to advantage. FIG. 4 shows schematically in a fragmental perspective view a exemplary structure of the metal O-ring suited advantageously for the use in the electron lens assembly according to the present invention. As will be seen in the figure, the metal O-ring is constituted by a cylindrical pipe 28 of aluminum having a slot-like cut-out of notch extending longitudinally of the pipe. A helical spring 29 of stainless steel is disposed within the cylindrical pipe 28. (Parenthetically, this type of metal O-ring is commercially available from Cefilac Company of France and Usui Kokusai Sangyo K.K. of Japan under designation "Helicoflex".) Turning back to FIG. 3, reference numeral 24 denotes a metal flanged pipe, and 11" denotes a transverse bore for mounting the objective aperture, cold fingers and others. In the case of the illustrative embodiment under consideration, the objective aperture or other member is mounted on the flange of the flanged pipe 24 with an O-ring or the like being interposed while the metal flanged pipe 24 is welded to the spacer 10 by Heliarc or the like welding as indicated by 25 for maintaining the vacuum. An O-ring 12' for sealing vacuum tightly the connection between the upper yoke member 1 and a portion of the column (not shown) disposed on the yoke member 1 is also constituted by a metal O-ring similarly to the O-ring 13'. On the other hand, the electron beam passage defining pipe 9' is realized in an integral structure with the lower yoke member 2. In that case, if the pipe 9' projects beyond the bottom surface of the lower yoke member 2, inconvenience will be involved in the handling. Accordingly, in the case of the illustrative embodiment, the electron beam passage defining pipe 9' is so implemented as not to project beyond the bottom surface of the lower yoke member 2, wherein the lower end of the pipe 9' is coupled to a lower spacer 27 with an O-ring 26 being interposed therebetween. For the purpose of degassing by heating, there is disposed a heater 18 at a location adjacent to the pipe 9', while another heater 19 is disposed on the bottom side of the spacer 10, as shown in FIG. 3. Additionally, cooling pipes 23 and 23' are disposed above and below the coil 3 for dissipating heat generated by the coil 3 when it is electrically energized. With the structure described above, the vacuum state within the specimen chamber in which a specimen 8 is disposed can be sustained satisfactorily by means of the welds formed by Heliarc or the like welding and the metal O-rings. Since the metal O-ring can well withstand a high temperature of 200.degree. C. or higher, it is possible to produce the vacuum on the order of 10.sup.-10 Torr. It should be mentioned that a Xe-lamp or the like may be disposed within the space adjacent to the spacer 10 for degassing purpose in addition to the degassing heaters 18 and 19. Further, the Heliarc welding for realizing the vacuum tight seal may be replaced by other method such as oven blazing or other. In particular, the bonding between the non-magnetic member 20 and the lower yoke member 2 can be advantageously realized by the oven blazing method according to which a blazing material such as silver or the like is disposed in a ring-like pattern on the portions to be bonded and heated subsequently within an oven filled with an inert gas such as argon or the like. In a modification, the electron beam passage defining pipe 9' may be formed integrally with the non-magnetic member 20 and welded to the lower yoke member 2 at the portion indicated by 22. Alternatively, the electron beam passage defining pipe 9' may be directly welded to the lower yoke 2 without resorting to the use of the interconnecting non-magnetic member 20. The electron lens assembly of the structure described above can well withstand the heating at a temperature of about 200.degree. C. and allows the ultra-high vacuum to be easily produced. |
claims | 1. A fuel assembly comprising: a plurality of fuel rods arranged in a square lattice array, said fuel rods including a plurality of short-length fuel rods each having a fuel active length shorter than that of said fuel rods which are non-short-length fuel rods; at least one water rod arranged in a region in which at least one of said fuel rods are arrangeable in said array; and a plurality of fuel spacers, provided at a plurality of positions in the axial direction, for holding said plurality of fuel rods and said at least one water rod with mutual radial intervals therebetween kept immovable; wherein said plurality of short-length fuel rods include at least one first short-length fuel rod arranged in the outermost peripheral region of said square lattice array, and at least one second short-length fuel rod arranged at a lattice position adjacent to said at least one water rod; each of said plurality of fuel spacers including a plurality of cylindrical members which are connected to each other and in which said fuel rods are to be inserted respectively, and first loop-shaped springs each of which is provided at a joined portion between a pair of adjacent cylindrical members for pressing two of said fuel rods inserted in said adjacent cylindrical members; said plurality of fuel spacers including first fuel spacers positioned above the upper ends of said first and second short-length fuel rods, and at least one of said first fuel spacers is configured such that cylindrical members are omitted at locations of first and second lattice positions associated with said first and second short-length fuel rods; each of said plurality of cylindrical members located at lattice positions adjacent to said second lattice position has on said second lattice position side a second loop-shaped spring for pressing said fuel rod in said cylindrical member; said plurality of second loop-shaped springs located at said lattice positions adjacent to said second lattice position are supported by a spring pressing member provided at said second lattice position, said spring pressing member having a transverse cross-section smaller than a transverse cross-section of said cylindrical member; and said spring pressing member in said at least one first fuel spacer includes a plurality of spring holding portions which are inserted in loops of said plurality of second loop-shaped springs for holding said plurality of second loop-shaped springs, respectively, and a plurality of spring pressing portions which are brought in contact with the loops of said plurality of second loop-shaped springs from an outer peripheral side for supporting said plurality of second loop-shaped springs such that said plurality of second loop-shaped springs generate pressing forces applied to associated fuel rods, respectively. 2. A fuel assembly according to claim 1 , wherein said plurality of spring holding portions include a plurality of spring holding projecting pieces and said plurality of spring pressing portions include a plurality of spring pressing projecting pieces; and claim 1 said plurality of spring holding projecting pieces all project in one direction, and at least one of said plurality of spring pressing projecting pieces projects in a direction opposite to said one direction. 3. A fuel assembly according to claim 1 , wherein said plurality of spring holding portions include a plurality of spring holding projecting pieces and said plurality of spring pressing portions include a plurality of spring pressing projecting pieces; and claim 1 at least one of said plurality of spring pressing projecting pieces is configured such that the leading end thereof is connected to a portion opposed to the leading end of said spring pressing projecting piece of a base plate portion of said spring pressing member. 4. A fuel assembly according to claim 1 , wherein said at least one first fuel spacer further includes a water rod holding member for holding said at least one water rod in the radial direction; and said spring pressing member is joined to said water rod holding member. claim 1 5. A fuel assembly according to claim 1 , wherein said spring pressing member of said at least one first fuel spacer serves as a water rod holding member for holding said at least one water rod in the radial direction. claim 1 6. A fuel assembly according to claim 1 , wherein said plurality of second loop-shaped springs are provided on said second lattice sides of two of said plurality of cylindrical members which are adjacent to each other on said second lattice position in the row direction and column direction. claim 1 7. A fuel assembly comprising: a plurality of fuel rods arranged in a square lattice array, said fuel rods including a plurality of short-length fuel rods each having a fuel active length shorter than that of said fuel rods which are non-short-length fuel rods; at least one water rod arranged in a region in which at least one of said fuel rods are arranged in said array; and a plurality of fuel spacers, provided at a plurality of positions in the axial direction, for holding said plurality of fuel rods and said at least one water rod with mutual radial intervals therebetween kept immovable; wherein said plurality of short-length fuel rods include at least one first short-length fuel rod arranged in the outermost peripheral region of said square lattice array; each of said plurality of fuel spacers including a plurality of cylindrical members which are connected to each other and in which said fuel rods are to be inserted respectively, a band member for surrounding the outermost peripheries of said plurality of cylindrical members, and first loop-shaped springs each of which is arranged at a joined portion between a pair of the adjacent ones of said plurality of cylindrical members for pressing two of said fuel rods inserted in said adjacent cylindrical members; said plurality of fuel spacers including first-fuel spacers positioned above the upper end of said at least one first short-length fuel rod and at least one of said first fuel spacers is configured such that at least one of said cylindrical members, located at at least one of a first lattice position associated with said at least one first short-length fuel rod is omitted, and, instead, at least one of: (a) a first supporting member for connecting two first cylindrical members of said plurality of cylindrical members, adjacently located on both sides of said first lattice position in the outermost peripheral region adjacent to said band member, is provided at said first lattice position, (b) a second supporting member for connecting two first cylindrical members, of said plurality of cylindrical members, adjacently located on both sides of said first lattice position in the outermost peripheral region, to a second cylindrical member, of said plurality of cylindrical members, located inwardly from and adjacently to said first lattice position, is provided at said first lattice position, and (c) said plurality of short-length fuel rods include at least one second short-length fuel rod arranged at at least one of a second lattice position adjacent to said at least one water rod, said first fuel spacers being positioned above the upper ends of said first and second short-length fuel rods, said at least one of said first fuel spacers being configured such that said plurality of cylindrical members located at said first and second lattice positions are omitted, each of said plurality of cylindrical members located at lattice positions adjacent to said second lattice position has on said second lattice position side a second loop-shaped spring for pressing said fuel rods in said cylindrical member, said plurality of second loop- shaped springs located at said lattice position adjacent to said second lattice position are supported by a spring pressing member provided at said second lattice position, said spring pressing member having a transverse cross-section smaller than a transverse cross-section of said cylindrical member, and said spring pressing member in said at least one first fuel spacer includes a plurality of spring holding portions which are inserted in loops of said plurality of second loop-shaped springs for holding said plurality of second loop-shaped springs, respectively, and a plurality of spring pressing portions which are brought in contact with the loops of said plurality of second loop-shaped springs from an outer peripheral side for supporting said plurality of second loop-shaped springs such that said plurality of second loop-shaped springs generate pressing forces applied to associated fuel rods, respectively. |
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048790907 | abstract | Integral vanes 4 with optimized size, shape and bend angles maximize coolant mixing and fuel rod 12 heat transfer downstream. Recessed weld nuggets 5 with no vane cutout are optimized for size, strength and corrosion resistance. Staggered arches 20',22' and springs 20,22 minimize turbulence and reduce grid pressure drop and promote coolant mixing. Crowned arches 20',22' and springs 20,22 decrease scoring of fuel rods 12 and are sized to minimize turbulence and pressure drop. Minimum cutouts in unslotted section of grid strip 46,46' give 15-20% strength increase. Intermediate weld and tapered end slots 48 give 15% strength increase. Ribbed and round dimple stiffeners 38,39 on outer strips increase buckling resistance, reduce handling damage and spreads accidental loading. Outer strips 32 are optimized for strength, handling, turbulence generation and pressure drop. They also divert enough flow to interior of fuel assembly to match thermal power distribution and eliminate fuel rod corrosion concerns. |
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abstract | The method for filling water into and changing the air of a main circuit of a water-cooled nuclear reactor includes a step of placing a connection and fluid isolation device which is connected to a hot leg of each cooling loop of the main circuit so as to substantially insulate, from inside the vessel, the assembly of hot legs. The method also includes a step of injecting water through an injection circuit on at least one hot leg until each cooling loop is filled with water having changed the air from a steam generator and until the water level in the vessel reaches above the side openings of the vessel that correspond to the loops, after which the connecting device is taken out of the vessel. The connecting device is capable of using telescopic connection elements. |
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059149949 | claims | 1. A storage basket, comprising: a plurality of inserts for receiving a fuel element of a nuclear power plant, said inserts defining a cruciform gap for receiving a control rod. a carrying structure for carrying fuel elements and control rods; and a storage basket having a plurality of inserts for receiving a fuel element, said inserts defining a cruciform gap for receiving a control rod. placing carrying wells on a carrying structure in a fuel element storage rack; and introducing, in a respective one of the carrying wells, both a control rod and a storage basket having a plurality of inserts, the inserts defining a cruciform gap for receiving the control rod; and introducing at least one fuel element in the storage basket. a plurality of inserts for receiving a fuel element of a nuclear power plant, said inserts defining a cruciform gap for receiving a control rod; and a supporting element for enclosing a foot part of the control rod and for supporting said inserts on a carrying structure of a fuel element storage rack, said supporting element having a device for fastening said inserts. a plurality of inserts for receiving a fuel element of a nuclear power plant, said inserts defining a cruciform gap for receiving a control rod; and a supporting element for enclosing a foot part of the control rod and for supporting said inserts on a carrying structure of a fuel element storage rack, said supporting element having a device for securing said inserts against rotation. a plurality of inserts for receiving a fuel element of a nuclear power plant, said inserts defining a cruciform gap for receiving a control rod and having an upper end with a cruciform reinforcing element for securing said inserts against rotation. a plurality of inserts for receiving a fuel element of a nuclear power plant, said inserts defining a cruciform gap for receiving a control rod; and a drop latch rotatable about a center of rotation for securing said inserts against unintentional lifting. 2. The storage basket according to claim 1, wherein said inserts have a rectangular cross-sectional area and extend along a main axis in a direction perpendicular to the cross-sectional area. 3. The storage basket according to claim 1, wherein said inserts have a quadratic cross-sectional area and extend along a main axis in a direction perpendicular to the cross-sectional area. 4. The storage basket according to claim 1, wherein said plurality of inserts includes four inserts. 5. The storage basket according to claim 1, including at least one connecting element fixedly connecting said adjacent inserts to form a unit. 6. The storage basket according to claim 1, including a connecting sheet fixedly connecting said adjacent inserts to form a unit. 7. The storage basket according to claim 1, including a supporting element for enclosing a foot part of the control rod and for supporting said inserts on a carrying structure of a fuel element storage rack. 8. The storage basket according to claim 1, including a base plate having a given thickness, said inserts placed on and fastened to said base plate, and said base plate having a cruciform gap tapering over the given thickness of said base plate toward said inserts. 9. The storage basket according to claim 1, including at least one releasable locking element for securing said inserts against unintentional lifting. 10. A fuel element storage rack for the compact storage of fuel elements and control rods of a nuclear power plant, comprising: 11. The storage rack according to claim 10, including carrying wells each receiving a control rod and a storage basket, said carrying wells disposed on said carrying structure. 12. A method for the storage of fuel elements and control rods of a nuclear power plant, which comprises: 13. The method according to claim 12, which comprises introducing four fuel elements in the storage basket. 14. The method according to claim 13, which comprises placing the carrying wells in a checkered pattern defining interspaces therebetween, and placing both the control rod and the fuel elements in a respective one of the interspaces. 15. A storage basket, comprising: 16. A storage basket, comprising: 17. A storage basket, comprising: 18. The storage basket according to claim 17, wherein said reinforcing element has a device for lifting and transporting. 19. The storage basket according to claim 18, wherein said device is at least one hook. 20. A storage basket, comprising: |
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051951207 | description | MORE DETAILED DESCRIPTION In order to eliminate the image of an antiscatter grid with a spatial frequency F.sub.g on a radiological film, the invention proposes the use of an X-ray tube having two focal spots with a width a and a spacing b such that the modulation transfer function has a cut-off frequency for the spatial frequency F.sub.g. The way in which an elimination such as this is obtained shall be explained by means of the graphs of FIGS. 4 to 9. The diagram of FIG. 4 indicates the respective positions of the focal spots A and B of an object to be examined such as the breast 16, of the antiscatter grid 19 and of a receiver such as the radiological film 14. If c is the distance between the focal spots and the breast 16, and d is the distance between the breast 16 and the film 14, a magnification G is defined as the ratio: ##EQU3## FIG. 5 shows the theoretical energy distribution of the focal spots A and B as a function of their x-axis value with respect to a midpoint 0. For each focal spot, this distribution has a rectangular shape that is assumed to be perfect. The modulation transfer function of the two focal spots A and B presenting the energy distribution of the graph of FIG. 5, which is given by the modulus of the optical transfer function, is represented by the solid line curve 33 of FIG. 6. In this FIG. 6, the y-axis is graduated in contrast values from 0 to 1 defined as the ratio C such that: ##EQU4## I.sub.max being the maximum value and I.sub.min being the minimum value of the luminous intensity. The x-axis is graduated in spectral frequency F and, more precisely, in pairs of lines per millimeter which, for an antiscatter grid, signifies the number of pairs of X-ray opaque strips per millimeter (p.1./mm). This curve 33 is the product of the two curves 34 and 35 shown in dashes. The curve 34 is of the type modulus of: ##EQU5## and depends only on the width a of the focal spots A and B. The curve 35 is of the type modulus of (cos y) and depends only on the spacing b of the focal spots A and B. Given that b is generally greater than a, the first value of the spatial frequency for which the contrast is cancelled out is given by: ##EQU6## i.e. by the .vertline.cos y.vertline. type curve. As a rule, the spatial frequencies for which .vertline.cos y.vertline.=0 are given by: ##EQU7## The second value of the spatial frequency for which the contrast gets cancelled out is given by ##EQU8## i.e. by a curve of the ##EQU9## type when b<3a. As a rule, the spatial frequencies for which ##EQU10## are given by: ##EQU11## The resultant curve 33, which is the product of the curves 34 and 35, shows that to eliminate an object having a spatial frequency F.sub.g from the image, it is necessary to choose, for a given magnification G, a value of the spacing b such that: ##EQU12## giving ##EQU13## The modulation transfer function of the focal spots A and B as represented by the curve 33 of FIG. 6 or the curve 33' of FIG. 7 does not take account of the modulation transfer function of the receiver represented by the curve 36 of FIG. 8. If this function is taken into account, the modulation transfer function of the image system is represented by the curve 37 of FIG. 9 which is the product of the curves 33' and 36. The curve 33' of FIG. 7 is similar to the curve 33 of FIG. 6, but its x-axis has been plotted to the same scale as in FIGS. 8 and 9. The X-ray tubes, which have two focal spots, are known, but the characteristics of each of the focal spots are generally different for it is desired to obtain X-rays having different characteristics. To make an X-ray tube having two simultaneous focal spots, it is necessary to use two electron transmission sources, i.e. two emitting filaments. This may be obtained in different ways, for example, by two separate cathodes each having an emitting filament. According to the invention, it is preferable to use only one cathode having two filaments. Three exemplary embodiments of a cathode such as this shall be described with reference to FIGS. 10, 11 and 12. In the exemplary embodiment of FIG. 10, a cathode 42 has two identical filaments 40 and 41 that are positioned in a concentration element, divided into two identical parts, one to focus the electrons emitted by the filament 40 and the other to focus the electrons emitted by the filament 41. It is thus that the filament 40 is positioned at the bottom of a stepped groove 43 made of two metal parts 44 and 45 attached to each other but insulated from each other by an insulating layer or partition 46. In the same way, the filament 41 is located at the bottom of a stepped groove 47 that is made of two metal parts 48 and 49 that are attached to but insulated from each other by an insulating layer 50. The central metal parts 45 and 48 are attached to but insulated from each other by an insulating layer 51. The concentration element thus has four metal parts 44, 45, 48 and 49 that are attached to but insulated from one another. This arrangement enables the application of different voltages to the different metal parts and hence the modification of the focusing of the electron beams 55 and 56, i.e. the dimensions of their point of impact 53 and 54 on the anode 52 and notably the dimension a. Furthermore, it also enables the modification of the angular position of the axis of emission 57 and 58 of said beams to modify the distance b between the points of impact 53 and 54 corresponding to the two focal spots. While the cathode described with reference to FIG. 10 is satisfactory for the implementation of the invention, it is not sophisticated enough to fulfil certain other functions implemented in radiological instruments, such as obtaining X-rays with different characteristics. To this effect, there is a known method of using cathodes comprising two cathodes that have different characteristics and are used successively, i.e. non-simultaneously. In order to continue to fulfil the requisite functions while at the same time implementing the present invention, these cathodes have to be modified, and FIGS. 11 and 12 show examples of such modifications, resulting from the exemplary embodiment described with reference to FIG. 10. In FIG. 11, a cathode 60 has a first filament 61, the dimensions of which are such that it gives rise, at an anode 62, to a first focal spot 63 known as the "large focal spot". The electrons emitted by the first filament 61 are focused and sent towards the focal spot 63 by a first focusing device that has two metal parts 64 and 65, which are attached to but insulated from each other by an insulating layer 66. These two parts 64 and 65 form, around the filament 61, a concentration element with a well-known three-stepped shape. The cathode 60 has a second filament 71, the dimensions of which are smaller than those of the first filament 61. This second filament 71 gives rise, at the anode 62, to a second focal spot 73 called a "small focal spot". The electrons emitted by this second filament 71 are focused and directed towards the focal spot 73 by a second focusing device that has two metal parts 74 and 75 which are attached to but insulated from each other by an insulating layer 76. These two metal parts 74 and 75 form, around the filament 71, a concentration element with a two-stepped shape. The two focusing devices are attached by the metal parts 65 and 64 which are insulated from each other by an insulating layer 72. The number of steps of each focusing device as well as the height of said steps are different in order to obtain the desired focusing and deflection of the electron beams, the intensity of which is generally different for the characteristics of the filaments (dimensions, lengths, resistance values, heating currents. The mechanical characteristics of the focusing devices are generally not sufficient to obtain the desired effects: there is therefore provision for the application of the different voltages to the metal parts 64, 65, 74 and 75, and this is the reason for their electrical insulation with respect to one another. The making of a cathode such as this is within the scope of those skilled in the art. The exemplary embodiment of FIG. 12 is similar to that of FIG. 11 with the difference that the axes of the focusing devices are not parallel to one another as in FIG. 11 but are secant, and what is called a dihedron-shaped cathode is obtained. In this FIG. 12, the elements similar to those of FIG. 11 bear the same references but have the added sign "'". In the X-ray tubes with two focal spots used in standard type radiological devices, only one focal spot is used at a time and it is enough to focus one of the two electron beams at a point 77 that will be the same for the two beams. To use an X-ray tube such as this in order to implement the present invention, it is necessary to simultaneously obtain two focal spots 63 and 73 that are as identical as possible and are separated by a determined distance b. To this effect, the mechanical and electrical characteristics of the focusing devices of each element as well as the values of the heating currents of said filaments are determined to enable the modification of the focusing of the beams so that they give rise to identical focal spots and to the modification of their deflection in order to obtain the distance b between the focal spots. It is important to be able to modify the distance in order to take account of the fact that the magnification G of the radiological apparatus changes according to the type of exposure. To determine the potentials to be applied to the metal parts of the different focusing and deflection elements of the cathodes of FIGS. 10, 11 and 12, it is necessary to carry out calibrations of the X-ray tube comprising said cathode by varying the voltages U applied to the different metal parts. These calibrations have also been described in the corresponding U.S. patent application Ser. No. 07/649,041 filed on Feb. 1st, 1991, but shall be described again herein in their application to the cathode 60 of FIG. 11. A first calibration consists in measuring the width f of the point of impact of the electron beam on the anode 62, i.e. the width of the focal spot of the X-ray beam as a function of the potential U.sub.64 applied to the part 64 for different values of U.sub.65 applied to the part 65. The curves 80 to 83 of FIG. 13a are obtained. These curves 80 to 83 correspond respectively to U.sub.65 =0 volts, -100 volts, -200 volts and -300 volts. A second calibration consists in measuring the deflection .delta. of the electron beam, namely the shifting of its focal spot 62 on the anode 63 with respect to a median axis or plane 67 as a function of the potential U.sub.64 and for the same values of U.sub.65. The curves 84 to 87 of FIG. 13b are obtained. These curves 80 to 83 correspond respectively to U.sub.65 0 volts, -100 volts, -200 volts and -300 volts. The combination of the curves of FIGS. 13a and 13b makes it possible to obtain the two systems of curves of FIG. 13c, i.e. to obtain U.sub.65 as a function of U.sub.64 for different values of the width f of the point of impact (network 88) and for different values of the deflection .delta. of the beam (network 89). In this FIG. 13c, the coordinates of the points of intersection of the curves of the two networks 88 and 89 give the values of U.sub.64 and U.sub.65 to obtain the width of the focal spot and of its deflection as indicated by the secant curves. Naturally, these calibrations described here above also have to be performed for the beam controlled by the parts 74 and 75. In brief, for a given antiscatter grid, i.e. for a given spatial frequency F.sub.g, the formula (1) makes it possible to determine the distances b to be set up between the two focal spots for different values of the magnification G which usually varies between 1.05 and 1.2 for, beyond this value, the scattered radiation is eliminated as a result of the path in the air so that the antiscatter grid is no longer used. Then, the curves of FIG. 13c make it possible to determine the potentials to be applied to the different parts of the focusing and deflection devices to obtain, firstly, each spacing b=2.delta. between the focal spots 63 and 73, giving a deflection of b/2 per device and, secondly, the same width a=f per focal spot. Furthermore, the characteristics of each filament make it possible to determine the heating current to obtain the same energy per focal spot if different filaments are used. For a given radiological apparatus, these different elements of information are, for example, recorded in the memory of a microprocessor which, for a certain value of magnification, gives the values of the potentials to be applied to the different metal parts of the cathode as well as the values of the heating currents of the filaments in order to obtain the two focal spots with a width a spaced out by the distance b. Hence, as shown in FIG. 14, the system, according to the invention, for the elimination of the image of the antiscatter grid, comprises an X-ray tube 90 with two focal spots A and B, the cathode of which has two identical or non-identical filaments that are each associated with a focusing and deflection device. The different supply and biasing voltages of this tube 90 are given by a device 91 which is under the control of a microprocessor 92. This microprocessor has at least one memory 93 which contains the values of the voltages to be applied to the tube 90 as well as the heating currents for the filaments in order to simultaneously obtain two identical focal spots with a width a and a spacing b that enable the elimination of the image of the fixed antiscatter grid for a given magnification G. To obtain the deflections required of the electron beams, exemplary embodiments have been described wherein the deflections are obtained by the application of the bias voltages to the different parts of the focusing elements. These deflections can also be obtained by other devices such as, for example, additional electrodes which would be placed on either side of each beam and would be electrically insulated from the focusing elements. These devices would be borne by said focusing elements and would constitute extensions of the steps. The use of such additional electrodes would enable the two functions to be separated into two, one being a focusing function reserved for the focusing elements in the vicinity of the filaments and the other being a deflection function reserved for these additional electrodes placed along the beams. |
045541280 | summary | BACKGROUND OF THE INVENTION The present invention relates generally to quality control inspection of nuclear reactor fuel rods, and more particularly, to an apparatus and method for inspecting TIG (tungsten inert gas) welds of an end plug on a sealed fuel rod. A typical nuclear fuel rod includes an elongated cladding tube containing nuclear fuel pellets, a bottom end plug girth welded to the bottom of the cladding tube, and a top end plug girth welded to the top of the cladding tube. The top end plug usually contains an axial bore through which the interior of the cladding tube is pressurized with a gas (such as helium). The axial bore is then closed with a seal weld. The girth and seal welds are typically TIG welds. Quality control inspection of the manufactured fuel rod has included inspection of the top end plug's girth and seal welds, and the bottom end plug's girth weld. In a conventional X-ray weld inspection technique, an X-ray film was developed for a weld area and was subjectively evaluated by an inspector. Relatively heavy shielding was required to minimize radiation exposure to personnel. The X-ray system detected tungsten inclusions in the top end plug's seal weld. Since the seal weld was made under pressure, a higher potential was required for the welding electrode to overcome the dielectric of the pressurized gas, and tungsten from the TIG welding tip could be ejected into the weldment. Tungsten inclusions were less of a problem with the unpressurized girth welds. The X-ray system also detected voids in the girth and seal welds. By a "void" is meant an area devoid of weldment which should have such weldment. It includes a cavity within a weldment as well as a weldment which is lacking in desired thickness. Conventional ultrasonic flaw detection techniques have been applied to material and weld inspection, typically in non-nuclear fuel rod areas. Existing ultrasonic inspection systems have included the use of multiple transducers for better void detection, the capability of automatic testing, and the use of flaw alarm monitors. Pipes have been tested in transit tanks according to the immersion technique, with spiral translatory motion of the pipes. The entrance and exit openings of the tank were sealed in such a way that, during the test run, the pipes remained submerged to a sufficient depth. Canning tubes for nuclear reactor fuel elements have been tested by the immersion method using a long tank with completely immersed pipes and traveling probes, or using a transit tank through which the pipe is fed in spiral motion, with the probes remaining stationary. Ultrasonic weld inspection systems detect the material/air boundary of voids (either the thin wall or cavity type), but cannot reliably detect tungsten inclusions (which give a material/material boundary) which can arise from TIG seal welding. The presence of a precalculated amount of tungsten is a weld defect which must be tested for by any nuclear fuel rod end plug weld inspection apparatus. SUMMARY OF THE INVENTION Briefly stated, the invention is directed towards apparatus for inspecting end plug TIG welds on a sealed nuclear fuel rod. The apparatus includes an X-ray fluorescent spectrograph to detect tungsten in the welds and an ultrasonic weld inspection system to detect voids in the welds. There also are devices for moving the end plug into and out from the X-ray fluorescent spectrograph, into and out from the ultrasonic weld inspection system, and between the X-ray fluorescent spectrograph and the ultrasonic weld inspection system. A controller regulates (in a predetermined manner) the three end plug moving devices, the X-ray fluorescent spectrograph, and the ultrasonic weld inspection system. The invention also is directed towards a second ultrasonic weld inspection system in combination with the apparatus described in the preceding paragraph in which the second ultrasonic weld inspection system tests the girth weld of the bottom end plug for voids, the X-ray fluorescent spectrograph tests the seal weld of the top end plug for tungsten, and the first ultrasonic weld inspection system tests the girth and seal welds of the top end plug for voids. The invention additionally is directed towards a method for inspecting end plug TIG welds on a sealed nuclear fuel rod which includes testing the welds for voids with an ultrasonic weld inspection system and testing the welds for tungsten with an X-ray fluorescent spectrograph. |
claims | 1. A nuclear fuel assembly spacer grid defining cells for receiving fuel rods, the spacer grid comprising:a peripheral band composed of at least one peripheral strip delimiting a portion of the peripheral contour of the spacer grid, and at least one spacer grid positioning spring elastically deformable and formed in the peripheral band; andinterlaced intermediate strips distributed within the peripheral contour,the at least one spacer grid positioning spring protruding outwardly relative to the at least one peripheral strip,the at least one spacer grid positioning spring being cantilevered by extending in a cantilever fashion horizontally away from or towards a corner of the spacer grid,the at least one spacer grid positioning spring including a flexible cantilever tab and a contact portion protruding outwardly from the flexible cantilever tab,the at least one spacer grid positioning spring being located vertically between an upper plane and a lower plane of the spacer grid, the upper plane being defined respectively by upper edges of the interlaced intermediate strips, the lower plane being defined by lower edges of the interlaced intermediate strips of the spacer grid. 2. The spacer grid according to claim 1, wherein the at least one spacer grid positioning spring is stamped in the peripheral band. 3. The spacer grid according to claim 1, wherein the at least one spacer grid positioning spring is delimited in the peripheral band by at least one elongated slot cut in the peripheral band. 4. The spacer grid according to claim 3, wherein the at least one elongated slot is a curved slot. 5. The spacer grid according to claim 1, wherein the contact portion is rigid. 6. The spacer grid according to claim 1, wherein the at least one spacer grid positioning spring is at a longitudinal end portion of one of the at least one peripheral strip. 7. The spacer grid according to claim 6, wherein the at least one spacer grid positioning spring includes two spacer grid positioning springs formed in one of the at least one peripheral strip each at a respective longitudinal end portion of the peripheral strip. 8. The spacer grid according to claim 1, wherein the peripheral band is composed of several peripheral strips each delimiting a side of the peripheral contour of the spacer grid, the at least one spacer grid positioning spring including a plurality of spacer grid positioning springs, each peripheral strip comprising two of the spacer grid positioning springs each at a respective longitudinal end portion of the corresponding peripheral strip. 9. The spacer grid according to claim 1, wherein the at least one spacer grid positioning spring is formed by at least one free cantilevered end portion of the peripheral band. 10. The spacer grid according to claim 9 further comprising at least one corner cell delimited by two peripheral strip free end portions of two peripheral strips of the peripheral band delimiting two adjacent sides of the peripheral contour of the spacer grid, said free end portions being separated from each other by an aperture such that the at least one corner cell is laterally opened. 11. The spacer grid according to claim 10, wherein each corner cell is delimited by two free end portions of two adjacent peripheral strips separated by an aperture. 12. The spacer grid according to claim 10, wherein the at least one spacer grid positioning spring includes at least two spacer grid positioning springs, each of the two free end portions delimiting the at least one laterally opened corner cell includes therein a respective one of the at least two spacer grid positioning springs. 13. The spacer grid according to claim 1 further comprising at least one motion limiter configured for limiting motion of the at least one peripheral strip and avoiding overstress of the at least one spacer grid positioning spring and formed in the peripheral band vertically aligned with the at least one spacer grid positioning spring and protruding outwardly. 14. The spacer grid according to claim 9 wherein the at least one peripheral strip includes a plurality of peripheral strips, the intermediate strips distributed between the peripheral strips forming an assembly of interlaced strips. 15. A nuclear fuel assembly comprising:a bundle of fuel rods,a fuel channel andat least one spacer grid according to claim 1, for laterally positioning the bundle of fuel rods within the fuel channel. |
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052280691 | summary | FIELD OF THE INVENTION This invention is concerned with computerized tomographic scanners and more particularly with tomographic scanning systems equipped to simultaneously acquire multiple slice data in a single scan. BACKGROUND OF THE INVENTION Early CT scanners that were used for scanning the brain had only a single detector and a single pencil beam X-ray source. The source and detector were repeatedly translated across the head rectilinearly a short distance and then rotated to acquire the plurality of views required to obtain an image. The early scanners required about 300 seconds to complete a 180 degree scan. Historically the next advance in scanners, known as "second generation" scanners also used a two motion system, but improved the data acquisition speed to below 20 seconds through the use of an array of detectors and a fan beam X-ray source. Twenty seconds is a normal breath holding period; and thus, the second generation tomographic scanners managed to reduce motion blurring and artifacts due to respiration. Third generation CT scanners also known as rotate-rotate scanners used fan beam X-ray sources and an array of detectors that rotated simultaneously about the subject. The scan time of the third generation scanners in general is under 5 seconds. The fourth generation CT scanners also use a fan beam X-ray source that rotate within a circle of stationary detectors occupying a full 360 degree circle around the subject. Hence, the successive generations of CT scanners increased the scan speed to decrease the scanning time. Each generation used more detectors in the detector arrays and thereby substantially increased the costs of the system. The increased number of detectors, of course, increased the spatial resolution. Thus in successive gnerations the speed of operation and the cost of the scanners were increased while the spatial resolution was improved. One method used to increase the speed of the earlier scanners; i.e., first generation scan CT scanners was the use of tandem detectors to obtain dual slices in a single scan. This practice was discontinued when detector arrays were used. Thus, after the scan speed improvement of the second generation it was generally assumed by those skilled in the art that there was no longer a need to acquire data for two slices simultaneously. An important factor mitigating against the simultaneous acquisition of dual slice data in a single scan is that to accomplish such dual slice imaging it is necessary to increase the number of detectors. Each detector, of course, normally requires a separate channel with all of the front end electronics and hardware to support the detector. Hence, each added detector substantially increases the cost of the tomographic equipment. Thus, while dual slice equipment saves time it does substantially increase the cost and in the past has increased artifacts caused by the scanning operation. Accordingly, those skilled in the art have not used simultaneous dual slice features since about the time of the introduction of the fan beam; i.e., the second generation scanners and certainly it is not known that any have been used in third generation scanners even though there have been suggestions for using simultaneous dual slice acquisition with fourth generation machines. See, for example, an article entitled "Theoretical Possibilities for a CT Scanner Development" by Dr. D. P. Boyd, which was published in Diagnostic Imaging in December, 1982. In general, the speed of scanning of computerized tomographic systems has increased from something like 5 minutes to less than a second. The increased speed has led to improved image quality; because among other things, of a reduction of motion caused artifacts. In addition the spatial resolution has improved due to increased computer power, and the number and density of the detectors. In the article, the problem of the additional cost of the detectors and hardware required for dual slice acquisition is addressed by the suggestion of the use of a plurality of X-ray sources displaced from each other in the Z direction rather than detectors displaced from each other in the Z direction. The Z direction is transverse to the longitudinal direction of the detector array or where the detector array is arcuate, the longitudinal direction of the top view planar projection of the detector array. As witnessed by the fourth generation scanners, however, those skilled in the art are still searching for methods and apparatus to further decrease motion caused artifacts in addition to increasing the throughput and decreasing the exposure of the subject to radiation. Accordingly, an object of the present invention is to provide a dual slice data acquisition system for use in third generation rotate-rotate computerized tomographic scanners. BRIEF DESCRIPTION OF THE INVENTION In accordance with the present invention a computerized tomographic system is provided, said system comprising: a gantry, PA1 said gantry including means for retaining X-ray source means on one side of a patient and X-ray detector means on the other side of said patient, PA1 means for simultaneously rotating said source means and said detector means about the patient, PA1 said detector means comprising means for simultaneously detecting X-rays that have traversed multiple plane sections in said patient, PA1 means for processing said detected X-rays to provide image data, and PA1 means for displaying images based on said image data. A feature of the present invention provides means for more efficiently using the X-ray beams to obtain dual slice imaging data per scan. This efficient use of the X-ray beams speeds throughput, reduces motion caused artifacts and also reduces the patient's exposure to X-ray radiation without any undue adverse effects on the image quality. The dual slicing can be accomplished at a minimum increase in cost due to extra detectors by using two modes of operation, a single slice mode and a dual slice mode. The dual slice mode may be limited to scans less than whole body scans. A related feature of the present invention provides means for shifting the detector means in the Z direction to assure that artifacts caused by beam divergence is readily correctable by using the familiar single slice scan geometry. A further feature of the invention includes detector means wherein said means for simultaneously detecting X-rays that have traversed multiple plane sections in the patient comprises a pair of abutting detectors extending in the Z direction with means for isolating each of the detectors from affecting the juxtaposed detectors. Where the Y direction is the direction between the source and the detectors and the X direction is the longitudinal direction of a detector array. The Z direction is perpendicular to both the X and the Y direction. A further feature of the invention comprises utilizing a source means that has a dimension in the Z direction and thus is not a point source in the Z direction, said source means providing a fan beam which extends from the source means to the detector means and encompasses the patient in the X direction. Another feature of the invention comprises utilizing multiple detectors extending in the Z direction only for a portion of the array in the X dimension. Thus, this utilization of limited extra detectors in the X direction extending in the Z direction minimizes the costs of extra detectors while providing the benefits of the dual slice capability in critical acquisition procedures, such as head scans. |
051951215 | description | DESCRIPTION OF A PREFERRED EMBODIMENT OF THE INVENTION FIG. 1 shows a system 1 for lower limb angiography, comprising an X-ray cone and tube assembly 3 of conventional structure, designed to generate an X-ray beam 5 in a given direction "D" towards a radiographic table 7 incorporating a radiographic film cassette 9 mounted in a changer. This kind of equipment wherein the X-ray cone and tube assembly 3 is always centered on the cassette changer is well known in the art and commonly used to take a sequence of radiographic pictures of the lower part of a patient "P". Accordingly, it does not need to be further described. As better shown in FIG. 2, the X-ray cone and tube assembly 3 incorporates an X-ray beam modulator 11 mounted in a large box 13 fixed to the bottom of the assembly 3. The box 13 has a small upper window opening 15 coaxially positioned with respect to the outlet of the X-ray tube, and a large lower window opening 17 that is also aligned with the outlet of the X-ray tube in order to let the X-ray beam 5 pass therethrough. In accordance with the invention, the X-ray beam modulator 11 located in the box 13 comprises a rotor 19 having a rotation axis "A" parallel to the direction "D". The rotor 19 is positioned close to the upper window opening 15 and thus extends adjacent the X-ray beam 5. Means are provided for driving rotor 19 about its rotation axis "A" at variable speeds. These means include an electric motor 23 whose shaft is connected via a belt 25 to the rotor 19 in order to drive the same. These means also include remote control means 21 including a rheostat or any other similar device known per se to control the rotation speed of the motor 23 and thus the rotation speed of rotor 19. These control means 27 are preferably associated to a programmable computer to allow preselection of one or more given speeds of rotation of the rotor, corresponding to some required compensation curves, as will be explained hereinafter. The X-ray beam modulator also comprises a set of at least three and preferably four blades 27 made of a material opaque to X-ray. The blades 27 are identical and each comprises a central hub 29 and a pair of symmetrical wings 31 extending away from the hub 29. The wings 31 are advantageously shaped to cause variation in the modulation obtained therewith according to their skew angle, and thus to generate different compensation curves as a function of the rotation speed of the rotor, as will be explained hereinafter. The blades 27 have their hubs 29 slidably mounted on a corresponding member of radially projecting shafts 33 symmetrically positioned about the rotor 19, each shaft being fixed to the rotor by screwing of one of its ends 35 into a receiving hole 37. Guiding means are provided on each pair of hub and shaft to cause the corresponding blade 27 to pivot about the shaft 33 on which it is mounted, when the speed of the rotor 19 increases and all the blades 27 are then moved in unison radially outwardly along the shafts 33 because of the centrifugal force. The guiding means preferably comprises a set of rollers 39 freely mounted on pins 41 symmetrically fixed to and projecting from any one of the hub and shaft in such a manner as to engage a corresponding set of threads 43 made in the other one of these hub and shaft. In the illustrated embodiment, the pins 41 projects inwardly from the hub 29 touward the shaft 33 and the threads 43 are made in a cylindrical member 45 coaxially fixed to the shaft 33 away from the rotor 19, this member having an external diameter wider than the shaft 33 so as to define a bearing flange 47. The hub 29 has an internal diameter substantially identical to the external diameter of the member 45 and an opening 49 adjacent the roto 19. This opening 49 is closed by a guiding ring 51 fixed to the hub by a key 53. Of course, the ring 51 is slidably mounted onto the shaft 33 to allow the blade to slide along this shaft. Return spring means are provided for urging each of the blades 27 radially inwardly towards the rotor 19 along their respective shafts 33. These return spring means preferably consist of compression springs 55 mounted about the shafts 33 within the hub 29, each spring 55 having one end bearing against the flange 47 and another end bearing against the ring 51. The pins 41 and their rollers 39 and the threads 43 are positioned in such a manner that the blades 27 extend at a small angle with respect to the direction "D" when the rotor 19 is stopped or driven at low speed, as is shown in FIG. 1, and extends substantially horizontally when the rotor is driven at high speed. In accordance with the invention, the blades and their shafts 33 are sized to cause the blades to intersect the X-ray beam 5 when the rotor is driven. Assuming that the exposure time is 1/10 sec., the rotation speed of the rotor 19 will be selected to be equal to at least 600 rpm, to prevent the blades from being "radiographed" and thus appear onto the film. As soon as the motor is driven, the blades 27 are moved radially outwardly by the centrifugal force, and start rotating about their own axis as a result of the combined action of the pins 41 and rollers 39 within the threads 43. The more they slide radially outwardly, the more they rotate and the more they extend horizontally across the beam 5, thereby reducing the amount of irradiation. Thus, it becomes very simple to modulate the intensity of the X-ray beam 5 as a function of the angular position of the blades 27 about their shafts 33, which itself depends on the speed of the rotor 19 and allow more or less radiation to pass between the blades. As aforesaid, shape of the winfs 31 of the blades 27 may be selected to achieve variation in the modulation obtained therewith according to the skew angle of the blades 27, and thus to generate different compensaton curves as a function of the rotation speed of the rotor. Moreover, the programmable control means may be used to preselect one or more given speeds of rotation of the rotor corresponding to some required compensation curves adapted to the patient's anatomy and fatness. The X-ray beam modulator may programmed to start simultaneously with the anode rotation of the X-ray tube and cone assembly and then to operate continuously so as to be totally independent from the selected exposition sequence and duration that can be very short depending on the speed of screens use. FIG. 3 shows the kind of compensation curves that may be obtained in a lower limb angiography. It must be understood however that the X-ray beam modulator according to the invention may be used in other systems and is not restricted exclusively to be used with a lower limb angiography system. |
039363490 | claims | 1. A nuclear reactor core comprising an array of closely packed elongate components, the components being arranged in groups with their longitudinal axes generally vertical, support means for the array of components, means for arranging at least some of the components of each group so that they tilt towards the centre of their respective group wherein at least some of the components of the group are urged laterally into abutment with one another, at least some of the components being nuclear fuel elements having interengaging splined bearing pad means intermediate their ends for resisting relative lateral movement of two adjacent fuel elements. a cluster of parallel fuel pins, a peripheral wrapper of hexagonal cross-section enclosing the fuel pins, the wrapper having bearing pads for abutment with corresponding pads of similar fuel elements, the bearing pads being disposed on each side of the wrapper intermediate the ends of the wrapper, the bearing pads being in the form of spline like ribs extending parallel to the longitudinal axis of the fuel element and interengagable with complementary ribs of pads of similar fuel elements. 2. A nuclear reactor core according to claim 1 wherein the components comprise fuel elements and control rods and are generally arranged in modules, each module comprising a cluster of four components at least three of which are fuel elements, one fuel element being rigidly supported whilst the remaining components are resiliently tilted towards the centre of the cluster to lean on the rigidly supported element. 3. A nuclear reactor core according to claim 1 wherein the fuel elements are arranged in modules, each module comprising a cluster of six fuel elements, each resiliently tilted towards a central void to form a circular arch. 4. A nuclear reactor core according to claim 3 wherein at least some of the modules have additional fuel elements disposed outside the clusters and resiliently tilted towards the central voids. 5. A nuclear reactor core according to claim 3 wherein the central voids of at least some of the modules accommodate control rods. 6. A nuclear reactor fuel element comprising: 7. A nuclear reactor fuel element according to claim 6 wherein the spline like ribs of the bearing pads have taper lead-in surfaces at each end at least on one side and on their faces. 8. A nuclear reactor fuel element according to claim 7 wherein the wrapper has a correspondingly hexagonal taper in an end region which is uppermost when the element is in a nuculear reactor core and a circular taper at a lower end region of the wrapper, and a groups of rib like corner features disposed above and adjacent the conical taper. |
abstract | The present invention provides a method of operating a Boiling Water Reactor, having the steps of analyzing LPRM signals for oscilliatory behavior indicative of neutron-flux-coupled density wave oscillations, determining if oscilliatory behavior is present in the signals; initiating a reactor protective corrective action if the oscilliatory behavior is determined, and in addition, initiating corrective actions if neutron uncoupled oscillations are possible. Detecting the later is performed through analytically determined exclusion zone on the power flow map or by on-line stability calculations for several high power channels. |
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041728076 | summary | BACKGROUND OF THE INVENTION In present day reprocessing of radioactive waste from nuclear reactors, the high-level waste is obtained in a strong nitric acid solution. The predominant radioactive substances in the waste during the first centuries are strontium-90 and cesium-137. The waste also contains, among other things, minor amounts of uranium, plutonium and transuranic elements which have considerably greater half-lives than strontium-90 and cesium-137. Those skilled in the art are generally of the opinion that it is advantageous, after a suitable period of cooling, to convert the liquid high-level waste into a solid product of good chemical resistance which is stable to leaching out of the contained radioactive substances by water and which is able to withstand heating produced by the fission products and stresses during management and transportation of the product. Materials that have been proposed for use in containing the waste include glasses such as boron silicate glass and phosphate glass, quartz, titanium dioxide, certain zeolites and other minerals existing in nature, particularly those having the ability to retain gases. In a known method of containing high-level waste in glass, the waste is evaporated, calcined and additives are added thereto which, when heated to 1000.degree.-1200.degree. C., result in a glass melt. Calcination of high-level waste may take place at a temperature of the order of magnitude of 300.degree.-500.degree. C. and results in the waste products being transformed into oxides. The melt is poured into tight steel containers which are then transferred to a cooled and supervised storage plant. SUMMARY OF THE INVENTION The present invention is directed to a method of containing high-level waste which allows for an extremely efficient and controllable management of the waste during the process of containment as well as an efficient containment in a resistant material. When the high-level waste has been isolated from the solution, which may, for example, take place in conventional manner by evaporation, possibly followed by calcination, according to the present invention the high-level waste is confined in a capsule at all times during the containment process, and neither gaseous nor liquid products are able to escape from this capsule. This confinement is obtained by sealing the capsule and subjecting the capsule to a hot-isostatic pressing. The isostatic pressure counteracts the formation of volatile constituents in the waste material contained in the capsule during the heating that is required for the contained material to be transformed into a coherent, tight unit, while at the same time achieving the necessary compression of the material. Another important advantage of the method is that it allows enclosing the material in a capsule when the material is at room temperature. A further important advantage of the invention is that it affords great freedom in choosing resistant materials for the containment of the high-level waste material. More particularly, the present invention is directed to a method of anchoring radioactive substances, particulary radioactive substances present in high-level radioactive waste, in a body which is resistant to leaching by water. The method comprises providing a mass containing radioactive substances and either materials which are resistant to leaching by water or materials which, when heated, form materials resistant to leaching by water, enclosing the mass in a capsule, isostatically pressing the capsule at a pressure and a temperature sufficient for the formation of a coherent, tight body of the mass. The capsule is preferably evacuated prior to sealing. Further advantages and features of the invention will become more apparent from a detailed consideration of the particular embodiments as set forth in the following specification in conjunction with the accompanying drawing. |
description | One example an embodiment of the present invention will be explained with reference to the drawings below. FIG. 1 schematically shows an entire structure of an exposure apparatus of the present example. In FIG. 1, the exposure apparatus of the present example is a reduction projection type exposure apparatus for carrying out a scanning exposure operation in a step-and-scan manner using, as an exposure beam (exposure light), Extreme Ultraviolet Light (xe2x80x9cEUV lightxe2x80x9d, hereinafter) EL in a soft X-ray region having a wavelength of 5 to 20 nm. As will be described later, the present example uses a projection optical system 200 which projects a main beam of a reflection luminous flux from a reticle 2 as a mask disposed on an object side substantially perpendicularly onto a wafer 10 disposed on an image side. The projection optical system 200 is a reflection system (projection magnification is xc2xc, ⅙, ⅙ and the like) including only a plurality of (e.g., three to eight, and four in FIG. 1) reflection optical devices, and the reflection system is non-telecentric on the object side and is telecentric on the image side. In the following explanation, a direction parallel to the main beam of the EUV light EL entering the wafer 10 from the projection optical system 200 is called as an optical axis direction of the projection optical system 200. It is defined that a Z-axis is parallel to the optical axis direction, a Y-axis is directed to a lateral direction of the paper sheet of FIG. 1 within a plane perpendicular to the Z-axis (corresponding to a substantially xe2x80x9chorizontal planexe2x80x9d in the present example), and an X-axis is perpendicular to the paper sheet of FIG. 1. The exposure apparatus of the present example comprises an illumination system 100 including a laser plasma light source as an exposure light source. The EUV light EL as an exposure beam is emitted to an exposure body substantially horizontally along the Y direction. The exposure body comprises a reflecting mirror 1 for reflecting the EUV light EL from the illumination system 100 to allow the EUV light EL to enter a pattern surface (a lower surface in FIG. 1) of the reticle 2 at a predetermined incident angle, a reticle stage 3 as a mask stage for holding the reticle 2, the projection optical system 200 including a refection system for projecting the EUV light EL reflected from the pattern surface of the reticle 2 onto a surface to be exposed of the wafer 10, oblique-incidence type autofocus sensors (xe2x80x9cAF sensorsxe2x80x9d hereinafter) 4 and 5 located on the reticle side, oblique-incidence type AF sensors 12 and 13 on the wafer side, an off-axis type alignment sensor ALG as a mark detecting system, and the like. First, the illumination system 100 of the present example will be explained with reference to FIGS. 2 to 5. FIG. 2 shows the illumination system 100. In FIG. 2, the gas jet cluster type laser plasma light source of the present example includes a high output laser light source LD such as a YAG laser light source utilizing semiconductor laser pumping or an excimer laser light source, a condenser lens CL for condensing laser light from the high output laser light source LD, a nozzle NZL for injecting xenon gas (Xe), krypton gas (Kr) and the like as a target of the laser plasma light source, and a condenser mirror CM. Here, the manner of generating EUV light will be explained simply. First, laser light emitted from the high output laser light source LD is condensed to one point by the condenser lens CL. Xenon gas, krypton gas and the like as the target of the laser plasma light source are injected from the nozzle NZL to this condensed point, and the target is optically pumped into a plasma state by energy of the laser light from the high output laser light source LD. When the target is changed to a low potential state (cooled), the target emits EUV light, ultraviolet light having a wavelength of 100 nm or longer, visible light and other light having other wavelengths. The EUV light and the like generated in this manner diverge to all directions. For the purpose of condensing or gathering the light, there is provided the condenser mirror CM having a spheroidal reflection plane which is formed such as to surround the target. The laser light from the high output laser light source LD is condensed in the vicinity of a first focus in the condenser mirror CM. An inner surface of the condenser mirror CM is formed with an EUV light reflection layer for reflecting the EUV light, and a cooling apparatus (not shown) is mounted to a back face of the condenser mirror CM. Since light having a wavelength which is not reflected is absorbed by a multilayered film and the like and changed into heat, the condenser mirror CM rises in temperature. A cooling apparatus for cooling the condenser mirror CM is necessary. As the cooling apparatus, one using a cooling liquid is preferable in terms of cooling efficiency, but the cooling apparatus is not limited to this. A metal is suitable as a material of the condenser mirror CM because the metal has excellent thermal conductivity. It is known that when a multilayered film including two kinds of materials laminated alternately is used as a reflection layer of EUV light formed on a surface of the condenser mirror CM, it is possible to reflect only light having a specific wavelength. For example, when molybdenum (Mo) and silicon (Si) are alternately coated several tens layers, it is possible to selectively reflect EUV light having a wavelength of about 13 nm. Further, when molybdenum (Mo) and beryllium (Be) are alternately coated several tens layers, it is possible to selectively reflect EUV light having a wavelength of about 11 nm. The EUV light EL selectively reflected by the condenser mirror CM of the present example is condensed in the vicinity of a second focus of the condenser mirror CM as an exposure beam. A transmission filter 18 (the detail of which will be described later) having a predetermined transmittance with respect to the EUV light EL is disposed in the vicinity of the second focus. The EUV light EL reflected and condensed by the condenser mirror CM passes through the transmission filter 18 and then, the EUV light EL is reflected and deflected by a concave reflection surface of reflecting mirror 17 and becomes a substantially parallel luminous flux, and enters the first fly eye mirror FEM1 as a reflection type optical integrator (homogenizer). The EUV light EL reflected by the first fly eye mirror FEM1 is reflected by the second fly eye mirror FEM2 and its illumination distribution is uniformed. In this state, the EUV light EL is reflected and condensed by a condenser mirror CDM, and deviated toward the reflecting mirror 1 shown in FIG. 1. Although it is not illustrated in FIG. 2, an EUV light selection plate is provided on the traveling direction side (right side in FIG. 2) of the EUV light EL reflected by the condenser mirror CDM for the purpose of filtering out ultraviolet light other than light having exposure wavelength and visible light and allowing only EUV light EL to pass. This is because the EUV reflection film including the multilayered film formed on the condenser mirror CM has considerably sharp wavelength selectivity with respect to a wavelength of near the EUV light, and selectively reflects only light of a specific wavelength used for exposure, but the EUV reflection film also adversely reflect ultraviolet light having a wavelength far from the exposure wavelength, visual light and the like as well. When such light having these wavelengths is introduced to the reticle 2 and the projection optical system 200, because of excessive energy, there is an adverse possibility that the mirrors 6 to 9 constituting the reticle 2 and the projection optical system 200 may be heated, and at the worst, unnecessary light may be transferred onto the wafer 10 and an image may be deteriorated. Therefore, the unnecessary light is eliminated by the EUV light selection plate. A thin metal film or the like can be used as the EUV light selection plate. The transmission filter 18 may be provided with a function of the EUV light selection plate. The above-described laser plasma light source, the transmission filter 18, the reflecting mirror 17, the first fly eye mirrors FEM1 and FEM2, the condenser mirror CDM and the like constitute the illumination system 100. The transmission filter 18 of the present example is disposed at the incident surface side of the reflecting mirror 17, i.e., between the first fly eye mirrors FEM1 and FEM2 and the plasma light source, and therefore, the first fly eye mirrors FEM1 and FEM2 will be explained first. As shown in FIG. 4A, the first fly eye mirror FEM1 includes a plurality of (three, in the present example) optical device groups GE11 to GE13. The optical device groups GE11, GE12 and GE13 respectively include reflection optical devices E11a to E11v, E12a to E12y and E13a to E13v respectively arranged along axes A1, A2 and A3 parallel to the Z axis and respectively having arc reflection surfaces which are elongated in the X direction. As shown in FIG. 4B, the second fly eye mirror FEM2 includes a plurality of (three, in the present example) optical device groups GE21 to GE23. The optical device groups GE21 to GE23 include a plurality of reflection optical devices E2 each having a substantially square reflection surface. The optical device groups GE21 to GE23 are arranged circular as a whole. As shown in FIG. 5, of luminous fluxes (exposure beams) entering the optical device group GE11 having the axis A1 of the first fly eye mirror FEM1, a luminous flux reflected by the reflection optical device E11a having an upper point C1a as its center enters reflection surfaces (expressed with a region Ia as a representative) of a plurality of upper reflection optical devices E2 of the optical device group GE2i (i=1 to 3) of the second fly eye mirror FEM2. Similarly, of luminous fluxes entering the optical device group GE11, a luminous flux reflected by the reflection optical devices E11f and E11k respectively having a center point C1f and a lower point C1k as their centers enter reflection surfaces (expressed with regions If and Ik as representatives) of a plurality of center and lower members of the reflection optical devices E2 of the optical device group GE2i. In this manner, each luminous flux from each of the reflection optical devices of the optical device groups GE11 to GE13 of the first fly eye mirror FEM1 enters the plurality of laterally arranged reflection optical devices E2 of the second fly eye mirror FEM2. In this case, the number of the reflection optical devices of the first fly eye mirror FEM1 arranged in the Z direction is about three times more than the number of those of the second fly eye mirror FEM2 arranged in the Z direction. Therefore, illuminance on each of the reflection optical devices E2 of the second fly eye mirror FEM2 is uniformed by integral effect. The condenser mirror CDM shown in FIG. 2 is irradiated, in a superimposing manner, with reflection light from each of the reflection optical device E2 of the second fly eye mirror FEM2, thereby further enhancing the uniformity of the illuminance distribution of the EUV light EL. When the laser plasma light source is used as in the present example, the nozzle NZL shown in FIG. 2 may be eroded by high temperature plasma and thus scattering particles, i.e., debris may be adversely generated from the nozzle NZL in some cases. In order to avoid ill effects such as deterioration in illuminance and deterioration in uniformity of illuminance distribution of the exposure beam which are caused by debris attaching to the optical members (such as the reflecting mirrors 1 and 17, the fly eye mirrors FEM1 and FEM2, the condenser mirror CDM, the reticle 2, the projection optical system 200) used in the exposure apparatus, and the like, the transmission filter 18 is disposed at the incident surface side of the reflecting mirror 17 in the illumination system 100 of the present example. The transmission filter 18 of the present example is formed in such a manner that a central portion of a silicon wafer which is a disk-like silicon (Si) is etched into a predetermined thickness to form a thin film portion 18a (membrane). When EUV light EL as the exposure beam passes through the thin film portion 18a corresponding to the transmission member, most of debris mixed in the EUV light EL is adsorbed or absorbed. For this reason, it is preferable that the transmittance of the thin film portion 18a is set in a range of 10 to 90% with respect to the EUV light EL. In FIG. 2, the transmission filter 18 for EUV light of the present example is mounted on a rotation plate 101. The rotation plate 101 is also provided with an unused replacement transmission filter 18A. Further, the rotation plate 101 is also provided with a cooling apparatus 103, which effectively dissipates heat accumulated in the transmission filter 18. The cooling apparatus 103 may be of a liquid-cooling type in which cooled liquid is circulated, of a heat-absorbing type in which heat is absorbed using Peltier device or the like, of a heat exchange type in which heat is released to a heat-dissipating plate using a heat pipe, or the like. When the transmission filter 18 is damaged by irradiation of EUV light EL or when debris is attached and the transmittance of the thin film portion 18a with respect to EUV light is lowered, a driving section 102 rotates the rotation plate 101 to replace the transmission filter 18 with the unused transmission filter 18A. By replacing the transmission filter 18 with the new one, the illuminance of the exposure beam is maintained at a high level. The transmission filter 18 of the present example is disposed in the vicinity of the second focus of the condenser mirror CM on the optical path between the condenser mirror CM and the reflecting mirror 17, i.e., disposed at a position where a beam diameter of the EUV light EL becomes thinnest. A thickness of the thin film portion 18a of the transmission filter 18 is set to 1 xcexcm or smaller in some cases, and since the area of the thin film portion 18a of the transmission filter 18 can be made minimum by the above-described disposition, there is a merit that the strength of the thin film portion 18a can be maintained at a high level. The transmission filter 18 may be disposed at a position slightly away from a position where the beam diameter of the EUV light EL becomes thinnest like a position 18B shown with a dotted line. At the position where the beam diameter of the EUV light EL becomes thinnest, energy of EUV light concentrates on substantially one point and thus, there is an adverse possibility that the transmission filter 18 is prone to be damaged. Therefore, when the transmission filter 18 is disposed at the position 18B slightly away from the position where the beam diameter of the EUV light EL becomes thinnest, it is possible to reduce a load of thermal stress applied to the transmission filter 18 to enhance the durability of the transmission filter 18 without increasing the area of the thin film portion 18a of the transmission filter 18 so much. As shown in FIG. 3, the transmission filter 18 for EUV light may be disposed between the reflecting mirror 17 and the first fly eye mirror FEM1, i.e., at a position 18C immediately in front of the first fly eye mirror FEM1. It is necessary to increase the area of the thin film portion 18a (see FIG. 2) of the transmission filter 18 as compared with a case in which the transmission filter 18 is disposed between the condenser mirror CM and the reflecting mirror 17, but it is possible to reduce the energy of the EUV light EL irradiated per unit area to enhance the durability of the transmission filter 18. Further, when the transmission filter 18 is disposed closer to the laser plasma light source than the first fly eye mirrors FEM1 and FEM2 as in the above-mentioned embodiment, even though the transmittance distribution of the transmission filter 18 becomes nonuniform and the illuminance distribution of the EUV light EL becomes nonuniform due to thickness variation of the thin film portion 18a of the transmission filter 18 or attachment of debris to the transmission filter 18 and the like, there is a merit that the illuminance distribution of the EUV light EL is uniformed by the first fly eye mirrors FEM1 and FEM2. Although silicon (Si) is used as the material for the transmission filter 18, the material of the transmission filter 18 is not limited to this, and any material may be used only if it has a predetermined transmittance with respect to the EUV light and it can prevent debris from entering the optical path of the EUV light. Referring back to FIG. 1, although the reticle stage 3 for holding the reticle 2 is omitted in FIG. 1, actually, the reticle stage 3 is supported in a floating manner by a two-dimensional linear actuator of a magnetic levitation type on a reticle base disposed along an XY plane. The actuator includes a permanent magnet (not shown) provided on a bottom of a peripheral portion of the reticle stage 3, and a coil running throughout on the reticle base two dimensionally of the X and Y directions. By controlling a current flowing through the coil, six-dimensional position and attitude of the reticle stage 3 are controlled. That is, the reticle stage 3 is driven in the Y direction by the actuator through a predetermined stroke (e.g., about 100 mm or more), and is driven in the X direction and xcex8 direction (direction of rotation around the Z axis) slightly. It is also possible to slightly drive the reticle stage 3 in an inclined direction with respect to the Z direction and the XY plane by the actuator. The reticle stage 3 comprises a reticle holder for adsorbing and holding the reticle 2 such as to be opposed to the reticle base, a stage body for holding a periphery of the reticle holder, and a temperature control section, provided at a back face side (upper surface side) of the reticle holder within the stage body, for controlling a temperature of the reticle holder. The reticle holder holds the reticle 2 in an electrostatic chuck manner. This is because EUV light EL is used as the exposure beam and thus, actually, the exposure apparatus of the present example is accommodated in a vacuum chamber (not shown), and it is difficult to use a vacuum chuck type reticle holder. The reticle holder may be made of material such as low-expansion glass, ceramics or the like used in a conventional exposure apparatus using deep ultraviolet light (DUV light) having a wavelength of about 200 to 300 nm. A plurality of temperature sensors are disposed on a reticle adsorbing surface of the reticle holder at predetermined distances from one another. The temperature sensors precisely measure a temperature of the reticle 2 and keep the temperature of the reticle 2 at a target temperature based on the measured value. A cooling apparatus constituting the temperature control section may be of a liquid-cooling type for introducing, therein, a cool liquid from outside through a flexible tube, of a type using an electronic device such as a Peltier device, or of a heat exchange type using a heat pipe. A side surface of the reticle stage 3 at the xe2x88x92Y direction side is mirror finished and formed with a reflection surface for reflecting light in a visible region. Although it is not illustrated in FIG. 1, a side surface of the reticle stage 3 at the xe2x88x92X direction side is also mirror finished and formed with a reflection surface for reflecting light in the visible region. In the exposure apparatus of the present example, like the conventional exposure apparatus using DUV light as the exposure beam, a position and a rotation amount (yawing amount, pitching amount and rolling amount) of the reticle stage 3 in the XY plane are monitored by an interferometer system which irradiates the reflection surface and the like with a measurement beam. In this case, Y coordinate and the rotation amount (pitching amount) of the reticle stage 3 around the X axis with respect to the position of the projection optical system 200 is measured by a measurement beam RIFYR supplied to a reference mirror 19 provided on a side surface of the projection optical system 200 and a measurement beam RIFYM supplied to the reflection surface of the reticle stage 3. Similarly, X coordinate the rotation angle (rotation amount (yawing amount) around the Z axis and the rotation amount (rolling amount) around the Y axis of the reticle stage 3 with respect to the position of the projection optical system 200 are measured. The position (focus position) in the Z direction and the inclination amount of the reticle 2 are measured by the AF sensors 4 and 5 including an illumination optical system 4 for diagonally projecting a slit image on a surface to be detected and a photoreceiver optical system 5 for re-forming the image of the slit image with luminous flux from the surface to be detected and outputting a signal which corresponds to a lateral deviation amount of the re-image-formed image. A reflection film for reflecting the EUV light EL is formed on a surface (pattern surface) of the reticle 2. This reflection film is, for example, a multilayered film including two kinds of material laminated alternately. When the wavelength of the EUV light EL is about 13 nm, when a multilayered film including alternately laminated molybdenum (Mo) and silicon (Si) is used, a reflectance of about 70% can be obtained. When the wavelength of the EUV light EL is about 11 nm, a multilayered film including alternately laminated molybdenum (Mo) and beryllium (Be) can be used as the reflection film. A master pattern is formed by applying an absorbing film which absorbs EUV light all over the reflection film and by patterning the absorbing film. If the reflection film itself such as the multilayered film is patterned, when a defect is caused in the pattern, it is difficult to repair the pattern, but if the absorbing film on the reflection film is patterned, the pattern can be reprocessed and thus, it is easy to repair the pattern. Since most of existent materials do not reflect EUV light, the materials can be used for the absorbing layer. In the present example, since the AF sensors 4 and 5 are used for measuring the position of the reticle 2 in the Z direction, the absorbing layer is made of material capable of obtaining a reflectance substantially equal to that of the reflection layer (reflection film). Other criteria for selecting material of the absorbing layer are that it is easy to form a pattern, the absorbing layer can be in tight contact with the reflection layer, and secular change by oxidation is small. Since the reticle 2 is formed on its surface with the reflection layer as described above, its material itself is not especially limited. Examples of the material of the reticle 2 are low-expansion glass, quartz glass, ceramics, and a silicon wafer. As a criteria for selecting the material, the same material as the reticle holder may be used as the material of the reticle 2 for example. Thermal expansion is caused in the reticle 2 and the reticle holder due to temperature increasing by the irradiation of exposure beam and the like. If materials of both the members are the same, they expand in the same amount and thus, there is a merit that a deviation force (thermal stress) does not act between both the members. The material of the reticle 2 is not limited to this, and even if the material of the reticle 2 is different from that of the reticle holder, when the two materials have substantially the same thermal coefficients of linear expansion, the same effect can be obtained. For example, a silicon wafer may be used as the material of the reticle 2, and silicon carbide (SiC) may be used as the material of the reticle holder. When the silicon wafer is used as the material of the reticle 2, there is a merit that a processing apparatus and the like used for producing semiconductor devices such as a pattern forming apparatus, a resist applying apparatus and an etching apparatus can be used as they are for producing the reticle 2. In the present example, silicon wafer is used as the material of the reticle 2 and the reticle holder is made of SiC for these reasons. Although it is not illustrated in FIG. 1, a movable blind and a slit plate as a field stop are disposed in proximity to and below the reticle 2 (on the incident side of the EUV light EL). More specifically, these movable blind and slit plate are disposed in the reticle base (not shown) on which the reticle stage 3 is disposed. The slit plate is for defining an arc illumination region. The slit plate may be secured to the projection optical system 200, but in the present example, the slit plate can be driven by a driving mechanism as a switching mechanism including a motor and the like. The slip plate comprises a first slit for defining an arc illumination region on the reticle 2 irradiated with the EUV light EL as the exposure beam, and a second slit for defining an illumination region including portions of alignment marks formed on both sides of a pattern region of the reticle 2. At the time of exposure, the slit plate is switched such that the arc illumination region is irradiated with light, and at the time of positioning of the reticle 2 (alignment), the slit plate is switched such that the region including the alignment mark is irradiated with light. When it is not desired to transfer, onto the wafer 10, a redundant circuit pattern (or portion other than a pattern region) formed in the same reticle, the movable blind prevents the redundant circuit portion from being included in the illumination region. Movement of the movable blind in the Y direction is controlled in synchronization with movement of the reticle stage 3 in the Y direction by a predetermined driving mechanism. In this case, the movable blind may start scanning in the same manner as the reticle 2 after the reticle 2 started scanning, or may start moving when a target pattern to be hidden comes to the illumination region. Next, the projection optical system 200 of the present example will be explained in detail. The projection optical system 200 is a reflection system including a reflection optical device (mirror) only, and its projection magnification from the reticle 2 to the wafer 10 is xc2xc times. Therefore, the EUV light EL reflected by the reticle 2 and including pattern information formed on the reticle 2 forms, onto the wafer 10, an image which is reduced to xc2xc of the pattern in the illumination region of the reticle 2 through the projection optical system 200. The projection optical system 200 includes four mirrors (reflection optical devices) in total, i.e., a first mirror 6, a second mirror 7, a third mirror 8 and a fourth mirror 9. The projection optical system 200 further includes a lens barrel 14 for holding these mirrors 6 to 9. The first mirror 6 and the fourth mirror 9 have concave aspherical reflection surfaces, the second mirror 7 has a flat reflection surface, and the third mirror 8 has a convex spherical reflection surface. It is preferable that a convex surface such as the reflection surface of the third mirror 8 is a spherical surface to suit the convenience of working and measuring operations. A material of each of the mirrors is a low-expansion glass or metal, and each mirror is formed on its surface with a reflection layer for reflecting EUV light EL. The reflection layer is formed of a multilayered film including two kinds of materials laminated alternately like the reticle 2. In this case, the fourth mirror 9 is formed with a through hole so that light (exposure beam) reflected by the first mirror 6 can reach the second mirror 7. Similarly, in order for light reflected by the fourth mirror 9 to reach the wafer 10, an outward appearance of the first mirror 6 is formed with a notch so that a luminous flux can pass therethrough. The first mirror 6 may be formed with a through hole. When exposure is performed using the EUV light EL, since space in which the projection optical system 200 is placed is also maintained under vacuum, heat caused by irradiation of the exposure beam can not be released by only disposing the mirrors 6 to 9. Thereupon, in the present example, the mirrors 6 to 9 and the mirror barrel 14 holding the mirrors 6 to 9 are connected to each other through heat pipes HP, and a cooling apparatus (not shown) for cooling the mirror barrel 14 is provided. That is, the mirror barrel 14 is formed into a double structure including an inner mirror holding portion and a cooling jacket mounted to its outer periphery. A helical pipe for flowing a cooling liquid is provided in the cooling jacket. Here, a cooling water is used as the cooling liquid. The cooling water flowing out from the cooling jacket through an outflow tube performs the heat exchange with a refrigerant in the cooling apparatus (not shown) and cooled down to a predetermined temperature and then, flows into the pipe in the cooling jacket, and the cooling water circulates in this manner. For this reason, in the projection optical system 200 of the present example, even if thermal energy is given to the mirrors 6 to 9 by the irradiation of the EUV light EL, the thermal exchange is performed by the heat pipes HP between the mirrors and the mirror barrel 14 whose temperature is adjusted to a constant value, and a temperature of each of the mirrors 6 to 9 is maintained at a constant value. In this case, in the present example, the heap pipes HP are attached not only to the back surface sides of the mirrors 6, 7, 9 and the like but also to portions of the front surface sides (at the reflection surface sides) thereof which are not irradiated with exposure beam. Therefore, the mirrors can be cooled more effectively as compared with a case in which only the back surface side is cooled. The heat pipes HP on the back surface side of the third mirror 8 and the front surface side of the first mirror 6 reach the inner peripheral surface of the mirror barrel 14 in a direction of depth on the paper sheet of FIG. 1. The mirror barrel 14 has a square prism like outer appearance. The wafer 10 is placed on a wafer stage 11. The wafer stage 11 is supported in a floating manner by a magnetic levitation type two-dimensional linear actuator 16 on a wafer base 15 disposed along an XY plane. The actuator 16 includes a permanent magnet provided on a bottom surface of the wafer stage 11, and a coil running throughout on the wafer base 15 two dimensionally of the X and Y directions. The wafer stage 11 is driven in the X and Y directions by the actuator 16 through a predetermined stroke (e.g., about 300 to 400 mm), and is driven also in the xcex8 direction (direction of rotation around the Z axis) slightly. It is also possible to slightly drive the wafer stage 11 in an inclined direction with respect to the Z direction and the XY plane by the actuator 16. An electrostatic chuck type wafer holder (not shown) is placed on the upper surface of the wafer stage 11, and the wafer 10 is adsorbed and held by the wafer holder. Although it is not illustrated in FIG. 1, a side surface of the wafer stage 11 in the +Y direction is mirror-finished and formed with a reflection surface for reflecting visible light. Similarly, a side surface of the wafer stage 11 in the xe2x88x92X direction is also formed with a reflection surface for reflecting visible light. A position and a rotation amount (yawing amount, pitching amount and rolling amount) of the wafer stage 11 in the XY plane with respect to the projection optical system 200, for example, are precisely measured by an interferometer system which irradiate the reflection surfaces with measurement beams. A position and an inclination amount of the wafer 10 in the Z direction with respect to the mirror barrel 14 shown in FIG. 1 are measured by the AF sensors 12 and 13 including an illumination optical system 12 for diagonally projecting a slit image onto a surface to be detected and a photoreceiver optical system 13 for re-forming the image of the slit image from a luminous flux from the surface to be detected and outputting a signal which corresponds to a lateral deviation amount of the re-formed image. As the FA sensors, for example, a multipoint focus position detecting system disclosed in Japanese Patent Application Laid-open No. 6-283403 and the corresponding U.S. Pat. No. 5,448,332, the disclosures of which are herein incorporated by reference, can be used. It is important that the AF sensors 12 and 13 are integrally secured to the mirror barrel 14. The wafer stage 11, i.e., the focus position and the inclination angle of the wafer 10 are corrected through the magnetic levitation type two-dimensional linear actuator 16 based on the measured values from the AF sensors 12 and 13. One end of the upper surface of the wafer stage 11 is provided with a space image measuring member FM for EUV light. The space image measuring member FM measures a relative positional relation (i.e. , a so-called base line amount) between a position where the pattern formed on the reticle 2 is projected onto the surface of the wafer 10 and a detection center of an alignment sensor ALG. The space image measuring member FM corresponds to a reference mark plate of the conventional exposure apparatus using DUV light. An upper surface of the space image measuring member FM is formed with a slit as an opening. This slit is formed by making a pattern in the EUV light reflection layer formed on a surface of a fluorescence generating material secured on the upper surface of the wafer stage 11 and having a predetermined thickness. A photoelectric conversion device such as a photo multiplier is disposed in the wafer stage 11 at the bottom surface side of the slit. With this disposition, when the space image measuring member FM is irradiated with EUV light EL from above through the projection optical system 200, the EUV light EL which has passed through the slit reaches the fluorescence generating material, and this fluorescence generating material emits light having a longer wavelength than that of the EUV light. This light is received by the photoelectric conversion device and converted into a detection signal corresponding to intensity of the light. With this operation, the projection position on the wafer stage 11 of the reticle pattern can be easily obtained. Instead of this reflection layer, an EUV light absorbing layer may be provided, and this absorbing layer may be formed with the slit. Next, the operation of the exposure procedure after the second layer by the exposure apparatus of the present example will be explained. The laser plasma light source of the illumination system 100 does not emit light until the alignment operation of the reticle or the exposure of the wafer is started. In FIG. 1, the reticle 2 is transferred by a reticle transfer system (not shown) and the reticle 2 is adsorbed and held by the reticle holder of the reticle stage 3 located on the loading position. The wafer 10 on which an EUV light EL-sensitive resist is applied is placed on the wafer stage 11 by a wafer transfer system (not shown) and wafer-delivery mechanism (not shown) on the wafer stage 11. Next, a position of the wafer alignment mark (one or more per one shot) which is a previously determined as a sample of the wafer alignment marks provided in each shot region of the wafer 10 on the wafer stage 11 is detected using the alignment sensor ALG while sequentially moving the wafer stage 11. When the detection of the positions of the wafer alignment marks of the sample shots was completed, coordinates of arrangement in all the shot regions on the wafer 10 are obtained using statistical technique utilizing least squares method disclosed in, e.g., Japanese Patent Application Laid-open No. 61-44429 and the corresponding U.S. Pat. No. 4,780,617, the disclosures of which are herein incorporated by reference, using the detected data (this alignment technique is called as xe2x80x9cEGA (enhanced global alignmentxe2x80x9d hereinafter). Alternatively, a variation amount including coordinates of arrangement in all the shot regions on the wafer 10 and magnification of each shot are obtained using statistical technique utilizing least squares method disclosed in, e.g., Japanese Patent Application Laid-open No. 6-275496, the disclosure of which is herein incorporated by reference, using the data of detected positions of the plurality of wafer alignment marks in the shots (this alignment technique is called as xe2x80x9cmultipoint EGA in shotsxe2x80x9d hereinafter). When the measurement of the alignment was completed in this manner, a variation in magnification of the shot (X, Y scaling) is calculated based on a shot distance found from the result of the above-described EGA, or based on a shot size found from the result of the above-described multipoint EGA in shots. Then, calculation is carried out for obtaining a control amount of projection magnification for making a size of the reticle pattern image in the X direction (second direction) coincident with a size of the shot region on the wafer 10 in accordance with the magnification variation amount, i.e., a driving amount of the reticle 2 in the Z direction, and the reticle 2 is driven in the Z direction (vertical direction) by the calculated amount. When the reticle 2 is driven in the Z direction, the projection magnification is varied and the position of the projection region of the reticle pattern image is deviated. Therefore, the base line and the projection magnification are measured using the space image measuring member FM as described above. Next, based on the result of the measurement of magnification, it is judged whether an adjustment remaining error of the projection magnification with respect to the target magnification adjustment amount is equal to or smaller than the allowance. If the result of the judgment was NO, i.e., if the adjustment remaining error of the projection magnification exceeded than the allowance, a procedure is return to the step for driving the reticle 2 for re-setting the projection magnification, and the reticle 2 is re-driven in the Z direction and then, the above-described procedure and judgment are repeated. On the other hand, if the result of the judgment was YES, i.e., if the adjustment remaining error of the projection magnification was equal to or smaller than the allowance, the slip plate below the reticle 2 is switched to a position where the illumination region is irradiated with the EUV light EL and then, the procedure is advanced to a next step. Next, scanning and exposure are carried out in the step-and-scan manner using the EUV light EL as the exposure beam. That is, in accordance with the position information, obtained by the above described procedure, on each regions on the wafer 10, the wafer stage 11 is positioned at a scan-starting position of a first shot and the reticle stage 3 is positioned at a scan-starting position while monitoring position information from the AF sensors 4, 5, the AF sensors 12, 13 and the interferometer system, thereby carrying out the scanning and exposure of the first shot. When the scanning and exposure are carried out, speeds of the following both stages are controlled such that the velocity ratio of the reticle stage 3 and the wafer stage 11 substantially coincide with the projection magnification of the projection optical system 200, and the exposure (transfer of reticle pattern) is carried out in a constant speed and synchronization state of the velocity ratio of the both stages. When the scanning and exposure of the first shot were completed, a stepping action is carried out between shots in which the wafer stage 11 is moved the scan-starting position of a second shot. The scanning and exposure of the second shot are carried out in the same manner as that described above. In this case, in order to omit an action for returning the reticle stage 3 to enhance the throughput, the scanning and exposure directions of the first and second shots are made opposite from each other, i.e., if the first shot was exposed to light from xe2x88x92side to +side on the Y axis, the second shot is exposed to light from +side to xe2x88x92side. That is, scanning is carried out alternately. In this manner, the stepping action between the shots and the scanning and exposure action of the shots are repeated, and the pattern of the reticle 2 is transferred in all the shot regions on the wafer 10 in the step-and-scan manner. A series of procedure with respect to one wafer 10 is completed in this manner. According to the present example, since the EUV light EL having a wavelength of 5 to 20 nm is used as the exposure beam, even if a numerical aperture of the reflection system including only the plurality of mirrors (reflection optical devices) 6 to 9 as the projection optical system 200 is not so high, it is possible to precisely transfer an extremely fine pattern such as a line-and-space pattern having a pitch of 100 to 70 nm, or an isolating line pattern of line width of 70 to 55 nm. At the time of such an exposure operation, according to the exposure apparatus of the present example, it is possible, by using the transmission filter 18, to suppress the reduction in a reflectance of the reflection member caused by debris generated from the laser plasma light source in the illumination system 100. Further, almost no debris reach the reflection members disposed closer to the reticle 2 than the transmission filter 18, it is possible to reduce a frequency of replacement of these reflection members. Therefore, there is a merit that a workload of maintenance operation of the exposure apparatus is lightened. The nozzle NZL may be made of material having high transmittance with respect to the EUV light EL (more specifically, material having higher transmittance than noble metal), or material selected from those used in the multilayered film formed on the reflection surface such as the condenser mirror CM and the reflecting mirror 17. In this case, even if debris generated by erosion of the nozzle NZL is attached to the condenser mirror CM or the reflecting mirror 17, since the debris have high transmittance with respect to the EUV light EL, the reflectance of the condenser mirror CM and the reflecting mirror 17 is not extremely lowered. Therefore, the number of maintenance operations (replacement and cleaning of mirrors) can be reduced, and the productivity of semiconductor devices can be enhanced. It is possible to prolong the life of the transmission filter 18. Although it is not illustrated in FIGS. 2 and 3, the nozzle NZL, the condenser mirror CM and the like are accommodated in an evacuated container. A transparent window which is provided on the evacuated container and which allows EUV light to pass therethrough may be formed with a thin film capable of selecting a wavelength, and only the EUV light EL may be allowed to pass through the transparent window. At that time, at least one element of the illumination system (the reflecting mirror 17, the fly eye mirrors FEM1, FEM2 and the condenser mirror CDM in the present example) is disposed in the evacuated container. Although the EUV light having a wavelength of 5 to 20 nm, especially about 11 nm or about 13 nm is used as the exposure beam in the above-described embodiment, the wavelength of the exposure beam is not limited to these values. For example, EUV light EL having a wavelength of about 100 nm or smaller is effective as the exposure beam. Known typical wavelength of the EUV light EL is 1.5 nm, 13.4 nm and the like. Recently, EUV light having a wavelength of about 50 nm becomes a promising candidate in order to lower the required surface precision of mirror. The exposure apparatus in the above embodiment is assembled by adjusting the illumination system 100 and the projection optical system 200, and connecting the constituent elements electrically, mechanically or optically. The wafer 10 exposed with light in the above-described manner is subjected to a developing step, a pattern-forming step, a bonding step and the like, thereby producing a device such as a semiconductor device. In the above-described embodiment, the present invention is applied to the scanning and exposure type and reduction projection type exposure apparatus. The present invention can also be applied to an illumination system of a proximity-type exposure apparatus which transfers a pattern of the reticle directly onto a wafer using EUV light without through a projection optical system, or to a full field exposure type projection exposure apparatus. According to the exposure method of the present invention, when a laser plasma light source is used as a light source for exposure, for example, even if scattering particles are generated from the light source, it is possible to suppress the deterioration in optical characteristics such as reduction in reflectance of a plurality of reflection members. According to the exposure apparatus of the invention, it is possible to carry out the exposure method of the invention. Further, since almost no scattering particles reach the reflection members disposed closer to the pattern than the transmission member, it is possible to reduce a frequency of replacement of these reflection members. Therefore, there is a merit that a workload of maintenance operation of the exposure apparatus is lightened. When the reflection type optical integrator is disposed between the light source and the pattern, and the transmission member is disposed between the light source and the optical integrator, even if the transmittance of the transmission member becomes uneven and the illumination of the exposure beam becomes uneven in some degree due to variation in thickness of the transmission member, due to the attachment of the scattering particles to the transmission members or the like, the illumination distribution of the exposure beam is uniformalized by the reflection type optical integrator. The entire disclosure of Japanese Patent Application No. 11-157635 filed on Jun. 4, 1999 including specification, claims, drawings and abstract are incorporated herein by reference in its entirety. Although various exemplary embodiments have been shown and described, the invention is not limited to the embodiments shown. Therefore, the scope of the invention is intended to be limited solely by the scope of the claims that follow. |
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abstract | An apparatus for generating medical isotopes provides for the irradiation of dry-phase, granular uranium compounds which are then dissolved in a solvent for separation of the medical isotope from the irradiated compound. Once the medical isotope is removed, the dissolved compound may be reconstituted in dry granular form for repeated irradiation. |
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047073233 | abstract | A closure for an inclined duct having an open upper end and defining downwardly extending passageway. The closure includes a cap for sealing engagement with the open upper end of the duct. Associated with the cap are an array of vertically aligned plug members, each of which has a cross-sectional area substantially conforming to the cross-sectional area of the passageway at least adjacent the upper end of the passageway. The plug members are interconnected in a manner to provide for free movement only in the plane in which the duct is inclined. The uppermost plug member is attached to the cap means and the cap means is in turn connected to a hoist means which is located directly over the open end of the duct. |
abstract | There is provided a particle beam therapy system that can rapidly change beam energy without increasing the size of a deflection electromagnet even in the case where the number of required beam-energy changes is large. |
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050193244 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS OF THE INVENTION Referring to FIG. 1, reference numeral 1 indicates a laterally guided vehicle for operating on rails and which is remotely controlled so as to be movable in the direction of a lock 2. The vehicle 1 has two side holders 3 for accommodating a Maltese-cross transmission (not shown) for pivotally journalling the carrier 4. The carrier 4 comprises a frame and accommodates a container 5 therein. The Maltese-cross transmission for pivoting the carrier 4 is actuated via a toothed disk 6, a toothed belt 7 and a toothed disk 8 via a drive (not shown). The container carrier 4 is equipped with an annular part 9 having an inner conical surface 10. Two pins 12 are provided in the base 11 of the container carrier 4 and each of these pins has a conical end portion. A part 13 is attached to the carrier 4 at the head region thereof and has a slot-like opening 14 which expands outwardly so that the opening 14 terminates in a V-cut 15. Three holders 16 are likewise provided at the head region of the container carrier 4 and are uniformly distributed about the periphery of the carrier. The holders 16 each include a clamping piece 18 and a counter clamp 19 pivotally mounted on the frame of the carrier 4 and a remotely-manipulated spindle 17 for tightening clamping piece 18 to hold the container 5 in place. The container 5 has two through openings 20 and two blind openings 21 all arranged at the lower end thereof. The container 5 is equipped with a holding loop 22 on its upper end so as to permit transport of the container via an overhead crane. The holding loop 22 is releasably attached to the container and in a position which cannot be mistaken. The holding loop 22 is provided with a part 23 configured as a rod having a circular cross section. The parts 23 and 13 are guide parts which conjointly define a container guide 24. The operation of the centering arrangement of the invention is described below. The container 5 is a terminal storage container for cut irradiated fuel rods and is positioned with a remotely-controlled crane into the carrier 4 with the latter in the vertical position. For centering, the container 5 first reaches the opening 14 of guide part 13 with its guide part 23 whereby a precentering of the container 5 occurs. This centering is in a very specific angular position with respect to the longitudinal axis of the container and is predetermined by the position of the lock openings. The container 5 then passes through the annular part 9 which centers the container 5 with its longitudinal axis to the longitudinal axis of the carrier 4. The longitudinal axis of the carrier is calibrated to the point M of the lock 2. Just before the container 5 meets up with the base 11, the two pins 12 on the carrier 4 engage corresponding ones of the blind openings 21 in the base of the container 5 which effects a fine centering. Then, when the container 5 comes into contact engagement with the base 11, an even finer centering is obtained which brings the container 5 into precisely a position which assures an accurate alignment of the container openings 20 with corresponding ones of the openings in the lock when the container 5 is later docked at the lock 2. The container 5 is held in this position to prevent a change in position by means of the holders 16 which are pivoted into place and tightened. The carrier 4 together with the container 5 is then pivoted out of the vertical position into a precise horizontal position by means of the Maltese-cross transmission. The container is then transported to the lock 2 so as to be docked thereat. One of the two container openings 20 is provided for filling the container with fuel rods which have been just previously cut to a length corresponding to the inner length of the container. The other container opening 20 is provided for a conduit which is connected to a vacuum device for drawing away the dust in the container 5 which was produced by cutting the fuel rods. It is understood that the foregoing description is that of the preferred embodiments of the invention and that various changes and modifications may be made thereto without departing from the spirit and scope of the invention as defined in the appended claims. |
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