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description | 1. Field of the Invention The present invention relates to a method and an apparatus for producing a radionuclide using an accelerator, and in particular, to a method and an apparatus for producing a radionuclide, capable of efficiently producing a radionuclide that is a source material of a nuclear medicine diagnostic pharmaceutical, using a small-sized, lightweight apparatus. 2. Description of the Related Art Conventionally, molybdenum 99 (Mo-99), which is used as a source material of a nuclear medicine diagnostic pharmaceutical, is obtained by collecting and purifying a product produced in a nuclear reactor as a fission product of a high or low concentration of uranium 235 (U-235). Facilities that use a nuclear reactor for production as described above are limited in number, and are also lopsidedly located in the world. Countries not having such a facility rely on importation by airfreight, and thus have a concern for stable supply. A facility that uses a nuclear reactor for production also has issues regarding facility operation, including temporal degradation of the nuclear reactor. Thus, the issue of stable supply also exists. Although Japan has many nuclear reactors, including commercial reactors and test reactors, Mo-99 is not produced domestically, and thus fully relies on importation. Need for huge investment and maintenance cost in a production facility that uses a nuclear reactor disrupts tangible progress with domestic production using a nuclear reactor. Meanwhile, a method for producing Mo-99 without using nuclear fission reaction has been extensively studied. First, a first method for producing a radionuclide is neutron activation of molybdenum 98 (Mo-98) [Mo-98(n, γ)Mo-99]. Since this reaction can be triggered using a neutron source, neutron generated by an accelerator may be used instead of neutron generated by a nuclear reactor. In addition, methods for producing a radionuclide with use of an accelerator include one that utilizes a reaction between molybdenum 100 (Mo-100) and neutron (Mo-100(n, 2n)Mo-99). Such methods may be able to solve issues such as the need for huge investment and maintenance cost in a production facility that uses a nuclear reactor. However, these methods using neutron require a large-scale accelerator, and also need providing a large screen around a Mo-98 or Mo-100 target, thus posing a problem of large overall size of the apparatus. Moreover, such methods each have a low yield, and provide a low specific activity due to a relatively high abundance ratio of Mo-98 or Mo-100 to required Mo-99, as compared with the method using nuclear fission described above, and therefore also pose problems in that, for example, the purification technique that has been established for a facility using a nuclear reactor cannot be applied. Nevertheless, purification technique is also being developed for the methods using an accelerator described above, and it is also becoming possible to obtain Tc-99m, which is ultimately needed as a source material of a nuclear medicine diagnostic pharmaceutical. WO 2011/132265 A discloses a method in which accelerated proton is emitted to molybdenum 100 (Mo-100) [Mo-100(p, pn)Mo-99 or Mo-100(p, 2n)Tc-99m] as another method for producing a radionuclide using an accelerator. However, emitting accelerated proton to molybdenum 100 (Mo-100) generates not only Mo-99 and Tc-99m, but also technetium 99 (Tc-99) (Mo-100(p, 2n)Tc-99), and thus poses a problem in that Tc-99m having a high specific activity cannot be obtained in principle. An object of the present invention is to provide a method and an apparatus for producing a radionuclide, capable of efficiently producing a radionuclide that is a source material of a nuclear medicine diagnostic pharmaceutical, using a small-sized, lightweight apparatus. To solve the above-described object, an aspect of the present invention lies in a method for producing a radionuclide that produces molybdenum trioxide 99 (Mo-99.O3) and technetium oxide 99m (Tc-99m2.O7) by emitting an electron beam accelerated by an electron linear accelerator to a molybdenum trioxide 100 (Mo-100.O3) powder sample, and which separates and purifies technetium oxide 99m from both the molybdenum trioxide 99 and the technetium oxide 99m by using a separation/purification unit, the method including: supplying temperature-regulated gas to the molybdenum trioxide 100 powder sample during an irradiation period during which the electron beam is emitted to the molybdenum trioxide 100 powder sample. According to the present invention, a radionuclide that is a source material of a nuclear medicine diagnostic pharmaceutical can be efficiently produced with a small-sized, lightweight apparatus. Molybdenum trioxide 99 (Mo-99.O3) is produced by emitting an electron beam accelerated by an electron linear accelerator to a radionuclide-producing source material (e.g., a powder sample of molybdenum trioxide 100 (Mo-100.O3)). By supplying gas (e.g., oxygen-containing gas) to the molybdenum trioxide 99 (Mo-100.O3) produced, technetium oxide 99m (Tc-99m2.O7) is produced after molybdenum 99 (Mo-99) in the molybdenum trioxide 99 (Mo-100.O2) decays. The radionuclide production section can be reduced in size due to a fact that an electron linear accelerator can be smaller in size than a proton accelerator or a heavy particle accelerator for a same acceleration energy, and that the production cross section of a (γ, n) reaction that generates Mo-99 from Mo-100 is comparable with the cross-sectional area of when Mo-99 is generated through another reaction, such as a method using a reaction between Mo-100 and neutron (Mo-100(n, 2n)Mo-99), or a method in which accelerated proton is emitted to Mo-100 (Mo-100(p, pn)Mo-99 or Mo-100(p, 2n)Tc-99m). The generated Mo-99 changes into technetium 99m (Tc-99m), which is a progeny nuclide, with a half-life of 66 hours. Technetium 99m (Tc-99m) is a nuclide that is required as a radiopharmaceutical in nuclear medicine, and molybdenum 99 (Mo-99) is a source material thereof. Technetium 99m (Tc-99m) changes into Tc-99, which is a progeny nuclide, with a half-life of 6.02 hours. Due to the fact that Tc-99 is not needed for a radiopharmaceutical, and that separation of Tc-99 from Tc-99m is difficult, it is necessary to separate and purify Tc-99m from Mo-99 in a condition in which Tc-99m has as high a specific activity as possible. Therefore, it is preferable that the production section that produces Mo-99, and the purification unit that separates and purifies Tc-99m from Mo-99 be located near each other, and/or that the separation and purification process can be performed during electron beam irradiation. The use of a powder sample of molybdenum trioxide 100 (Mo-100.O3) as the radionuclide-producing source material permits an increased surface area of the radionuclide-producing source material to be provided. This can expedite freeing of Tc-99m2.O7 generated in the sample from the source material, and can reduce the content of Tc-99m2.O7 remaining in the Mo-100.O3 source material. By supplying temperature-regulated (heated or cooled) gas to the radionuclide-producing source material (Mo-100.O3 powder sample) when an electron beam is being emitted to the Mo-100.O3 powder sample, freeing of Tc-99m2.O7 generated in the sample from the source material can be further expedited, and the content of Tc-99m2.O7 remaining in the Mo-100.O3 source material can be further reduced. Moreover, application of vibration to the radionuclide-producing source material (Mo-100.O3 powder sample) can further expedite freeing of Tc-99m2.O7 generated in the sample from the source material, and can further reduce the content of Tc-99m2.O7 remaining in the Mo-100.O3 source material. Preferred embodiments of the present invention will be described below with reference to the accompanying drawings, in which like reference numerals refer to like parts throughout the drawings. The configuration of an apparatus for producing a radionuclide according to a first embodiment, which is one preferred embodiment of the present invention, will be described below with reference to FIG. 1. As shown in FIG. 1, an apparatus for producing a radionuclide of this embodiment includes an electron linear accelerator 1, a sample container 4 for containing radionuclide-producing source material 3, a radionuclide separation/purification unit 5, a thermometer 10, a heating/cooling control unit 11, a heater/cooler 12, a heating/cooling section 13, a gas supply unit 14, a vibration device 20, and a control system (not shown). The radionuclide-producing source material 3 contains a source nuclide from which a radionuclide is generated. This embodiment will be described using the powder sample 3 of molybdenum trioxide 100 (Mo-100.O3) as an example of the radionuclide-producing source material. The thermometer 10 measures the temperature of the molybdenum trioxide 100 (Mo-100.O3) powder sample 3 contained in the sample container 4. The heating/cooling control unit 11 adjusts the temperature of the heater/cooler 12 based on the temperature information of the molybdenum trioxide 100 (Mo-100.O3) powder sample 3 measured by the thermometer 10. The heater/cooler 12 adjusts the temperature of the heating/cooling section 13 to control the temperature of the gas supplied from the gas supply unit 14. The vibration device 20 applies vibration to the sample container 4 to vibrate the molybdenum trioxide 100 (Mo-100.O3) powder sample 3 in the sample container 4. The vibration device 20 may be an ultrasonic vibration device that uses a heat-resistant ultrasonic vibrator. Referring to FIG. 7, a configuration of the gas supply unit 14 will be described. The gas supply unit 14 includes a gas mixture container 70, an oxygen gas cylinder 81 that retains oxygen gas (O3 gas), a massflow controller 71A that controls the flow rate of the oxygen gas supplied from the oxygen gas cylinder 81, a regulator 72A that controls the pressure of the oxygen gas supplied from the oxygen gas cylinder 81, a noble gas cylinder 82, a massflow controller 71B that controls the flow rate of the noble gas supplied from the noble gas cylinder 82, and a regulator 72B that controls the pressure of the noble gas supplied from the noble gas cylinder 82. In this embodiment, the gas supply unit 14 will be described, by way of example, as supplying a gas mixture 30 of the oxygen gas and the noble gas, but may be configured to supply only the oxygen gas. The apparatus for producing a radionuclide of this embodiment emits an electron beam 2 accelerated by the electron linear accelerator 1 to the molybdenum trioxide 100 (Mo-100.O3) powder sample 3 to produce molybdenum trioxide 99 (Ma-99.O3). By supplying oxygen gas to the molybdenum trioxide 99 (Mo-100.O3) that has been produced, technetium oxide 99m (Tc-99m2.O7) is produced after molybdenum 99 (Mo-99) in the molybdenum trioxide 99 (Mo-100.O3) decays. The thermometer 10 measures the temperature of the molybdenum trioxide 100 (Mo-100.O3) powder sample 3, and the heating/cooling control unit 11 controls the heater/cooler 12, and thus the oxygen gas, or a gas mixture 30 of the oxygen gas and the noble gas, before passing through the heating/cooling section 13 is heated or cooled to adjust the temperature of the oxygen gas, or the gas mixture 31 of the oxygen gas and the noble gas, after being heated or cooled. FIG. 2 illustrates vapor pressure curves of MoO3 and Tc2O7. The melting point of molybdenum trioxide 100 (Mo-100.O3) or molybdenum trioxide 99 (Mo-99.O2) is 795° C. The boiling point of technetium oxide 99m (Tc-99m2.O7) or technetium oxide 99 (Tc-992.O7) is 310.6° C. Therefore, the temperature of the oxygen gas, or of the gas mixture 31 of the oxygen gas and the noble gas, is adjusted to a temperature within a range between 310.6° C., which is the boiling point of technetium oxide 99m (Tc-99m2.O7) or of technetium oxide 99 (Tc-992.O7), and 795° C., which is the melting point of molybdenum trioxide 100 (Mo-100.O3) or of molybdenum trioxide 99 (Mo-99.O3), inclusive. By adjusting the temperature of molybdenum trioxide 100 (Mo-100.O3) powder sample 3 to a temperature within the range between 310.6° C. and 795° C. inclusive, gaseous technetium oxide 99m (Tc-99m2.O7) or gaseous technetium oxide 99 (Tc-992.O7) is separated and purified from molybdenum trioxide 100 (Mo-100.O3) or molybdenum trioxide 99 (Mo-99.O3), and the oxygen gas, or the gas mixture of the oxygen gas and the noble gas, or gaseous technetium oxide 99m and gaseous technetium oxide 99 generated are transferred to the radionuclide separation/purification unit 5. During an irradiation period during which the electron beam 2 is emitted to the molybdenum trioxide 100 (Mo-100.O3) powder sample 3, the vibration device 20 applies vibration to the molybdenum trioxide 100 (Mo-100.O3) powder sample 3. Application of vibration expedites freeing of technetium oxide 99m (Tc-99m2.O7) generated in the sample from the sample, and can thus reduce the content of the technetium oxide 99m (Tc-99m2.O7) remaining in the molybdenum trioxide 100 (Mo-100.O3) powder sample 3. According to this embodiment, a radionuclide having a short half life, such as molybdenum 99 (Mo-99) or technetium 99m (Tc-99m) that is in great demand as a source material of a nuclear medicine diagnostic pharmaceutical can be efficiently produced with a small-sized, lightweight apparatus. The configuration of an apparatus for producing a radionuclide according to a second embodiment, which is one preferred embodiment of the present invention, will be described below with reference to FIG. 3. The apparatus for producing a radionuclide of this embodiment has a fundamental configuration similar to that of the apparatus for producing a radionuclide of the first embodiment, but differs from that of the first embodiment in that the pipe that carries the oxygen gas, or the gas mixture 31 of the oxygen gas and the noble gas, after being heated or cooled, and the pipe connected to the separation/purification unit 5 are each connected to the sample container 4 which contains the molybdenum trioxide 100 (Mo-100.O3) powder sample 3 via a joint 40 of the sample container 4 using a bellows-type pipe. The second embodiment will be described below mainly with respect to the configuration different from that of the first embodiment. The vibration device 20 controlled by a vibration device control unit 21 vibrates the sample container 4 containing the molybdenum trioxide 100 (Mo-100.O3) powder sample 3. Since the vibration can be damped at the joint portions connected by the bellows-type pipes, the pipe that carries the oxygen gas, or the gas mixture 31 of the oxygen gas and the noble gas, and the pipe connected to the radionuclide separation/purification unit 5 are not subjected to unnecessary stress, so that the integrity of the apparatus can be maintained even during the application of vibration. The vibration device 20 may be an ultrasonic vibration device that uses a heat-resistant ultrasonic vibrator. According to this embodiment, a radionuclide having a short half life, such as molybdenum 99 (Mo-99) or technetium 99m (Tc-99m) that is in great demand as a source material of a nuclear medicine diagnostic pharmaceutical can be efficiently produced with a small-sized, lightweight apparatus. The configuration of an apparatus for producing a radionuclide according to a third embodiment, which is one preferred embodiment of the present invention, will be described below with reference to FIG. 4. The apparatus for producing a radionuclide of this embodiment has a fundamental configuration similar to that of the apparatus for producing a radionuclide of the second embodiment, but differs from that of the second embodiment in that a spirally-wound heating/cooling pipe 60 is provided in the heating/cooling section 13 relating to the pipe that carries the oxygen gas, or the gas mixture 31 of the oxygen gas and the noble gas, that has been temperature regulated. Providing the spirally-wound heating/cooling pipe 60 in the heating/cooling section 13 as in the apparatus for producing a radionuclide of this embodiment can increase the time period during which the oxygen gas, or the gas mixture 30 of the oxygen gas and the noble gas, flows in the heating/cooling section 13. In addition, the apparatus for producing a radionuclide of this embodiment includes a mesh filter 50 in each of a pipe section that connects the sample container 4 which contains the molybdenum trioxide 100 (Mo-100.O3) powder sample 3 to the separation/purification unit 5, and a pipe section that connects the sample container 4 which contains the molybdenum trioxide 100 (Mo-100.O3) powder sample 3 to the pipe that carries the oxygen gas, or the gas mixture of the oxygen gas and the noble gas. The mesh filter 50 has a mesh opening size smaller than the particle size of the powder sample, and large enough to permit passage of the oxygen gas and the noble gas, where applicable, or the gaseous technetium oxide 99m (Tc-99m2.O7) and the gaseous technetium oxide 99 (Tc-992.O7). Providing the mesh filter 50 as in this embodiment prevents the molybdenum trioxide 100 (Mo-100.O3) powder sample 3 from entering the separation/purification unit 5, and permits only the oxygen gas and the noble gas, where applicable, or the gaseous technetium oxide 99m (Tc-99m2.O7) and the gaseous technetium oxide 99 (Tc-992.O7) to enter the separation/purification unit 5. According to this embodiment, a radionuclide having a short half life, such as molybdenum 99 (Mo-99) or technetium 99m (Tc-99m) that is in great demand as a source material of a nuclear medicine diagnostic pharmaceutical can be efficiently produced with a small-sized, lightweight apparatus. An apparatus for producing a radionuclide according to a fourth embodiment, which is one preferred embodiment of the present invention, will be described below with reference to FIG. 5. FIG. 5 is a configuration diagram illustrating an example of the radionuclide-producing sample section included in an apparatus for producing a radionuclide of this embodiment. In this embodiment, the sample container 4 is formed of a bellows-type pipe. The molybdenum trioxide 100 (Mo-100.O3) powder sample 3 is placed in the bellows-type pipe, and an electron beam 2 is emitted thereto. During an irradiation period during which the electron beam 2 is emitted, the vibration device 20, controlled by the vibration device control unit 21, vibrates the sample container 4 which contains the molybdenum trioxide 100 (Mo-100.O3) powder sample 3. The temperature of the sample during the irradiation is measured by the thermometer 10 having a thermocouple or other means. A gas mixture feed pipe 61 for feeding the oxygen gas, or the gas mixture of the oxygen gas and the noble gas, that has been temperature regulated, and a stationary portion of the sample container 4 are connected together, and a mesh filter 50 is provided in this connection portion. The mesh filter 50 has a mesh opening size smaller than the particle size of the powder sample 3, and large enough to permit passage of the oxygen gas and the noble gas, where applicable, or the gaseous technetium oxide 99m (Tc-99m2.O7) and the gaseous technetium oxide 99 (Tc-992.O7). A similar filter is provided in the pipe section that connects the sample container 4 to the separation/purification unit 5. The apparatus for producing a radionuclide of this embodiment allows the sample configuration to become more compact, and moreover, permits only the O2 gas and the noble gas, where applicable, or the gaseous technetium oxide 99m (Tc-99m2.O7) and the gaseous technetium oxide 99 (Tc-992.O7) to enter the separation/purification unit 5 without allowing the molybdenum trioxide 100 (Mo-100.O3) powder sample 3 to enter the separation/purification unit 5. According to this embodiment, a radionuclide having a short half life, such as molybdenum 99 (Mo-99) or technetium 99m (Tc-99m) that is in great demand as a source material of a nuclear medicine diagnostic pharmaceutical can be efficiently produced with a small-sized, lightweight apparatus. An apparatus for producing a radionuclide according to a fifth embodiment, which is one preferred embodiment of the present invention, will be described below with reference to FIG. 6. FIG. 6 is a configuration diagram illustrating an example of the radionuclide-producing sample section included in an apparatus for producing a radionuclide of this embodiment. In this embodiment, the gas mixture feed pipe 61 for feeding the oxygen gas, or the gas mixture of the oxygen gas and the noble gas, that has been temperature regulated, is connected to a bottom portion of the sample container 4 which contains the molybdenum trioxide 100 (Mo-100.O3) powder sample 3. A mesh filter 50 is provided in this connection portion. The mesh filter 50 has a mesh opening size smaller than the particle size of the powder sample 3, and large enough to permit passage of the oxygen gas and the noble gas, where applicable, or the gaseous technetium oxide 99m (Tc-99m2.O7) and the gaseous technetium oxide 99 (Tc-992.O7). The apparatus for producing a radionuclide of this embodiment has a similar mesh filter provided in the pipe section that connects the sample container 4 to the radionuclide separation/purification unit 5. The introduction of the oxygen gas, or the gas mixture of the oxygen gas and the noble gas, from a bottom portion causes the molybdenum trioxide 100 (Mo-100.O3) powder sample 3 to blow upward, and thus the flow of the oxygen gas, or of the gas mixture of the oxygen gas and the noble gas, creates a stirring effect, which achieves an effect similar to applying vibration to the molybdenum trioxide 100 (Mo-100.O3) powder sample 3. The apparatus for producing a radionuclide of this embodiment needs no particular vibration device, and thus further size reduction can be obtained. According to this embodiment, a radionuclide having a short half life, such as molybdenum 99 (Mo-99) or technetium 99m (Tc-99m) that is in great demand as a source material of a nuclear medicine diagnostic pharmaceutical can be efficiently produced with a small-sized, lightweight apparatus. |
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060118253 | claims | 1. A method for producing a radionuclide from a target nuclide using an accelerator capable of generating a beam of charged particles at energies of at least about 5 MeV, the method comprising loading a solid target comprising the target nuclide in a target holder mounted in line with the charged-particle beam generated by the accelerator and adapted to releasably hold the target in position for irradiation by the charged-particle beam, irradiating the target held by the target holder with the charged-particle beam at energies of at least about 5 MeV to form the radionuclide, removing the irradiated target from the target holder, transferring the removed irradiated target to an automated separation system, and separating the radionuclide from unreacted target nuclide using the automated separation system. transferring the irradiated target from the target holder to a fluid conveyance system, conveying the irradiated target through the conveyance system, and transferring the irradiated target from the conveyance system to the separation system. loading a solid target comprising the target nuclide in a target holder, irradiating the target with the charged-particle beam at energies of at least about 5 MeV to form the radionuclide, transferring the irradiated target from the target holder to a fluid conveyance system comprising a transfer fluid moving through a transfer line, conveying the irradiated target using the conveyance system, and separating the radionuclide from unreacted target nuclide. loading a solid target in a target holder adapted for use with the accelerator, the target comprising a substrate consisting essentially of an inert material and a target layer electroplated on a surface of the substrate, the target laver consisting essentially of a target nuclide capable of reacting with charged particles generated by the accelerator at energies ranging from about 5 MeV to about 25 MeV to form the radionuclide and having a projected thickness that will produce at least about 50% of the thick target yield for the reaction, irradiating the target with a beam of charged particles generated by the accelerator for at least about one hour to form the radionuclide, the beam having an energy ranging from about 5 MeV to about 25 MeV and a current sufficient to produce a clinically significant yield of the radionuclide, removing the irradiated target from the target holder, transferring the removed irradiated target to an automated separation system, and separating the radionuclide from unreacted target nuclide using the automated separation system. loading a solid target in a target holder adapted for use with the accelerator, the target comprising a substrate consisting essentially of an inert material and a target layer electroplated on a surface of the substrate, the target layer consisting essentially of a target nuclide capable of reacting with charged particles generated by the accelerator at energies ranging from about 5 MeV to about 25 MeV to form the radionuclide and having a projected thickness that will produce at least about 50% of the thick target yield for the reaction, irradiating the target with a beam of charged particles generated by the accelerator for at least about one hour to form the radionuclide, the beam having an energy ranging from about 5 MeV to about 25 MeV and a current sufficient to produce a clinically significant yield of the radionuclide, transferring the irradiated target from the target holder to a fluid conveyance system conveying the irradiated target using the conveyance system, and transferring the irradiated target from the conveyance system to the separation system. loading the target in a target holder suitable for use with an accelerator capable of generating the proton beam at energies greater than about 5 MeV, irradiating a target comprising isotonically enriched .sup.64 Ni with a proton beam to produce .sup.64 Cu according to the reaction .sup.64 Ni(p,n).sup.64 Cu in an amount which is at least sufficient for preparing a radioimaging agent, the proton beam having an energy of at least about 5 MeV and a current at least sufficient to produce an amount of .sup.64 Cu sufficient for clinical use in a radioimaging agent during the period of irradiation, removing the irradiated target from the target holder, transferring the removed irradiated target to an automated separation system suitable for separating .sup.64 Cu from unreacted .sup.64 Ni, and separating .sup.64 Cu from unreacted .sup.64 Ni, the separated .sup.64 Cu having a specific activity at least sufficient for clinical use in a radioimaging agent. loading the .sup.64 Ni target in a target holder adapted for use with a proton accelerator capable of generating a proton beam at energies ranging from about 5 MeV to about 25 MeV, the target holder including an elongated body and a cooling head, the body having an irradiation chamber and a front seat adapted to sealingly engage the target, the front seat having an aperture for allowing fluid communication between the irradiation chamber and the target, the cooling head having a cavity and a back seat adapted to sealingly engage the target, the back seat having an aperture for allowing fluid communication between the cavity and the target, the cooling head being retractable from the body to allow for loading and unloading the target from the target holder and being engageable with the body to hold the target against the body during irradiation, the target being loaded in the target holder by positioning the target against the front seat of the body or the back seat of the cooling head and drawing a vacuum in the chamber or in the cavity, respectively, to hold the target in such position before the cooling head is engaged, engaging the cooling head to hold the target against the body, after the target is irradiated, retracting the cooling head from the body and holding the irradiated target in place against the cooling head or against the body by vacuum after the cooling head is retracted, and unloading the irradiated target from the target holder by pressurizing the chamber or the cavity, the pressure being effective to act through the aperture in the front seat or back seat, respectively, to separate the target from the front seat or back seat and eject the target for further processing. exposing the target to an acidic solution in the dissolution vessel to dissolve the target layer off of the substrate, thereby forming a target-layer solution comprising .sup.64 Cu, .sup.64 Ni and other radionuclides, passing the target-layer solution through the anion-exchange column and collecting a first eluate therefrom, the first eluate being substantially enriched in nickel relative to copper, and passing water or an acidic solution having a normality of about 0.5 N through the anion-exchange column and collecting a second eluate therefrom, the second eluate being substantially enriched in .sup.64 Cu relative to other radionuclides or impurities. 2. The method as set forth in claim 1 wherein the step of transferring the removed irradiated target to the separation system includes conveying the irradiated target through a fluid conveyance system comprising a transfer fluid moving through a transfer line. 3. The method as set forth in claim 2 wherein the transfer fluid contacts the irradiated target to transfer the irradiated target through the transfer line without using a transfer capsule. 4. The method as set forth in claim 1 wherein the step of transferring the removed irradiated target to the separation system includes 5. The method as set forth in claim 1 wherein the target holder comprises an elongated body adapted to sealingly engage the accelerator and a cooling head, the body having an irradiation chamber and a front seat adapted to sealingly receive the target, the front seat having an aperture for allowing fluid communication between the irradiation chamber and the target, the cooling head including a cavity and a back seat adapted to sealingly receive the target, the back seat having an aperture for allowing fluid communication between the cavity and the target, the head being retractable from the body to allow for loading and unloading the target from the target holder and being engageable with the body to hold the target against the front seat of the body during irradiation, and wherein the step of loading the target in the target holder comprises positioning the target against the front seat of the body or the back seat of the cooling head and drawing a vacuum in the irradiation chamber or in the cavity, respectively, to hold the target in such position at least until the head is engaged with the chamber. 6. The method as set forth in claim 1 wherein the target holder comprises an elongated body and a cooling head, the body including an irradiation chamber and a front seat adapted to sealingly receive the target, the front seat having an aperture for allowing fluid communication between the irradiation chamber and the target, the cooling head including a cavity and a back seat adapted to sealingly receive the target, the back seat having an aperture for allowing fluid communication between the cavity and the target, the cooling head being retractable from the body to allow for loading and unloading the target from the target holder and being engageable with the body to hold the target against the front seat of the body during irradiation, and wherein the step of removing the irradiated target from the target holder comprises retracting the cooling head from the body after the target is irradiated, the irradiated target being held in place against the cooling head seat or against the body seat by vacuum after the cooling head is retracted, and pressurizing the chamber or the cavity, the pressure being effective to act through the aperture in the front seat or back seat, respectively, to separate the target from the front seat or back seat and eject the target for further processing. 7. The method as set forth in claim 1 wherein the target is irradiated with a charged particle beam generated in a low or medium energy accelerator at a beam energy ranging from about 5 MeV to about 25 MeV. 8. The method as set forth in claim 1 wherein the target nuclide is .sup.64 Ni and the target is irradiated with protons to form .sup.64 Cu according to the reaction .sup.64 Ni(p,n).sup.64 Cu. 9. A method for producing a radionuclide from a target nuclide using an accelerator capable of generating a beam of charged particles at energies of at least about 5 MeV, the method comprising 10. The method as set forth in claim 9 wherein the irradiated target is removed from the target holder prior to being conveyed to the conveyance system. 11. The method as set forth in claim 9 wherein the transfer fluid contacts the irradiated target to transfer the irradiated target through the transfer line without using a transfer capsule. 12. A method for producing a radionuclide from a target nuclide using an accelerator capable of generating a beam of charged particles at energies ranging from about 5 MeV to about 25 MeV, the method comprising 13. A method for producing a radionuclide from a target nuclide using an accelerator capable of generating a beam of charged particles at energies ranging from about 5 MeV to about 25 MeV, the method comprising 14. A method for producing .sup.64 Cu suitable for use in preparing a radiopharmaceutical agent for clinical applications, the method comprising 15. The method as set forth in claim 12 wherein the target layer has a projected thickness that will produce at least about 75% of the thick target yield. 16. The method as set forth in claim 12 wherein the target layer has dimensions that define a target area and the charged-particle beam impinges the target over an area which substantially matches the target area. 17. The method as set forth in claim 12 wherein the charged-particles generated by the accelerator travel unimpeded from the accelerator to the target during irradiation without passing through an attenuating foil or window. 18. The method as set forth in claim 12 wherein the target nuclide is .sup.64 Ni and the target is irradiated with protons to form .sup.64 Cu according to the reaction .sup.64 Ni(p,n).sup.64 Cu. 19. The method as set forth in claim 13 wherein the target layer has a projected thickness that will produce at least about 75% of the thick target yield. 20. The method as set forth in claim 13 wherein the target layer has dimensions that define a target area and the charged-particle beam impinges the target over an area which substantially matches the target area. 21. The method as set forth in claim 13 wherein the charged-particles generated by the accelerator travel unimpeded from the accelerator to the target during irradiation without passing through an attenuating foil or window. 22. The method as set forth in claim 13 wherein the target nuclide is .sup.64 Ni and the target is irradiated with protons to form .sup.64 Cu according to the reaction .sup.64 Ni(p,n).sup.64 Cu. 23. The method as set forth in claim 14 wherein the amount of .sup.64 Cu produced is at least an amount sufficient for preparing a radiotherapeutic agent and the specific activity of the separated .sup.64 Cu is sufficient for clinical use in a radiotherapeutic agent. 24. The method as set forth in claim 14 wherein the target is irradiated for at least about 1/2 hour with a proton beam having a current sufficient to produce at least about 100 mCi of .sup.64 Cu in less than about 24 hours. 25. The method as set forth in claim 14 wherein the .sup.64 Ni comprises less than about 250 ppm by weight carrier copper, and the target is irradiated for at least about 1 hour with a proton beam having an energy ranging from about 5 MeV to about 25 MeV and a current sufficient to produce at least about 200 mCi of .sup.64 Cu in less than about 12 hours. 26. The method as set forth in claim 14 wherein the amount of .sup.64 Cu produced is at least about 10 mCi. 27. The method as set forth in claim 14 wherein the amount of .sup.64 Cu produced is at least about 100 mCi. 28. The method as set forth in claim 14 wherein the separated .sup.64 Cu has a specific activity of at least about 15 mCi/.mu.g Cu. 29. The method as set forth in claim 14 wherein the separated .sup.64 Cu has a specific activity of at least about 100 mCi/.mu.g Cu. 30. The method as set forth in claim 14 wherein the beam energy ranges from about 5 MeV to about 25 MeV. 31. The method as set forth in claim 30 wherein the beam current ranges from about 1 .mu.A to about 1 mA at about 5 MeV, to about 150 .mu.A at about 8 MeV, to about at 100 .mu.A at about 11 MeV, to about 60 .mu.A at about 25 MeV and to about 45 .mu.A at about 25 MeV. 32. The method as set forth in claim 14 wherein the target comprises a substrate and a target layer formed on a surface of the substrate, the target layer consisting essentially of isotopically enriched .sup.64 Ni and having a projected thickness of at least about 20 .mu.m, the substrate consisting essentially of an inert material having a thermal conductivity which is about equal to or greater than the thermal conductivity of .sup.64 Ni. 33. The method as set forth in claim 32 wherein the target layer is an electroplated target layer. 34. The method as set forth in claim 32 wherein the target layer consists essentially of .sup.64 Ni enriched to at least about 95% and has a projected thickness ranging from about 20 .mu.m to about 500 .mu.m, and the substrate consists essentially of gold and has a front surface, a back surface substantially parallel to and opposing the front surface and a thickness ranging from about 0.5 mm to about 2 mm. 35. The method as set forth in claim 32 wherein the .sup.64 Ni target is irradiated with a proton beam having an energy ranging from about 5 MeV to about 25 MeV, the method further comprising 36. The method as set forth in claim 14 wherein the target comprises a target layer formed over a surface of a substrate, the target layer including, after irradiation, .sup.64 cu, unreacted .sup.64 Ni and other radionuclides, and .sup.64 Cu is separated from unreacted .sup.64 Ni and from the substrate layer using a separation unit, the separation unit including a shielded housing that encloses components arranged to facilitate automatic and remote separation of the .sup.64 Cu, the components being selected from the group consisting of one or more fluid containers, an ion exchange column, and one or more pipetters in isolable fluid communication with the containers or the column, the .sup.64 Cu being separated by 37. The method as set forth in claim 36 wherein the pipetters include a plunger and the acid solution in the dissolution vessel is agitated while the irradiated target is exposed to the acid solution by effecting repetitive upward and downward movements of the pipetter plunger in fluid communication with the dissolution vessel. 38. The method as set forth in claim 14 further comprising, after the step of separating the radionuclide from unreacted target nuclide, recycling the unreacted .sup.64 Ni for use in preparing another target. |
abstract | A laser induced thermal imaging (LITI) apparatus, an LITI method, and an organic light emitting display (OLED) device. An LITI apparatus for forming a light emitting layer of an OLED device includes a substrate stage adapted to receive an accepter substrate and a donor film to be laminated, the accepter substrate having a pixel area of the OLED device and a magnet, the donor film having the light emitting layer transferred to the pixel area; a laser oscillator for radiating a laser to the donor film; a contact frame adapted to be placed between the substrate stage and the laser oscillator, the contact frame having at least one transmission portion corresponding to the light emitting layer transferred to the acceptor substrate and having a magnet for forming a magnetic force with the accepter substrate; and a contact frame transferring mechanism for moving the contact frame toward the substrate stage. |
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abstract | A blocking device for preventing the actuation of an automatic depressurization system in a pressurized nuclear reactor system due to spurious signals resulting from a software failure. The blocking signal is removed when the coolant level within the core makeup tanks drop below a predetermined level. |
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description | The present disclosure relates generally to X-ray backscatter imaging systems, and more particularly to generating three dimensional image data of features within a composite structure. Known ultrasound systems do not provide information about the thickness of features, such as wrinkles, within a composite structure because they cannot “see” past the feature from one side. There is a need to supplement ultrasonic detection of wrinkles with a method that can quantify wrinkle thickness and better determine the shape of the wrinkles. Addressing this need would provide more accurate structural performance models and predictions. X-ray laminography provides sequential layers of information for the interior of a structure. However, existing X-ray laminography systems require access to both sides of a structure, in order to operate. X-ray backscatter systems provide a tool for seeing into structures from one side only. However, images collected for inspection or evaluation are typically two dimensional representations of a superposition of depth information. Various methods to collect three dimensional information have been attempted, but they have been impractical or slow with relatively poor accuracy, because of single X-ray insertion and collection angles. X-ray computed tomography can obtain accurate three dimensional X-ray attenuation data, but requires access to all sides of the structure being analyzed. Accordingly, computed tomography cannot be used for inspecting large aircraft parts, such as a wing or fuselage. A portable system is needed that combines the best of the above methods: a one-sided inspection system for three dimensional imaging of hidden features within a composite structure. In one aspect, an imaging system for generating three dimensional image data using X-ray backscattering from one side of a structure is provided. The imaging system includes at least one X-ray source, at least one rotating collimator coupled to the at least one X-ray source, an X-ray detector, and a controller coupled to the at least one rotating collimator and the X-ray detector. The controller is configured to emit X-rays from the at least one X-ray source through the at least one rotating collimator towards the one side of the structure. Additionally, the controller is configured to detect backscattered X-rays from the one side of the structure, using the X-ray detector, at a plurality of depths within the structure. Additionally, the controller is configured to generate three dimensional image data of the structure based on the detected backscattered X-rays. In another aspect, a method for generating three dimensional image data using X-ray backscattering from one side of a structure is provided. The method includes emitting X-rays from at least one X-ray source through at least one rotating collimator towards the one side of the structure. The method additionally includes detecting backscattered X-rays from the one side of the structure, using an X-ray detector, at a plurality of depths within the structure and generating three dimensional image data of the structure based on the detected backscattered X-rays. Described herein are implementations of a rotating collimator X-ray backscatter system for creating 3-D images (single and dual) through a laminography-type reconstruction. The disclosed system is portable, and scanning can be performed using a robotic arm, a stages system vacuum-mounted to a structure, a rotating stage mounted on the structure, or a crawling robot. Scanning is performed in a plane perpendicular to the structure, with each scatter point in space receiving a rotating pencil beam from two angles. These crossing beams can be made from two aligned rotating collimators (using shuttered or alternating beams), or a single rotating collimator that is precisely re-oriented (180 degrees) to scan a beam from the other side of the centerline, but in the same scan plane. Combining of backscatter images from the two angles facilitates maximizing the scatter information from that location, and facilitates minimizing the scatter information from the surrounding area, similar to laminography. Collecting and combining measurements at each point (based on the equations defined herein), can produce a scatter intensity line, which can be combined with parallel adjacent lines to create image layers (“slices”). These layers can be created from the same depth (d) data, collected through incremental rotation or translation of the system relative to the structure, from data collected in the plane of the pencil beam scans, or from the planes perpendicular to both. A method of linear reconstruction with the dual sources, as described herein, enables a crawling robot or scanner mounted on a track to continuously collect data on extended paths autonomously. The layers of scatter intensity data described above can be analyzed individually or stacked to create 3-D scatter intensity data sets that can be analyzed and viewed using 3-D imaging software. Accordingly, features such as wrinkles, fiber waviness, voids, and gaps can be identified and quantified in 3-D space for analysis, structural model performance predictions, and decision-making. Elements that serve a similar, or at least substantially similar, purpose are labeled with like numbers in each of FIGS. 1-15, and these elements may not be discussed in detail herein with reference to each of FIGS. 1-15. Similarly, all elements may not be labeled in each of FIGS. 1-15, but reference numerals associated therewith may be utilized herein for consistency. Elements, components, and/or features that are discussed herein with reference to one or more of FIGS. 1-15 may be included in and/or utilized with any of FIGS. 1-15 without departing from the scope of the present disclosure. FIG. 1 is a diagram of an example environment 100 in which an imaging system 102 generates image data of a structure 110, for example a composite structure such as an aircraft wing or fuselage. Imaging system 102 includes a controller 104, for example a computing device that controls operations of a backscatter imaging system 106. Backscatter imaging system 106 includes one or more X-ray sources, one or more rotating collimators that collimate X-rays generated by one or more X-ray sources, and at least one detector, for example a scintillating detector, that receives backscattered X-rays from an object to be imaged (e.g., structure 110) and generates corresponding electrical signals representing intensities of the backscattered X-rays at each of a plurality of pixels. In at least some implementations, the one or more X-ray sources are stationary, and do not rotate with the respective collimator. In at least some implementations, backscatter system 106 is coupled to a platform 108, which moves imaging system 102 relative to structure 110. In some implementations, platform 108 is a robotic arm, a crawling robot, and/or one or more rotating or translating stages, which are components that rotate and/or translate about a structure to be analyzed (e.g., structure 110), as described in more detail herein. In some implementations, platform 108 can be mounted to structure 110, for example with vacuum mounts. In some implementations, structure 110 is a portion of a vehicle, such as an aircraft. In operation, imaging system 102 is configured to emit X-rays into structure 110 at one or more depths and receive backscattered X-rays that provide information about features within structure 110 at each depth. For any given point at a given depth within structure 110, imaging system 102 emits separate X-ray beams at the point from two different angles and receives backscattered X-rays from each respective X-ray beam. Imaging system 102 generates imaging data from the backscattered X-rays from each respective X-ray beam and combines the imaging data to form three dimensional imaging data. Importantly, while structure 110 includes a first side 112 (e.g., a front side) and a second side 114 (e.g., back side), imaging system 102 only requires access to one of the sides (e.g., first side 112) to generate three dimensional imaging data of the inside of structure 110. Accordingly, imaging system 102 provides a convenient way to perform laminography of structure 110, without removing structure 110 from its current environment. FIG. 2 is an elevation view of a first example imaging system 200. More specifically, imaging system 200 is an implementation of imaging system 102. Imaging system 200 includes a first X-ray source 202 and a second X-ray source 204. First X-ray source 202 is coupled to a first rotating collimator 206 having a first plurality of apertures 205. First rotating collimator 206 has N (e.g., six) apertures 205, wherein each of the apertures are approximately equidistantly spaced around a center point or extend from a center point radially outward. By including multiple apertures in first rotating collimator 206, rather than just one aperture, imaging system 200 is able to collect data much more quickly. As the first collimator 206 rotates, for example in a counter-clockwise direction, one of the first plurality of apertures 205 directs a collimated pencil beam 218 towards structure 110. More specifically, first X-ray source 202 emits a fan beam, which is then collimated by first collimator 206. The output from first collimator 206 is collimated pencil beam 218. Likewise, second X-ray source 204 is coupled to a second rotating collimator 208 having a second plurality of apertures 207, arranged similarly to the apertures 205 of first collimator 206. As the second collimator 208 rotates, for example in a clockwise direction, one of the second plurality of apertures 207 directs a second collimated pencil beam 220 towards structure 110. Specifically, second X-ray source 204 emits a fan beam, which is then collimated into second collimated pencil beam 220 by second collimator 208. For a given point 222 along a line 216 within structure 110, imaging system 200 impinges point 222 with X-rays, generated by two different X-ray sources emitting X-rays from two different angles (e.g., first collimated pencil beam 218 and second collimated pencil beam 220). A detector 210 receives the backscattered X-rays (shown in FIG. 6) and generates corresponding electrical signals that are used by controller 104 (FIG. 1) to generate image data (shown in FIG. 4). As described in more detail herein, imaging system 200 includes shutters 226 and 228 that control the emission of X-rays from each collimator (e.g., first rotating collimator 206 and second rotating collimator 208) such that detector 210 receives backscattered X-rays from only one pencil beam at a time. In some implementations, imaging system 102 additionally or alternatively uses electronic shuttering of a power supply (e.g., alternatingly activating and deactivating the power supply) and/or synchronization of the rotating collimators 206 and 208 to avoid overlap of the pencil beams 218 and 220. As stated above, every point (e.g., point 222) along line 216 has two angles of impingement, one angle of impingement (α) from the pencil X-ray beam 218 generated by the first X-ray source/collimator and a second angle of impingement (13) generated by the second X-ray source/collimator. Reconstruction of scatter intensities on a point by point basis is therefore associated with the angles of the X-rays impinging on the structure 110, as follows:tan(α)+tan(β)=L/d (Equation 1)Pn=tan(α)×d (Equation 2)Pn=L−tan(β)×d (Equation 3) When both α and β are at a zero angle, the ray is substantially perpendicular to the surface of the object being inspected (e.g., structure 110). Thus, the angle α or β is the angle defined between a ray that would be normal to the surface and the ray currently being emitted. Controller 104 combines image data collected at angle α, representing the angle the first pencil beam 218 impinges on the structure 110, and angle β, representing the angle of second pencil beam 220 emanating from second rotating collimator 208, to generate an image of an nth point (Pn) (e.g., point 222) along length L (e.g., length 217), representing a length of a line segment that backscatter imaging system 106 scans across, at a depth d (e.g., depth 224). Controller 104 causes backscatter imaging system 106 to scan (emit X-ray pencil beams and detect backscattered X-rays) at a plurality of positions (e.g., position 222) along lines (e.g., line 216) at different depths (e.g., depth 224) within structure 110. Furthermore, imaging system 200 includes a ring bearing 212, for example mounted to or incorporated into platform 108 (FIG. 1) that enables imaging system 200 to rotate relative to structure 110, as imaging system 200 scans structure 110. FIG. 3 is a top-down view showing rotation of imaging system 200. More specifically, the first rotating collimator 206 and the second rotating collimator 208 may be rotated 180 degrees on ring bearing 212 around a center point (e.g., a point halfway along length 217). Accordingly, for a given depth (e.g., depth 224), imaging system 200 generates a circular slice of image data. As described in more detail with reference to FIGS. 5 and 6, controller 104 combines the generated image data to generate three dimensional image data of structure 110. FIG. 4 is a diagram of first image data 400 of structure 110 and second image data 402 of structure 110 being combined into three dimensional image data 404. More specifically, controller 104 receives, from detector 210, electrical signals for each pixel of detector 210. For example detector 210 transmits a voltage representing an intensity of backscattered X-rays for each respective pixel. Controller 104 generates first image data 400 based on intensities of X-rays backscattered from pencil beams (e.g. first pencil beam 218) emitted from first rotating collimator 206. Likewise, controller 104 generates second image data 402 based on intensities of X-rays backscattered from pencil beams (e.g. second pencil beam 220) directed out of second rotating collimator 208. Controller 104 then combines first image data 400 with second image data 402 to form three dimensional image data 404 (e.g., a stereoscopic image). For example, when viewed by a human, first image data 400 is displayed to a first eye and second image data 402 is displayed to a second eye to enable the viewer to perceive depth in three dimensional image data 404. FIG. 5 illustrates a timeline 500 of synchronization of sets of first image data 400 and sets of second image data 402 to generate three dimensional image data 404. More specifically, a synchronization system 501, for example controller 104 in combination with shutters (e.g., shutters 226 and 228 shown in FIG. 6) coupled to first rotating collimator 206 and second rotating collimator 208, causes imaging system 200 to generate a first image datum 502 for a first point (e.g., point 222) from X-rays backscattered from first rotating collimator 206, then receive a second image datum 504 for the first point (e.g., point 222) from X-rays backscattered from second rotating collimator 208. Further, imaging system generates a third image datum 506 corresponding to a second different point along the same line from the first rotating collimator 206, a fourth image datum 508 of the second point from the second rotating collimator 208, etc. Thus, at each different point, two image datums are acquired, one from backscatter information acquired from backscatter associated with the first rotating collimator 206 and a second from the backscatter associated with the second rotating collimator 208. In the exemplary embodiment, images datums 502 . . . 516 are acquired. Imaging system 200 then combines corresponding data (e.g., first image datum 502 with second image datum 504, third image datum 506 with fourth image datum 508, etc.) to form stereoscopic image data (e.g., image data 518) along each scanned line at a given depth. Further, imaging system 200 combines the stereoscopic image data for each line at a given depth to produce a set of three dimensional image data 520 for a given plane or “slice” at the given depth. Further, the imaging system 200 may repeat the above process to generate parallel slices of image data 520, each slice corresponding to a different depth within structure 110. FIG. 6 is a diagram of backscattered X-rays being detected by imaging system 200. A first shutter 226, for example a mechanical shutter, is coupled to aperture 205 of first rotating collimator 206. Likewise, a second shutter 228 is coupled to aperture 207 of second rotating collimator 208. In the exemplary embodiment, each aperture has its own shutter. More specifically, if the collimators 206 and 208 each have N apertures, the imaging system 200 will include N shutters, wherein each respective shutter is positioned in front of the aperture. First shutter 226 and second shutter 228 are synchronized, for example by electrical signals (not shown) from controller 104, such that first rotating collimator 206 outputs first pencil beam 218 while first shutter 226 is open and second shutter 228 is closed. Detector 210 then detects first backscattered X-rays 606 caused by first pencil beam 218 impinging on point 222 of structure 110. Subsequently, second rotating collimator 208 outputs second pencil beam 220 while first shutter 226 is closed and second shutter 228 is open. Detector 210 then detects second backscattered X-rays 608 caused by second pencil beam 220 impinging on the same point (e.g. point 222) of structure 110. Imaging system 200 continues operating in this manner, alternately emitting pencil beams from each X-ray source and detecting backscattered X-rays for each point along each line (e.g., line 216) within a plane or slice at each of a plurality of depths (e.g., depth 224) within structure 110. FIG. 7 is an elevation view of a second example imaging system 700. FIG. 8 is a top-down view showing rotation of imaging system 700. Imaging system 700 is an implementation of imaging system 102. Imaging system 700 is similar to imaging system 200 (shown in FIG. 3) except rather than having two X-ray sources and two rotating collimators, imaging system 700 includes only a single X-ray source 202 and a single rotating collimator 206. Imaging system 700 rotates 360 degrees on a ring bearing 212. Accordingly, rotating collimator 206 impinges point 222 with a pencil beam when rotating collimator 206 is at a first position 710, and when ring bearing has rotated 180 degrees from first position 710 such that rotating collimator 206 is at second position 712. Accordingly, Equations 1, 2, and 3, described with reference to imaging system 200 also apply to reconstruction of scatter intensities detected with imaging system 700. FIG. 8 is a top-down view showing rotation of imaging system 700. FIG. 9 is an elevation view of a third example imaging system 800. FIG. 10 is a top-down view of imaging system 800, showing stages that enable translation of imaging system along two axes. Imaging system 800 is an implementation of imaging system 102 and is similar to imaging system 200. Imaging system includes the first X-ray source 202, the second X-ray source 204, the first rotating collimator 206, the second rotating collimator 208, and the detector 210. Additionally, imaging system 800 includes shutters (not shown) coupled to first rotating collimator 206 and second rotating collimator 208, similar to first shutter 226 (FIG. 6) and second shutter 228 (FIG. 6), to synchronize the emission of pencil beams, as described above. However, unlike imaging system 200, platform 108 of imaging system 800 includes a first stage 812 that is coupled to a first rail 816 and a second rail 818. First rail 816 and second rail 818 are coupled to a first vacuum mount 820, a second vacuum mount 822, a third vacuum mount 824, a fourth vacuum mount 826, a fifth vacuum mount 828, and a sixth vacuum mount 830. Vacuum mounts 820, 822, 824, 826, 828, and 830 attach to structure 110. First stage 812 includes a first stepper motor (not shown) that is configured to translate first stage 812 along a first axis 832 (e.g., an X-axis). In some implementations, platform 108 includes a second stage 814 that is coupled to first stage 812. Second stage 814 includes a second stepper motor (not shown) that is configured to translate second stage 814 along a second axis 834 (e.g., a Y-axis) that is perpendicular to first axis 832. Accordingly, in at least some implementations, controller 104 transmits electrical signals to first stage 812 and second stage 814 while imaging system 800 is scanning structure 110 to, for example, translate first stage 812 along rails 816 and 818 (i.e., first axis 832) while second stage 814 is at a first position along second axis 834, then translate second stage 814 to a second position along second axis 834, and then translate first stage 812 along rails 816 and 818 (i.e., first axis 832) while second stage 814 is at the second position along second axis 834. By including second stage 814, imaging system 800 is able to move incrementally between linear scans, such as across wrinkles or stiffener features, for additional scatter data around the feature, which can improve three dimensional imaging of the feature. FIG. 11 is an elevation view of a fourth example imaging system 900. FIG. 12 is a top-down view of imaging system 900. Imaging system 900 is an implementation of imaging system 102. Further, imaging system 900 is similar to imaging system 800, except instead of having two rotating collimators, imaging system 900 includes only one rotating collimator 206. Additionally, instead of synchronizing the emission of pencil beams from each collimator to impinge point 222 from two different angles, imaging system 900 scans structure 110 by impinging each point (e.g., point 222) with X-rays from when first collimator 206 is in a first position 904, then first collimator 206 is moved (e.g., remounted) to second position 906 and rescans structure 110. Accordingly, imaging system 900 takes twice as long as imaging system 800 to scan structure 110, but requires only a single rotating collimator (e.g., rotating collimator 206). FIG. 13 is an elevation view of a fifth example imaging system 1000. Imaging system 1000 is an implementation of imaging system 102 and is similar to imaging system 800, except as described herein. Rather than performing reconstruction of scatter intensities along a horizontal line (e.g., line 216 in FIG. 2), imaging system 1000 performs reconstruction along a vertical line 1002. Every point (Pn) along vertical line 1002 will have two angles of impingement (e.g., α and β) from pencil X-ray beams. More specifically, imaging system 1000 performs reconstruction according to the following equations:tan(α)=Pn/d (Equation 4)tan(β)=(L−Pn)/d (Equation 5) Imaging system 1000 combines images collected at α and β during scanning, similar to the process described with reference to FIGS. 4 and 5 to generate three dimensional image data. FIG. 14 is a diagram of an example computing device 1102. Computing device 1102 is representative of controller 104. Computing device 1102 includes one or more processors 1105 for executing instructions. In some implementations, executable instructions are stored in a memory device 1110. Processor 1105 may include one or more processing units (e.g., in a multi-core configuration). One or more memory devices 1110 are any one or more devices allowing information such as executable instructions and/or other data to be stored and retrieved. One or more memory devices 1110 may include one or more computer-readable media. Computing device 1102 also includes at least one media output component 1115 for presenting information to a user 1101. Media output component 1115 is any component capable of conveying information to user 1101. In some implementations, media output component 1115 includes an output adapter such as a video adapter and/or an audio adapter. An output adapter is operatively coupled to processor 1105 and operatively couplable to an output device such as a display device (e.g., a liquid crystal display (LCD), organic light emitting diode (OLED) display, cathode ray tube (CRT), or “electronic ink” display) or an audio output device (e.g., a speaker or headphones). In some implementations, computing device 1102 includes an input device 1120 for receiving input from user 1101. Input device 1120 may include, for example, a keyboard, a pointing device, a mouse, a stylus, a touch sensitive panel (e.g., a touch pad or a touch screen), a gyroscope, an accelerometer, a position detector, or an audio input device. A single component such as a touch screen may function as both an output device of media output component 1115 and input device 1120. Computing device 1102 additionally includes a communication interface 1125, which is communicatively couplable to another device such as backscatter system 106 and/or platform 108. Communication interface 1125 may include, for example, a wired or wireless network adapter or a wireless data transceiver for use with a mobile phone network (e.g., Global System for Mobile communications (GSM), 3G, 4G or Bluetooth) or other mobile data network (e.g., Worldwide Interoperability for Microwave Access (WIMAX)). Stored in one or more memory devices 1110 are, for example, computer-readable instructions for providing a user interface to user 1101 via media output component 1115 and, optionally, receiving and processing input from input device 1120. A user interface may include, text, graphics, and/or sound that enable user 1101 to interact with computing device 1102, for example to control operations of computing device 1102 and/or view output (e.g., three dimensional image data 404). The computer-readable instructions additionally cause computing device 1102 perform the processes for scanning structure 110, moving backscatter system 106, and combining image data into three dimensional image data. FIG. 15 is a flowchart of an example process 1200 for generating three dimensional image data using imaging system 102. Initially, imaging system 102 emits 1202 X-rays (e.g., first pencil beam 218 and second pencil beam 220) from at least one X-ray source (e.g., first X-ray source 202 and second X-ray source 204) through at least one rotating collimator (e.g., first rotating collimator 206 and second rotating collimator 208) towards one side (e.g., first side 112) of the structure 110. Additionally, imaging system 102 detects 1204 backscattered X-rays (e.g., first backscattered X-rays 606 and second backscattered X-rays 608) from the one side (e.g., first side 112) of the structure 110, using an X-ray detector (e.g., detector 210), at a plurality of depths (e.g., depths 224) within the structure 110. Further, imaging system 102 generates 1206 three dimensional image data (e.g., three dimensional image data 404) of the structure 110 based on the detected backscattered X-rays (e.g., first backscattered X-rays 606 and second backscattered X-rays 608). In some implementations, imaging system 102 emits a first X-ray beam (e.g., first pencil beam 218) through the first rotating collimator (e.g., first rotating collimator 206) at a first point (e.g., point 222) in structure 110 while the second shutter (e.g., second shutter 228) is closed, detects backscattered X-rays (e.g., first backscattered X-rays 606) from the first X-ray beam (e.g., first pencil beam 218), emits a second X-ray beam (e.g., second pencil beam 220) through the second rotating collimator (e.g., second rotating collimator 208) at the first point (e.g., point 222) while the first shutter (e.g., first shutter 226) is closed, and detects backscattered X-rays (e.g., second backscattered X-rays 608) from the second X-ray beam (e.g., second pencil beam 220). In some implementations, imaging system 102 emits a first X-ray beam (e.g., first pencil beam 218) at a first point (e.g., point 222) in the structure (e.g., structure 110) from a first angle (e.g., α), detects first backscattered X-rays (e.g., first backscattered X-rays 608) from the first X-ray beam (e.g., first pencil beam 218), generates first image data (e.g., first image data 400) from the first backscattered X-rays (e.g., first backscattered X-rays 608), emits a second X-ray beam (e.g., second pencil beam 220) at the first point (e.g., point 222) in the structure (e.g., structure 110) from a second angle (e.g., β), detects second backscattered X-rays (e.g., second backscattered X-rays 608) from the second X-ray beam (e.g., second pencil beam 220), generates second image data (e.g., second image data 402) from the second backscattered X-rays (e.g., second backscattered X-rays 608), and combines the first image data (e.g., first image data 400) with the second image data (e.g., second image data 402). In some implementations, imaging system 102 includes a platform (e.g., platform 108) that is coupled to the structure (e.g., structure 110) and imaging system 102 is configured to translate across at least a portion (e.g., first side 112) of the structure (e.g., structure 110) while emitting the X-rays (e.g., first pencil beam 218 and second pencil beam 220) and detecting the backscattered X-rays (e.g., first backscattered X-rays 606 and second backscattered X-rays 608). In some implementations, imaging system 102 includes a platform (e.g., platform 108) that is coupled to the (e.g., structure 110) and imaging system 102 rotates relative to the structure (e.g., structure 110) while emitting the X-rays (e.g., first pencil beam 218 and second pencil beam 220) and detecting the backscattered X-rays (e.g., first backscattered X-rays 606 and second backscattered X-rays 608). In some implementations, imaging system 102 generates parallel slices of image data (e.g., image data 520) corresponding to a plurality of depths 224 within structure 110. In some implementations, imaging system 102 combines image data from two angles (e.g., angles α and β) along a plane 1002 that is perpendicular to the structure (e.g., structure 110). In some implementations, imaging system 102 includes a platform (e.g., platform 108) that includes a first stage (e.g., first stage 812) that is adapted to translate along a first axis (e.g., first axis 832), a second stage (e.g., second stage 814) that is coupled to the first stage (e.g., first stage 812) and is adapted to translate along a second axis (e.g., second axis 834) that is perpendicular to the first axis (e.g., first axis 832), and at least one rail (e.g., first rail 816) that is coupled to at least the first stage (e.g., first stage 812) and is adapted to mount to the structure (e.g., structure 110). In some implementations, imaging system 102 generates first image data 520 for a plurality of first points (e.g., line 216) at a first depth (e.g., depth 224), then generates second image data for a plurality of second points at a second depth in the structure (e.g., structure 110). In some implementations, imaging system 102 generates first image data 520 at a first plurality of depths 224 along a first plane 216 within the structure (e.g., structure 110), then generates second image data at a second plurality of depths along a second plane that is adjacent to the first plane. A technical effect of systems and methods described herein includes at least one of: (a) emitting X-rays from at least one X-ray source through at least one rotating collimator towards one side of a structure; (b) detecting backscattered X-rays from the one side of the structure, using an X-ray detector, at a plurality of depths within the structure; and (c) generating three dimensional image data of the structure based on the detected backscattered X-rays. As compared to known methods and systems for performing imaging of a structure, the systems and methods described herein enable generating three dimensional images of internal features of a structure while only accessing a single side of the structure. Accordingly, users of the systems and methods described herein may view and evaluate three dimensional image data of features within a structure without the need to disassemble the structure or have access to more than one side of the structure. More specifically, the systems and methods described herein solve an important problem pertaining to non-destructive evaluation and quantification of wrinkles in composite structures. Namely, the system and methods described herein provide the ability to see under a wrinkle in a composite structure to determine its depth, thickness, and magnitude. The description of the different advantageous implementations has been presented for purposes of illustration and description, and is not intended to be exhaustive or limited to the implementations in the form disclosed. Many modifications and variations will be apparent to those of ordinary skill in the art. Further, different advantageous implementations may provide different advantages as compared to other advantageous implementations. The implementation or implementations selected are chosen and described in order to best explain the principles of the implementations, the practical application, and to enable others of ordinary skill in the art to understand the disclosure for various implementations with various modifications as are suited to the particular use contemplated. This written description uses examples to disclose various implementations, which include the best mode, to enable any person skilled in the art to practice those implementations, including making and using any devices or systems and performing any incorporated methods. The patentable scope is defined by the claims, and may include other examples that occur to those skilled in the art. Such other examples are intended to be within the scope of the claims if they have structural elements that do not differ from the literal language of the claims, or if they include equivalent structural elements with insubstantial differences from the literal languages of the claims. |
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claims | 1. A method for exposing one or more areas on a semiconductor wafer to charged particles, comprising:aligning a wafer mask, having one or more mask patterns thereon, with the semiconductor wafer;bonding the semiconductor wafer with the wafer mask; andpassing the charged particle through then mask patterns to land on one or more selected areas on the semiconductor wafer. 2. The method of claim 1 wherein the charged particle transform semiconductor materials in the selected areas to have a higher resistivity than other areas not exposed to the charged particle. 3. The method of claim 2 wherein the selected areas become semi-insulating areas after receiving the charged particles with a collective energy level exceeding a predetermined threshold. 4. The method of claim 1 wherein the wafer mask has one or more alignment patterns thereon for aligning with the semiconductor wafer. 5. The method of claim 1 wherein the mask patterns are openings with their center portions unremoved. 6. The method of claim 5 wherein the openings are in rectangle shape. 7. The method of claim 1 wherein the wafer mask is made from a silicon wafer having a thickness between 100 um to 800 um. 8. The method of claim 1 wherein an angle between sidewalls of the mask patterns on the wafer mask and the surface of the wafer mask is between 80 to 100 degrees. 9. The method of claim 1 further comprising:determining the one or more selected areas on the semiconductor wafer for receiving the charged particles; andgenerating the wafer mask having the mask patterns containing thereon for passing the charged particles to land on the selected areas when the semiconductor wafer is aligned with the wafer mask. 10. The method of claim 1 wherein the charged particles are protons. 11. The method of claim 1 wherein the charged particles have an energy level between 0.5 to 5 MeV. 12. A method of exposing one or more areas on a semiconductor wafer to charged particles for making semi-insulating areas, comprising:aligning a wafer mask, having one or more mask patterns thereon, with the semiconductor wafer, with the mask being a predetermined proximity to the semiconductor wafer;bonding the semiconductor wafer with the wafer mask;generating the charged particles; anddirecting the charged particles through the mask patterns to land on one or more selected areas on the semiconductor wafer,wherein semiconductor materials in the selected areas of the semiconductor wafer are transformed to have a higher resistivity than those not exposed to the charged particles. 13. The method of claim 12 wherein the selected areas become semi-insulating areas. 14. The method of claim 12 further comprising:determining one or more selected areas on the semiconductor wafer for receiving the charged particles; andgenerating a wafer mask having one or more mask patterns contained thereon for passing the charged particles. 15. The method of claim 12 wherein the wafer mask has one or more alignment patterns thereon for aligning with the semiconductor wafer. 16. The method of claim 12 wherein the mask patterns are rectangular openings. 17. The method of claim 12 wherein the mask patterns are rectangular openings with their center portions of wafer material unremoved. 18. The method of claim 12 wherein the wafer mask has a thickness between 110 um to 800 um. 19. The method of claim 12 wherein an angle between sidewalls of the mask patterns on the wafer mask and the surface of the wafer mask is between 80 to 110 degrees. 20. The method of claim 12 wherein a fluence of the charge particles is between 1E14 ea/cm2 to 1E17 ea/cm2. 21. The method of claim 12 wherein the charged particles land substantially perpendicularly to the selected areas. 22. A system for exposing one or more areas on a semiconductor wafer to charged particles, comprising:means for aligning a wafer mask, having one or more mask patterns thereon with the semiconductor wafer;means for bonding the semiconductor wafer with the wafer mask:means for generating the charged particles; andmeans for directing the charged particles through the mask patterns to land on selected areas on the semiconductor wafer. 23. The system of claim 22 further comprising means for determining one or more selected areas on the semiconductor wafer for receiving the charged particles. 24. The system of claim 22 further comprising means for generating the mask patterns for passing the charged particles. 25. The system of claim 24 wherein the means for generating the mask further includes means for generating one or more alignment patterns thereon for aligning with the semiconductor wafer. 26. The system of claim 24 wherein the mask patterns are rectangular openings. 27. The system of claim 24 wherein the mask patterns are rectangular openings with their center portions of wafer material unremoved. 28. The system of claim 22 wherein the means for generating charged particles generates the charged particles with an energy level above 0.5 MeV. 29. The system of claim 22 wherein the means for generating charged particles generates the charged particles with an energy level below 5 MeV. |
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051986809 | claims | 1. A method for manufacturing a single focus collimator, comprising the steps of: forming grooves on a surface of a bulk block member; casting a metallic material having a sufficient .gamma. ray shielding property into said grooves formed on said bulk block member; and immersing said bulk block member with said metallic material casted into said grooves into a solvent capable of dissolving said bulk block member but not said metallic material, such that a collimator body formed by said metallic material in a shape of said grooves is obtained as said bulk block member is dissolved by said solvent. forming grooves on a surface of a bulk block member; casting a metallic material having a sufficient .gamma. ray shielding property into said grooves formed on said bulk block member; and immersing said bulk block member with said metallic material casted into said grooves into a solvent capable of dissolving said bulk block member but not said metallic material, such that a collimator body formed by said metallic material in a shape of said grooves is obtained, as said bulk block member is dissolved by said solvent. 2. The method of claim 1, wherein said metallic material is made of lead. 3. The method of claim 1, wherein at the forming step, said grooves are formed by using a disk cutter device. 4. The method of claim 1, wherein at the forming step, said grooves are formed by using a wire cut electric spark manufacturing process. 5. The method of claim 1, wherein said bulk block member is made of aluminum. 6. The method of claim 5, wherein said solvent is made of strong acid. 7. The method of claim 5, wherein said solvent is made of strong alkali. 8. The method of claim 1, wherein said bulk block member is made of sodium chloride. 9. The method of claim 8, wherein said solvent is made of water. 10. The method of claim 1, wherein each of said grooves is formed with a width in a range of 0.1 to 1 mm. 11. The method of claim 1, wherein said grooves are formed with such intervals that holes formed on said collimator body have larger size toward a center of said collimator body. 12. The method of claim 1, wherein said grooves are formed with such intervals that holes formed on said collimator body have smaller size toward a center of said collimator body. 13. The method of claim 1, wherein said grooves are formed with such widths that septa thicknesses of different sections of said collimator body are different. 14. A single focus collimator manufactured by a process comprising the steps of: |
048083186 | claims | 1. A process for separating and recovering cesium ions from a feed solution comprising: contacting the feed solution containing cesium together with other metal ions with a modified phlogopite which is a hydrated, sodium phlogopite mica whereby the cesium ions are selectively taken up by the modified phlogopite, and separating the phlogopite containing the cesium ions from the feed solution, thereby recovering the cesium ions. contacting the liquid waste with a modified phlogopite which is a hydrated sodium phlogopite mica having a c-axis spacing of about 12.23.ANG. whereby the cesium ions are selectively taken up by the modified phlogopite, and reducing the c-axis spacing an amount sufficient to lock-in the cesium ions, thereby immobilizing the cesium ions for long-term storage. 2. The process of claim 1 wherein the feed solution also contains sodium and calcium ions. 3. The process of claim 2 wherein the modified phlogopite mica has a c-axis spacing of about 12.23.ANG.. 4. A method of recovering and fixing radioactive cesium ions contained in a liquid waste feed solution for long-term storage comprising: 5. The method of claim 4 wherein sufficient cesium ions are taken up by the modified phlogopite to reduce the c-axis spacing. 6. The method of claim 4 wherein the cesium-containing phlogopite is heated to about 150.degree. C. for a period of time sufficient to reduce the c-axis spacing. |
claims | 1. A containment cooling apparatus, comprising:a cooling water tank disposed above a containment and containing cooling water therein;a spray header connected to the cooling water tank via a first communicating pipe, wherein the spray header is disposed on an outside of the containment for spraying the cooling water to an outer wall of the containment;a bell shaped shield covering the containment, wherein the cooling water tank is disposed on a top portion of the shield;a space formed between an inner wall of the shield and the outer wall of the containment, wherein the spray header is disposed in the space;an exhaust hole disposed on the top portion of the shield;a water separator disposed in the exhaust hole and/or the space; anda gas tank disposed in the containment, wherein the gas tank is connected to an upper portion of the cooling water tank via a second communicating pipe, and the cooling water tank is a closed container. 2. The containment cooling apparatus according to claim 1, wherein the cooling water tank is annular, and an axis of the cooling water tank is collinear with an axis of the shield. 3. The containment cooling apparatus according to claim 1, wherein a cooling water outlet is disposed on a bottom portion of the shield. 4. The containment cooling apparatus according to claim 1, wherein the spray header is axisymmetrically disposed above the containment. 5. The containment cooling apparatus according to claim 1, further comprising a rupture disk disposed in the gas tank and/or the second communicating pipe, wherein the rupture disk is ruptured during an increase of pressure in the gas tank, and the rupture disk in an intact state is capable of isolating a space on both sides thereof. |
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055740774 | abstract | A microwave-absorbing material composed of blends of polar icosahedral molecular units with a variety of host matrices, or with polymers with the units covalently bonded in a pendant manner to the polymer chain. Both blends and polymers must impart a high degree of orientational mobility to the units so that they can absorb microwave radiation. These materials employ orientationally mobile, polar icosahedral molecular units as the source of dielectric loss at microwave frequencies. Examples of these units are the polar carboranes (ortho- and meta-carborane), polar carboranes with electronegative and/or electropositive substitutes, and polar "buckminsterfullerenes." |
039789578 | claims | 1. For a charging device of a core reactor apparatus for guiding a gripper body internally of a horizontal hollow mast consisting of sections placed against each other, each section consisting of a plate forming a mantle which is subdivided into two halves and only two horizontal spars inserted between these halves, each spar having two legs which converge toward the inside of the mast and guide means supported thereby inside the mast, the halves of the plate mantle being welded to the spars toward the inside of the mast, the gripper body having only two diagonally oppositely located pairs of horizontal guide surfaces, the guide surfaces of each pair converging in a direction toward one of the spars and engaging against the guide means supported thereby. 2. A device according to Claim 1, which includes bores in said legs, bolts in said bores having eccentric pivot means toward the inside of the mast, rollers which form said guide means being journalled on said pivot means, and means for clamping said bolts in said plates in adjusted rotary positions. 3. A device according to Claim 2, in which said rollers are journalled by means of anti-friction bearing means having inner race means on each pivot means, said means for clamping said bolts in adusted rotary positions comprising nuts on each bolt clamping the respective inner race means while simultaneously clamping the respective bolt against rotation in the said plate. |
summary | ||
039742695 | claims | 1. A radioimmunoassay method for determining the presence of Neisserria gonorrhoeae antibodies in human serum which comprises the steps of: A. adding anti-human IgG to the serum to be tested in a buffered aqueous medium, B. thereafter adding a heat labile antigen which has been produced by a growth culture of Neisseria gonorrhoeae, isolated therefrom, and labelled with a detectable radioactive element, C. incubating resulting mixture at from about 4.degree.C to 45.degree.C for from about 24 to 2 hours at a pH of from about 6.5 to 8.5 to form an antigen-antibody conjugate when said antibodies are present, and D. determining the level of radioactivity as a measure of the presence of said antigen-antibody conjugate. A. adding anti-human IgG to the serum to be tested in a buffered aqueous medium, B. thereafter adding a heat labile antigen which has been produced by a growth culture of Neisseria gonorrhoeae, isolated therefrom, and labelled with a detectable radioactive element, C. incubating resulting mixture at from about 4.degree.C to 45.degree.C for from about 24 to 2 hours at a pH of from about 6.5 to 8.5 to form an antigen-antibody conjugate when said antibodies are present, D. filter resulting mixture through a filter which has been previously washed with a reagent selected from the group consisting of bovine serum albumin, human serum immunoglobulin, ovalbumin, and hemoglobin to separate said antigen-antibody conjugate, and E. determining the level of radioactivity in resulting precipitate as a measure of the presence of said antigen-antibody conjugate. A. adding anti-human IgG to the serum to be tested in a buffered aqueous medium containing a reagent selected from the group consisting of bovine serum albumin, human serum immunoglobulin, ovalbumin, and hemoglobin, B. thereafter adding a heat labile antigen which has been produced by a growth culture of Neisseria gonorrhoeae, isolated therefrom, and labelled with a detectable radioactive element together with bovine serum albumin, C. incubating resulting mixture at from about 4.degree.C to 45.degree.C for from about 24 to 2 hours at a pH of from about 6.5 to 8.5 to form an antigen-antibody conjugate when said antibodies are present, D. diluting resulting mixture with an aqueous buffer and centrifuging, and E. determining the level of radioactivity in resulting precipitate as a measure of the presence of said antigen-antibody conjugate. 2. A method as in claim 1 wherein the radioactive element is .sup.125 I. 3. A radioimmunoassay method for determining the presence of Niesseria gonorrhoeae antibodies in human serum which comprises the steps of: 4. A method as in claim 3 wherein the radioactive element is .sup.125 I. 5. A method as in claim 3 wherein the reagent utilized in Step D is bovine serum albumin. 6. A method as in claim 3 wherein the radioactive element is .sup.125 I and the reagent utilized in Step D is bovine serum albumin. 7. A radioimmunoassay method for determining the presence of Neisseria gonorrhoeae antibodies in human serum which comprises the steps of: 8. A method as in claim 7 wherein the reagent is bovine serum albumin. 9. A method as in claim 7 wherein the radioactive element is .sup.125 I. 10. A method as in claim 7 wherein the buffer is phosphate buffered saline. 11. A method as in claim 7 wherein the reagent is bovine serum albumin, the radioactive element is .sup.125 I and the buffer is phosphate buffered saline. |
053274717 | summary | Cross-Reference to Related Application: This application is a continuation of International Application PCT/DE91/00168, filed Feb. 27, 1991. The invention relates to a nuclear reactor fuel assembly, having a bundle of fuel rods being aligned parallel to one another and containing nuclear fuel; a fuel assembly head being disposed above the bundle and having a top with a handle thereon; a bottom plate being disposed under the bundle and being permeable to liquid coolant; a coolant tube being disposed in the bundle, being parallel to the fuel rods and being open on both of its ends, the coolant tube having a lower end with a lower end piece and being retained by the lower end piece on the bottom plate; and the coolant tube having an upper end piece protruding from below into a recess of the fuel assembly head. Such a fuel assembly is known from Published European Application No. 0 307 705 A1. The end piece at the upper end of the coolant tube in that known fuel assembly is a platelike end screen with flow openings for the liquid coolant. A bolt which is parallel to the coolant tube is located on the end screen and protrudes from underneath into a recess in the fuel assembly head, where it is loosely guided without any further connection. The end piece at the lower end of the coolant tube is also an end screen. Disposed on the outside of that end screen is a threaded bolt which is also parallel to the longitudinal direction of the coolant tube, but it also reaches through a leadthrough in a gridlike perforated bottom plate in the fuel assembly foot, where it is screwed firmly to the fuel assembly foot with a nut. That nuclear reactor fuel assembly has gridlike spacers in a plurality of cross-sectional planes between its head and its foot. The spacers have meshes or holes, through each of which one fuel rod containing nuclear fuel extends. Some of the fuel rods are so-called bearing rods having threaded bolts on both ends, with which they are firmly screwed to the fuel assembly head with a nut and to the fuel assembly foot directly. The other fuel rods reach loosely through leadthroughs in the fuel assembly head and foot with bolts located on both ends of the rods. All of the fuel rods are fixed with a compressionally biased helical spring, and each helical spring is seated on the threaded bolt or bolt on the inside of the fuel assembly head and is supported on that end in the applicable fuel rod. The invention also relates to a fuel assembly for a boiling water reactor with fuel rods of different lengths. One such fuel assembly is known from U.S. Pat. No. 4,675,154 and includes a fuel assembly case that laterally surrounds the fuel assembly and is open at the top and bottom. A coolant tube extends axially inside the case and has at least one opening on each of its top and bottom end for the flow of liquid coolant. A cover plate provided with coolant outlets on the upper end of the case and a bottom plate provided with coolant inlets on the lower end of the case cover the two open ends of the case. Spacers that are at right angles to the water tube and include support ribs, extend over the cross section of the fuel assembly case in predetermined axial positions. T spacers are held on the coolant tube through suitable means (such as stops). A number of fuel rods is also provided, each of which is laterally supported on the support ribs of a plurality of spacers and need not be of uniform length. In the case of such boiling water fuel assemblies, it is customary to construct some fuel rods as retaining rods and to screw them to the bottom and cover plates with corresponding upper and lower closure caps. If fuel rods of different lengths are provided, then the shorter fuel rods are screwed only to the bottom plate by their lower end, while their upper end is held in position by corresponding spacers. However, at least some of the retaining rods are provided on their upper closure caps with retainers for a handle, which bear the entire weight of the fuel assembly. Additional helical springs on such upper closure caps can press the cover plate against the screw connections on the upper end pieces of the retaining rods. Such long fuel assemblies that are constructed as retaining rods are part of the load-bearing or supporting fuel assembly skeleton and are essentially under tension. Between the handle, the cover plate and the bottom plate, they produce a force-locking, torsion-proof connection, they carry the weight of the fuel assembly and also absorb the various forces that arise upon expansion of material from the radiation and heating that prevail during operation. A force-locking connection is one which connects two elements together by force external to the elements, as opposed to a form-locking connection which is provided by the shapes of the elements themselves. Since the plates are pierced by bores, some of which form the water inlets and water outlets and some of which receive the closure caps of the fuel rods, they must be dimensioned with a correspondingly great thickness, and therefore there is only limited space available for the flow cross section of the inlets and outlets. As a result, there is a considerable pressure difference, especially at the upper cover plate, since some of the water that is pumped axially through the fuel assembly during operation evaporates, so that the mixture of water and steam would require larger outlet cross sections in order to maintain an unimpeded flow. If the fuel assembly case is removed axially upward from the load-bearing or supporting skeleton, then the fuel rods become accessible and can be inspected from the side. However, in the event of more extensive inspection, maintenance and possible replacement of fuel rods, the cover plate must be unscrewed and the load-bearing or supporting skeleton must accordingly be dismantled. Since the screw connections generally seize after relatively long service, considerable forces are needed to release the fuel rods. Such forces can be brought to bear only at the upper closure caps of the retaining rods and must be transmitted to the screw connections of the lower closure caps by torsion on the retaining rods. Great mechanical stability and accordingly a certain minimum thickness are therefore necessary for the rods. However, if a change is to be made from a fuel assembly having rods which are disposed in eight rows and eight columns (which is known as an 8.times.8 assembly) to an 11.times.11 fuel assembly with the same external dimensions, then thinner rods must be used, having a mechanical stability that cannot meet the requirements made for retaining rods, especially with respect to the forces necessary to release the screw connection. For economical reasons, it is generally advantageous to use a larger number of thinner rods, instead of fewer but thicker fuel rods. It is accordingly an object of the invention to provide a nuclear reactor fuel assembly with a load-bearing or supporting coolant tube, which overcomes the hereinafore-mentioned disadvantages of the heretofore-known devices of this general type and which improves such nuclear reactor fuel assemblies, in particular by avoiding load-bearing or supporting rods with threaded bolts located on both ends that are firmly screwed to both the head and foot of the fuel assembly. In the novel structure for the load-bearing or supporting parts of the fuel assembly skeleton, loading and forces are better distributed among the load-bearing or supporting parts, which creates new possibilities for retaining the fuel rods (and in particular shorter and/or thinner fuel rods), for simplifying assembly and disassembly of the fuel assembly, and/or for creating more favorable water inlets and water outlets with optimized pressure drops. The invention takes the fact that the coolant tube is a relatively stable, low-wear element as its point of departure. It is suitable for transmitting relatively strong forces and practically need not be removed for inspecting or repairing the fuel assembly. With an adequately stable connection to the bottom plate, the entire weight of the fuel assembly can therefore be borne, which makes retaining rods superfluous and makes load-bearing or supporting screw connections to fuel rods unnecessary for either the bottom plate or the top plate. A rigid connection for force transmission between the coolant tube and the bottom plate can therefore be non-detachably constructed, and a seized screw connection of the tube need not be loosened for inspections. The cover plate can be practically relieved of load-bearing or supporting forces and then serves only to cover the water channel, as an upper stop for unscrewed fuel assemblies, which are then retained only by the spacers in the coolant flow, and to maintain the desired flow. With the cover plate removed, the fuel rods can be inserted or replaced without further screwing operations. With the foregoing and other objects in view there is provided, in accordance with the invention, a nuclear reactor fuel assembly, comprising at least one coolant tube with an upper end having an opening formed therein and an upper end piece and a lower end having an opening formed therein and a lower end piece; a bottom plate being supported by said coolant tube and joined to said lower end piece in a dimensionally rigid manner, said bottom plate having inlet openings formed therein for liquid coolant; a cover plate retaining said upper end piece and having outlet openings formed therein for a liquid/steam mixture of coolant; gridlike spacers defining meshes therein, means for retaining and attaching said spacers on said coolant tube between said plates; and fuel rods each being guided through a respective one of said meshes, being filled with nuclear fuel and being joined to at most one of said plates. In accordance with another feature of the invention, there is provided a stop body mounted on said upper end piece, said cover plate being detachably retained on said stop body, and a lower closure cap disposed on said bottom plate, said fuel rods including longer and shorter fuel rods, and only said shorter fuel rods being retained by said lower closure cap on said bottom plate. As compared with known fuel assembly structures, the upper end piece of the coolant tube accordingly gains special significance, because it is no longer freely or only loosely guided in a recess relative to the fuel assembly head, but instead is retained on the fuel assembly head. The entire weight of the fuel assembly is then supported by the coolant tube when the handle on the fuel assembly head is raised. If additional forces arise at the fuel assembly head during manipulation (such as additional forces of gravity due to the application of the hoisting tool), they can be absorbed by springs between the head and the foot part, without uncontrollably loading the coolant tube. With the objects of the invention in view, there is also provided a nuclear reactor fuel assembly, comprising a bundle or cluster of mutually parallel aligned fuel rods containing nuclear fuel; a fuel assembly head being disposed above the bundle and having a top with a handle; a bottom plate being disposed under the bundle and being permeable to liquid coolant; and a coolant tube being disposed in the bundle, being parallel to the fuel rods and having upper and lower open ends; the coolant tube having a lower end piece on the lower end being supportingly retained on the bottom plate for retaining the coolant tube on the bottom plate; and the coolant tube having an upper end piece on the upper end being supportingly retained on the fuel assembly head and protruding from below into a recess formed in the fuel assembly head. In accordance with another feature of the invention, the bottom plate and the fuel assembly head are supports for the coolant tube, one of the end pieces is longitudinally displaceably supported on one of the supports, and including a stop body on the fuel assembly head defining a maximal spacing between the supports, and at least one spring disposed between the supports for pressing one of the supports against the stop body. In accordance with a further feature of the invention, there is provided a stop shoulder formed onto the one displaceably supported end piece, the shoulder defining a minimal spacing between the fuel assembly head and the bottom plate. In accordance with an added feature of the invention, there is provided a spring is disposed on one of the fuel rods. In accordance with an additional feature of the invention, the spring is disposed on one of the pieces, preferably on the upper end piece. In accordance with yet another feature of the invention, the fuel assembly head has a part being formed onto the handle, being secured against rotation, and being screwed on from above, and the upper end piece is passed through the part. In accordance with yet a further feature of the invention, the upper end piece carries a socket pin, the fuel assembly head carries a cover plate having coolant outlet openings formed therein, and the socket pin extends through the cover plate and carries a detachably mounted stop body for the cover plate. In accordance with yet an added feature of the invention, the cover plate is mounted on the upper end piece, and including a supporting mechanical connection securing the bottom plate to the lower end piece, a spring under pressure pressing the cover plate against the stop body, and the fuel rods having upper ends with upper closure caps, the upper closure caps being loosely guided on or not touching the cover plate in an appropriate receiving position. As a special advantage, at least all of the fuel rods that extend practically up to the cover plate in the head of the fuel assembly carry lower closure caps on their lower ends, with which these fuel assemblies then stand, unscrewed, on the bottom plate. Accordingly, the closure caps need merely be loosely introduced into suitable plug-type connections on the bottom plate or they need merely be placed on the surface of the bottom plate pointing toward the cover plate, without using special retainers. A lower end piece of the coolant tube is also provided, which is rigidly joined to the bottom plate so that all of the forces exerted upon the bottom plate are practically transmitted to the coolant tube. A corresponding upper end piece of the coolant tube, which also supports the cover plate, is detachably joined to the cover plate and/or to a handle secured to the cover plate. Therefore, with the objects of the invention in view, there is additionally provided a fuel assembly of a boiling water reactor, comprising a fuel assembly case laterally surrounding the fuel assembly and having an interior and open upper and lower ends at the top and bottom; a cover plate covering the upper end of the case and having coolant outlets formed therein; a bottom plate covering the lower end of the case and having coolant inlets formed therein; a coolant tube extending in axial direction in the interior of the case and having upper and lower ends, each of the upper and lower ends having at least one opening formed therein for the passage of water, the coolant tube having a lower end piece joined to the bottom plate with a rigid connection transmitting substantially all forces exerted upon the bottom plate to the lower end piece and to the coolant tube, and the coolant tube having an upper end piece, and a releasable connection supporting the upper end piece at the cover plate; a plurality of spacers standing practically or substantially perpendicular to and retained on the coolant tube at predetermined axial positions, the spacers containing support ribs; a plurality of fuel rods being parallel to the case, being respectively supported on the support ribs of a plurality of the spacers and having lower ends, some of the fuel rods extending practically or substantially as far as the cover plate; and lower closure caps standing unscrewed on the bottom plate and being disposed on the lower ends of at least all of the fuel rods extending substantially as far as the cover plate. In accordance with another feature of the invention, at least two of the spacers are penetrated by all of the fuel rods, the fuel rods include first and second groups of fuel rods, each of the fuel rods at least in the first group extend from the bottom plate as far as the cover plate and carry one of the lower closure caps on the lower end, the bottom plate has an upper surface facing toward the cover plate, the lower closure caps stand on the upper surface of the bottom plate, the fuel rods of the first group have upper ends, and including upper closure caps on the upper ends being loosely disposed on the cover plate at given receiving positions. In accordance with a further feature of the invention, each of the fuel rods in the second group is shorter than a fuel rod in the first group and carries one of the lower closure caps on the lower end being retained on the bottom plate, the fuel rods of the second group have upper ends, and including upper closure caps on the upper ends being disposed in the vicinity of one of the spacers and being spaced apart from the cover plate. In accordance with an added feature of the invention, there is provided a plug connection joining the lower closure caps of the fuel rods of the second group to the bottom plate. In accordance with an additional feature of the invention, the coolant outlets in the cover plate are outlet openings disposed between the given receiving positions of the fuel rods of the first group and enlarged outlet openings disposed in a projection of the shorter fuel rods. In accordance with again another feature of the invention, the fuel rods of the second group are each disposed along diagonals of the cross section of the case. In accordance with again a further feature of the invention, the case has walls, some of the fuel rods are adjacent the walls, and each of the fuel rods adjacent the walls belongs to the first group. In accordance with again an added feature of the invention, the support ribs of the spacers define first through eleventh rows and first through eleventh columns of meshes between the support ribs, counting from the wall inward, and the fuel rods of the second group are each disposed in the third row or the third column. In accordance with again an additional feature of the invention, the upper end piece of the coolant tube passes through the cover plate from below and is screwed to the cover plate and/or to the handle from above. In accordance with still another feature of the invention, the handle is undetachably secured to the cover plate. In accordance with still a further feature of the invention, there is provided a part formed onto the bottom plate into which the lower end piece of the coolant tube is introduced, and a securing bolt securing the lower end piece to the part, to protect against relative rotation of the bottom plate and the coolant tube. In accordance with still an added feature of the invention, the cover plate is resiliently supported against the bottom plate. In accordance with still an additional feature of the invention, the coolant inlets in the bottom plate are flow openings having cross sections creating a uniform flow through the fuel assembly with a pressure loss being negligible as compared to a pressure loss at the coolant outlets in the cover plate. In accordance with a concomitant feature of the invention, threadless closure caps are disposed on the upper and lower ends of all of the fuel rods. Other features which are considered as characteristic for the invention are set forth in the appended claims. Although the invention is illustrated and described herein as embodied in a nuclear reactor fuel assembly with a load-bearing or supporting coolant tube, 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. |
description | This application is a National Stage Application, filed under 35 U.S.C. §371, of International Application No. PCT/SE2007/050563, filed Aug. 21, 2007, which claims priority to Swedish Application No. 0601872-5, filed Sep. 12, 2006; the contents of both of which are hereby incorporated by reference in its entirety. The present invention refers generally to fuel assemblies for nuclear plants and handling of fuel assemblies in connection with shutdown and maintenance of nuclear plants. Especially, the present invention refers to a device for handling fuel assemblies, and a method for handling fuel assemblies. The invention especially refers to light water reactors and in particular boiling water reactors, BWR. In such reactors, water, the purpose of which is to function as coolant and moderator in the nuclear reactor in the nuclear plant, circulates. It is important to keep this water clean from debris. If debris particles are allowed to accompany the water in the core of the reactor, these may cause defects to the cladding of the fuel rods, which can lead to such defects that nuclear fuel, i.e. uranium, leaks out into the water. The debris particles can of course also cause defects on other components in the nuclear plant, for instance pumps. With debris particles is to be understood in this application different particles such as for instance metal chips, formed in connection with various repairs of components of the plant, metal wires or other foreign particles which have entered the plant from outside, objects such as nuts, screws, smaller tools etc. Particularly difficult debris particles are such with an elongated shape, i.e. thin wires or chips which may have a length of down to about 7-8 mm. Such debris particles tend to get attached higher up in the fuel assembly, for instance in spacers. The particles vibrate in the coolant stream and may wear the cladding of the fuel rod so that a hole arises. In order to solve this problem, it is known to provide some kind of debris filter in the lower part of the fuel assemblies, which comprise a number of fuel rods and which form the core of the reactor. One example of such a debris filter is disclosed in WO02/058075. The water, which circulates through the reactor during the operation of the plant, passes through this lower part of the fuel assemblies upwardly from below. Possible debris particles may thus be caught by one such debris filter. The debris particles are prevented at least to a certain extent from being released from the debris filter and conveyed back to the water due to the flowing of the water through the debris filter. Debris filters are normally also designed to let through smaller debris particles and in such a way reduce the risk of being stopped up. The flow through the lower part of the fuel assembly is during normal operation about 2 m/s but may at reduced power be decreased to about 1 m/s. In connection with shutdown of the reactor, the flow of water is maintained at least initially due to the rest heat which is present in the fuel assembly. It is also possible, during an initial phase of the revision, to operate the main circulation pumps of the plant with a relatively low power sufficient for maintaining a flow of water through the fuel assemblies. Each fuel assembly is standing in a seat and water is guided during normal operation into the fuel assembly via a sealing. When the fuel assembly is to be removed and lifted out from the reactor, the flow of water will, however, be reduced and it can even be reversed depending on the speed with which the fuel assembly is lifted upwardly. This lifting speed can during a normal shutdown amount to about 1 m/s. This means that the force which is created by the flow and retains the debris particles in or immediately beneath the debris filter, is reduced or ceases and that the debris particles may fall out from the debris filter. Consequently, the tendency for the particles to be released and fall back to the reactor vessel increases. This means that there is no longer any mechanism for removing the debris particles from the reactor and the primary system, but more and more debris particles will be accumulated in the reactor vessel. U.S. Pat. No. 5,383,226 discloses a device for handling a fuel assembly, which comprises a number of fuel rods extending between a lower part and an upper part of the fuel assembly and a casing surrounding the fuel rods. The device comprises a lifting device which is arranged during a lifting operation to engage a fuel assembly located in a reactor vessel and to lift the fuel assembly upwardly and out from the reactor vessel. A conduit member is connected to the upper part of the fuel assembly and a pump is arranged to create a flow of liquid through the fuel assembly and the conduit member to a detecting equipment. The detecting equipment is adapted to detect fission products in the water, which can indicate that one or several fuel rods are defect. The object of the present invention is to reduce the accumulation of debris particles in the reactor vessel. This object is achieved by the device initially defined, which is characterized in that the pump is arranged to provide a flow of such a size that possible debris particles, which are contained in and immediately beneath the debris filter at least are retained in and/or immediately beneath the debris filter during the lifting operation. By means of such a device it may thus be ensured that the quantity of debris particles which are released during the lifting operation is very small. The quantity of accumulated debris particles in the reactor vessel may thus be reduced. The applicant has during experiments concluded that the floating speed of the debris particles, i.e. the speed with which they sink in the water, amounts to approximately 0.1-0.3 m/s. This means that the created flow of water through the fuel assembly at the inlet of the fuel assembly ought to amount to at least 0.5 m/s in order to be able to ensure that the debris particles do not fall down from the debris filter. Consideration is then to be given to the lifting speed used, see above. The inlet of the fuel assembly is its thinnest section and consequently determines if the debris particles may fall out. According to an embodiment of the invention, the device is arranged to control the pump in such a way that the flow is maintained during the whole lifting operation. Advantageously, the device may also be arranged to control the pump in such a way that the flow is initiated as soon as the lifting operation has been initiated. According to a further embodiment of the invention, the lifting device comprises an elongated grip element, which is arranged to be submerged in the reactor vessel and to engage the fuel assembly, wherein the pump is provided on the grip element. Such a grip element is available at the most of the present nuclear plants and is well adapted also to carry the defined pump. Advantageously, the pump may be provided adjacent to the upper part of the fuel assembly. According to a further embodiment of the invention, the conduit member is arranged to convey the flow to a position outside the fuel assembly. The flow of water may thus be conveyed back to the water in the reactor vessel, since the debris particles are retained in the debris filter. According to a further embodiment of the invention, the device comprises a collecting member, which is connected to the conduit member and arranged to collect the debris particles accompanying the flow of water through the fuel assembly. By means of such a collecting member, which may be provided upstream or downstream the pump, the debris particles, which for any reason are present in the flow channel of the fuel assembly when the lifting operation is initiated or pass through the debris filter, are caught and collected. Such a collection is especially advantageous in the case that the flow of water is conveyed back to the reactor vessel. The collecting member may advantageously comprise a filter and/or a container. According to a further embodiment of the invention, the conduit member comprises a cover, which is arranged to be provided on the upper part of the fuel assembly and which defines a passage for said flow. The cover suitably has such dimensions that it sealingly encloses the fuel assembly so that the whole flow through the fuel assembly is conveyed into and through the conduit member. According to a further embodiment of the invention, the debris filter is provided in such a way that all water flowing into the flow channel flows through the debris filter. According to a further embodiment of the invention, the lifting device is arranged to transport the lifted fuel assembly to a water pool at a distance from the reactor vessel, wherein the device is arranged to control the pump in such a manner that the flow is maintained during the whole of this transport. In such a way, debris particles are prevented from falling out from the fuel assembly also during the transport and from contaminating the transport path. Advantageously, the device may be arranged to control the pump in such a manner that the flow of water through the fuel assembly is reversed in a position where the debris particles are depositable, for instance when the transport is finished. Consequently, at least a part of the debris particles, which are retained in and/or immediately beneath the debris filter, may be permitted to fall down and at any suitable occasion be collected and removed. The object is also achieved by the method initially defined, which is characterized in that the flow has such a size that possible debris particles which are contained in and/or immediately beneath the debris filter at least are retained in and/or immediately beneath the debris filter during the lifting operation. Advantageous further developments of the method are defined in the dependent claims. FIG. 1 discloses schematically a nuclear plant with a boiling water reactor, BWR. The plant comprises a reactor vessel 1 having a core 2 of fuel assemblies 3. A fuel assembly 3 is disclosed schematically in FIG. 2. Each fuel assembly 3 in the core 2 comprises a number of fuel rods 4, which extend between a lower part and an upper part of the fuel assembly 3. A casing 5 surrounds the fuel rods 4 and forms a flow channel 6 in which the fuel rods 4 are located. A debris filter 7 is provided in the lower part of the fuel assembly 3 at an inlet 8 of the fuel assembly 3. The debris filter 7 is provided in such a manner that all water flowing through the inlet 8 into the flow channel 6 flows through the debris filter 7. A handle 9 is provided at the upper part of the fuel assembly 3. The plant disclosed in FIG. 1 is shut down and the cover of the reactor vessel 1 is removed. The reactor vessel 1 is thus open at the top. Moreover, the plant comprises a water pool 11 which is provided at a distance from the reactor vessel 1. A transport path 12 extends between the reactor vessel 1 and the water pool 11. The reactor vessel 1, the transport path 12 and the water pool 11 form in the shut down state a common space which is filled with water up to a water surface. Furthermore, the plant comprises a device for handling a fuel assembly. This device comprises a lifting device 15 which is provided on a horizontal rail device 16 extending over the reactor vessel 1, the transport path 12 and the water pool 11. The lifting device 15 comprises a carriage 17 which is displaceable on and along the rail device 16. On the carriage 17 a grip element 18 is suspended, which is designed as a telescopic pipe and which thus can be extended and shortened. At the lower end of the grip element 18 there is a gripping tool 19 which is arranged to engage a handle 9 of the fuel assembly 3. By means of the lifting device 15, a fuel assembly may be engaged in the handle 9 during a lifting operation and lifted up from the core 2 and the reactor vessel 1. When the fuel assembly 3 has reached an upper position, which still is situated beneath the water surface, the lifting device 15 may be displaced along the rail device 16, wherein the fuel assembly 3 is transported under water from the reactor vessel 1 to the water pool 11 via the transport path 12. In the water pool 11, the fuel assembly 3 may be submerged and put into a suitable device (not disclosed) at the bottom of the water pool 11. The device for handling a fuel assembly 3 also comprises a conduit member 31, which is connectable to the upper part of the fuel assembly 3 and a pump 32 which is attached to the grip element 18 in the proximity of the griping tool 19 and the upper part of the fuel assembly 3. The pump 32 is provided on the conduit member 31 and arranged to permit pumping of a flow of water through the conduit member 31 and the fuel assembly 3. In the embodiment disclosed, the conduit member 31 is arranged to convey the flow of water to a position outside the fuel assembly and close to the fuel assembly 3. This means that when the fuel assembly 3 is located in or immediately above the reactor vessel, the pumped water will be conveyed back to the reactor vessel 1. In the same way, the pumped water will when the fuel assembly 3 is located along the transport path 12 or in the water pool 11, be conveyed back to the transport path 12 and the water pool 11, respectively. Furthermore, the device may comprise a collecting member 33, which is connected to the conduit member 31 and arranged to collect the debris particles accompanying the flow of water through the fuel assembly 3. The collecting member 33 is in the embodiment disclosed provided downstream the pump 32, but it is also possible to provide the collecting member 33 upstream the pump 32. In the embodiment disclosed, the collecting member 33 comprises a filter 34 catching possible debris particles in the flow of water. Furthermore, the collecting member 33 may comprise a container 35 which can be provided upstream and/or downstream the filter 34 and arranged to collect the debris particles caught by the filter 34. The collecting member 33 may be designed in many different ways, for instance as a cyclone where debris particles are caught by means of the centrifugal force. Furthermore, the conduit member 31 comprises a cover 36 which is arranged to be attached to the upper part of the fuel assembly 3 and which defines a passage for the flow of water into the conduit member 31. The cover 36 is adapted in such a way that it sealingly encloses the upper part of the fuel assembly 3 and prevents inflow of water between the cover 36 and the upper part of the fuel assembly 3. Furthermore, the device comprises a control unit 40, which is connected to the pump 32 and arranged to control the pump 32. By means of the control unit 40, the pump 32 can be controlled so that it creates a flow of water during the above-mentioned lifting operation. The flow upwardly through the debris filter 7 and the fuel assembly 3 ought to have such a size during the lifting operation that possible debris particles which are contained in and/or immediately beneath the debris filter 7 at least are retained in and/or immediately beneath the debris filter 7. This means that the upwardly directed flow at the inlet of the fuel assembly 3 beneath the debris filter 7 ought to have a flow speed which is at least 0.4 m/s, preferably at least 0.5 m/s. Advantageously, the pump 32 is controlled in such a way that the flow is maintained during the whole lifting operation and in such a way that the flow is initiated as soon as the lifting operation has been initiated. Furthermore, the pump 32 may be controlled in such a way that the flow is maintained during the whole transport from the reactor vessel 1 via the transport path 12 to a position where the debris particles are depositable, such as in the disclosed water pool 11. When the fuel assembly 3 has been transported to the water pool 11, the pump 32 may be controlled so that the flow of water through the fuel assembly 3 is reversed. In such a way, at least a part of the debris particles which are contained in the debris filter 7 may be deposited by being flushed out from the debris filter 7 and falling down into the water pool 11. It is to be noted that it is possible to provide the pump 32 at a distance from the fuel assembly 3, for instance higher up at the grip element 18 or at a wail of the common space formed by the reactor vessel 1, the transport path 12 and the water pool 11. However, it is advantageous if the pump 32 is submergible and located beneath the water surface since possible debris particles transported through the pump 32 may be radioactively contaminated. It is also to be noted that it is possible to convey the water from the pump 32 to any other place than back to the reactor vessel, for instance to any collecting vessel outside the reactor vessel 1. The water in such a collecting vessel may be cleaned and/or filtered and thereafter recycled to the reactor vessel 1 or removed from the plant. The invention is not limited to the embodiments described and disclosed but may be varied and modified within the scope of the following claims. |
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description | The present invention relates to a fuel assembly and a reactor loaded with the fuel assembly, and more particularly to a fuel assembly suitable for application to a boiling water reactor, and a reactor loaded with the fuel assembly. A plurality of fuel assemblies is loaded in a reactor of a boiling water reactor. The fuel assembly includes a plurality of fuel rods enclosing a plurality of fuel pellets including a nuclear fuel material (for example, uranium oxide), an upper tie plate that supports upper end portions of the fuel rods, a lower tie plate that supports lower end portions of the fuel rods, a plurality of fuel spacers that holds spacing between the fuel rods, and a square tubular channel box. The channel box has an upper end portion attached to the upper tie plate and extends toward the lower tie plate, and surrounds the plurality of fuel rods bundled by the plurality of fuel spacers. A plurality of control rods is inserted in the reactor to control reactor power. Further, some of the fuel rods in the fuel assembly contain burnable poison (for example, gadolinium: Gd) in the fuel pellets. The control rod and the burnable poison absorb extra neutrons generated by nuclear fission of the nuclear fuel material. The burnable poison turns into a substance that hardly absorbs neutrons due to absorption of neutrons. Therefore, the burnable poison contained in a new fuel assembly (a fuel assembly with the burnup of 0 GWd/t) loaded in the reactor disappears when a certain operation period of the nuclear reactor has passed since the new fuel assembly was loaded in the reactor. In the fuel assembly from which the burnable poison has disappeared, the reactivity is monotonously decreased as the nuclear fuel material is burned. Since a plurality of fuel assemblies having different operating cycle numbers staying in the reactor is loaded in the reactor, the critical state is maintained throughout the operation period of the whole reactor. A fuel assembly fabricated using a material as a fuel such as plutonium (Pu) obtained by reprocessing a uranium fuel taken out of the nuclear reactor is called a MOX fuel. Burnable poison is loaded even in the MOX fuel. However, since the average energy of neutrons becomes high in the reactor using the MOX fuel (the neutron energy spectrum becomes hard), the neutron absorption effect of the burnable poison becomes small. Further, since plutonium is taken out by reprocessing, a concentration process to increase the plutonium enrichment is unnecessary, unlike the uranium fuel. That is, since the increase in the enrichment of the MOX fuel is possible at a relatively low cost, an increase in the burnup by the MOX fuel with high enrichment is effective for cost reduction. However, the amount of loading of plutonium is decreased if the burnable poison is added to and loaded in the uranium fuel rod, which hinders the increase in the enrichment. In the case of loading fissile plutonium to flatten the power distribution in the fuel assembly, the enrichment of the outermost periphery of the fuel assembly is decreased, similarly to uranium. That is, the disadvantage of the decrease in the enrichment can be minimized by loading the fuel rod to which the burnable poison is added on the outermost periphery. Here, the concentration of the burnable poison needs to be increased when considering a long-term cycle operation, in addition to the increase in the burnup. The burnable poison can be maintained without being burned out for a long period of time by loading the fuel rod at a position with hard neutron spectrum except the outermost periphery portion. However, it is favorable to arrange the fuel rod on the outermost periphery for the above reason. However, since the neutron moderation effect is large on the outermost periphery and the neutron energy spectrum becomes soft, the burnable poison is burned out at an early stage. The addition amount of the burnable poison has an upper limit due to problems such as fuel manufacturability and a decrease in thermal conductivity at loading. Therefore, for example, the technology described in PTL 1 has been proposed. PTL 1 discloses a configuration in which fuel rods containing natural uranium and gadolinium (Gd) and not containing plutonium (hereinafter referred to as gadolinium fuel rods) are arranged at four corners on an outermost periphery in a horizontal section of a fuel assembly, and two gadolinium fuel rods are arranged vertically and horizontally adjacent to the gadolinium fuel rod at one corner. A configuration to arrange a plutonium high enriched fuel rod, a plutonium low enriched fuel rod, and a plutonium lowest enriched fuel rod is described in addition to the above configuration. Since gadolinium as burnable poison has large absorption of neutrons, a neutron flux around gadolinium is small. When the gadolinium fuel rod is arranged there, neutrons reacting with gadolinium apparently become small and the burnable poison can be maintained for a long period of time without burning out the burnable poison. PTL 2 discloses a configuration in which gadolinium-containing uranium fuel rods (hereinafter referred to as gadolinium fuel rods) are arranged at four corner positions on an outermost periphery in a horizontal section of a fuel assembly, and two gadolinium fuel rods are arranged at positions vertically and horizontally adjacent to two corner positions except a corner position on a control rod side and a corner position diagonally located across the corner position on a control rod side. PTL 1: JP 05-008398 B PTL 2: JP 2000-180574 A PTL 1 discloses the configuration in which the two gadolinium fuel rods are arranged vertically and horizontally adjacent to the gadolinium fuel rod at one corner in the horizontal section of the fuel assembly, and the gadolinium fuel rods arranged at the other three corners do not have adjacent gadolinium fuel rods. Here, in the two gadolinium fuel rods arranged vertically and horizontally adjacent to the gadolinium fuel rod at one corner, an effect to decrease the infinite multiplication factor is reduced due to suppression of the neutron absorption effect (neutron shielding) at the initial stage of burnup. On the other hand, the gadolinium fuel rods arranged at the other three corners decrease the infinite multiplication factor at the initial stage of burnup without suppression of the neutron absorption effect. However, in the configuration of PTL 1, the number of the gadolinium fuel rods arranged at the three corners without being adjacent to one another is three, and the number of the gadolinium fuel rods adjacent to only one gadolinium fuel rod (the gadolinium fuel rod at one corner) is two. The infinite multiplication factor is excessively decreased at the initial stage of burnup, and flattening of the excess reactivity is impaired. Further, PTL 2 discloses the configuration in which two gadolinium fuel rods are arranged vertically and horizontally adjacent to the gadolinium fuel rods arranged at the two corner positions except the corner position on a control rod side and the corner position diagonally located across the corner position on a control rod side, and one gadolinium fuel rod is arranged at the other two corner positions. Therefore, the number of the gadolinium fuel rods arranged at the corner positions without being adjacent to each other is two whereas the number of the gadolinium fuel rods adjacent to only one gadolinium fuel rod (the gadolinium fuel rods arranged at two corner positions except the corner position on a control rod side and the corner position diagonally located across the corner position on a control rod side) is four (the number of the gadolinium fuel rods adjacent to two gadolinium fuel rods is two). The excessive decrease in the infinite multiplication factor at the initial stage of burnup like PTL 1 can be prevented. However, the gadolinium fuel rods adjacent to the two gadolinium fuel rods of PTL 2 impede the flattening of the excess reactivity due to a large shielding effect, and the number is the same as the number of the gadolinium fuel rods for flattening the excess reactivity (two), and thus the flattening is impaired. Therefore, the present invention provides a fuel assembly capable of linearizing change of the infinite multiplication factor of the fuel and flattening the excess reactivity while increasing average fissile plutonium enrichment of the MOX fuel, and a reactor loaded with the fuel assembly. To solve the above problem, a fuel assembly of the present invention includes at least, a first fuel rod containing plutonium and not containing burnable poison, a second fuel rod containing uranium and burnable poison and not containing plutonium, a water rod, and a channel box having a rectangular shape in horizontal section and accommodating the first fuel rod, the second fuel rod, and the water rod in a bundle, wherein the second fuel rod is disposed on an outermost periphery and/or adjacent to the water rod, of a fuel rod array in the horizontal section, N2<N1 (N2 is a positive integer including zero) is satisfied where the number of the second fuel rods arranged on the outermost periphery is N1 and the number of the second fuel rods arranged adjacent to the water rod is N2, and W2<N2+W0<W1 (W2 is a positive integer including zero) is satisfied where the number of the second fuel rods arranged without being vertically and/or horizontally adjacent to each other in the horizontal section is W0, the number of the second fuel rods arranged vertically and/or horizontally adjacent to only one second fuel rod in the horizontal section is W1, and the number of the second fuel rods arranged vertically and/or horizontally adjacent to two second fuel rods in the horizontal section is W2, of the second fuel rods arranged on the outermost periphery. Further, a reactor of the present invention is a reactor of a nuclear reactor loaded with a plurality of fuel assemblies, the fuel assembly including at least, a first fuel rod containing plutonium and not containing burnable poison, a second fuel rod containing uranium and burnable poison and not containing plutonium, a water rod, and a channel box having a rectangular shape in horizontal section and accommodating the first fuel rod, the second fuel rod, and the water rod in a bundle, wherein the second fuel rod is disposed on an outermost periphery and/or adjacent to the water rod, of a fuel rod array in the horizontal section, N2<N1 (N2 is a positive integer including zero) is satisfied where the number of the second fuel rods arranged on the outermost periphery is N1 and the number of the second fuel rods arranged adjacent to the water rod is N2, and W2<N2+W0<W1 (W2 is a positive integer including zero) is satisfied where the number of the second fuel rods arranged without being vertically and/or horizontally adjacent to each other in the horizontal section is W0, the number of the second fuel rods arranged vertically and/or horizontally adjacent to only one second fuel rod in the horizontal section is W1, and the number of the second fuel rods arranged vertically and/or horizontally adjacent to two second fuel rods in the horizontal section is W2, of the second fuel rods arranged on the outermost periphery. According to the present invention, a fuel assembly capable of linearizing change of an infinite multiplication factor of a fuel and flattening excess reactivity while increasing average fissile plutonium enrichment of a MOX fuel, and a reactor loaded with the fuel assembly can be provided. Problems, configurations, and effects other than those described above will be clarified from description of the following embodiments. The present inventors and the like have repeated various examinations and have found a new configuration that increases plutonium enrichment and prolongs an operation cycle in a MOX fuel. The examination result and the outline of the newly found MOX fuel configuration will be described below. As in the prior technologies, to increase the plutonium enrichment of the MOX fuel of a boiling water reactor (BWR), the number of burnable poison-added fuel rods needs to be decreased as many as possible. Furthermore, to prolong the operation cycle, the concentration of the burnable poison needs to be increased. Gadolinium (Gd), which is used as burnable poison, is generally used with the concentration of 10 wt % as an upper limit. This upper limit is set due to the problem of fuel manufacturability and thermal conductivity. To decrease the number of fuel rods to which burnable poison is added, the fuel rods are arranged at positions where the neutron absorption effect is large. Therefore, the present inventors and the like have examined neutron absorption characteristics with respect to a position of burnable poison, using the verification system illustrated in FIG. 7. FIG. 7 is a horizontal sectional view of a fuel assembly used for verification of an effect of burnable poison. The fuel assembly illustrated in FIG. 7 includes fuel rods F having uniform fissile plutonium enrichment except burnable poison-added fuel rods, and burnable poison-added fuel rods Rod 1 to Rod 4 in which the nuclear fuel material of the burnable poison-added fuel rod is configured only by natural uranium and to which 10 wt % of gadolinium is added as the burnable poison. The burnable poison-added fuel rod Rod 1 is disposed on an outermost periphery, and the burnable poison-added fuel rod Rod 4 is disposed adjacent to a water rod WR. Further, the burnable poison-added fuel rod Rod 2 is disposed on a layer that is one layer inner than the burnable poison-added fuel rod Rod 1 arranged on the outermost periphery, and the burnable poison-added fuel rod Rod 3 is disposed on a layer that is one layer inner than the burnable poison-added fuel rod Rod 2. FIG. 8 is a diagram illustrating a relationship of a neutron absorption rate of Gd157 with respect to an average fissile plutonium enrichment in the verification system illustrated in FIG. 7. In FIG. 8, the horizontal axis represents average fissile plutonium enrichment (wt %) and the vertical axis represent a neutron absorption rate of Gd157, and change of the respective neutron absorption rates of Gd 157 of the burnable poison-added fuel rods Rod 1 to Rod 4 are illustrated. Here, Gd 157 is a nuclide that is easy to react among gadolinium nuclides. As illustrated in FIG. 8, in the horizontal section of the fuel rod, the neutron absorption rate of the burnable poison-added fuel rod Rod 1 is highest, and the neutron absorption rate is decreased with the increase in the fissile plutonium enrichment. FIG. 9 illustrates a relationship between the average fissile plutonium enrichment in the verification system illustrated in FIG. 7 and the neutron absorption rate of Gd157 with respect to the burnable poison-added fuel rod Rod 1. In FIG. 9, the horizontal axis represents the average fissile plutonium enrichment (wt %) and the vertical axis represents the neutron absorption rate of Gd157 with respect to the burnable poison-added fuel rod Rod 1, and change of the neutron absorption rates of the burnable poison-added fuel rods Rod 2, Rod 3, and Rod 4 with respect to the neutron absorption rate of the burnable poison-added fuel rod Rod 1 are illustrated. While the difference from the neutron absorption rate of the burnable poison-added fuel rod Rod 1 is increased with the increase in the fissile plutonium enrichment, the difference of the burnable poison-added fuel rod Rod 4 nearly becomes constant at the fissile plutonium enrichment of 4 wt % or more, and the neutron absorption rate at that time is about half the neutron absorption rate of the burnable poison-added fuel rod Rod 1. That is, to increase the neutron absorption effect of the burnable poison, loading of the fuel rod onto the outermost periphery is effective and the effectiveness becomes noticeable with the increase in the fissile plutonium enrichment. Further, although the neutron absorption rate around the water rod WR is about half of that of the outermost periphery, the neutron absorption rate is constant at the fissile plutonium enrichment of 4 wt % or more. This position is superior to other positions other than the outermost periphery, that is, the position where the burnable poison-added fuel rod Rod 2 and the burnable poison-added fuel rod Rod 3 are disposed. Next, the present inventors and the like have paid attention to the controllability of a reactor loaded with a MOX fuel. The controllability of the reactor described here refers to, when design of continuity of energy generation during a predetermined operation period at the time of operation is made, operating the reactor with a margin on design at the time of operation in consideration of an error of the design. If there is no such margin, the operation cannot be continued if the design has an error. Further, control of the reactor becomes difficult if the margin is too small. To maximize the margin throughout the operation period, the maximum margin can be obtained throughout the period if surplus reactivity (excess reactivity) at the time of operation is nearly constant from the beginning to the end of the operation. Note that the economic efficiency is improved when the margin is designed to become almost 0 at the end of the operation. To achieve such design, the fuel to which the burnable poison is added and the fuel of the burned added burnable poison are mixed and used in the existing reactor. The fuel that is decreased in the infinite multiplication factor by burnup and the fuel that is increased in the infinite multiplication factor by the burnable poison are arranged in the reactor, whereby the excess reactivity is flattened. To flatten the excess reactivity, a straight line in which the infinite multiplication factor is increased with the burnup is required to offset a straight line in which the infinite multiplication factor is decreased with the burnup. However, if the increasing straight line becomes nonlinear, the excess reactivity becomes non flat and the margin on design becomes small. FIG. 10 illustrates a linearization effect of the infinite multiplication factor obtained by a fuel assembly. As illustrated in FIG. 10, in the case where the fuel rods to which burnable poison is added are vertically and/or horizontally adjacent to each other in the horizontal section of the fuel assembly, the increasing straight line becomes nonlinear. This nonlinearity is due to suppression of neutron absorption at the initial stage of burnup. To make the nonlinear line be a straight line, a configuration to increase the neutron absorption effect at the initial stage of burnup and to make the neutron absorption effect zero in a middle stage of burnup may just be added. In the case where the burnable poison is not arranged at adjacent positions of the fuel rods to which the burnable poison is added on the outermost periphery or at the position adjacent to the water rod of the fuel assembly, the neutron absorption effect is large, and thus the neutron absorption effect is large at the initial stage of burnup. Furthermore, in the case where the concentrations of the burnable poison are similar between the adjacent burnable poison-added fuel rod and a nonadjacent burnable poison-added fuel rod, the burnable poison in the latter rod is burned up in an about half period of that in the former rod. That is, from the viewpoint of the adjacent burnable poison-added fuel rod, the neutron absorption effect of the nonadjacent burnable poison-added fuel rod becomes almost zero in the middle of burnup. However, if the number of the nonadjacent burnable poison-added fuel rods becomes too large, the neutron absorption effect at the initial stage becomes large, and the straight line in which the infinite multiplication factor is increased is not obtained. If the nonadjacent burnable poison-added fuel rod is added for this purpose, it is counterproductive unless the number of the nonadjacent burnable poison-added fuel rods is smaller than the number of the adjacent burnable poison-added fuel rods. Embodiments of the present invention reflecting the above examination result will be described below with reference to the drawings. Note that, hereinafter, an advanced boiling water reactor (ABWR) will be described as an example. However, an embodiment is not limited to the example. For example, the present invention can be similarly applied to other nuclear reactors such as an ordinary boiling water reactor (BWR) provided with recirculation pumps, and which circulates a coolant (also functioning as a moderator for neutrons) to the outside of a reactor pressure vessel and causes the coolant to flow back to a downcomer in the reactor pressure vessel, thereby to circulate the coolant, or an economic simplified boiling water reactor (ESBWR) that eliminates the recycling pumps in BWR and internal pumps in ABWR, by use of a natural circulation system of cooling water by chimney. FIG. 1 is an overall schematic configuration view of a fuel assembly of a first embodiment according to an embodiment of the present invention, FIG. 2 is a sectional view taken along line AA (horizontal sectional view) of the fuel assembly illustrated in FIG. 1, and is a view illustrating the enrichment of each fuel rod and addition of burnable poison, and FIG. 3 is a schematic configuration view of an advanced boiling water reactor provided with a reactor loaded with the fuel assembly illustrated in FIG. 2. As illustrated in FIG. 3, in the advanced boiling water reactor (ABWR), a cylindrical reactor shroud 102 is provided in a reactor pressure vessel (reactor vessel) 103, and a reactor 105 loaded with a plurality of fuel assemblies (not illustrated) is disposed in the reactor shroud 102. Further, a steam-water separator 106 extending upward of the reactor 105 and a steam dryer 107 arranged above the steam-water separator 106 are provided in the reactor pressure vessel (hereinafter referred to as RPV) 103. An annular downcomer 104 is formed between the RPV 103 and the reactor shroud 102. An internal pump 115 is disposed inside the downcomer 104. Cooling water discharged from the internal pumps 115 is supplied to the reactor 105 via a lower plenum 122. When passing through the reactor 105, the cooling water is heated and becomes gas-liquid two-phase flow containing water and steam. The steam-water separator 106 separates the gas-liquid two-phase flow into steam and water. Moisture is further removed from the separated steam by the steam dryer 107 and the steam is led to a main steam pipe 108. The steam from which moisture has been removed is led to a steam turbine (not illustrated) to rotate the steam turbine. A generator connected to the steam turbine is rotated to generate power. The steam discharged from the steam turbine is condensed in a condenser (not illustrated) to become water. This condensed water is supplied as cooling water into the RPV 103 through a water supply pipe 109. The water separated by the steam-water separator 106 and the steam dryer 107 falls and reaches the downcomer 104 as cooling water. Although not illustrated in FIG. 3, a control rod guide pipe is provided in the lower plenum 122 of the RPV 103. The control rod guide pipe enables a plurality of control rods CR having a cross shape in cross section to be inserted into the reactor 105 to control a nuclear reaction of the fuel assembly. A control rod drive mechanism is provided in a control rod drive mechanism housing installed below a bottom portion of the RPV 103. The control rods are connected to the control rod drive mechanism. FIG. 1 illustrates an overall schematic configuration view of the fuel assembly 1. The fuel assembly 1 of the present embodiment includes a plurality of fuel rods 2, partial length fuel rods 3, an upper tie plate 5, a lower tie plate 6, a plurality of fuel spacers 8, a plurality of water rods WR, and a channel box 7. The fuel rods 2 (so-called full length fuel rods) and the partial length fuel rods 3 have a plurality of fuel pellets (not illustrated) filled in a sealed cladding tube (not illustrated). The lower tie plate 6 supports lower end portions of the fuel rods 2 and the partial length fuel rod 3, and the upper tie plate 5 holds upper end portions of the fuel rods 2. Lower end portions of the water rods WR are supported by the lower tie plate 6, and upper end portions of the water rods WR are held by the upper tie plate 5. The plurality of fuel spacers 8 is arranged at predetermined intervals in an axial direction of the fuel assembly 1, and holds the fuel rods 2 and the water rods WR to form flow paths in which the cooling water flows between the fuel rods 2 (including the partial length fuel rod 3) and between the fuel rod 2 and the water rod WR. The channel box 7, which is a square tube having a square shape in cross section, is attached to the upper tie plate 5 and extends downward. The fuel rods 2 bundled by the fuel spacers 8 are arranged in the channel box 7. A handle is fastened to an upper end portion of the upper tie plate 5, and when the handle is lifted up, the entire fuel assembly 1 can be pulled up. FIG. 2 is a sectional view taken along line AA (horizontal sectional view) of the fuel assembly 1 illustrated in FIG. 1, and is a view illustrating the enrichment of each fuel rod and addition of burnable poison. As illustrated in the upper view of FIG. 2, fuel rods 21a to 21c, a partial length fuel rod 31a, a water rod WR, and a gadolinium-containing fuel rod 41 that is a fuel rod containing gadolinium that is burnable poison are arranged in a nine-row by nine-column square lattice formed in the channel box 7 in the horizontal section of the fuel assembly 1. Two water rods WR having a cross-sectional area occupying a region where four fuel rods 2 are arrangeable are arranged in a central portion of the horizontal section (cross section) of the fuel assembly 1. The water rod WR is a large-diameter water rod having a cross-sectional area occupying a region where at least two fuel rods 2 are arrangeable. The length of a region where a fuel pellet containing fissile uranium is loaded in the fuel rod 2 in the present embodiment, that is, the effective fuel length of the present embodiment is 3.7 m. Further, when the fuel assembly 1 is loaded in the reactor 105 of the advanced boiling water reactor (ABWR), the fuel assembly 1 is arranged to have one corner face the control rod CR having a cross shape in cross section inserted in the reactor 105. The channel box 7 is attached to the upper tie plate 5 by a channel fastener (not illustrated). The channel fastener functions to hold a gap of a width necessary between the fuel assemblies 1 so that the control rod CR can be inserted into between the fuel assemblies 1 when the fuel assemblies 1 are loaded in the reactor 105. For this purpose, the channel fastener is attached to the upper tie plate 6 to be located at a corner facing the control rod CR. The corner portion of the fuel assembly 1, the corner portion facing the control rod CR, is in other words the corner portion to which the channel fastener is attached. Each fuel pellet filled in each fuel rod 2 is manufactured using uranium dioxide and plutonium oxide which are nuclear fuel materials, and contains uranium −235 and plutonium −239 and 241 that are fissile materials, and the like. As illustrated in the upper and lower views of FIG. 2, the fissile plutonium enrichment of the fuel rod 21a is 6.1 wt %, and thirty four fuel rods 21a are accommodated at lattice positions in the horizontal section of the fuel assembly 1. Further, the fissile plutonium enrichment of the fuel rod 21b is 4.2 wt %, and twenty fuel rods 21b are accommodated at lattice positions in the horizontal section of the fuel assembly 1. The fissile plutonium enrichment of the fuel rod 21c is 2.5 wt %, and five fuel rods 21c are accommodated at lattice positions in the horizontal section of the fuel assembly 1. Further, the fissile plutonium enrichment of the partial length fuel rod 31a is 6.1 wt %, and eight partial length fuel rods 31a are accommodated at lattice positions in the horizontal section of the fuel assembly 1. The gadolinium-containing fuel rod 41 does not contain plutonium but is constituted only by a uranium fuel, the enrichment of uranium is 0.2 wt % and the concentration of gadolinium (Gd), which is burnable poison, is 10 wt %. Seven gadolinium-containing fuel rods 41 are accommodated at lattice positions within the horizontal section of the fuel assembly 1. Note that the concentration of gadolinium (Gd) is not limited to 10 wt % and may be appropriately set to a desired value within a range of several wt % to 10 wt %, for example, and the enrichment of uranium is also not limited to 0.2 wt %. The fissile plutonium enrichment of a horizontal section average of the fuel assembly 1 is 4.8 wt %. In the horizontal section of the fuel assembly 1, the number of the gadolinium-containing fuel rods 41, which are burnable poison-containing fuel rods arranged on the outermost periphery, is N1, and the number of the gadolinium-containing fuel rods, which are burnable poison-containing fuel rods arranged adjacent to the water rod WR, is N2 (N2 is a positive integer including zero). In this case, in the upper view of FIG. 2, N1 is seven and N2 is zero, and a relationship of N1>N2 is satisfied. Further, the number of gadolinium-containing fuel rods, which are burnable poison-containing fuel rods arranged on the outermost periphery without being vertically and/or horizontally adjacent to each other, is W0, the number of gadolinium-containing fuel rods, which are burnable poison-containing fuel rods vertically and/or horizontally adjacent to each other and arranged on the outermost periphery, and adjacent to only one burnable poison-containing fuel rod, is W1, and the number of gadolinium-containing fuel rods, which are burnable poison-containing fuel rods vertically and/or horizontally adjacent to each other and arranged on the outermost periphery, and adjacent to two burnable poison-containing fuel rods, is W2 (W2 is a positive integer including zero). In this case, in the upper view of FIG. 2, W0 is three, W1 is four, W2 is zero, and a relationship of W2<N2+W0<W1 is satisfied. By arranging the gadolinium-containing fuel rods 41 as illustrated in the upper view of FIG. 2, change of an infinite multiplication factor of the fuel can be linearized, as illustrated in FIG. 10, and the average fissile plutonium enrichment of the MOX fuel can be increased because the fissile plutonium enrichment of the horizontal section average of the fuel assembly 1 is 4.8 wt %. As described above, according to the present embodiment, the change of the infinite multiplication factor of the fuel can be linearized and excess reactivity can be flattened while increasing the average fissile plutonium enrichment of the MOX fuel. FIG. 4 is a horizontal sectional view of a fuel assembly of a second embodiment according to another embodiment of the present invention, and is a view illustrating the enrichment of each fuel rod and addition of burnable poison. In the present embodiment, the fissile plutonium enrichment in fuel rods and arrangement positions of gadolinium-containing fuel rods, which are burnable poison-containing fuels, are different from those in the first embodiment. The other points are similar to those of the first embodiment, and hereinafter description overlapping with the first embodiment will be omitted. As illustrated in the upper view of FIG. 4, fuel rods 22a to 22d, a partial length fuel rod 32a, a partial length fuel rod 32b, a water rod WR, and a gadolinium-containing fuel rod 42 that is a fuel rod containing gadolinium that is burnable poison are arranged in a nine-row by nine-column square lattice formed in a channel box 7 in a horizontal section of a fuel assembly 1a in the fuel assembly 1a of the present embodiment. Two water rods WR having a cross-sectional area occupying a region where four fuel rods are arrangeable are arranged in a central portion of the horizontal section (cross section) of the fuel assembly 1a. The water rod WR is a large-diameter water rod having a cross-sectional area occupying a region where at least two fuel rods are arrangeable. The length of a region where a fuel pellet containing fissile uranium is loaded in the fuel rod in the present embodiment, that is, the effective fuel length of the present embodiment is 3.7 m. As illustrated in the upper and lower views of FIG. 4, the fissile plutonium enrichment of the fuel rod 22a is 9.3 wt %, and thirty two fuel rods 22a are accommodated at lattice positions in the horizontal section of the fuel assembly 1a. Further, the fissile plutonium enrichment of the fuel rod 22b is 6.5 wt %, and twenty one fuel rods 22b are accommodated at lattice positions in the horizontal section of the fuel assembly 1a. The fissile plutonium enrichment of the fuel rod 22c is 3.0 wt %, and one fuel rod 22c is accommodated at a lattice position in the horizontal section of the fuel assembly 1a. The fissile plutonium enrichment of the fuel rod 22d is 5.5 wt %, and two fuel rods 22d are accommodated at lattice positions in the horizontal section of the fuel assembly 1a. Further, the fissile plutonium enrichment of the partial length fuel rod 32a is 8.0 wt %, and four partial length fuel rods 32a are accommodated at lattice positions in the horizontal section of the fuel assembly 1a. The fissile plutonium enrichment of the partial length fuel rod 32b is 9.3 wt %, and four partial length fuel rods 32b are accommodated at lattice positions in the horizontal section of the fuel assembly 1a. The gadolinium-containing fuel rod 42 does not contain plutonium but is constituted only by a uranium fuel, the enrichment of uranium is 0.2 wt % and the concentration of gadolinium (Gd), which is burnable poison, is 10 wt %. Ten gadolinium-containing fuel rods 42 are accommodated at lattice positions within the horizontal section of the fuel assembly 1a. Note that the concentration of gadolinium (Gd) is not limited to 10 wt % and may be appropriately set to a desired value within a range of several wt % to 10 wt %, for example, and the enrichment of uranium is also not limited to 0.2 wt %. The fissile plutonium enrichment of a horizontal section average of the fuel assembly 1a is 6.8 wt %. As illustrated in the upper view in FIG. 4, in the horizontal section of the fuel assembly 1a, the number (N1) of the gadolinium-containing fuel rods 42, which are burnable poison-containing fuel rods arranged on the outermost periphery, is nine, and the number (N2) of the gadolinium-containing fuel rods 42, which are burnable poison-containing fuel rods arranged adjacent to the water rod WR, is one, and N1>N2 is satisfied. Further, the number (W0) of gadolinium-containing fuel rods, which are burnable poison-containing fuel rods arranged on the outermost periphery without being vertically and/or horizontally adjacent to each other, is two, the number (W1) of gadolinium-containing fuel rods, which are burnable poison-containing fuel rods vertically and/or horizontally adjacent to each other and arranged on the outermost periphery, and adjacent to only one burnable poison-containing fuel rod, is six, and the number (W2) of gadolinium-containing fuel rods, which are burnable poison-containing fuel rods vertically and/or horizontally adjacent to each other and arranged on the outermost periphery, and adjacent to two burnable poison-containing fuel rods, is one, and W2<N2+W0<W1 is satisfied. By arranging the gadolinium-containing fuel rods 42 as illustrated in the upper view of FIG. 4, change of an infinite multiplication factor of the fuel can be linearized, as illustrated in FIG. 10, and average fissile plutonium enrichment of a MOX fuel can be increased because the fissile plutonium enrichment of the horizontal section average of the fuel assembly 1a is 6.8 wt %. As described above, according to the present embodiment, the average fissile plutonium enrichment can be further improved as compared with the first embodiment in addition to the effect of the first embodiment. FIG. 5 is a horizontal sectional view of a fuel assembly of a third embodiment according to another embodiment of the present invention, and is a view illustrating the enrichment of each fuel rod and addition of burnable poison. In the present embodiment, the fissile plutonium enrichment in fuel rods and arrangement positions of gadolinium-containing fuel rods, which are burnable poison-containing fuels, are different from those in the first embodiment. The other points are similar to those of the first embodiment, and hereinafter description overlapping with the first embodiment will be omitted. As illustrated in the upper view of FIG. 5, fuel rods 23a to 23c, partial length fuel rods 33a to 33c, a water rod WR, and a gadolinium-containing fuel rod 43 that is a fuel rod containing gadolinium that is burnable poison are arranged in a nine-row by nine-column square lattice formed in a channel box 7 in a horizontal section of a fuel assembly 1b in the fuel assembly 1b of the present embodiment. Two water rods WR having a cross-sectional area occupying a region where four fuel rods are arrangeable are arranged in a central portion of the horizontal section (cross section) of the fuel assembly 1b. The water rod WR is a large-diameter water rod having a cross-sectional area occupying a region where at least two fuel rods are arrangeable. The length of a region where a fuel pellet containing fissile uranium is loaded in the fuel rod in the present embodiment, that is, the effective fuel length of the present embodiment is 3.7 m. As illustrated in the upper and lower views of FIG. 5, the fissile plutonium enrichment of the fuel rod 23a is 10.9 wt %, and thirty two fuel rods 23a are accommodated at lattice positions in the horizontal section of the fuel assembly 1b. Further, the fissile plutonium enrichment of the fuel rod 23b is 7.5 wt %, and twenty fuel rods 23b are accommodated at lattice positions in the horizontal section of the fuel assembly 1b. The fissile plutonium enrichment of the fuel rod 23c is 2.5 wt %, and one fuel rod 23c is accommodated at a lattice position in the horizontal section of the fuel assembly 1b. The fissile plutonium enrichment of the partial length fuel rod 33a is 10.0 wt %, and three partial length fuel rods 33a are accommodated at lattice positions in the horizontal section of the fuel assembly 1b. Further, the fissile plutonium enrichment of the partial length fuel rod 33b is 10.9 wt %, and four partial length fuel rods 33b are accommodated at lattice positions in the horizontal section of the fuel assembly 1b. The partial length fuel rod 33c is a gadolinium-containing fuel rod, the enrichment of uranium is 0.2 wt %, the concentration of gadolinium is 10 wt %, and one partial length fuel rod 33c is accommodated at a lattice position in the horizontal section of the fuel assembly 1b. The gadolinium-containing fuel rod 43 does not contain plutonium but is constituted only by a uranium fuel, the enrichment of uranium is 0.2 wt % and the concentration of gadolinium (Gd), which is burnable poison, is 10 wt %. Thirteen gadolinium-containing fuel rods 43 are accommodated at lattice positions within the horizontal section of the fuel assembly 1b. Note that the concentration of gadolinium (Gd) is not limited to 10 wt % and may be appropriately set to a desired value within a range of several wt % to 10 wt %, for example, and the enrichment of uranium is also not limited to 0.2 wt %. The fissile plutonium enrichment of a horizontal section average of the fuel assembly 1b is 7.8 wt %. As illustrated in the upper view of FIG. 5, in the horizontal section of the fuel assembly 1b, the number (N1) of the gadolinium-containing fuel rods 43, which are burnable poison-containing fuel rods arranged on the outermost periphery, is eleven, and the number (N2) of the gadolinium-containing fuel rods 43, which are burnable poison-containing fuel rods arranged adjacent to the water rod WR, is two, and N1>N2 is satisfied. Further, the number (W0) of gadolinium-containing fuel rods, which are burnable poison-containing fuel rods arranged on the outermost periphery without being vertically and/or horizontally adjacent to each other, is two, the number (W1) of gadolinium-containing fuel rods, which are burnable poison-containing fuel rods vertically and/or horizontally adjacent to each other and arranged on the outermost periphery, and adjacent to only one burnable poison-containing fuel rod, is six, and the number (W2) of gadolinium-containing fuel rods, which are burnable poison-containing fuel rods vertically and/or horizontally adjacent to each other and arranged on the outermost periphery, and adjacent to two burnable poison-containing fuel rods, is three, and W2<N2+W0<W1 is satisfied. By arranging the gadolinium-containing fuel rods 42 as illustrated in the upper view of FIG. 5, change of an infinite multiplication factor of the fuel can be linearized, as illustrated in FIG. 10, and average fissile plutonium enrichment of a MOX fuel can be increased because the fissile plutonium enrichment of the horizontal section average of the fuel assembly 1b is 7.8 wt %. According to the present embodiment, the average fissile plutonium enrichment can be further improved as compared with the first and second embodiments in addition to the effect of the first embodiment. FIG. 6 is a horizontal sectional view of a fuel assembly of a fourth embodiment according to another embodiment of the present invention, and is a view illustrating the enrichment of each fuel rod and addition of burnable poison. The present embodiment is different from the second embodiment in arranging the fuel rods in a ten by ten array in a square lattice manner. As illustrated in the upper view of FIG. 6, fuel rods 24a to 24d, a partial length fuel rod 34a, a partial length fuel rod 34b, a water rod WR, and a gadolinium-containing fuel rod 44 that is a fuel rod containing gadolinium that is burnable poison are arranged in a ten-row by ten-column square lattice formed in a channel box 7 in a horizontal section of a fuel assembly 1d in the fuel assembly 1d of the present embodiment. Two water rods WR having a cross-sectional area occupying a region where four fuel rods are arrangeable are arranged in a central portion of the horizontal section (cross section) of the fuel assembly 1d. The water rod WR is a large-diameter water rod having a cross-sectional area occupying a region where at least two fuel rods are arrangeable. The length of a region where a fuel pellet containing fissile uranium is loaded in the fuel rod in the present embodiment, that is, the effective fuel length of the present embodiment is 3.7 m. As illustrated in the upper and lower views of FIG. 6, the fissile plutonium enrichment of the fuel rod 24a is 9.3 wt %, and forty fuel rods 24a are accommodated at lattice positions in the horizontal section of the fuel assembly 1d. Further, the fissile plutonium enrichment of the fuel rod 24b is 6.5 wt %, and twenty four fuel rods 24b are accommodated at lattice positions in the horizontal section of the fuel assembly 1d. The fissile plutonium enrichment of the fuel rod 24c is 3.0 wt %, and one fuel rod 24c is accommodated at a lattice position in the horizontal section of the fuel assembly 1d. The fissile plutonium enrichment of the fuel rod 24d is 5.5 wt %, and two fuel rods 24d are accommodated at lattice positions in the horizontal section of the fuel assembly 1d. Further, the fissile plutonium enrichment of the partial length fuel rod 34a is 8.0 wt %, and eight partial length fuel rods 34a are accommodated at lattice positions in the horizontal section of the fuel assembly 1d. The fissile plutonium enrichment of the partial length fuel rod 34b is 9.3 wt %, and six partial length fuel rods 34b are accommodated at lattice positions in the horizontal section of the fuel assembly 1d. The gadolinium-containing fuel rod 44 does not contain plutonium but is constituted only by a uranium fuel, the enrichment of uranium is 0.2 wt % and the concentration of gadolinium (Gd), which is burnable poison, is 10 wt %. Eleven gadolinium-containing fuel rods 44 are accommodated at lattice positions within the horizontal section of the fuel assembly 1d. Note that the concentration of gadolinium (Gd) is not limited to 10 wt % and may be appropriately set to a desired value within a range of several wt % to 10 wt %, for example, and the enrichment of uranium is also not limited to 0.2 wt %. The fissile plutonium enrichment of a horizontal section average of the fuel assembly 1d is 7.2 wt %. As illustrated in the upper view of FIG. 6, in the horizontal section of the fuel assembly 1d, the number (N1) of the gadolinium-containing fuel rods 44, which are burnable poison-containing fuel rods arranged on the outermost periphery, is nine, and the number (N2) of the gadolinium-containing fuel rods 44, which are burnable poison-containing fuel rods arranged adjacent to the water rod WR, is two, and N1>N2 is satisfied. Further, the number (W0) of gadolinium-containing fuel rods, which are burnable poison-containing fuel rods arranged on the outermost periphery without being vertically and/or horizontally adjacent to each other, is two, the number (W1) of gadolinium-containing fuel rods, which are burnable poison-containing fuel rods vertically and/or horizontally adjacent to each other and arranged on the outermost periphery, and adjacent to only one burnable poison-containing fuel rod, is six, and the number (W2) of gadolinium-containing fuel rods, which are burnable poison-containing fuel rods vertically and/or horizontally adjacent to each other and arranged on the outermost periphery, and adjacent to two burnable poison-containing fuel rods, is one, and W2<N2+W0<W1 is satisfied. By arranging the gadolinium-containing fuel rods 42 as illustrated in the upper view of FIG. 6, change of an infinite multiplication factor of the fuel can be linearized, as illustrated in FIG. 10, and average fissile plutonium enrichment of a MOX fuel can be increased because the fissile plutonium enrichment of the horizontal section average of the fuel assembly 1d is 7.2 wt %. As described above, according to the present embodiment, the average fissile plutonium enrichment can be further improved as compared with the second embodiment in addition to the effect of the second embodiment. Further, an average output per fuel rod is decreased as the number of fuel rods in the fuel assembly 1d is increased, and thus heat removal becomes easy and thermal margin can be improved as compared with the second embodiment. Note that the present invention is not limited to the above-described embodiments and includes various modifications. For example, the above embodiments have been described in detail for easy understanding of the present invention, and the present invention is not necessarily limited to one including all the described configurations. Further, a part of the configuration of a certain embodiment can be replaced with the configuration of another embodiment. Further, the configuration of another embodiment can be added to the configuration of a certain embodiment. Further, the configuration of another embodiment can be added to/deleted from/replaced with a part of the configurations of each embodiment. 1, 1a, 1b, 1d fuel assembly 2, 21a to 21c, 22a to 22d, 23a to 23c, 24a to 24d fuel rod 3, 31a, 32a, 32b, 33a, 33b, 34a, 34b partial length fuel rod 4, 41, 42, 43, 44 gadolinium-containing fuel rod 5 upper tie plate 6 lower tie plate 7 channel box 8 spacer WR water rod 102 reactor shroud 103 reactor pressure vessel 104 downcomer 105 reactor 106 steam-water separator 107 steam dryer 108 main steam pipe 109 water supply pipe 115 internal pump 122 lower plenum |
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claims | 1. A nuclear fuel pellet, comprising:a porous substrate; anda plurality of layers of a fuel containing material deposited via atomic layer deposition upon said porous substrate, wherein the total porosity of the nuclear fuel pellet is between about 3 and 30 percent, wherein the porous substrate underlies the deposited layers of the fuel. 2. The nuclear fuel pellet of claim 1, wherein the fuel containing material is fissile material selected from the group consisting of uranium dioxide, plutonium dioxide, uranium-zirconium, uranium molybdenum and uranium nitride, and combinations thereof. 3. The nuclear fuel pellet of claim 1, wherein said porous substrate is an aerogel which defines pores which extend through the substrate. 4. The nuclear fuel pellet of claim 1, wherein the fuel overlays substantially all of the surfaces of the substrate which define pores. 5. The nuclear fuel pellet of claim 1, wherein the porous substrate has a density less than the density of the fuel. 6. The nuclear fuel pellet of claim 5, wherein the porous substrate is selected from the group consisting of carbon, tungsten, beryllium oxide, and aluminum nitride. 7. The nuclear fuel pellet of claim 1, wherein the fuel pellet has an overall density of about 90% of the density of the fuel material. 8. The nuclear fuel pellet of claim 1, wherein the pores within the pellet have a uniform size and distribution. 9. The nuclear fuel pellet of claim 1, wherein the pores at the surface of the pellet are smaller than the pores at the center of the pellet and wherein the size of the pores tapers between the center and the surface of the pellet. 10. The nuclear fuel pellet of claim 1, wherein the fuel pellet has closed pores. 11. The nuclear fuel pellet of claim 1, wherein the pores are uniformly distributed throughout the pellet and wherein the pores are spaced between about 0.02 mm and about 0.2 mm apart from each other. 12. The nuclear fuel pellet of claim 1 wherein the porosity of the pellet is between about 15 and 30 percent. 13. A nuclear reactor fuel cladding, said cladding comprising:a porous substrate of silicon carbide, wherein the substrate is substantially a hollow tube, and wherein the porous substrate surrounds nuclear fuel without an underlying liner between said fuel and said porous substrate; andat least one layer of silicon carbide deposited on the substrate via atomic layer deposition. 14. The nuclear reactor fuel cladding of claim 13, wherein the porous substrate is a silicon carbide aerogel. 15. The nuclear reactor fuel cladding of claim 13, wherein the porous substrate is a silicon carbide cloth. 16. The nuclear reactor fuel cladding of claim 13, wherein the at least one layer of silicon carbide deposition is one atom thick. 17. The nuclear reactor fuel cladding of claim 13, wherein the coating of silicon carbide is conformal and substantially defect free. |
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abstract | A lithographic apparatus includes a projection system including a movable optical element. The movable optical element is capable, by a displacement thereof to influence a position quantity of a radiation beam projected by the projection system. A control device is provided to drive the optical element actuator to influence a position quantity of the movable optical element, thereby influencing a position quantity of the radiation beam as projected by the projection system. The control device is adapted to move the movable optical element to position the radiation beam as projected by the projection system with respect to the substrate, or to correct a position quantity of the radiation beam as projected by the projection system caused by any type of disturbance on the projection system. |
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056129823 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT Referring now to the drawings in detail and in particular to FIG. 1 there is shown a pressurized water nuclear power plant 10 including a nuclear reactor vessel 12 interconnected with one or more steam generators 14 (illustrated as one steam generator) by a hot leg 16 and at least one cold leg 18. A typical nuclear power plant 10 will have from two to four steam generators 14 coupled with a reactor vessel 12. FIG. 1 generally represents a passive pressurized water reactor facility like that disclosed by U.S. Pat. Nos. 4,753,771 and 5,049,353, which are incorporated by this reference for their discussion of the structure and operation of such plants. The present invention may be employed with other passive plant designs or with conventional active plant designs using air fans, water pumps and like active components. It may also be employed in connection with boiling water reactor plants in addition to pressurized water reactor plants and in connection with spherically shaped containment structures. The reactor vessel 12 and steam generator 14 of FIG. 1 are located in a double containment 20 including an inner shell 22 generally concentrically spaced from an outer shell 24. The inner shell 22 generally includes a sidewall 26 disposed between a domed top end 28 and a domed bottom end 30. Similarly, the outer shell 24 generally includes a sidewall 32 disposed between a domed top end 34 and a domed bottom end 36. The containment shells 22 and 24 may be fabricated of a metal such as steel or of concrete. The containment structure 20 has various conventional penetrations through the containment shells 22 and 24 for piping, instrumentation and electrical lines and various hatches (not shown) which permit fuel assemblies, personnel and the like to enter and to exit the containment structure 20. As is shown in FIG. 1, the reactor vessel 12 is generally located below the grade 38 of the terrain. The steam generator 14 generally extends about four feet or more above a grade level operating floor 40. A containment cooling system 50 is piped with an out-of-containment heat sink such as a heat exchanger 52 for transferring heat from within the containment structure 20. The heat exchanger 52 transfers heat from water circulating in the cooling system 50 to water from a plant service water system 54 which then transfers the heat to an ultimate heat sink (not shown) such as an air-cooled cooling tower or a heat exchanger cooled by river water and the like. The containment cooling system 50 generally includes a circulating pump 60 such as a centrifugal pump (which is normally paired with an installed spare pump) located outside of the containment structure 20. The cooling system 50 also has check valves 61 to limit backflows through the circuit. The circulating pump 60 has a suction side connected with the heat sink 52 and a discharge side connected via an inlet header 62 to one or more in-containment heat exchangers 64, and preferably, to a plurality of in-containment heat exchangers 64 substantially surrounding the reactor vessel 12 and steam generators 14. The circulating water then flows from the in-containment heat exchangers 64 to an outlet header 66 and returns to the out-of-containment heat exchanger 52. The containment cooling system 50 may include a cooling water supply tank 68. FIG. 1 shows a supply tank 68 which is connected via a pipe 70 with the suction of the circulating pump 60 and with the outlet header 66 by a return pipe 72. The supply tank 68 of a passive plant normally will be sized to provide cooling water for up to about seventy-two hours or more in order to provide water to the in-containment heat exchangers 64 in the event the service water system 54 is unavailable. The supply tank 68 may alternatively be designed to help meet initial blowdown requirements to minimize, the heat exchanger requirements. In a plant 10 such as that shown in FIG. 1, the cooling system 50 is designed to continue functioning under all circumstances, even if the circulating pump 60 is unavailable because of mechanical malfunction, power unavailability and the like. Thus, the various heat exchangers 52 and 64 and other parts of the cooling system 50 are preferably sized and otherwise designed to facilitate natural circulation of cooling water through the system 50 and transfer heat from the containment atmosphere for up to about seventy-two hours or more. Accordingly, the normally pumped circulation through the in-containment heat exchangers 64 is upwardly through the heat exchangers 64 to facilitate natural circulation of the water as it is being heated. The in-containment heat exchangers 64 preferably extend from about the level of the operating floor 40 into the dome top 28 for inducing convective air flow throughout the containment structure 20. Most preferably, the heat exchangers 64 below the operating floor 40 and are configured to curve with the dome top 28 to induce mixing and circulation. Advantageously, this circulation restricts the development of high localized concentrations of hydrogen or other gases in the dome or other areas of the containment structure 20 under all operating conditions. As is best shown in FIG. 2, the heat exchangers 64 extend vertically adjacent the inner containment wall 26 and, most preferably, the heat exchangers 64 are no more than about three feet from the wall 26, for inducing natural flow of the air and condensed steam along the heat exchangers 64 and along the sidewall 26. As is best shown in FIGS. 2 and 3, the in-containment heat exchangers 64 preferably includes several subassemblies 74, 76, 78 which may be assembled at the plant site. The heat exchangers 64 may alternatively be constructed of one assembly (not shown). The heat exchangers 64 may have substantially parallel lengths of water conducting pipes 82 connected in series by 180.degree. bends 83 in a serpentine design. In another design (shown in FIG. 4) the substantially parallel pipes may be manifolded for parallel water flow through the pipes. The pipes 82 may be connected with each other and to the inlet and outlet headers 62, 66 by unions 84 (as is shown by FIG. 2) or by welds 85 (as is shown by FIG. 4). The pipes 82 may be fabricated of carbon or stainless steel or other suitable metal. Also, the pipes 82 may have an inorganic zinc coating for improving the wettability of the heat exchanger 64 in order to condense the steam. The in-containment heat exchangers 64 have cooling fins 90 extending vertically from the pipes 82 for providing sufficient surfaces for heat transfer and for directing the flow of air and condensed steam downwardly around the periphery of the containment structure 20. The cooling fins 90 are preferably shrunk fit on the pipes 82 and retain their fit under all thermal conditions. The fins 90 may be fabricated of a material such as steel, copper or other suitable material which satisfies the structural and corrosion requirements of the heat exchanger 64. In addition, the fins 90 may have an inorganic zinc coating to enhance their wettability. The fins 90 preferably adjoin a backing plate 92 which structurally reinforces the fins 90 and extends their effective surface. Preferably, the fins 90 are welded to the backing plate 92 and both have an inorganic zinc coating for improving the wettability of the heat exchanger 64 in order to condense the steam. As is best shown in FIG. 3, the backing plates 92 may have configured ends 93, 94 to fit with adjacent backing plates 92. FIGS. 4 and 5 generally show an in-containment heat exchanger 102 comprised of subassemblies 104, 106 and 108. Cooling water flows upwardly from the inlet header 62 into manifold 112 and then through substantially parallel vertical pipes 114. The cooling water then flows out through manifold 116 and into the outlet header 66. The pipes 114 have vertically extending cooling fins 120. As is best seen in FIG. 5, the pipes 114 may also have cooling fins 122 extending vertically between the pipes 114. Advantageously, the cooling fins 122 may also function as the backing plate 92 shown in FIGS. 2 and 3. Alternatively, and as is shown in FIG. 6, the pipes 114 may have vertical fins 124 and vertical fins 126 oriented at substantially 180.degree. relative to each other. A backing plate 128 adjoining the vertical fins 128 may be employed to mechanically support the fins 128 and to extend the heat transfer surfaces. As is shown in FIGS. 4-6, there may also be one or more rows of baffles 130 extending horizontally of the fins 120, 124 (and, although not shown, fins 90 in FIGS. 2-3). These baffles 130 slope downwardly toward the pipes 114 for directing the flow of air in the containment structure 20 downwardly along the heat exchanger 104. Preferably, the baffles 130 are oriented so as to permit visual inspection of the pipes 114 and yet block a horizontal view of the pipes 114 for protecting the pipes 114 from damage. FIG. 6 shows an arrangement where both a backing plate 128 and baffles 130 may be employed to structurally reinforce the heat exchanger 102 against hydraulic and thermal transient flows. A nuclear power plant embodying the present invention is designed to cool the atmosphere and the apparatus in a containment structure 20 in all situations. Studies have shown that nuclear power plants embodying the present invention will experience lower pressure increases upon the occurrence of certain design basis events than will other designs. Referring to FIG. 3, the air in the containment structure 20 naturally circulates downwardly along the heat exchangers 64 and containment sidewall 26 as is shown by flow indicators 150, 152 and then inwardly as is shown by flow indicators 154, 156. The cooled air then contacts equipment and other apparatus located in the interior portions of the containment structure 20, which are at high temperatures, is entrained by rising steam jets or plumes and rises toward the dome area to complete the cycle. In situations where large amounts of steam are released into the atmosphere, such as in the event of a guillotine pipe break, cooled condensate falls from the bottom of heat exchangers 64 and 102 toward a collection gutter 162 as is shown by the condensate flow indicator 160 in FIG. 3. The collection gutter 162 or other return system may be employed to continue cooling the primary system. For example, the flow of air 156 from between the heat exchangers 64, 114 and the sidewall 26 of the containment structure 20 may entrain at least some of the condensate and carry it to the interior portions of the containment structure 20 where it can contact and cool the apparatus and piping. The containment cooling system 50 passively circulates water through the system if necessary. If, for example, the circulating pump 60 shown on FIG. 1 malfunctions or there is a power interruption, the water in the containment cooling system 50 circulates by natural convection. If the out-of-containment heat sink 52 is also unavailable, water from supply tank 68 may be circulated through the cooling system 50. While the preferred embodiments described herein set forth the best mode to practice this invention presently contemplated by the inventors, numerous modifications and adaptations of this invention will be apparent to others skilled in the art. Therefore, the embodiments are to be considered as illustrative and exemplary and it is understood that the claims are intended to cover such modifications and adaptations as they are considered to be within the spirit and scope of this invention. |
claims | 1. A system for regulating a liquid in a circuit, with the system comprising:a plug valve comprising at least one inlet and one outlet, the plug comprising an internal passage through which is intended to pass the liquid flowing from the inlet to the outlet of the valve when the valve is open at least partially,an expansion reservoir in communication with the liquid flowing in the circuit and intended to contain liquid and a compensating gas,wherein the plug comprises at least a part of an expansion channel which has at least one lateral opening located on a lateral face of the plug and which is conformed to provide in operation a communication between said lateral opening and the expansion reservoir, the valve being conformed in such a way that:at least when the valve is closed: the lateral opening is in direct communication with the liquid coming from the inlet or from the outlet of the valve;when the valve is partially open and when the valve is fully open, the lateral opening cooperates with an inner wall integral with a body of the valve in such a way as to form a conduit in communication on the one hand with the expansion reservoir and on the other hand with the internal passage. 2. The system according to claim 1, wherein the conduit opens on the one hand into the expansion reservoir and opens on the other hand into a space formed by a lower face of the plug and a bottom of the body of the valve, with this space being in communication with the internal passage by said part of the expansion channel, said part of the expansion channel being made in the plug. 3. The system according to claim 1, wherein the lateral opening is a recess, with the cooperation of the recess and the inner wall forming the conduit when the valve is open at least partially. 4. The system according to claim 2, wherein the lateral opening is a recess, with the cooperation of the recess and the inner wall forming the conduit when the valve is open at least partially and wherein the recess extends from the expansion reservoir to the lower face of the plug. 5. The system according to claim 1, wherein the plug is a spherical plug. 6. The system according to claim 1, wherein the plug is a cylindrical plug. 7. The system according to claim 6, wherein the conduit opens on the one hand into the expansion reservoir an opens on the other hand into a space formed by a lower face of the plug and a bottom of the body of the valve, with this space being in communication with the internal passage by a part of the expansion channel, said part of the expansion channel being made in the plug, the system being conformed in such a way that when the valve is open, the expansion reservoir communicates with the liquid passing through the valve solely through the recess, of said space and of the lower channel. 8. The system according to claim 1, wherein the inlet and the outlet of the valve form an angle between 130° and 180°. 9. The system according to claim 1, wherein the valve is a straight valve. 10. The system according to claim 1, wherein the inlet and the outlet of the valve forming an angle less than 130°. 11. The system according to claim 1, wherein the valve comprises a cover forming with the body an enclosure and wherein the expansion reservoir is housed in the enclosure. 12. The system according to claim 1, configured in such a way as to orient the direction of closing of the plug according to the direction of the circulation of the liquid in the circuit. 13. The system according to claim 1, wherein the plug is actuated by a control device comprising a reduction gear housed inside the expansion reservoir. 14. The system according to claim 13, wherein the reduction gear is immersed in the compensating gas. 15. The system according to claim 14, comprising an overflow in order to limit the level of liquid in the expansion reservoir and wherein the reduction gear is arranged above the overflow. 16. The system according to claim 15, comprising an aerator device arranged in the expansion reservoir, under the overflow and configured to break the jets of liquid coming from the expansion channel. 17. The system according to claim 13, comprising a thermal protection device housed inside the expansion reservoir and conformed to thermally insulate the reduction gear from the heat of the liquid. 18. The system according to claim 1, comprising a rotational guiding bearing of the plug and wherein the bearing is housed inside the expansion reservoir. 19. The system according to claim 18, configured in such a way that in operation the bearing is immersed in the liquid contained in the expansion reservoir. 20. The system according to claim 19, wherein the bearing comprises a passage allowing the free circulation of the liquid contained in the expansion reservoir through the bearing. 21. The system according to claim 1, wherein the valve is a throttle valve. 22. The system according to claim 1, wherein the plug is movable with respect to the expansion reservoir. 23. The system according to claim 1, wherein the expansion reservoir is vertically arranged higher than the plug. 24. The system according to claim 1, wherein the expansion reservoir surmounts the plug valve. 25. The system according to claim 1, wherein the expansion reservoir is fixed with respect to the body of the valve during the displacement of the plug valve. 26. The system according to claim 25, wherein the expansion reservoir is formed at least partially by the inner wall of the valve. 27. The system according to claim 26, wherein the expansion reservoir is connected to the valve by being arranged at a distance from the latter. 28. The system according to claim 1, wherein the valve comprises a cover, wherein the cover comprises an inner wall and wherein the expansion reservoir is formed by the inner walls of the body of the valve, by the inner wall of the cover and by an upper face of a body of the plug valve. 29. A circuit comprising the system according to claim 1 and a pump able to deliver in two opposite directions. 30. The circuit according to claim 29, wherein the plug comprises the expansion channel for the passing of the liquid opening into the inner passage of the plug in order to place in communication the expansion reservoir with the circuit, the circuit being configured in such a way as to orient the direction of closing of the plug according to the direction of the circulation of the liquid in the circuit. 31. The circuit according to claim 29, configured in such a way that, during the closing of the valve, the plug is turned in such a way that the expansion channel remains in communication with a portion of the circuit separating the valve from an inlet of the pump. 32. A method of using a system according to claim 1 comprising a step of providing the circuit with the liquid to be regulated, the liquid having a temperature greater than or equal to 350° C. 33. The method according to claim 32 wherein the liquid to be regulated is liquid sodium intended to provide for the heat transfer in a circuit of a sodium cooled nuclear reactor. |
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description | The present application is a U.S. national stage application under 35 U.S.C. § 371 of PCT Application No. PCT/US2014/062094, filed Oct. 24, 2014, which claims the benefit of U.S. Provisional Patent Application Ser. No. 61/895,267 filed Oct. 24, 2013, the entireties of which are incorporated herein by reference. The present invention relates nuclear steam supply systems, and more particularly to a steam generator used in a modular reactor system having natural gravity driven coolant flow circulation. Pressurized water reactors (PWRs) for nuclear power generation facilities utilize both pumped and natural circulation of the primary coolant (water) to both cool the reactor core and heat the secondary coolant (water) to produce steam which may be working fluid for a Rankine power generation cycle. The existing natural circulation PWRs suffer from the drawback that the heat exchange equipment is integrated with and located within the reactor pressure vessel. Such an arrangement not only makes the heat exchange equipment difficult to repair and/or service, but also subjects the equipment to corrosive conditions and results in increased complexity and a potential increase in the number of penetrations into the reactor pressure vessel. In addition, locating the heat exchange equipment within the reactor pressure vessel creates problems with respect to radiation levels encountered for crews to repair the heat exchange equipment in proximity to the radioactively hot components of the reactor vessel. The general view has also been that the heat exchangers should be located in the reactor vessel to achieve natural circulation in those systems which may utilize this type of flow circulation. The steam generator (SG) is a vitally important tubular heat exchanger in a pressurized water reactor (PWR). It serves to boil the purified Rankine cycle secondary coolant water (also called the “secondary” side water or feedwater) into steam using the heat energy from the reactor primary coolant heated by its circulation through the reactor's core (called the “primary” side). Because of the high operating pressure (typically over 2200 psi) of the reactor coolant, the steam generator is a massive piece of vertically arrayed equipment. The transfer of heat energy occurs from the primary fluid flowing inside the tubes to the secondary water located in the space outside the tubes. Improvements in nuclear steam generators are desired. The present invention provides an improved steam generator for a nuclear steam supply system. According to one embodiment, a nuclear steam supply system with natural gravity-driven coolant circulation includes: a vertically-oriented reactor vessel comprising an elongated cylindrical shell forming an internal cavity configured for containing primary coolant and a nuclear reactor fuel core; a vertically-oriented steam generating vessel comprising an elongated cylindrical shell defining an internal cavity, a top tubesheet, and a bottom tubesheet; a vertical riser pipe extending vertically between the top and bottom tubesheets, the riser pipe fluidly connected to the reactor vessel; a plurality of heat transfer tubes extending vertically between the top and bottom tubesheets; and a fluid coupling comprising an eccentric cone section forming a flow conduit for exchanging primary coolant between the steam generating vessel and reactor vessel. A closed primary coolant loop is formed in which primary coolant flows from the reactor vessel through the eccentric cone into the steam generator vessel and returns from the steam generating vessel to the reactor vessel through the eccentric cone. According to another embodiment, a nuclear steam supply system with natural gravity-driven coolant circulation includes: a vertically-oriented reactor vessel comprising an elongated cylindrical shell forming an internal cavity configured for containing primary coolant and a nuclear reactor fuel core; a vertically-oriented steam generating vessel comprising an elongated cylindrical shell defining an internal cavity configured for containing secondary coolant, a top tubesheet, and a bottom tubesheet; a plurality of heat transfer tubes extending vertically between the top and bottom tubesheets, the tubes including a preheater section, a steam generator section, and a superheater section, wherein secondary coolant in a liquid state enters a shell side of the preheater section at a bottom of the steam generating vessel and flows upward to the steam generator section where a portion of the secondary coolant boils to produce steam which in turn flows upward into the superheater section at a top of the steam generating vessel; a vertical riser pipe extending vertically between the top and bottom tubesheets, the riser pipe fluidly connected to the reactor vessel; a fluid coupling forming a flow conduit for exchanging primary coolant between the steam generating vessel and reactor vessel; and a tubular recirculation shroud surrounding the tubes in the steam generator section, the shroud configured to recirculate a portion of the liquid secondary coolant in the steam generator section to the preheater section. The primary coolant flows upward through the riser pipe and downward through the tubes on the tube side of the steam generating vessel to heat the secondary coolant. According to one embodiment, a steam generator for a nuclear steam supply system includes: a vertically-oriented steam generating vessel comprising an elongated cylindrical shell defining an internal cavity configured for containing secondary coolant, a top tubesheet, a secondary coolant outlet nozzle below the top tubesheet, a bottom tubesheet, and a secondary coolant inlet nozzle above the bottom tubesheet; a plurality of heat transfer tubes extending vertically between the top and bottom tubesheets, the tubes including a preheater section, a steam generator section, and a superheater section, wherein secondary coolant in a liquid state enters a shell side of the preheater section via the inlet nozzle and flows upward to the steam generator section where a portion of the secondary coolant boils to produce steam which in turn flows upward into the superheater section and exits the steam generating vessel through the outlet nozzle; a vertical riser pipe extending vertically between the top and bottom tubesheets, the riser pipe in fluid communication with the tubes and configured for fluid coupling to a reactor vessel containing primary coolant; a double-walled fluid coupling forming a flow conduit for exchanging primary coolant between the steam generating vessel and reactor vessel, the fluid coupling configured so that primary coolant from the reactor vessel flows through the fluid coupling into the steam generator vessel and returns from the steam generating vessel to the reactor vessel through the fluid coupling; a bottom collection plenum formed below the bottom tubesheet by the fluid coupling and configured for fluid coupling to the reactor vessel, the collection plenum in fluid communication with the tubes; a top distribution plenum formed above the top tubesheet, the distribution plenum in fluid communication with the riser pipe and tubes; and a tubular recirculation shroud surrounding the tubes in the steam generator section, the shroud configured to recirculate a portion of the liquid secondary coolant in the steam generator section to the preheater section. Advantages and aspects of the present invention include the following: Core deep underground: The reactor core resides deep underground in a thick-walled Reactor Vessel (RV) made of an ASME Code material that has decades of proven efficacy in maintaining reactor integrity in large PWR and BWR reactors. All surfaces wetted by the reactor coolant are made of stainless steel or Inconel, which eliminates a major source of corrosion and crud accumulation in the RV. Gravity-driven circulation of the reactor coolant: The nuclear steam supply system according to the present disclosure does not rely on any active components (viz., a Reactor Coolant pump) for circulating the reactor coolant through the core. Instead, the flow of the reactor coolant through the RV, the steam generator heat exchangers, and other miscellaneous equipment occurs by the pressure head created by density differences in the flowing water between the hot and cold segments of the primary loop. The reliability of gravity as a motive force underpins its inherent safety. The movement of the reactor coolant requires no pumps, valves, or moving machinery of any kind. Black-start capable (no reliance on off-site power): Off-site power is not essential for starting up or shutting down the nuclear steam supply system. The rejection of reactor residual heat during the shutdown also occurs by gravity-driven circulation. Thus, the need for an emergency shutdown power supply at the site—a major concern for nuclear plants—is eliminated. Indeed, the nuclear steam supply system uses gravity (and only gravity) as the motive force to meet its operational imperatives under both normal and accident conditions. Assurance of a large inventory of water around and over the reactor core: The present nuclear steam supply system reactor vessel (RV) has no penetrations except at its very top, which means that the core will remain submerged in a large inventory of water even under the hypothetical postulated event under which all normal heat rejection paths are lost. No large penetrations in the Reactor Vessel (RV): All penetrations in the RV are located in the top region of the RV and are small in size. The absence of large piping in the reactor coolant system precludes the potential of a “large break” Loss of Coolant Accident (LOCA) event. Easy accessibility to all critical components: In contrast to the so-called “integral” reactor systems, the steam generator and the control rod drive system are located outside the RV at a level that facilitates easy access, making their preventive maintenance and repair a conveniently executed activity. The steam generator consists of a single loop that includes in some embodiments a preheater, steam generator, and a superheater topped off by a pressurizer. The heat exchangers in the loop, namely the preheater, the steam generator, and the superheater have built-in design features to conveniently access and plug tubes such as appropriate placed manholes that provide access to the heat exchanger tubesheets and/or tube bundles. The decision to deploy the heat exchange equipment outside of the harsh environment of the reactor cavity in the nuclear steam supply system has been informed by the poor reliability of PWR steam generators over the past 3 decades and the colossal costs borne by the industry to replace them. The RV flange features a reverse joint to minimize its projection beyond the perimeter of the RV cylinder. This design innovation makes it possible to connect the Stack directly to the RV nozzle-gorging to forging connection-eliminating any piping run between them. This design features eliminates the risk of a large pipe break LOCA. Demineralized water as the reactor coolant: The reactor coolant is demineralized water, which promotes critical safety because of its strong negative reactivity gradient with rise in temperature. Elimination of borated water also simplifies the nuclear steam supply system (NSSS) by eliminating the systems and equipment needed to maintain and control boron levels in the primary coolant. Pure water and a corrosion-resistant primary coolant loop help minimize crud buildup in the RV. Improved steam cycle reliability: The reliability of the steam cycle is improved by dispensing with the high pressure turbine altogether. Rather, the cycle steam is superheated before it is delivered to the low pressure turbine. The loss in the Rankine efficiency is less than 0.5 percent; the rewards in terms of enhanced reliability and simplification of the power cycle are quite substantial. Pressure Control: The pressurizer contains a conventional heating/quenching element (water/steam for pressure control). A bank of electric heaters are installed in the pressurizer section which serve to increase the pressure by boiling some of the primary coolant and creating a steam bubble that resides at the top of the pressurizer near the head. A spray column is located near the top head of the pressurizer which sprays water into the steam bubble thereby condensing the steam and reducing the steam bubble. The increase/decrease in size of the steam bubble serves to increase/decrease the pressure of the primary coolant inside the reactor coolant system. In one exemplary embodiment, the primary coolant pressure maintained by the pressurizer may be without limitation about 2,250 psi. In alternative embodiments, a nitrogen type pressurizer system may be used. In this embodiment, the pressurizer serves to control the pressure in the reactor vessel by the application of controlled nitrogen pressure from external high pressure nitrogen tanks fluidly coupled to the pressurizer. Nitrogen pressure controlled reactors have been used in other reactor types and have years of successful operating experience with a quick response profile. Preventing fuel failures in the reactor: Over 70 percent of all fuel failures in operation are known to occur from fretting (erosion from repetitive impact) damage, which is the result of “pinging” of the fuel rods by the grid straps. The vibration of the grid straps is directly related to the level of turbulence around the fuel. In the present nuclear steam supply system, the Reynolds number is approximately 20 percent of that in a typical operating PWR today. A lower Reynolds number translates into an enfeebled pinging action (erosion rate varies approximately as 4.8 power of velocity of impact!) on the rods and thus a drastically reduced fretting damage rate. Lower burn-up levels selected for present nuclear steam supply system (in the 45 GWD per MTU range) in comparison to around 60 in the presently operating reactors) will also help ameliorate embrittlement of the fuel cladding and thus prevent rod wastage. Increased Self-shielding: The gravity-driven circulation of the primary fluid in the present nuclear steam supply system (NSSS) accrues another significant dividend in the form of a dramatically reduced radiation dose emanating from the NSSS. This is because the Nitrogen (N-16) isotope, produced by the neutron bombardment of oxygen in the reactor water in the core, generates high gamma energy emitting N-16 isotope which is largely responsible for the radiation emanating from the Containment. N-16, however, has a half-life of only 7.4 seconds which is less than one-fourth of the time needed for the primary water to travel to the top of the steam generators. Therefore, the quantity of N-16 is attenuated by over 7 half-lives, which means it is in effect depopulated down to minuscule values. Scoping calculations suggest that the radiation dose from the top of the steam generator in the NSSS can be 3 or more orders of magnitude less than that in a pumped-water PWR of a similar size. Thus, it is not necessary to build a thick concrete containment for present NSSS for radiation shielding. In lieu of building and in situ reinforced concrete containment, a shop fabricated steel containment capable of withstanding a crashing airplane is deployed which is more suitable, and more economical. All drawings are schematic and not necessarily to scale. 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. 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. Steam generators used in modern nuclear power plants fall into two categories: Recirculating Type and Once-Thru Type. Recirculating Type Steam Generators: The recirculating steam generator is the most commonly used design in the industry. It features a vertical U-bundle with a hemi-head integrally welded to the tubesheet and the primary fluid entering the up flow leg of the U-tubes and exiting through the same hemi-head after flowing down the other leg of the tubes. The hemi-head space is divided into two compartments to separate the “hot” and “cooled” primary streams. The secondary side features a shroud that enables the boiling water to recirculate by the thermo-siphon action. Most world suppliers of operating PWRs, including Westinghouse, Siemens, Combustion Engineering, Framatome, and Mitsubishi utilized the Recirculating type U-bundle steam generators in their PWRs. However, the recirculating type steam generators have suffered from massive tube failures in PWRs around the world well before the end of their design life, forcing utilities to spend billions of dollars in replacement costs. Some plants such as Maine Yankee. Trojan, and Connecticut Yankee have shutdown permanently because of the high cost of steam generator replacement. Once-Thru Type Steam Generators: This design employs straight tubes fastened to the tubesheets located at the top and bottom extremities of the tube bundle. The primary water (reactor coolant) and the secondary water (boiler water) flow in a counter-current arrangement with the latter boiling outside the tubes as it extracts heat from the former across the tube walls. The mixture of water and steam in the lower produced in the lower reaches of the tube bundle progressively becomes more dry steam as the secondary flow stream rises inside the steam generator. Babcock & Wilcox was the only reactor supplier who used the once-thru steam generator configuration in its Pressurized Water Reactor (PWR) designs. The disaster at Three Mile Island Unit 2 supplied by B&W in 1979 exposed the flaws of this design. As shown in FIGS. 1 and 2, a nuclear steam supply system 100 (NSSS) of a safe modular underground reactor (SMR) according to the present disclosure comprises a vertical subterranean reactor vessel 200 with the nuclear fuel cartridge (or core) 210 containing the nuclear fuel source standing upright near its bottom. The basic flow or circulation path of the primary coolant or water (i.e. closed primary coolant loop) contained in the reactor vessel through the steam generating vessel 300 and reactor vessel 200 is shown schematically in FIG. 12 and functions as follows. The pressurized reactor primary coolant (at about 2250 psig) is heated by the fission in the core 210, which reduces the coolant density causing it to rise within the tubular reactor shroud 220. The heated reactor coolant (@ about 600 deg. F.) exits the reactor vessel and enters the steam generator 301. Once inside the steam generator, the primary coolant or water flows upward and is delivered to the top distribution plenum 391 by a centrally located tubular riser shell or pipe 337. The primary coolant fluid reverses direction and descends the steam generator flowing inside the bank of tubes (tube side), progressively transmitting its heat energy to the secondary water that flows vertically upwards on the shell side of the steam generator in a countercurrent arrangement to the primary flow stream. The cooled (and thus densified primary stream) re-enters the reactor vessel and flows downward in the annular space between the shroud and reactor vessel walls, reaching the bottom of the fuel core. The primary fluid reverses direction and flows upward resuming its circulation in the primary coolant circuit or loop of the nuclear steam supply system as it is heated by nuclear fission occurring in the reactor. The rate of flow of the primary fluid is set by the balance in the hydrostatic head between its hot up-flowing segment and the cool down-flowing segment (both of which span the reactor vessel and the steam generator) and the hydrodynamic pressure loss governed by what engineers know as the classic Bernoulli's equation. The closed primary coolant flow loop is therefore configured to produce and maintain natural gravity-driven circulation of primary coolant without the assistance of pumps. As the above summary indicates, the design objectives of the steam generator to fulfill the needs of primary side (i.e. primary coolant side) in one embodiment include: Provide for the flow of the primary fluid in it with as little friction loss as possible. Reduced pressure loss will increase the rate of circulation and improve the in-tube heat transfer coefficient, which are both salutary effects. Minimize the lateral distance between the reactor vessel and the steam generator so that the joint between them does not require a pipe (failure of such piping connections from events such as earthquakes, thermal fatigue, and other adverse mechanisms is a matter of safety concern in nuclear plants, which drives the decision to eliminate any large piping in the present SMR). The secondary side (i.e. secondary coolant side) of the steam generator also has several needs that should be addressed to avoid reliability problems that have afflicted prior designs. The design objectives of the secondary side include: In recirculating steam generators, the boiling of water in the shell side space has led to the accumulation of aggressive species on the tubesheet surface and in the crevices where the tubes emerge from the tubesheet, which over time, attack the tube walls causing leakage. The accumulation of solids in the tube support plate crevices has resulted in choking of the tubes in recirculating steam generators. In recirculating, the steam leaving the steam generator must be stripped of its entrained moisture by mechanical means. The effectiveness of the moisture separator (installed above the tube bundle of the steam generator) has been a source of operation problems in PWRs. The once-through steam generators suffered from the demerit of having too little water inventory in the shell side space. In case of an interruption in the in-flow from the feedwater (secondary coolant) line, the steam generator would tend to go dry in a very short time, making it a safety risk. This risk actually materialized at Three Mile Island nuclear generating plant in April 1979. The above deficiencies in the present day steam generator designs have guided the development of a new design disclosed herein. While the novel design features of the new present embodiment of a steam generator can be used in any steam producing boiler (i.e. non-nuclear), its anatomy and features are explained in the context of a nuclear SMR. Referring initially to FIGS. 1-6, a steam supply system 100 for a nuclear pressurized water reactor (PWR) according to the present disclosure is shown. From the thermal-hydraulic standpoint, the system includes reactor vessel 200 and steam generator 301 fluidly coupled thereto. The steam generating vessel 300 and reactor vessel 200 are vertically elongated and separate components which hydraulically and physically are closely coupled, but discrete vessels in themselves that are thermally isolated except for the exchange of primary loop coolant (i.e. reactor coolant) flowing between the vessels. The steam generating vessel 300 is laterally adjacent to, but vertically offset from the reactor vessel 200 to provide access to the reactor vessel internal components and control rods. As further described herein, the steam generating vessel 300 in one embodiment includes three heat transfer zones comprising (from bottom to top) a preheater section 320, main steam generator section 330, and a superheater section 350 which converts a fluid such as water flowing in a secondary coolant loop from a liquid entering the steam generating vessel 300 at feedwater inlet 301 to superheated steam leaving the steam generating vessel at steam outlet 302. The secondary coolant loop water may be a Rankine cycle fluid used to drive a turbine-generator set for producing electric power in some embodiments. Other uses of the steam are possible. The assemblage of the foregoing three heat exchangers may be referred to as a “stack.” Pressurizer 380 maintains a predetermined pressure of the primary coolant fluid. The pressurizer is an all-welded pressure vessel mounted atop the steam generating vessel 300 and engineered to maintain a liquid/gas interface (i.e. primary coolant water/inert gas) that operates to enable control of the primary coolant pressure in the steam generator. Pressurizer 380 may be mounted directly on top of the steam generating vessel 300 above the top tubesheet 333a and hydraulically seals the vessel at the top end. The pressurizer 380 is in fluid communication with the top tubesheet 333a and primary coolant pooling above the tubesheet in the top distribution plenum 391. The top head of the pressurizer 390 may have a hemispherical or an ellipsoidal profile in cross section. The pressurizer 380 is an all-welded pressure vessel with an open flange at its bottom, a curved “false bottom” plate, and a combination of conical and cylindrical shell courses and a top head in the form of a surface of revolution. In one embodiment, as shown, pressurizer 380 has an open flange at its bottom and is removably mounted to the steam generating vessel 300 via a bolted and flanged connection 390 to provide access to the top tubesheet 333a for maintenance, inspection, and/or repair of the tubes (e.g. plugging tubes, checking tubesheet-to-tube joints, etc.). The pressurizer 380 in some embodiments includes a convexly curved false bottom plate 412 that separates the turbulated primary flow underneath it in the top distribution plenum 391 from the water mass in pressurizer space keeping the latter relatively quiescent (see, e.g. FIG. 2). Suitably located small holes or perforations in the false bottom plate 412 provide for fluid communication between the water inventories in the two spaces. Referring to FIG. 11, reactor vessel 200 and steam generating vessel 300 may be housed in a containment vessel 110. Containment vessel 110 may be formed of a suitable shop-fabricated steel comprised of a top 111, bottom 112, and cylindrical sidewall 113 extending therebetween. In some embodiments, portions of the containment vessel which may be located above ground level may be made of ductile ribbed steel to help withstand aircraft impact. A missile shield 117 which is spaced above the top 111 of the containment vessel 110 may be provided as part of the containment vessel or a separate containment enclosure structure (not shown) which encloses the containment vessel 110. A horizontal partition wall 114 divides the containment vessel into an upper portion 114a and a lower portion 114b. Partition wall 114 defines a floor in the containment vessel. In one embodiment, a majority of reactor vessel 200 may be disposed in lower portion 114b and steam generating vessel 300 may be disposed in upper portion 114a as shown. Partition wall 114 may be formed of any suitable material, including without limitation for example concrete or metal. In various embodiments, the containment vessel 110 may be mounted above ground, partially below ground, or completely below ground. In certain embodiments, the containment vessel 110 may be positioned so that at least part or all of lower portion 114b that contains the nuclear fuel reactor core (i.e. fuel cartridge 230) is located below ground level. In one embodiment, the entire reactor vessel 200 and a portion of the steam generating vessel 300 are located entirely below ground level for maximum security. The cylindrical shell or sidewall 113 of containment vessel 110 may be horizontally split into an upper section and lower section which are joined together by a circumferential welded or bolted flanged joint 119 as shown in FIG. 11 to provide a demarcation for portions of the containment vessel which are located essentially above and below ground level. In other embodiments, the upper and lower sections may be welded together without use of a flange. In one embodiment, for example without limitation, the containment vessel 110 may have a representative height of approximately 200 feet or more for a 160 MW (megawatt) modular nuclear electric generation facility. A non-limiting representative diameter for this power generation facility is about 45 feet. Any suitable height and diameter for the containment vessel may be provided depending on system component configuration and dimensions. Containment vessel 110 further includes a wet reactor well 115 defined in one embodiment by a cylindrical circumscribing walled enclosure 116 which is flooded with water to provide enhanced radiation shielding and a back-up reserve of readily accessible coolant for the reactor core. In one embodiment, the walled enclosure 116 may be formed of stainless steel cylindrical walls which extend circumferentially around the reactor vessel 200 as shown. Other suitable materials may be used to construct enclosure 116. The wet reactor well 115 is disposed in the lower portion 114b of the containment vessel 110. Lower portion 114b may further include a flooded (i.e. water) used fuel pool 118 adjacent to the enclosure 116. In one embodiment, both the used fuel pool 118 and walled enclosure 116 are disposed below horizontal partition wall 114 as shown in FIG. 1. Both the reactor vessel 200 and steam generating vessel 300 preferably may be vertically oriented as shown to reduce the footprint and diameter of the containment vessel 110. The containment vessel 110 has a diameter large enough to house both the reactor vessel, steam generating vessel, and any other appurtenances. The containment vessel 110 preferably has a height large enough to completely house the reactor vessel and steam generating vessel to provide a fully contained steam generator with exception of the water and steam inlet and outlet penetrations for second coolant loop fluid flow associated with the Rankine cycle for driving the turbine-generator set for producing electric power. FIG. 12 shows the circulation path of primary coolant (e.g. water) in the primary coolant flow loop (see directional arrows). In one embodiment, the primary coolant flow is gravity-driven relying on the change in temperature and corresponding density of the coolant as it is heated in the reactor vessel 200, and then cooled in the steam generating vessel 300 as heat is transferred to the secondary coolant loop of the Rankine cycle which drives the turbine-generator (T-G) set. The pressure head created by the changing different densities of the coolant (i.e. hot—lower density and cold—higher density) induces flow or circulation through the reactor vessel-steam generating vessel system as shown by the directional flow arrows. Advantageously, the gravity-driven primary coolant circulation requires no coolant pumps or machinery thereby resulting in cost (capital, operating, and maintenance) savings, reduced system power consumption thereby increasing energy conversion efficiency of the PWR system, in addition to other advantages as described herein. Reactor Vessel Reactor vessel 200 may be similar to the reactor vessel with gravity-driven circulation system disclosed in commonly-owned U.S. patent application Ser. No. 13/577,163 filed Aug. 3, 2012, which is incorporated herein by reference in its entirety. Referring to FIGS. 1-3, reactor vessel 200 in one non-limiting embodiment is an ASME code Section III, Class I thick-walled cylindrical pressure vessel comprised of a cylindrical sidewall shell 201 with an integrally welded hemispherical bottom head 203 and a removable hemispherical top head 202. Shell 201 defines an internal cavity 208 configured for holding the reactor core, reactor shroud, and other appurtenances as described herein. In one embodiment, the upper extremity of the reactor vessel shell 201 may be equipped with a slightly tapered hub flange 204 (also known as “welding neck” flange in the art) which is bolted to a similar mating flange 205 welded to the top head 202. The bolted connection of the top head 202 provides ready access to the reactor vessel internals such as the reactor core. Two concentric self-energizing gaskets 206 compressed between the two mating flanges 204, 205 provide leak tightness of the reactor vessel 200 at the connection between the top head 202 and shell 201. The leak tightness under operating conditions is assured by an axisymmetric heating of the flanged joint that is provided by the fluid flow arrangement of the primary coolant in the system, as further described herein. The top head 202 may contain vertical penetrations 207 for insertion of the control rods and further may serve as a base for mounting the associated control rod drives, both of which are not depicted but well known in the art without further elaboration. With continuing reference to FIGS. 1-3, the reactor vessel 200 includes a tubular cylindrical reactor shroud 220 which contains the reactor core defined by fuel cartridge 230. Reactor shroud 220 transversely divides the shell portion of the reactor vessel into two concentrically arranged spaces: (1) an outer annulus 221 defining an annular downcomer 222 for primary coolant entering the reactor vessel which is formed between the outer surface of the reactor shroud and the inner surface of the shell 201; and (2) an inner passageway 223 defining a riser column 224 for the primary coolant leaving the reactor vessel heated by fission in the reactor core. The reactor shroud 220 is elongated and extends in an axial direction along vertical axis VA1 of the reactor vessel which defines a height and includes an open bottom 225 and top 226. In one embodiment, the bottom 225 of reactor shroud 220 is vertically spaced apart by a distance from the bottom head 203 of reactor vessel 200 and defines a bottom flow plenum 228. Bottom flow plenum 228 collects primary coolant from annular downcomer 222 and directs the coolant flow into the inlet of the riser column 224 formed by the open bottom 225 of reactor shroud 220 (see, e.g. FIG. 2). On the opposite top end, the top hub flange 204 of reactor vessel 200 ensures that the hot primary coolant water exiting the reactor vessel through outlet nozzle 271 cannot flow back into the downcomer 222 and mix with the incoming cooled primary coolant from the steam generator 301. Both the fuel cartridge 230 and reactor shroud 220 are supported by a core support structure (“CSS”), which in one embodiment includes a plurality of lateral support members 250 that span between and are attached to the reactor shroud and the shell 201 of the reactor vessel 200. Two support members 250 are shown in FIG. 10 for brevity. A suitable number of supports members spaced both circumferentially and vertically apart are provided as needed to support the combined weight of the fuel cartridge 230 and reactor shroud 220. In one embodiment, the bottom of the reactor shroud 220 is not directly attached to the reactor vessel 200 to allow the shroud to grow thermally in a vertical axial direction (i.e. parallel to vertical axis VA1) without undue constraint. A plurality of circumferentially spaced apart flow baffles 251 may be attached to the bottom of shroud 220 which further support the dead weight of the shroud. Baffles 251 are vertically oriented and have a shape configured to complement the curvature of the hemispherical bottom head 203 of the reactor vessel 200 as shown (see, e.g. FIG. 10) on which the baffles are seated. A plurality of lateral perforations 252 may be provided in the baffles 251 to aid in mixing the cooled primary coolant flow descending in the downcomer 222 before rising to enter the fuel cartridge 230. The reactor shroud 220 may be a single-walled open cylinder in one embodiment as shown. In an alternative construction, shroud 220 may be a double-walled cylinder comprising two radially spaced apart single shells with a sealed air gap or insulating material therebetween. This double-wall construction of reactor shroud 220 forms an insulated structure designed to retard the flow of heat across it and forms a smooth vertical riser column 224 for upward flow of the primary coolant (i.e. water) heated by the fission in the fuel cartridge 230 (“core”), which is preferably located at the bottom extremity and inside passageway 224 of the shroud in one embodiment as shown in FIGS. 1-3. The reactor shroud 220 is laterally supported by the reactor vessel by the lateral support members 250 to prevent damage from mechanical flow-induced vibrations resulting in metal fatigue over a period of time. Shroud 220 and other wetted parts of reactor vessel 200 may be made of a corrosion resistant material, such as without limitation stainless steel. Other materials and/or corrosion resistant coatings may be used. Referring to FIGS. 2 and 10, fuel cartridge 230 in one embodiment is a unitary autonomous structure containing upright fuel assemblies, and is situated in a region of the reactor vessel 200 that is spaced above bottom head 203 so that a relatively deep plenum of water lies underneath the fuel cartridge. Fuel cartridge 230 may be located inside reactor shroud 230 at the bottom end of the shroud as shown. The fuel cartridge 230 is insulated by reactor shroud 220 so that a majority of the heat generated by the fission reaction in the nuclear fuel core is used in heating the primary coolant flowing through the fuel cartridge and adjoining upper portions of the riser column 224. Fuel cartridge 230 is an open cylindrical structure including cylindrically shaped sidewalls 231, open top 233, and open bottom 234 to allow the primary coolant to flow upward completely through the cartridge (see directional flow arrows). In one embodiment, the sidewalls 231 may be formed by multiple arcuate segments of reflectors which are joined together by suitable means. The open interior of the fuel cartridge 230 is filled with a support grid 232 for holding the nuclear fuel rods and for insertion of control rods into the core to control the fission reaction as needed. Briefly, in operation, the hot reactor primary coolant exits the reactor vessel 200 through a low flow resistance outlet nozzle 270 to be cooled in the adjacent steam generating vessel 300 (see, e.g. FIGS. 2, 3, and 12). The cooled reactor primary coolant leaves the steam generating vessel 300 and enters the reactor vessel 200 through the inlet nozzle 271. The internal plumbing and arrangement in the reactor vessel directs the cooled reactor coolant down through to the annular downcomer 222. The height of the reactor vessel 200 is preferably selected to support an adequate level of turbulence in the recirculating reactor primary coolant by virtue of the density differences in the hot and cold water columns which is commonly known as the thermo-siphon action (density difference driven flow) actuated by gravity. In one embodiment, the circulation of the reactor primary coolant is driven by over 8 psi pressure generated by the thermo-siphon action, which has been determined to ensure (with adequate margin) a thoroughly turbulent flow and stable hydraulic performance. Referring to FIGS. 2 and 4, the top of the reactor vessel shell 201 is welded to a massive upper support forging which may be referred to as a reactor support flange 280. Support flange 280 supports the weight of the reactor vessel 200 and internal components above the wet reactor well 115. In one embodiment, the support flange is structurally stiffened and reinforced by a plurality of lugs 281 which are spaced circumferentially apart around the reactor vessel and welded to both the reactor vessel and flange, as shown. Support flange 280 contacts and engages horizontal partition wall 114 which transfers the dead weight of the reactor vessel 200 to the containment vessel 110. The reactor vessel's radial and axial thermal expansion (i.e. a majority of growth being primarily downwards from horizontal partition wall 114) as the reactor heats up during operation is unconstrained. However, the portion of containment vessel 110 which projects above partition wall 114 is free to grow upwards in unison with the upwards growth of the steam generating vessel 30 to minimize axial differential expansion between the steam generating vessel and reactor vessel. Because the reactor vessel and steam generating vessel are configured and structured to thermally grow in height at substantially the same rate when heated, this arrangement helps minimize potential thermal expansions stress in the primary coolant fluid coupling 273 between the reactor vessel and steam generating vessel. The support flange 280 is spaced vertically downwards on reactor vessel shell 201 by a distance from top head 202 of reactor vessel 200 sufficient to allow a fluid connection to be made to the steam generating vessel 300 which is above partition wall 114, as shown in FIGS. 2 and 11. When the reactor vessel 200 is mounted inside containment vessel 110, top head 202 of the reactor vessel and the primary coolant fluid coupling 273 (collectively formed by combined inlet-outlet flow nozzle 270/271) are located above reactor well 115. This provides a location for connection to the steam generator plenums and for the engineered safety systems (e.g. control rods, etc.) to deal with various postulated accident scenarios. A majority of the reactor vessel shell 201, however, may be disposed below partition wall 114 and immersed in the wet reactor well 115 as shown in FIG. 1. The bottom region of the reactor vessel 200 is restrained by a lateral seismic restraint system which may be comprised of a plurality of circumferentially and vertically spaced apart lateral restraint members 260 (one of which is shown schematically in FIG. 11 for brevity). Restraint members 260 span the space between the reactor shell 201 and the reactor well 115 inside surface of the cylindrical enclosure 116 to withstand seismic events. The seismic restraint design is configured to allow for free axial (i.e. longitudinal along vertical axis VA1) and diametrical thermal expansion of the reactor vessel 200. The reactor well 115 is flooded during power operations to provide defense-in-depth against a (hypothetical, non-mechanistic) accident that is assumed to produce a rapid rise in the enthalpy of the reactor's contents. Because the reactor is designed to prevent loss of core water by leaks or breaks and the reactor well is flooded, burn-through of the reactor vessel by molten fuel (corium) is not likely. Referring to FIGS. 2-4, the reactor vessel combined inlet-outlet flow nozzle 270/271 (primary coolant fluid coupling 273) is comprised of two concentric flow conduits including an outer inlet nozzle 270 and inner outlet nozzle 271. The outlet nozzle 271 has one end welded to the reactor shroud 220 (internal to the reactor vessel shell 201) and an opposite end welded to the inlet nozzle 371 of the steam generating vessel 300 (at the bottom of riser pipe 337). The reactor vessel inlet nozzle 270 has one end welded to the reactor vessel shell 201 and an opposite end welded to steam generator outlet nozzle 370 defined at least in part by the bottom tubesheet 333b of the steam generating vessel 300. Accordingly, reactor vessel inlet nozzle 270 is essentially welded to the perimeter of bottom tubesheet 333b of the steam generator 301 (best shown in FIG. 4). It should be noted that the inlet nozzle 371 of the steam generating vessel 300 and riser pipe 337 are contiguous in structure. The inlet nozzle 371 is further contiguous with the outlet nozzle 271 of the reactor vessel. Accordingly, the riser pipe 337 may also be viewed from one perspective as physically extending and fluidly connected directly to the internal shroud 220 of the reactor vessel as a single flow conduit in lieu of three separate flow conduit sections. In one embodiment, therefore, the riser pipe 337 may have a constant diameter including portions which are considered to form the primary coolant inlet nozzle 371 and reactor vessel outlet nozzle 271. An annular bottom collection plenum 373 is formed between the inlet and outlet nozzles 270, 271 of primary coolant fluid coupling 273 below the bottom tubesheet 333b (see, e.g. FIG. 4). Collection plenum 373 serves to collect the cooled primary coolant exiting the bottom ends of the tubes 332 through the bottom tubesheet 333b which flows back to the reactor vessel 200 through inlet nozzle 270 into the annular downcomer 222. In the present embodiment, the outlet nozzle 271 of the reactor vessel and inlet nozzle 371 of the steam generating vessel each have a smaller diameter than the inlet nozzle 270 of the reactor vessel and outlet nozzle 370 of the steam generating vessel. The combined inlet-outlet flow nozzle 270/271 is located above partition wall 114 of the containment vessel 110. The inlet nozzle 371 and outlet nozzle 370 of the steam generating vessel 300 collectively define a mating concentrically arranged combined primary coolant inlet/outlet nozzle 371/370 for the steam generating vessel. In one embodiment, the inlet flow nozzle 270 and outlet flow nozzle 271 of the reactor vessel 200 are configured as 90 degree flow conduits or elbows as shown. This allows extremely close horizontal spacing of the reactor vessel and steam generator shells due to the closely coupled primary coolant combined inlet-outlet flow nozzle 270/271 to the steam generator, thereby eliminating a straight horizontal section of piping between the reactor vessel and steam generator. Advantageously, such close coupling of the reactor vessel 200 and steam generator vessel 300 avoids need for long loops of large piping in the reactor primary coolant which creates the potential for a “large break” Loss of Coolant Accident (LOCA) event. Close coupling of the reactor vessel 200 and steam generating vessel 300 are achieved by the minimal radial projection of the combined inlet-outlet flow nozzle 270/271 beyond the reactor vessel shell. The total horizontal length of the inlet/outlet nozzle connection between the reactor vessel 200 and steam generating vessel 300 in certain embodiment is less than or equal to the diameter of the reactor vessel 200, and/or the steam generating vessel 300 to eliminate long runs of large coolant piping between the reactor and steam generating vessels. Concomitantly, the vertical centerline VA2 of the steam generating vessel 300 may be less than two steam generator diameters apart horizontally from the vertical centerline VA1 of the reactor vessel 200 in some embodiments. To achieve the closest possible coupling of the reactor vessel 200 and steam generating vessel 300, the outer reactor vessel inlet nozzle 270 of the primary coolant fluid coupling 273 may be a mitered 90 degree elbow or bend comprising an eccentric cone section 274 joined to a short horizontal stub pipe section 275 using a miter joint 276 (best shown in FIG. 4). The miter joint 276 minimizes the lateral distance between the reactor vessel 200 and steam generating vessel 300. Miter joint 276 is disposed an angle between 0 and 90 degrees (e.g. about 30-60 degrees in some embodiments) with respect to the horizontal plane at the joint. The stub pipe section 275 is disposed at a 90 degree angle to the eccentric cone section 274. The outer reactor vessel inlet nozzle 270 therefore forms an asymmetrically-shaped outer flow jacket which surrounds the inner reactor vessel outlet nozzle 271 as shown. The eccentric cone section 274 has a circular cross section and inside diameter which varies (i.e. narrows) from its inlet end at steam generator outlet nozzle 370 adjacent bottom tubesheet 333b to the stub pipe section 275. Cone section 274 is formed by a straight inner sidewall 274b and an opposing inclined sidewall 274a which is angled with respect to the inner sidewall as shown in FIG. 4. The outlet end of the eccentric cone section at the miter joint 276 has a circular cross section as does the inlet to the stub pipe section 275 which is coupled to the reactor vessel wall 201 (e.g. welded). Steam Generator The steam generator 301 will now be described in further detail. Referring to FIGS. 1-9, the steam generating vessel 300 in one embodiment may a vertically oriented and elongated structure which defines a vertical axis VA2. In one embodiment, the vertical axis VA2 of the steam generating vessel is horizontally offset from the vertical axis VA2 of the reactor vessel 200 so that the steam generating vessel is arranged laterally adjacent to the reactor vessel. In one embodiment, the steam generating vessel 300 has a height which is at least as high as the height of the reactor vessel 200 to achieve the thermo-hydraulic conditions necessary to induce gravity-driven primary coolant circulation through the steam generating vessel 300 and reactor vessel 200. Structurally, steam generating vessel 300 includes a top 310, bottom 311, and a vertically extending hollow cylindrical shell 312 extending therebetween which defines an internal cavity 393 for holding a plurality of heat exchange components. Steam generating vessel 300 further includes a top tubesheet 333a, bottom tubesheet 333b, a plurality of heat transfer tubes 332 extending vertically between the tubesheets, an internal riser pipe 337, and pressurizer 380 disposed on the top 310 of the vessel. The top and bottom tubesheets 333a, 333b are circular in top plan view and of suitable thickness to withstand the operating pressure within the steam generating vessel 300 without undue deformation which could adversely affect the integrity of the joints between the tubes 332 and tubesheets. In one embodiment, the bottom tubesheet 333b may have a convexly rounded top so that any debris accumulating within the steam generating vessel 300 settles to the outside perimeter of the tubesheet inside the shell 312. The tubesheets are preferably formed a thick corrosion resistant steel such as stainless steel in one embodiment. In one embodiment, riser pipe 337 is concentrically aligned with shell 312 and lies on the vertical axis VA2 of the vessel. The tubes 332 are circumferentially arranged around the outside of the riser pipe 337 in any suitable pattern between the riser pipe and shell 312 of steam generating vessel 300. In one embodiment, the tubes 332 of the steam generating vessel 300 may define three heat transfer zones arranged vertically for converting secondary coolant feedwater entering the bottom of the vessel from a liquid phase to a steam phase exiting the top of the vessel. In one embodiment, the steam phase is superheated steam. The three heat transfer zones may include (from bottom up) a preheater section 320 for initial heating of the liquid secondary coolant, main steam generator section 330 which serves as the boiler for heating the secondary coolant to the boiling point temperature where it changes phase to steam, and superheater section 350 for healing the steam to superheated conditions. In certain arrangements and configurations of the steam generator 300, the preheater 320 may be omitted depending on the thermo-hydraulic design of the system. The preheater section 320, steam generator section 330, and superheater section 350 are tubular heat exchangers each including a plurality of parallel straight tubes 332 (i.e. tube bundles) with the top and bottom tubesheets 333a, 333b disposed at the uppermost and lowermost extremities or ends of each tube 332. In one embodiment, the tube bundles are contiguous in structure from top to bottom so that there are no intermediate structures formed between the three different heat transfer sections on the tubeside. Primary coolant therefore flows downwards through each of the tubes 332 which have a continuous structure and height from the top tubesheet 333a to bottom tubesheet 333b. The preheater section 320, steam generator section 330, and superheater section 350 therefore are defined by the phase of the secondary coolant within the three different heat transfer zones as the feedwater changes phase from a liquid state entering the steam generating vessel 300 at the bottom to steam exiting from the top of the vessel. The internal cavity 393 of the steam generating vessel 300 may be contiguous and open between the tubesheets 333a and 333b on the shell side of the steam generating vessel 300 without any intermediate structures which may interrupt the upward flow of secondary coolant. The preheater 320, steam generator 330, and superheater 350 are configured to form a parallel counter-flow type heat exchanger arrangement in which the secondary coolant (Rankine cycle) flows in an opposite, but parallel direction to the reactor primary coolant (see, e.g. FIGS. 4, 5, and 9B). In a certain embodiment, the preheater section 320 may be configured to provide a combination of parallel counter-flow and cross-flow of the secondary coolant with respect to the primary coolant flow via the provision of flow baffles 394 on the shell side of the steam generating vessel 300. Referring to FIGS. 4, 7B, and 9B, two different configurations and sizes of baffle plates may be provided comprising a circular outer baffle 394a attached to steam generator shell 312 and having a central opening 394c, and a circular inner baffle 394b attached to riser pipe 337 and having a central opening 394d. Outer baffle 394a has a central opening 394c (i.e. circular) with a diameter larger than the diameter of the riser pipe 337 forming a lateral outside gap between the riser pipe and baffle, and an outside diameter slightly smaller than the inside diameter of the shell 312 for attachment thereto. Inner baffle 394b has a central opening 394d with a diameter slightly larger than the riser pipe 337 for attachment thereto, and an outside diameter smaller than the inside diameter of the steam generator shell 312 forming a lateral outside gap between the outer baffle. This arrangement advantageously causes the secondary coolant to flow through the preheater section 320 in the circuitous path shown (see directional flow arrows in FIG. 4) which maximizes contact time and heat transfer between the tubes 332 heated on the tube side by the primary coolant and the secondary coolant feedwater flowing on the shell side of the steam generating vessel 300. The inner and outer baffles 394b, 394a are arranged in an alternating pattern in the vertical direction to produce a combination of a perpendicular cross-flow pattern and parallel counter-flow pattern of the liquid secondary coolant with respect to the primary coolant through the preheater section 320 (see, e.g. directional arrows FIGS. 4 and 9B). The baffle plates 394a, 394c in the shell side space are therefore shaped to promote a combination of either radially symmetric cross flow or axially symmetric longitudinal flow of the shell side fluid. In certain embodiments, the steam generator section 330 and/or superheater section 350 may include baffles similar to baffles 394a and 394b shown in FIGS. 4, 7B, and 9B. The tube support system (baffles) in each zone is configured to promote radially symmetric flow. Radially symmetric flow fields are desired to prevent bowing or bending of the steam generator shell 312 from circumferential thermal gradients. Referring to FIGS. 9, 9B, and 9C, the interface between the preheater and the steam generator section 320, 330 zones in one embodiment may be demarcated by a relatively thick interface plate 410 which has a plurality of drilled and polished holes to form an extremely tight fit around the tubes (e.g. radial gap to the tube less than 1/64 inch). In other configurations, the interface plate may be omitted. Both the bottom tubesheet 333b and the interface plate 410 may have slightly convex top surfaces so that any contaminants or debris produced by boiling the secondary coolant that may tend to settle on them are swept to the outer periphery of the steam generating vessel 300 from which they can be evacuated through suitably sized “blow down” openings in the steam generator shell 312 (not shown) periodically. The foregoing tubular heat exchangers (i.e. preheater, steam generator, and superheater) are hydraulically connected in series on both the tube side (reactor primary coolant) and the shellside (the secondary coolant forming the working fluid of the Rankine Cycle which changes phase from liquid to superheated gas). The top 310 of the steam generating vessel 300 may be terminated with flanged connection 390 which couples the pressurizer 380 to the vessel (see, e.g. FIGS. 7A, 7B, and 8). The bottom tubesheet 333b forms the bottom 311 of steam generating vessel 300 and is directly connected to the steam generating vessel shell 312 (see, e.g. FIG. 4). Pressurizer 380 is mounted to top 310 of steam generating vessel 300 and is in fluid communication with both the top or outlet of riser pipe 337 and the inlet to superheater tubes 332. Pressurizer 380 which features a cylindrically-curved shell of revolution includes internal features to maintain a quiescent mass of water therein while ensuring a communicable relationship with the primary coolant water coursing through the top of the steam generating vessel 300 in the top distribution plenum 391 (see, e.g. FIG. 2). The pressurizer 380 has conventional electric heaters and spray nozzles to control primary coolant pressure. The pressurizer 380 may therefore generally include a heating/quenching element (i.e. water/steam) for pressure control of the reactor primary coolant. The element is comprised of a bank of electric heaters which are installed in the pressurizer section that serve to increase the pressure by boiling some of the primary coolant and creating a steam bubble that resides at the top of the pressurizer near the head (above the liquid/gas interface 392 of the primary coolant). A water spray column is located near the top head of the pressurizer which sprays water into the steam bubble thereby condensing the steam and reducing the size of the steam bubble. The increase/decrease in size of the steam bubble serves to increase/decrease the pressure of the primary coolant inside the reactor coolant system. In one exemplary embodiment, a representative primary coolant pressure maintained by the pressurizer 380 and heating/quenching element 381 may be without limitation about 2,250 psi. In alternative embodiments, a liquid/gas interface may be formed between an inert gas, such as nitrogen (N2) supplied by supply tanks connected to the pressurizer 380, and the liquid primary coolant. The pressurizer 380 defines a top distribution plenum 391 which collects reactor primary coolant rising through riser pipe 337 and distributes the primary coolant to the inlet of each of the tubes 332 penetrating the top tubesheet 333a. Plenum 391 resides above the top tubesheet 333a within the pressurizer forming a liquid reserve of primary coolant. Top tubesheet 333a may be recessed below the top 310 of steam generating vessel 300 (best shown in FIG. 8) to facilitate formation of the plenum. The depth of the plenum 391 may vary depending on the exact location of the liquid/gas interface 392; however, the depth of primary coolant in the plenum is preferably sufficient to cover the tubes 332 and tubesheet 333a and evenly distribute the primary coolant from the riser pipe 337 to the inlet ends of each of the tubes 332 penetrating the tubesheet. Referring to FIGS. 1, 4, and 7-9, steam generating vessel 300 includes a secondary coolant inlet nozzle 395 which is fluidly connected to steam generator shell 312 for introducing liquid secondary coolant feedwater into the bottom of the preheater section 320. In one embodiment, the inlet nozzle 395 may be attached to shell 312 at one of two radially projecting expansion joints 396a, 396b formed integrally with the shell in the preheater section 320 as best shown in FIG. 4. The expansion joints may have a box-like configuration in cross-section as shown and encircle the shell 312 of the steam generating vessel 300 for accommodating thermal growth in length/height of the steam generating vessel 300. The risk of high tube stresses due to differential expansion between the tubes 332 and the steam generator shell 312 advantageously is mitigated by the flanged and flued expansion joints 396a, 396b located near the top and bottom tubesheets 333a. 333b. Steam generating vessel 300 also includes a secondary coolant outlet nozzle 397 which is fluidly connected to steam generator shell 312 for withdrawing secondary coolant superheated steam from the superheater section 350. In one embodiment, the outlet nozzle 397 may be attached to shell 312 at the second radially projecting expansion joints 396b formed integrally with the shell in the superheater section 350 as best shown in FIG. 8. Although steam generator 301 includes straight heat transfer tubes 332, the steam generator vessel 300 may be configured to form a recirculating type steam generator. Referring to FIGS. 2 and 7-9, the steam generator section 330 in one embodiment of a steam generator 301 includes a tubular recirculation shroud 398 having a diameter smaller than the inside diameter of the steam generator vessel shell 312 forming an annular downcomer 399 between the shell and shroud for recirculating liquid secondary coolant. The bundle of heat transfer tubes 332 is disposed inside the shroud 398. The top 401 of the shroud is spaced below the water level W in the steam generator 301 forming the steam-liquid interface at the superheater section 350 of the tube bundle (see, e.g. FIG. 9A). Accordingly, the shroud 398 is wetted at all times during normal operation of the steam generator. The water level W may be maintained within a narrow range by a conventional level controller (not shown) such that the shroud 398 in the steam generator section 330 is submerged in water (primary coolant) at all times. The heat transfer surfaces and flow areas are sized such that the re-circulation ratio (ratio of the re-circulation flow rate to the steam generation rate) is approximately 5 in one non-limiting embodiment. On the opposite end, the bottom 402 of the recirculation shroud is disposed above and proximate to the top of the preheater section 320 of the tube bundle above the interface plate 410 (see, e.g. FIG. 9B). In operation, liquid secondary coolant flows upward on the shell side inside the shroud 398 towards the water lever W as it is heated by the tubes 332 (primary coolant flowing downwards therein on the tube side). The fluid rises as it becomes less dense from heating and boils producing steam. The reserve of secondary coolant not converted into steam cools further and flows radially outwards into the top of the annular downcomer 399 and flows downward towards the preheater section 320. The secondary coolant in the downcomer 399 then reverses direction and re-enters the bottom of the shroud mixing and flowing upwards again with the secondary coolant leaving the preheater section 320 to complete the recirculation flow loop. The steam generating vessel 200 may be supported by a gusseted cylindrical flanged support skirt 400. FIGS. 4 and 5 show the support skirt in greater detail. The support skirt 400 is attached to the bottom 311 of the steam generator vessel 300 in one arrangement. Support skirt 400 is structurally robust and may have a double-flanged arrangement comprising a radially projecting top bearing flange 405, radially projecting bottom base flange 404, and a circumferentially extending vertical wall 407 extending between the flanges. Wall 407 forms a circular enclosure (in transverse cross section) at least partially or fully surrounding the primary coolant fluid coupling 273 as shown. In various configurations, the support skirt 400 may be circumferentially continuous for 360 degrees or extend circumferentially less than 360 degrees. The bearing flange 405 and base flange 404 are diametrically enlarged with respect to the wall 407 thereby projecting beyond the wall. Base flange 404 is configured for seating on and attachment to divider wall 114 of the containment vessel 110 to transfer the dead weight of the steam generator 301 to the vessel (see also FIG. 11). Base flange 404 may be attached to divider wall 114 by any suitable means. In one embodiment, the base 404 may be attached with bolting such as a plurality of anchor bolts 408 spaced circumferentially apart. The base flange 404 and vertical wall 407 form an angled flanged arrangement. In one embodiment, the bottom tubesheet 333b includes a diametrically enlarged and radially projecting flange 406 which is configured and dimensioned to engage the top bearing flange 405 of the support skirt 400. Flange 406 is an integral unitary structural part of the tubesheet 333b. Accordingly, the bottom tubesheet 33b serves a dual function as a flow and support device. The flange 406 forms an annular stepped surface 409 around the perimeter of tubesheet 333b to positively engage the top bearing flange 405 and prevent lateral movement of the bottom of the steam generating vessel 300 during a seismic event. The bottom tubesheet flange 406 is therefore machined or formed to serve as the transmission path for the weight of the steam generator unit to the support foundation (e.g. divider wall 114) via the flanged support skirt 400. In other possible embodiments, the tubesheet flange 406 may be formed separately on the steam generating vessel 300 from the tubesheet 333b. The steam generator support skirt 400 further includes a plurality of vertically oriented stiffeners 403 extending between the bearing and base flanges 405, 404. The stiffeners 403 are circumferentially spaced apart and formed of structure plate which may be cut an angle as shown (see, e.g. FIGS. 4 and 5). The support skirt 400 including stiffeners 403, flanges 404, 405, and wall 407 are preferably made of structural steel plate of suitable thickness to bear the weight of a steam generator 301 containing secondary coolant during operating conditions. In one non-limiting embodiment, the steam generating vessel 300 and other components herein described exposed to moisture may be made of a corrosion resistant metal such as stainless steel and/or steel with a corrosion resistant liner or coating. Other types of metals may be used. The flow path of the reactor primary coolant and secondary coolant for the Rankine cycle will now be described. FIG. 12 shows the reactor primary coolant flowpath via directional flow arrows (i.e. primary coolant flow loop). FIGS. 1-4 and 6-9 show the secondary coolant flowpath of the Rankine cycle through steam generating vessel 300 via directional arrows. Primary coolant flows on the tube side of the steam generating vessel 300 and secondary coolant flows on the shell side. Cooled primary coolant (“cold”) leaves steam generating vessel 300 through outlet nozzle 370 and enters reactor vessel 200 through outer inlet nozzle 270. The primary coolant flows downwards through annular downcomer 222 enters the bottom of riser column 224. The primary coolant flows upwards through fuel cartridge 230 and is heated by convection and conduction in the fuel core. The now heated or “hot” primary coolant exits the reactor vessel 200 through outer inlet nozzle 270 and enters steam generating vessel 300 through inlet nozzle 371. The hot primary coolant flows vertically upwards in riser pipe 337 and is directed to the top of the “stack” into the top distribution plenum 391 formed by the pressurizer 380. The hot primary coolant enters the tubes 332 through penetrations in top tubesheet 333a and reverses direction to begin the downwards journey through steam generating vessel 200 in the tubes. The hot primary coolant first flows down through the superheater 350 on the tube side of the tube bundle which has wet saturated steam (secondary coolant) flowing upwards on the shell side from the steam generator 230 below in the stack. The saturated steam becomes superheated and is dried by the primary water inside the tubes, which is flowing in counter flow to the rising steam mass. The counter-flow arrangement permits the steam to be superheated to within a few degree Fahrenheit of the reactor coolant's peak temperature, resulting in maximized thermodynamic efficiency. The superheated steam then leaves the steam generating vessel 300 via outlet nozzle 397. Continuing the process, the now less hot coolant continues to flow down through the steam generating vessel 300 next proceeding through the steam generator 330 on the tube side. On the shell side, liquid secondary coolant undergoes a phase change and is turned to steam as the primary coolant is further cooled in giving up heat to the secondary coolant. The now further cooled primary coolant flows down through the preheater 320 on the tube side which encounters and preheats the cold (e.g. sub-cooled) liquid secondary coolant entering the shell side through the feedwater inlet nozzle 395 of the steam generator. The now cooled primary coolant has completed the closed flow loop through the steam generating vessel 300 and reactor vessel 200, and re-enters the reactor vessel through inlet nozzle 270 to repeat the foregoing flow process in the closed primary coolant flow loop. In one embodiment, an exemplary non-limiting reactor vessel “hot” outlet temperature may be in a range of about and including 575 to 600 degrees F. An exemplary non-limiting reactor vessel “cold” inlet temperature may be in a range of about and including 350 to 385 degrees F. An exemplary reactor vessel operating pressure may be about 2,250 psi (pounds per square inch) which is maintained by pressurizer 380. Other suitable flow temperatures and pressures may be used depending on the heat transfer requirements of the specific application and Rankine cycle side steam production operating parameters. In one embodiment, the reactor vessel primary coolant may be unborated demineralized water. In one exemplary embodiment, the shell 312 of steam generating vessel may be made of steel such as type 508 carbon steel. Tubesheets 333a, 333b may be made of the same steel with an Inconel cladding when the tubes 312 are made of Inconel. In other embodiments, these components may be formed of other suitable metal materials including stainless steel. Other features and aspects of the steam generator 301 may include the following: a. The tubes 332 and the riser shell or pipe 337 may be fastened to the two tubesheets 333a, 333b by conventional methods such as edge welding, butt welding, hydraulic expansion, roller expansion, or a combination thereof. In non-limiting preferred embodiments, the tubes 332 are fastened to the two tubesheets 333a, 333b by a high integrity joining process such as hydraulic expansion or explosion bonding. Roller expansion is not necessarily favored in all situations because it has an adverse effect on the service life of the tubes due to work hardening of the tube material in the rolled zone. b. Either or both the steam generating vessel shell 312 and the riser pipe 337 may incorporate one or more “flexible shell elements” to acquire axial flexibility. c. The tubes 332 and/or the riser pipe 337 may be installed in the tubesheets 333a, 333b such that they are in a prescribed state of pre-tension. d. The shell side inlet and outlet nozzles 301, 302 are located close to the bottom and top tubesheets 333a, 333b, preferably in the shell 312 course of the “flexible shell elements” or expansion joints 396a, 396b. e. A perforated impingement shell 411 is installed in each of the two expansion joints 396a, 396b wherein the inlet and outlet nozzles are situated to provide for an essentially radially symmetric entrance of feedwater secondary coolant and exit of heated steam from the steam generating vessel 300, respectively (see, e.g. FIG. 4). The steam generator vessel 300 and pressurizer 380 may be laterally restrained at the four locations in one embodiment including proximate to the bottom tubesheet 333b, tope tubesheet 333a, near the mid-elevation of the steam generator shell 312, and the top of the pressurizer by lateral supports 420 (see, e.g. FIG. 6). In one embodiment, the support skirt 400 may provide the lateral restraint near the bottom tubesheet 333b. The lateral restraints 420 may be lined with an insulating material at their interface with the steam generating vessel shell 312 so as to prevent excessive heating of the structural material in the body of the restraints. The lateral restraints 420 may be equipped with a spring/damper material to reliably distribute the load on each during a seismic or mechanical loading event. The lateral supports 420 at mid-height of the steam generating vessel 300 and at the top tubesheet 333a location adjacent the flanged joint 390 shown advantageously help increase the beam mode frequency of the steam generator 301 in the rigid range. The lateral restraints further do not interfere with the axial vertical movement of the steam generator 301 along vertical axis VA2 due to thermal expansion. While the invention has been described and illustrated in sufficient detail that those skilled in this art can readily make and use it, various alternatives, modifications, and improvements should become readily apparent without departing from the spirit and scope of the invention. |
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052415694 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT FIG. 1 illustrates a schematic preferred embodiment of the present invention in which the sample is positioned behind a vacuum wall 8 surrounding the charged particle optics 10 for electron extraction and imaging of secondary electrons arising when energetic .beta.-particles pass out of the sample. The charged particle optics include a dual microchannel plate electron multiplier 12 from which the electron cascade 14 passes to the resistive anode encoder and detector 16 from which the X,Y position of the secondary electrons emitted from the sample is determined. The imaging process depends on accelerating the secondary electrons in an electric field, focusing them at a crossover point, and projecting the image onto the image intensifier 12 and detector 16. A germanium .gamma.-ray detector 18 is positioned below the sample S. The resistive anode encoder 16 is connected through the vacuum wall 8 to a position computer 22 and the dual microchannel plate electron multiplier 12 is connected through the vacuum wall 8 to a power supply 24. The .gamma.-ray detector 18 is connected to a pulse height and computer based multi-channel analyzer 26, and both the analyzer 26 and position computer 22 are connected to a coincidence detector 28 for determining which nuclide disintegrations give both position and energy information. The data from the position computer 22 and the analyzer 26 are connected to data storage sections 22A and 26A of computer memory buffers 30 of a computer 31 (not shown in detail). The coincidence detector 26 is also connected to the computer memory buffers whereby the computer system produces a coincidence .beta.-electron image buffer 22B and coincidence .gamma.-spectrum buffer 26B for providing the combined information to a coincidence file 32 which is stored on a hard disk 34. A data logger is provided to associate the energy and position data with a time stamp in the computer records, and computer software is provided to control and monitor the detection and to organize the data into elemental images and local area .gamma.-ray spectra. Referring now to FIGS. 2 and 3 there are shown the structures for the charged particle optics and secondary electron extraction and imaging optics 10 with its associated vacuum wall 8 and the germanium detector 18. The main vacuum chamber 8 consists of a stainless steel tube 40 with the sample holder 42 mounted on a vacuum flange 43. The extraction optics 44 are supported at one end of the tube 40 and the combination dual microchannel plate electron multiplier array and resistive anode encoder detector 46 mounted in a vacuum housing 48 on the other end. The detector housing 48 has a port connecting to a bleed valve 50 for venting the system and an ion gauge for monitoring the vacuum. A tee 54 and an elbow 56 connect a turbomolecular pump 58 to the main vacuum chamber tube 40. The vacuum system which typically operates at about 4.times.10.sup.-7 torr includes a second tee 60 having high voltage feed through 62 and 64 for the sample and focusing electrode. The sample holder 42 consists of an insulating sample support and vacuum sealing plate 66 mounted on the vacuum flange 43 with a sealing o-ring 68. The plate 66 includes a well 70 to provide a distance of approximately 3.7 mm from the sample to the extraction lens 44. The bottom of well 70 is covered with a disk 72 such as of stainless steel to which is applied a suitable high voltage via the feedthrough 64. A sample stage, not shown, moveable in all three directions can impart desired movement between the sample and the extraction optics. The extraction optics include an extraction lens opening 74 behind which are successively mounted a focus electrode 76 and a second electrode 78 spaced apart by insulators 80 and 82. The second electrode 78 is connected to a field free flight path tube 84. Operation of the focusing electrode 76 in an accelerating mode confers lower chromatic aberration on the system. However, higher voltages eventually lead to arcing among components in the vacuum system. In the accelerating mode the focusing electrode 76 requires about 2.5 times the acceleration voltage. In an operative prototype embodiment of the invention, the focusing lens 76 was operated in deceleration mode at about minus 2.85 KV with the sample at minus 4.0 KV and the field free region of tube 85 at ground potential. In the detector 46, the two plates of the dual microchannel plate electron multiplier arrays can have gains of approximately 1000 each, and when the electron cascade from the electron multiplier strikes the resistive anode encoder, the relative currents flowing from the four corner electrodes of the encoder can provide the position of the electron cascade and thus the position of the original electrons emitted from the sample. The calculated position is based on the centroid of the electron cascade. The cascade current is variable and the resistive anode encoder electronics have both lower and upper thresholds for the electron multiplier cascade current. Several electrons arriving simultaneously (within about 100 ns.) at the detector surface produce a larger cascade roughly in proportion to the number of impinging secondary electrons. However, a central position is calculated. The position is the average position (centroid) of the impinging secondary electrons weighted according to the magnitude of their individual multiplier cascades. Since most nuclide decompositions produce more than one secondary electron, the resistive anode encoder lower threshold can be set relatively high to eliminate stray electrons, cosmic rays and other background signals. Many factors will contribute to the lateral resolution of secondary electron images. Resolution of the detector. The analog electronic circuits used for the position calculations can provide resolution of about one part in 400. However, digitization reduces this to one part in 256. Thus, at a magnification of 30, the smallest resolvable image feature is about 6 .mu.m (two pixels). Smaller features can still generate measurable signals, but without information regarding feature size. PA1 Direction of .beta.-emission. A .beta.-electron travelling with a velocity component parallel to the sample surface would likely produce secondary electrons laterally displaced from the nuclide disintegration. The worst case would be a .beta.-electron travelling exactly parallel to the sample surface. Secondary electrons (relatively few) that eventually work their way to the surface have a centroid displaced from the original disintegration by half of the .beta.-particle stopping distance. Thin samples minimize this effect. Relatively high values of the RAE lower discriminator level also help reduce the effect. PA1 Energy spread of secondary electrons (chromatic aberration). Those .beta.-particles passing out of the sample with large energies usually strike an instrument surface and are stopped before reaching the detector. However, off-axis components of secondary electron energy lead to loss of image resolution. The propensity to form several secondary electrons at the sample surface also tends to minimize chromatic aberration because the detector provides a centroid of the positions of the several secondary electrons. Relatively high detector lower discriminator levels favor detection of those events producing multiple secondary electrons. PA1 Counting statistics. In high contrast images, features often emerge when only a few thousand counts are spread over relatively few pixels. However, a low contrast image requires a larger number of counts spread over many pixels. For example, if two regions within an image vary by 10% in signal intensity for a specified .gamma.-ray energy, signal-to-noise ratios of 10 are required to distinguish the differences. This necessitates about 100 recorded disintegrations for each pixel. If the low contrast features extend over the entire image (about 50,000 pixels), a total of about 5.times.10.sup.6 disintegrations must be recorded for each specified .gamma.-ray energy. This could take a long time. Several types of .gamma.-ray detectors can be used. However, the best available .gamma.-ray detectors consist of high purity n-type germanium. The outer contact is made by ion implantation of lithium and the inner contact by way of diffused lithium. An EG&G-Ortec model 8011-10185-5 GE (Li) detector (59 cm.sup.3 germanium volume), a Nuclear Data Model 475 amplifier, and a Nuclear Data model 582 analog-to-digital converter can be used for the detector 18 and analyzer 26. The model 581 analog-to-digital converter is designed for high resolution processing of the amplitude modulated signals typical of solid state radiation detectors. It provides 8 to 14-bit digitizations at a rate of 1.6.times.10.sup.5 s.sup.-1. The conversion time is fixed at 5 .mu.s, regardless of the input amplitude or the conversion accuracy. The constant conversion time allows the data ready signal from the analog-to-digital converter to be used in the coincidence circuitry. The electronic coincidence detection circuit 28 accepts x- and y- position data from the position detectors analog-to-digital converters in position computer 22, an .gamma.-ray energy data from the germanium detectors analog-to-digital converter. In the operative embodiment the output of the coincidence detector 28 is 32 bits of data. Since 16 bits of position information and 13 bits (8192 channels) of energy information are used, a 32 bit long word can accommodate three indicator bits. One indicator bit each is dedicated to indicating valid position, energy, and coincidence data. The position information arrives at the coincidence detector 28 first because the position detector electronics have a shorter dead time than the .gamma.-detector. The position detector dead time is constant (3.0 .mu.s) and is fixed on the position computer by an electronic timer (a one-shot). When position data arrive at the coincidence detector 28, the valid position bit is set and a delay timer is started. The end of the delay period starts the coincidence window timer. If the .gamma.-ray energy data arrives from the analog-to-digital converter in the coincident window, then the valid coincident bit is set. If the .gamma.-ray energy data arrive before or during the delay period, the valid energy bit is set. It is possible to have valid position and energy data in the same long word without having a valid coincidence. When the timers have timed out, the three indicator bits are added to the data lines, a strobe (handshake) pulse is sent to the computer parallel interface, and the energy and position data are passed through to the parallel interface. The delay and window times are adjustable and are set at 7.6 and 1.8 82 s, respectively in the operative embodiment. The total time between the nuclide decomposition and clocking the data into the computer is typically 11.5 .mu.s (7.6+3.0+1.8/2). Since the .gamma.-ray analog-to-digital converter has a fixed conversion time of 5 .mu.s, the germanium detector, preamplifier and amplifier combination must require about 6.5 .mu.s. This setting is variable over a range of about 1 .mu.s. Different energies require slightly different setting times. This can result in distortion of the .gamma.-ray spectrum if the coincidence delay and internal times are incorrectly set. The .beta.-particle position and the .gamma.-ray energy data come into the computer 31 via a parallel interface. In an operative embodiment, the computer 31 is a CompuAdd 386 SX with an 80387 SX arithmetic co-processor 2 Mbyte of random access memory, a 40 Mbyte hard disk drive, two floppy disk drives, a streaming tape mass storage unit, and a high resolution graphics interface. As 32-bit data words arrive at the computer parallel interface, an interrupt service routine (ISR) places the new data into a buffer. When the computer is not busy with the ISR, it can process the data out of the buffer in the first in first out order. If the computer goes to other tasks such as keyboard service routines or disk input-output, the ISR continues to store new data into the buffer. Since the buffer has 64 Kbytes of computer memory, the computer can be away at those other tasks for considerable time without missing any input data. The data is processed out of the input buffer 30 into five separate areas of computer memory. For the total spectrum buffer 26A the energy channel specified in each of the 32-bit input data word is incremented if that data word has a valid energy or coincidence bit. Energy channels use four bytes each. Thus, the total spectrum requires a data buffer of 32 Kbytes. This is the standard multi-channel analyzer function in which the resulting spectrum is independent of any coincidence condition. For the coincidence spectrum buffer 26B, the energy channel specified in each 32-bit input word is incremented if that data word has a valid coincidence bit. This spectrum also requires 32 Kbytes of computer memory. It is mainly useful for monitoring the operation of the coincidence detector 28 and computer interface. For the total electron image buffer 22A the image position specified in each 32-bit input data word is incremented if that data word has a valid position or coincidence bit. Images require a 128 Kbyte data buffer in computer memory. Like the total spectrum buffer 26A, the total image buffer 22A is independent of the coincidence condition. For coincidence image buffer 22B the energy channel specified in each 32-bit input data word is incremented if the data word has a valid coincidence bit. Before data acquisition, the operator is polled for the upper and lower boundaries of an energy window. Only those image positions associated with energies in the energy window are incremented. If the selected energy window contains a single peak of the .gamma.-spectrum, then the coincidence image constitutes a position map of the isotope responsible for the .gamma.-rays. During data acquisition, both the total image buffer 22A and the coincidence image buffer 22B are available in real time for monitoring the experiment. For the coincidence file buffer 32, whenever the valid coincidence bit is set, the energy and position data and a 16-bit time stamp are associated and logged into the computer memory buffer 32. This memory buffer occupies 256 Kbyte. When it is full, the contents are dumped to a hard disk file 34. The ISR continues to place new data into the input buffer, even during 15 disk operations. Thus, no data are lost under normal operation. As soon as data acquisition begins, the computer displays the total .gamma.-ray spectrum in real time. During data acquisition, the operator can examine any region of the total or coincidence .gamma.-ray spectrum with a linear or logarithmic display including any amount of y-axis offset. All of the spectral displays are continuously updated as new data are acquired. If the operator chooses image display, the system furnishes continuously updated total and coincidence images with either linear or logarithmic scale. All of these display features are available after the data acquisition is complete. In addition, there are options for labelling the displays and entering identification comments. The operator can store any or all of the images and spectra to a computer hard disk and can reaccess any stored spectra. The coincidence files are automatically stored if the operator selects this feature at the start of data acquisition. There are utilities for integration of spectral peaks and for printing spectra on a color printer. Finally, element maps and local area .gamma.-ray spectra can be prepared from stored coincidence file data. Thus, the operator need not specify or even know before beginning the analysis what coincidence images or spectra will be needed. Positions from those data records having energies within an operator selected window can be organized into an element map. Similarly, energies from those data records having positions within a operator selected area window can be organized into a local area spectrum. These two possibilities are implemented as menu choices in the imaging neutron activation analysis software. These two options bring the total number of available menu operations to 38. An example of operation of the present invention is illustrated in FIGS. 4, 5A and 5B. The sample from which these figures were derived consisted of particles of irradiated gold and nichrome dispersed on a piece of silicon wafer. The .gamma.-ray spectrum of the particle mixture is shown in FIG. 4. FIG. 4 shows the .gamma.-signals arising from radioactive chromium (.sup.51 Cr) gold (.sup.198 Au) and nickel (.sup.58 Co). Neutron irradiation of .sup.58 Ni produces .sup.58 Co, and when .sup.198 Au emits an electron, it becomes .sup.198 Hg. The distributions of gold and nickel on a portion of the particle sample are shown in FIGS. 5A and 5B, respectively. The area covered by the images is about 0.9 mm. There are various potential applications for the imaging neutron activation analysis of the present invention. The distribution of trace elements among mineral phases is a prime concern of geochemists. Better analysis techniques spur development of sophisticated models for partitioning of trace elements between minerals, magmas, and hydrothermal solutions. The models infer the evolution of igneous rocks and associated ore deposits from trace element signatures, especially rare earth signatures. Many interplanetary dust particles could be simultaneously counted with a single imaging detector using the present invention. Additionally, high purity ceramic materials find increasing use in many modern applications, perhaps most significantly in the electronics industry. For such materials imaging neutron activation analysis can be microanalytical, sensitive, accurate, unsusceptible to sample charging and independent of sample matrix. Materials containing radioactive nuclides for reasons other than neutron activation analysis will also benefit from the imaging .gamma.-ray detector of this invention. Natural and nuclear industry induced radionuclides are often heterogeneously distributed within a sample. For example, imaging secondary ion mass spectrometry suggests the presence of transuranic elements and the possibility of radioactive rare earth elements concentrated in various tissues in marine organisms living near French nuclear power facilities. Imaging with the present invention can provide much better sensitivity for the transuranic elements and radioactive fission fragments, free from the extensive interferences found in secondary ion mass spectrometry analysis of complex samples. Charged particle activation often complements neutron activation insensitivity and provides a unique set of experimental advantages. High energy ions of .sup.1 H, .sup.2 H, .sup.3 He, .sup.4 He, as well as heavier species can induce nuclear processes. For example, 5.5 MeV protons activate iron by the .sup.56 Fe (pn) .sup.56 Co reaction. The .sup.56 Co undergoes .beta..sup.+ (positron) decay with a half-life of 74 days and a characteristic 747 KeV .gamma.-ray. Ion beam activation occurs in the top several microns of a sample surface which is convenient because the secondary electrons can escape from about the implantation depth of the activating ion beam. The terms and expressions which have been employed here are used as terms of description and not a limitation, and there is no intention, in the use of such terms and expressions, of excluding equivalence of the features shown and described, or portions thereof, it being recognized that various modifications are possible within the scope of the invention claimed. |
claims | 1. A system for determining the radiological composition of material layers within a conduit, the system comprising:a collimator-probe comprising a probe contained within a collimator;a spectrometer operatively connected to the probe;a phantom setup comprising a vessel containing a reactor water test standard, a plurality of removable plates, a removable nuclear insulation layer, and a collimator probe attachment point; anda semi-logarithmic plot of spectrometer readings taken with various geometries of removable plates. 2. The system of claim 1, wherein the probe is a cadmium zinc telluride probe. 3. The system of claim 1, wherein the collimator comprises a generally cylindrical shield made from a high density material that surrounds the probe. 4. The system of claim 3, further comprising a grid attached to an opening in the collimator. 5. The system of claim 4, wherein the grid comprises a matrix of lead plates aligned to create a plurality of openings. 6. The system of claim 3, wherein the generally cylindrical shield comprises a first half and a second half. 7. The system of claim 3, wherein the high density material is selected from the group consisting of lead, tungsten, and depleted uranium. 8. The system of claim 1, further comprising a computer having a processor, memory and computer readable media, the computer configured to compare the semi-logarithmic plot of spectrometer readings stored on the computer readable media with spectrometer field readings stored on the computer readable media. 9. The system of claim 1, wherein the collimator is made from a high density material and covered with a layer of copper. 10. The system of claim 9, further comprising a layer of cadmium between the high density material and the layer of copper. 11. The system of claim 10, wherein the layer of cadmium is placed between the high density material and the layer of copper on the inner surface of the collimator. 12. The system for determining the radiological composition of material layers within a conduit of claim 1, wherein the collimator-probe comprises:a generally cylindrical shield comprising a high density material to contain the probe;the probe having a sensing face and placed within the generally cylindrical shield;an opening within a face of the generally cylindrical shield containing a grid of high density material to prevent low energy lateral photons from hitting the sensing face of the probe; anda removable cover comprising a high density material that is removably attached to the generally cylindrical shield to allow placement of the probe within the generally cylindrical shield. 13. The collimator-probe of claim 12, further comprising a spacer fitted to the shielded housing to keep the probe contained within the shielded housing a constant distance front the opening. 14. The collimator-probe of claim 12, further comprising a probe retainer placed between the probe and an interior wall of the generally cylindrical shield. 15. The system for determining the radiological composition of material layers within a conduit of claim 1, wherein the collimator-probe comprises:a generally cylindrical shield comprising a first half made from a high density material and a second half made from a high density material wherein the first half has protrusions that mate with channels of the second half;the probe having a sensing face and placed within the generally cylindrical shield; andan opening within a face of the generally cylindrical shield containing a grid of high density material to prevent low energy lateral photons from hitting the sensing face of the probe. 16. The collimator-probe of claim 15, further comprising a spacer fitted to the shielded housing to keep the probe contained within the shielded housing a constant distance from the opening. 17. A method of determining the radiological composition of a corrosion layer within a pipe using a system including a collimator-probe comprising a probe contained within a collimator; a spectrometer operatively connected to the probe; a phantom setup comprising a vessel containing a reactor water test standard, a plurality of removable plates, a removable nuclear insulation layer, and a collimator probe attachment point; and a semi-logarithmic plot of spectrometer readings taken with various geometries of removable plates, the method comprising the steps of:providing the wall thickness of a pipe being tested;taking readings of the pipe being tested with said collimator-probe;comparing the readings of the pipe being tested to said semi-logarithmic plot; andproviding sediment layer values. 18. The method of claim 17, further comprising the step of determining the detriment of corrosion layer. 19. The method of claim 18, further comprising the step of determining the source of the detriment. |
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046541837 | abstract | A process for selectively neutralizing H.sup.- ions in a magnetic field to produce an intense negative hydrogen ion beam with spin polarized protons. Characteristic features of the process include providing a multi-ampere beam of H.sup.- ions that are intersected by a beam of laser light. Photodetachment is effected in a uniform magnetic field that is provided around the beam of H.sup.- ions to spin polarize the H.sup.- ions and produce first and second populations or groups of ions, having their respective proton spin aligned either with the magnetic field or opposite to it. The intersecting beam of laser light is directed to selectively neutralize a majority of the ions in only one population, or given spin polarized group of H.sup.- ions, without neutralizing the ions in the other group thereby forming a population of H.sup.- ions each of which has its proton spin down, and a second group or population of H.sup.o atoms having proton spin up. Finally, the two groups of ions are separated from each other by magnetically bending the group of H.sup.- ions away from the group of neutralized ions, thereby to form an intense H.sup.- ion beam that is directed toward a predetermined objective. |
abstract | A bottom nozzle includes a skirt, support blocks, transverse blades and longitudinal blades. The skirt is a hollow structure and a bottom thereof is provided with corner legs which are protruded downwards, a cavity is defined in the hollow structure, the transverse blades are configured in the cavity, the longitudinal blades are configured in the cavity, the transverse blades and longitudinal blades are firmly connected with the skirt, projections of the transverse blades and the longitudinal blades in a level plane are intersectant to form interleaved grids, and the support blocks run through and are fixed on the transverse blades and the longitudinal blades. In such a way, the bottom nozzle forms a three-dimensional gridded water passage, thereby improving the filter capacity and generating small water pressure drop. |
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summary | ||
abstract | An imaging system (500) includes a focal spot (510) that rotates along a path around an examination region (506) and emits radiation. A collimator (512) collimates the radiation, producing a radiation beam (516) that traverses a field of view (520) of the examination region and a subject or object therein. A detector array (522), located opposite the radiation source, across the examination region, detects radiation traversing the field of view and produces a signal indicative of the detected radiation. A beam shaper (524), located between the radiation source and the collimator, rotates in coordination with the focal spot and defines an intensity profile of the radiation beam. The beam shaper includes a plurality of elongate x-ray absorbing elements (606) arranged parallel to each other along a transverse direction with respect to a direction of the beam, separated from each other by a plurality of material free regions (604). |
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044902875 | description | Referring now to FIG. 1 of the drawings there is shown a process vessel 1, composed of a heat-resistant material (e.g. stoneware clay, zircon or zirconia) surrounded with Vermiculate thermal insulation 2 and silica/alumina thermal insulating bricks 3 and enclosed within a microwave oven 4. The microwave oven 4 is provided with a wave guide 5 for the introduction of microwave radiation and an oven mode stirrer 6. An inlet pipe 7, fabricated from stainless steel and earthed to the oven 4 is provided to connect a feed pump 8 with the process vessel 1. The process vessel 1 is also provided with an outlet pipe 9 for the removal of gas/vapour therefrom, and a product outlet 10 for the passage of fused product to a collector 11 via aperture 12 in the thermal insulation 2 and 3. Microwave chokes 13 are provided at apertures in the microwave oven 4. Additionally there is provided a pressure relief device 14 in the process vessel 1 and a thermocouple temperature indicator 15. In operation solution or slurry to be treated to produce a dried product which is subsequently fused, is drawn from a supply (not shown) by pump 8 and delivered to the process vessel 1 via inlet pipe 7. Microwave radiation from a microwave source (e.g. a Magnetron) is introduced to the oven 4 by means of wave guide 5 and is distributed with the aid of oven mode stirrer 6. Due to coupling of the microwave radiation with solution or slurry in the process vessel 1 heat is generated therein and thermal energy is inhibited from escaping by thermal insulation 2 and 3. Consequently the temperature in the process vessel 1 rises to provide conditions in which the slurry or solution is converted to a dried product and the dried product is fused and runs out of the product outlet 10 to be collected in the collector 11 wherein it may solidify. Vapour and gases (from drying and possibly decomposition of constituents of the solution or slurry) are withdrawn through pipe 9. Due to the fact that inlet pipe 7 is earthed the solution or slurry passing therein is protected from microwave radiation. Thus it is only when solution or slurry leaves the inlet pipe 7 to enter the process vessel 1 that it is subjected to heating. This reduces the risk of blockage due to premature solidification of solution or slurry on its way from the pump 8. It should be noted that during start-up of the apparatus, a solid starting charge of a fusible material capable of coupling with microwave radiation may be placed in the process vessel 1 and subjected to microwave radiation to provide initial heating. The starting charge may be prepared by drying and fusing a sample of solution or slurry to be treated. If desired it can be arranged to permit microwave radiation to couple to the collector 11 to provide heating thereof thereby to promote efficient filling of the collector 11, reduce stresses in the solidifying product by preventing too rapid cooling, and to anneal the product. Referring now to FIG. 2 of the drawings there is shown diagrammatically a fluidised bed vessel 21 having a wave guide 22 for the introduction of microwaves, a solution/slurry inlet 23, a fluidising gas inlet 24, a dried product outlet 25, connected to a melter/receiver 26, and an off-gas outlet 27. An insulating window of microwave transparent material (not shown) may be placed between wave guide 22 and vessel 21. The off-gas outlet 27 is connected to a scrubber bed 28, for containing a particulate solid material, which has means 29 for discharging particulate solid material to the fluidised bed vessel 21. Off-gas outlet 27 and means 29 may be provided by a single piece of apparatus (e.g. a pipe). A particulate solid material inlet 30 is provided for charging the scrubber bed 28 and an off-gas outlet 31 is provided to connect the scrubber bed 28 to a condenser 32. The condenser 32 has cooling-fluid inlet 33 and outlet 34, an outlet 35 for condensate, and a gas outlet 36 which can be connected to a gas clean-up plant (not shown). In operation, particulate solid material is introduced to the fluidised bed vessel 21 via means 29 and is maintained as a fluidised bed (represented as 37) by use of fluidising gas inlet 24. Solution or slurry to be treated is introduced via inlet 23 and microwave radiation (e.g. from a Magnetron source not shown) is directed into the vessel 21 via wave guide 22. Due to the coupling of the microwave radiation with the contents of the fluidised bed 37 the temperature rises thereby to form particles of the solid material coated with a dried product formed from the solution or slurry. The coated solid particles are discharged by means of dried product outlet 25 to melter/receiver 26 wherein they may be fused by heating (e.g. with microwave energy or other means). Off-gases leave vessel 21 via outlet 27 and pass to scrubber bed 28 wherein contaminents in the off-gases are scrubbed out by contact with fresh particulate solid material. Particulate solid material can be passed counter-current to the off-gas and into the vessel 21 via means 29 thereby carrying back contaminants scrubbed from the off-gases. The scrubber-bed 28 can contain a fluidised bed of particulate solid material or a vibrating bed thereof. Fresh particulate solid material is introduced via inlet 30. Off-gases from scrubber bed 28 are passed to condenser 32 (cooled by passing a cooling fluid via 33 and 34) to give a condensate at outlet 35, and gas at outlet 36 for processing in a clean-up plant. In a particular example of the present invention the particulate solid material may comprise spheres (0.01-0.1 mm diameter) of glass formers (e.g. Na, Li, B.sub.2 O.sub.3 and SiO.sub.2) and the solution or slurry may contain radioactive waste, so that in the fluidised bed vessel 21 spheres of glass-formers are produced having a coating of dried product formed from the solution or slurry containing radioactive waste. Thus, after fusing in the melter/receiver (26) a glass-like solid incorporating radioactive waste is produced. Referring now to FIG. 3 of the drawings, there is shown a tube 41 a portion of which is located within a microwave oven 42. The tube 41 is provided with an inlet pipe 43 and a gas/vapour outlet 44 and is adapted to contain slugs of glass fibres 45. To permit tube 41 to extend out of the microwave oven 42 apertures 46 and 47 are provided. It will be appreciated that, in accordance with microwave technology, microwave chokes (not shown) may be provided as necessary at apertures 46 and 47 and also where inlet pipe 43 and gas/vapour outlet 44 penetrate the walls of the oven 42. In operation, the slugs of glass fibre 45 are introduced into the tube 41 from the direction 48. Subsequently, solution to be treated is introduced onto a slug 45 via inlet 43, is absorbed therein and subsequently converted to a dried product thereon by application of microwave radiation in the microwave oven 42. (It will be appreciated that microwave radiation is introduced into the microwave oven 42 in a known manner through a wave guide (not shown)). Off-gases produced during the production of the dried product pass through the tube 41 in the direction 49 and therefore pass through, and are filtered by, the "fresh" slugs 45 located in the tube 41 before being discharged therefrom through the gas/vapour outlet 44. Off-gases removed through the outlet 44 can be passed to other treatment apparatus, for example a condensate system, for further treatment. Subsequently a fresh slug 45 is introduced into the tube 41 from the direction 48 with the result that all of the slugs 45 move along the tube in that direction such that "loaded" slugs 45 carrying dried product are thereby moved out of the microwave oven 42 through the aperture 47 and are ultimately discharged from the tube 41. "Loaded" slugs 45 can be discharged from the tube 41 directly to a melting apparatus which may comprise a ceramic melting vessel surrounded by a microwave transparent thermal insulation located in a microwave oven. It will be appreciated that an automatic loading mechanism can be used to introduce fresh glass fibre slugs 45 to the tube 41 in a continuous or semi-continuous manner. It will be appreciated that the present invention is not limited to the treatment of radioactive wastes and that solutions of salts or slurries of non-radioactive substances can be subjected to drying, decomposition and fusion in accordance with the present invention to give a glass-like or ceramic material containing a non-radioactive substance (e.g. in the production of glasses). It will be appreciated that the use of microwave radiation enables the energy applied to be almost wholly absorbed in the matter to be treated thus avoiding the need to pass heat through the walls of containment vessels. The invention will now be further described with reference to the following Examples: EXAMPLE 1 In this example a feed solution simulating a radioactive waste solution was subjected to microwave radiation. The feed solution was a solution/suspension containing nitric acid, 25.7% by weight simulated "waste oxides" (containing some uranium but composed mainly of rare earths, aluminium, iron and magnesium) and the following glass forming components: Na.sub.2 O: 8.3 wt%, Li.sub.2 O: 4.0 wt%, B.sub.2 O.sub.3 : 11.1 wt%, SiO.sub.2 : 50.9 wt%. 126 g of the feed solution were placed in a Pyrex (Reg. Trade mark) beaker and subjected to microwave radiation (from a Magnetron source) in a microwave oven until a dried product was obtained. It was noted that 40 ml of liquid were evaporated in 5 minutes using a power of 750 watts. The beaker and dried product therein were returned to the oven and with the power still set at 750 watts the dried product underwent further decomposition, with the release of nitrous fumes. The temperature rose to bright red heat and the heating was stopped. It was found, after cooling, that the dried product had been converted to a vitreous, glass-like mass. EXAMPLE 2 In this example an apparatus of the type disclosed in FIG. 1 was used to treat a feed solution having the same composition as given in Example 1. For start-up 252 g of preformed fusible dried product (prepared from the solution to be treated) was placed in a vessel, said vessel being surrounded by thermal insulation and situated in a microwave oven (see FIG. 1). Microwave power was applied and increased to a maximum of .about.1.4 KW over a period of 1 hour, and the vessel and fusible product brought up to a temperature of 1020.degree. C. Feed solution as in Example 1 was fed to the vessel initially at 6 ml/min and microwave power maintained at .about.1.4 Kw. Glass flowed from an outlet in the base of the vessel irregularly and was collected in a beaker of water beneath the oven. It is believed that the irregularity of flow was due to the effects of surface tension at the low flow rates used. The main portion of the experiment was conducted with a feed solution flow rate of 7.5 ml/min. It was convenient to end the experiment after .about.9 hours although there was no reason to suppose the process could not have been operated indefinitely. During the experiment the oven was maintained at 1000.degree.-1050.degree. C. and 4.84 liters of feed solution treated to enable 1.344 Kg of glass to be collected (glass production rate 2.14 g/min). EXAMPLE 3 A volume of 400 ml of a suspension of magnesium basic carbonate in water (containing the equivalent of 36 g oxide) was introduced into an alumina tube having a closed bottom end and mounted vertically in thermal insulation. The suspension was subject to microwave radiation (power 1-1.5 Kw) and evaporated to give a fusible dried product. The temperature rose to 970.degree. C. in 80 minutes. At 970.degree. C. glass-forming components were added in the form of a glass frit (200 g) and 20 minutes further application of microwaves took the temperature to 1110.degree. C. at which the contents of the tube was molten. A glass-like solid was obtained on cooling. |
description | This application is a divisional application of U.S. patent application Ser. No. 11/642,920, filed on Dec. 21, 2006 now U.S. Pat. No. 7,634,043, and claims the associated benefit under 35 U.S.C. §121. The entire contents of parent U.S. patent application Ser. No. 11/642,920, entitled “PROTECTION SYSTEMS FOR AND METHODS OF OPERATING NUCLEAR BOILING WATER REACTORS”, are incorporated herein by reference. 1. Field Example embodiments relate to protection systems for and methods of operating nuclear boiling water reactor (“BWR”) power plants. 2. Description of Related Art FIG. 1 illustrates a related art BWR. As shown, a pump 100 supplies water to a reactor vessel 102 housed within a containment vessel 104. The core 106 of the reactor vessel 102 includes a number of fuel bundles such as those described in detail below with respect to FIG. 2. The controlled nuclear fission taking place at the fuel bundles in the core 106 generates heat that turns the supplied water into steam. This steam is supplied from the reactor vessel 102 to turbines 108 that power a generator 110. The generator 110 then outputs electrical energy. The steam supplied to the turbines 108 is recycled by condensing the steam from turbines 108 back into water at a condenser 112, and supplying the condensed steam back to the pump 100. FIG. 2 illustrates a typical fuel bundle 114 in the core 106. A core 106 may include, for example, anywhere from about 200 to about 900 of these fuel bundles 114. As shown in FIG. 2, the fuel bundle 114 may include an outer channel 116 surrounding a plurality of fuel rods 118 extending generally parallel to one another between upper and lower tie plates 120 and 122, respectively, and in a generally rectilinear matrix of fuel rods as illustrated in FIG. 3, which is a schematic representation of a cross-section or lattice of the fuel bundle 114 of FIG. 2. The fuel rods 118 may be maintained laterally spaced from one another by a plurality of spacers 124 vertically spaced apart from each other along the length of the fuel rods 118 within the outer channel 116. Referring to FIG. 3, there is illustrated in an array of fuel rods 118 (i.e., in this instance, a 10×10 array) surrounded by the outer channel 116. The fuel rods 118 are arranged in orthogonally related rows and also surround one or more “water rods,” two water rods 126 being illustrated. The fuel bundle 114 may be arranged, for example, in one quadrant of a control blade 128 (also known as a “control rod”). It will be appreciated that other fuel bundles 114 may be arranged in each of the other quadrants of the control blade 128. Movement of the control blade 128 up and/or down between the fuel bundles 114 controls the amount of reactivity occurring in the fuel bundles 114 associated with that control blade 128. The total number of control blades 128 utilized varies with core size and geometry, and may be, for example, between about 50 and about 200. The axial position of the control blades 128 (i.e., fully inserted, fully withdrawn, or somewhere in between) is based on the need to control excess reactivity and to meet other operational constraints. For each control blade 128, there may be, for example, 24, 48, or more possible axial positions or “notches.” The BWR may include several related art closed-loop control systems that control various individual operations of the BWR in response to demands. For example, a related art recirculation flow control system (“RFCS”) may be used to control core flowrate that, in turn, help to determine the output power of the reactor core. A control blade drive system affects the position of the control blades, the control blade density within the core, and core reactivity. A turbine control system controls steam flow from the BWR to the turbines based on load demands and pressure regulation. The operation of all of these systems, as well as other related art systems, is controlled utilizing various monitoring parameters of the BWR. Exemplary monitoring parameters include core flow and flowrate effected by the RFCS, reactor vessel dome pressure (which is the pressure of the steam discharged from the pressure vessel to the turbines), neutron flux or core power, feedwater temperature and flowrate, steam flowrate provided to the turbines, and various status indications of the BWR systems. Many monitoring parameters are measured directly by related art sensors, while others, such as core thermal power, are typically calculated using measured parameters. These status monitoring parameters are provided as output signals from the respective systems. Nuclear reactors are conservatively specified to minimize any risks from the hazardous materials involved in their use. The materials used in BWRs must withstand various loading, environmental, and radiation conditions. For example, operating pressures and temperatures for the reactor pressure vessel are about 7 MPa and 290° C. for a BWR. Reactor vessel walls are thus several inches thick and very strong materials are used for reactor components. Nonetheless, contingencies are required for failure as components are subjected to operational stress for decades. These contingencies involve not only many layers of preventive systems, but also procedures for rectifying problems that arise. Related art reactor control systems have automatic and manual controls to maintain safe operating conditions as the demand is varied. The several control systems control operation of the reactor in response to given demand signals. Computer programs are used to analyze thermal and hydraulic characteristics of the reactor core. The analysis is based on nuclear data selected from analytical and empirical transients and accidents, and from reactor physics and thermal-hydraulic principles. In the event of an abnormal transient, the reactor operator usually is able to diagnose the situation and take corrective action based on applicable training, experience, and/or judgment. Whether the manual remedial action is sufficient depends upon the transient and upon the operator's knowledge and/or training. If the transient is significant (i.e., challenges any of the reactor safety limits), reactor trip (also referred to as reactor shutdown, scram, or full insertion of all control blades) may be required (the term “scram” is alleged to have originated in the early years of reactor development and operation as an acronym for “super-critical reactor ax-man”). Some transients may occur quickly (i.e., faster than the capability of a human operator to react). In such a transient, a reactor trip will be initiated automatically. Safety analyses generally show that no operator action is necessary within 10 minutes of a postulated transient. A related art nuclear reactor protection system (“RPS”) comprises a multi-channel electrical alarm and actuating system that monitors operation of the reactor and, upon sensing an abnormal transient, initiates action to prevent an unsafe or potentially unsafe condition. At minimum, the related art RPS typically provides three functions: (1) reactor trip that shuts down the reactor when certain monitored parameter limits are exceeded; (2) nuclear system isolation that isolates the reactor vessel and all connections penetrating a containment barrier; and (3) engineered safety feature actuation that actuates related art emergency systems such as cooling systems and residual heat removal systems. Core power protection schemes are typically employed in BWRs when the reactor is operating in its normal operating domain (i.e., after startup and heatup of the reactor). FIG. 4 is a typical BWR power-to-flow operating map showing an operating domain of the reactor. Such operating domains are discussed, for example, in U.S. Pat. No. 5,528,639 (“the '639 patent”). After startup and heatup, the permissible operating domain for the BWR typically is above the cavitation region, below the maximum operating line, and bounded by the minimum normal flow line and the maximum normal flow line. In related art RPSs, when the BWR is operating within the operating domain, an unplanned transient that does not increase the power level (i.e., neutron flux) above a setpoint or setpoints associated with the maximum operating line will not cause a reactor overpower protection trip. FIG. 5 is a BWR power-to-flow operating map showing an operating domain of the reactor with expanded limits. Such operating domains are discussed, for example, in U.S. Pat. Nos. 6,721,383 B2 (“the '383 patent”) and 6,987,826 B2 (“the '826 patent”). FIG. 6 is a BWR power-to-flow operating map showing another operating domain of the reactor with expanded limits. Such operating domains are also discussed, for example, in the '383 patent and the '826 patent. The disclosures of the '639 patent, the '383 patent, and the '826 patent are incorporate in this application by reference. A reactor overpower protection trip is initiated for certain transients that could cause an increase in power above the maximum safe operating level. Generally, an overpower equal to about 120% of the rated power can be tolerated without causing damage to the fuel rods. If thermal power should exceed this limiting value (the maximum safe operating level) or if other abnormal conditions should arise to endanger the system, the RPS will cause a reactor trip. An essential requirement of an RPS is that it must not fail when needed. Therefore, unless the operator promptly and properly identifies the cause of an abnormal transient in the operation of the reactor, and promptly effects remedial or mitigating action, related art RPS will automatically effect reactor trip. However, it is also essential that reactor trip be avoided when it is not desired or necessary (i.e., when there is an error in the instrumentation or when the malfunction is small enough that reactor trip is unnecessary). As discussed in U.S. Pat. No. 5,528,639 (“the '639 patent”), for example, four power-related methods may be used to ensure that acceptable fuel and reactor protection are maintained. Each method uses monitored neutron flux to sense when an increase in power occurs, but each employs a different approach to initiate reactor trip. The first method of protection causes a reactor overpower protection trip if the monitored neutron flux exceeds a preselected and fixed first setpoint. This first setpoint may be, for example, about 120%-125% of rated power. The second method of protection causes a reactor overpower protection trip if the monitored neutron flux exceeds a preselected, but flow-referenced, second setpoint. In this method, the second setpoint is equal to the first setpoint when the reactor core flow is high. However, when reactor core flow is reduced, the second setpoint is also reduced. The third method of protection involves electronically filtering the measured neutron flux signal to produce a signal that has been called simulated thermal power (“STP”). Usual practice employs a single-time-constant filter that approximates the thermal response of the reactor fuel rods. A reactor overpower protection trip is initiated when the STP signal exceeds the flow-referenced second setpoint. This second setpoint may be, for example, about 110%-115% of rated power. The third method is usually used in combination with the first method. In the three methods discussed above, the reactor overpower protection trip setpoints are above the normal operating domain of the reactor to avoid undesired trips during operation in the upper portion of the operating domain. If more protection is required due to partial core power and flow conditions, the reactor overpower protection trip setpoints are manually adjusted. These manual adjustments are a cumbersome nuisance for reactor operators. However, if the reactor overpower protection trip setpoints are not adjusted, complex and restrictive core operating limits are required to ensure acceptable protection at all operating power and flow conditions. Slow transients have been postulated in the partial power and flow range that challenge the effectiveness of these three related art protection methods. These slow transients have been postulated to avoid the protection provided by the associated reactor overpower protection trip setpoints. As discussed in the '639 patent, a fourth method of protection involves automatically adjusting reactor overpower protection trip setpoints to be a controlled margin above the operating power level of the BWR. The fourth method provides enhanced reactor protection when the reactor is operating at less than the maximum operating level. However, alternate and/or supplemental methods of protection may be desired. Example embodiments may provide protection systems for operating nuclear BWR power plants. Also, example embodiments may provide methods of operating nuclear BWR power plants. In an example embodiment, a protection system for a nuclear BWR may include a power-dependent high reactor pressure setpoint. The high reactor pressure setpoint that corresponds to at least one value of percent power in an operating domain of the reactor may be less than the high reactor pressure setpoint that corresponds to 100% power. In another example embodiment, a protection system for a nuclear BWR may include a first high reactor pressure setpoint that corresponds to 100% power and at least one second high reactor pressure setpoint that corresponds to one or more values of percent power in an operating domain of the reactor. The at least one second high reactor pressure setpoint may be less than the first high reactor pressure setpoint. In yet another example embodiment, a method of operating a nuclear BWR may include implementing, in a protection system for the reactor, a power-dependent high reactor pressure setpoint. The high reactor pressure setpoint that corresponds to at least one value of percent power in an operating domain of the reactor may be less than the high reactor pressure setpoint that corresponds to 100% power. In a further example embodiment, a method of operating a nuclear BWR may include implementing, in a protection system for the reactor, a first high reactor pressure setpoint that corresponds to 100% power, and implementing, in the protection system for the reactor, at least one second high reactor pressure setpoint that corresponds to one or more values of percent power in an operating domain of the reactor. The at least one second high reactor pressure setpoint may be less than the first high reactor pressure setpoint. Example embodiments will now be described more fully with reference to the accompanying drawings. Embodiments, however, may be embodied in many different forms and should not be construed as being limited to the example embodiments set forth herein. Rather, these example embodiments are provided so that this disclosure will be thorough and complete, and will fully convey the scope to those skilled in the art. As used herein, the term “and/or” includes any and all combinations of one or more of the associated listed items. It will be understood that, although the terms first, second, third, etc., may be used herein to describe various elements, components, regions, layers, and/or sections, these elements, components, regions, layers, and/or sections should not be limited by these terms. These terms are only used to distinguish one element, component, region, layer, or section from another element, component, region, layer, or section. Thus, a first element, component, region, layer, or section discussed below could be termed a second element, component, region, layer, or section without departing from the teachings of the example embodiments. The terminology used herein is for the purpose of describing particular example embodiments only and is not intended to be limiting. As used herein, the singular forms “a,” “an,” and “the” are intended to include the plural forms as well, unless the context clearly indicates otherwise. It will be further understood that the terms “comprises,” “comprising,” “includes,” and/or “including,” when used in this specification, specify the presence of stated features, integers, steps, operations, elements, and/or components, but do not preclude the presence or addition of one or more other features, integers, steps, operations, elements, and/or components. Unless otherwise defined, all terms (including technical and scientific terms) used herein have the same meaning as commonly understood by one of ordinary skill in the art to which example embodiments belong. It will be further understood that terms, such as those defined in commonly used dictionaries, should be interpreted as having a meaning that is consistent with their meaning in the context of the relevant art and should not be interpreted in an idealized or overly formal sense unless expressly so defined herein. Reference will now be made to example embodiments, which are illustrated in the accompanying drawings, wherein like reference numerals refer to the like components throughout. As discussed above, a RPS may monitor operation of the reactor. Items monitored may include, for example, main steam isolation valve position(s), turbine stop valve position(s), fast closure of turbine control valve(s), drywell pressure, reactor dome pressure (referred to below as “reactor pressure”), reactor water level, main steam line radiation, and/or reactor neutron flux—possibly including STP—referred to below as “reactor power.” One or more warnings, alarms, and/or mitigating actions may be initiated, for example, in the event of abnormal main steam isolation valve position(s), turbine stop valve position(s), fast closure of turbine control valve(s), high drywell pressure, high reactor pressure, low reactor water level, main steam line high radiation, and/or reactor overpower. In addition or in the alternative, a reactor scram may be initiated, for example, in the event of abnormal main steam isolation valve position(s), turbine stop valve position(s), fast closure of turbine control valve(s), high drywell pressure, high reactor pressure, low reactor water level, main steam line high radiation, and/or reactor overpower. The reactor scram may be initiated, for example, by a lone scram signal or by more than one scram signal (possibly including redundancy and/or “voting,” both known to one of ordinary skill in the art). The scram signal or signals may be routed via a common scram bus. RPS monitoring of valve closure(s) related to main steam isolation valve position(s), turbine stop valve position(s), and/or fast closure of turbine control valve(s) can be complex because a reactor plant may include, for example, eight or more main steam isolation valves, four or more turbine stop valves, and/or four or more turbine control valves. Other issues for RPS monitoring of valve closure(s) may include: (1) below about 25% reactor power, for example, reactor plants may not be required to monitor thermal limits; (2) below about 30%-40% reactor power, for example, a low-power bypass (commonly referred to as “Pbypass”) or equivalent setting may disable scrams due to valve closure(s) because steam flow from the reactor is low enough to be bypassed directly to the condenser, obviating the need for a scram (use of and settings for such a low-power bypass or equivalent setting may be plant-dependent); (3) below about 40%-60% reactor power, for example, power-to-load unbalancing (“PLU”) may prevent fast closure—typically on the order of about 100 milliseconds—of the turbine control valves, allowing only slow closure—typically on the order of about 5 seconds—that may limit and/or prevent turbine overspeed when the slow closure does not generate a direct scram signal (use of and settings for PLU may be plant-dependent); (4) maintenance may affect the monitoring of valve closure(s) due to the respective valve(s) being out of normal operating position; and/or (5) other plant-specific needs and design features. As a result, some reactor plant transients may not result in a direct scram due to valve closure(s). However, in a BWR, such valve closure(s) may result in transients that raise both reactor pressure (i.e., loss of steam demand may cause reactor pressure to rise) and reactor power (i.e., the higher reactor pressure may collapse some voids in the core, adding positive reactivity). Although a high reactor pressure scram typically may be a backup scram to a valve closure(s) scram and/or a reactor overpower scram, there may be situations in which the high reactor pressure scram should be the primary scram mechanism. Pressurization transients, for example, that may require a scram for reactor protection—but that do not have a direct scram on valve closure(s)—may scram on reactor overpower and/or high reactor pressure. However, the time required to reach a scram on reactor overpower starting from a reduced level of reactor power may be longer, due to the larger difference between the reduced level of reactor power and the setpoint for scram on reactor overpower. As a result, a scram on high reactor pressure may need to be relied upon. At the same time, the reactor pressure associated with the reduced level of reactor power may be lower as well. As a result, the time required to reach a scram on high reactor pressure starting from a reduced level of reactor power may also be longer, due to the larger difference between the reduced reactor pressure and the setpoint for scram on high reactor pressure. Related art solutions to this problem may involve calculating more rigorous thermal limits for reactor operation at reduced levels of reactor power. However, these more rigorous thermal limits—often referred to as thermal limit “penalties”—may complicate reactor operation at reduced levels of reactor power, may severely limit maneuverability of the reactor at the reduced levels of reactor power, and/or may increase reactor power ascension times. FIG. 7 is a graph showing a typical curve of reactor pressure (psia) versus reactor power (% of rated power) for a BWR. As may be seen, reactor pressure tends to go up as reactor power goes up, but the relationship is not linear. Although the typical curve of FIG. 7 extends only from about 25% reactor power to about 100% reactor power, the typical curve would extend down and to the left and/or up and to the right according to the calculations discussed below. The operating pressure of the reactor at given values of reactor power, or “operating pressure,” may be calculated approximately as:operating pressure=control value+P/P0*PCB+(P/P0)2*SLPD where: the control value at rated conditions (“control value”)—which is independent of reactor power—is a pressure setting in the turbine control system calculated to achieve rated steam pressure at rated steam flow; P represents actual reactor power; P0 represents rated reactor power; PCB represents the pressure control band; and SLPD represents the steam line pressure drop. PCB may be calculated as:PCB=turbine inlet pressure−control value The turbines may be controlled, for example, by a turbine control system. The turbine control system may use turbine inlet pressure as a pressure it tries to maintain. Maintaining the turbine inlet pressure constant may result, for example, in lower reactor pressures at low steam flow rates. The turbine inlet pressure may be measured, for example, in a common steam header upstream of the turbine stop valves and the turbine control valves. PCB may be converted to percent steam flow by multiplying the PCB pressure value by the pressure regulator gain. Because the PCB may have a range of about 0 psi-30 psi and because the pressure regulator gain may be approximately linear over the full range of rated steam flow (0%-100%), the pressure regulator gain may be approximately constant at a value of about 3.33% steam flow per psi. Thus, at rated conditions—30 psi PCB—the control system produces 100% steam flow demand. SLPD may be calculated as:SLPD=operating pressure−turbine inlet pressure Values for SLPD, for example, may be greater than or equal to about 30 psi and less than or equal to about 100 psi. In an example embodiment, the SLPD may be greater than or equal to about 55 psi and less than or equal to about 70 psi. In another example embodiment, the SLPD may be about 65 psi. Because of the dependence of operating pressure on, for example, the control value, rated reactor power, and turbine inlet pressure, the values of operating pressure may be plant-dependent. Values for operating pressure corresponding to 100% power, for example, may be greater than or equal to about 1,000 psia and less than or equal to about 1,075 psia. In FIG. 7, the operating pressure corresponding to 100% power is about 1,050 psia. Steam flow may not be exactly proportional to reactor power due, in part, to changes in feedwater temperature (water supplied from the pump 100 to the reactor vessel 102) as reactor power changes. A more exact calculation may substitute the mass flow rate of steam divided by the rated mass flow rate of steam for the P/P0 term. The difference between the two calculations may be on the order of 2%-5%, with the approximate calculation of operating pressure generally being closer to the more exact calculation of operating pressure as reactor power increases. FIGS. 8-11 are graphs showing typical curves of reactor pressure (psia) versus reactor power (% of rated power) for a BWR, related art pressure setpoint lines, and examples of a new pressure setpoint curve (“NPSC”) or NPSCs according to some of the example embodiments. Although the typical curves, setpoint lines, and NPSCs of FIGS. 8-11 extend only from about 25% reactor power to about 100% reactor power, the typical curves, setpoint lines, and/or NPSCs could be continued to lower and/or higher values of reactor power. Generally, the NPSC at higher values of reactor power may not be higher than the high reactor pressure setpoint that corresponds to 100% power. FIG. 8 is a graph showing a typical curve of reactor pressure (psia) versus reactor power (% of rated power) for a BWR, a related art pressure setpoint line, and an example NPSC that may include zero-order, first-order, and/or second-order components. Related art BWR RPSs use a fixed high reactor pressure setpoint for all levels of reactor power. The fixed high reactor pressure setpoint may be, for example, determined based upon a reactor pressure corresponding to 100% power. As discussed above, values for reactor pressure corresponding to 100% power, for example, may be greater than or equal to about 1,000 psia and less than or equal to about 1,075 psia. In an example embodiment, the reactor pressure corresponding to 100% power may be about 1,050 psia. At least partially as a result, values for the fixed high reactor pressure setpoint in related art reactors, for example, may be greater than or equal to about 1,040 psia and less than or equal to about 1,125 psia. In FIG. 8, the fixed high reactor pressure setpoint is about 1,115 psia. If the setpoint line was continued to lower and/or higher values of reactor power, the setpoint at those lower and/or higher values of reactor power also would be about 1,115 psia. In an example embodiment, a protection system for a nuclear BWR may include a power-dependent high reactor pressure setpoint (“PDHRPS”). The PDHRPS may take on two or more values in an operating domain of the reactor, depending on the value of power that is, for example, measured, calculated, or measured and calculated. Values of power not expressed in percent power may be so expressed by dividing, for example, the measured, calculated, or measured and calculated power by the rated power. The high reactor pressure setpoint that corresponds to at least one value of percent power in the operating domain of the reactor may be less than the high reactor pressure setpoint that corresponds to 100% power. The example PDHRPS/NPSC of FIG. 8 maintains a substantially constant pressure difference between itself and the operating pressure curve. The PDHRPS may be calculated as:PDHRPS=control value+P/P0*PCB+(P/P0)2*SLPD+HRPS100−RP100 where: HRPS100 represents the high reactor pressure setpoint that corresponds to 100% power; and RP100 represents the rated reactor pressure that corresponds to 100% power. As can be seen, PDHRPS may include a zero-order (constant) component (control value+HRPS100−RP100), a first-order (linear) component (P/P0*PCB), and/or a second-order (quadratic) component ((P/P0)2*SLPD). This equation can be rewritten as:PDHRPS=operating pressure+HRPS100−RP100 The PDHRPS may initiate a reactor scram. In addition or in the alternative, the PDHRPS may initiate one or more warnings, alarms, or mitigating actions. The high reactor pressure setpoint that corresponds to at least one value of percent power in an operating domain of the reactor also may initiate a reactor scram. Additionally, the HRPS100 may initiate a reactor scram. The PDHRPS initiating a reactor scram may result in earlier pressure scrams to improve thermal limits due to anticipated operational transients, particularly slow pressurization transients and/or transients that do not have a direct scram from valve closure(s). In another example embodiment, a protection system for a nuclear BWR may include a first high reactor pressure setpoint that corresponds to 100% power and at least one second high reactor pressure setpoint that corresponds to one or more values of percent power in an operating domain of the reactor. The at least one second high reactor pressure setpoint may be less than the first high reactor pressure setpoint. In yet another example embodiment, a method of operating a nuclear BWR may include implementing, in a protection system for the reactor, a PDHRPS. The high reactor pressure setpoint that corresponds to at least one value of percent power in an operating domain of the reactor may be less than the high reactor pressure setpoint that corresponds to 100% power. In a further example embodiment, a method of operating a nuclear BWR may include implementing, in a protection system for the reactor, a first high reactor pressure setpoint that corresponds to 100% power, and implementing, in the protection system for the reactor, at least one second high reactor pressure setpoint that corresponds to one or more values of percent power in an operating domain of the reactor. The at least one second high reactor pressure setpoint may be less than the first high reactor pressure setpoint. FIG. 9 is a graph showing a typical curve of reactor pressure (psia) versus reactor power (% of rated power) for a BWR, a related art pressure setpoint line, and two example NPSCs that may be linear. The PDHRPS may be calculated as:PDHRPS=HRPS100−S*P/P0 where S represents a slope determined, for example, to optimize the margin between the operating pressure and the PDHRPS through at least some portion of the operating domain of the reactor. FIG. 10 is a graph showing a typical curve of reactor pressure (psia) versus reactor power (% of rated power) for a BWR, a related art pressure setpoint line, and an example NPSC that may be a series of constant values, each covering a range of reactor powers, in the manner of a step function. The ranges may be, for example, of similar or dissimilar span of percent powers. For the range of reactor powers that includes 100% reactor power, PDHRPS may equal HRPS100. FIG. 11 is a graph showing a typical curve of reactor pressure (psia) versus reactor power (% of rated power) for a BWR, a related art pressure setpoint line, and an example NPSC that may be a combination of one or more, for example, zero-order, first-order, second-order, higher-order, geometric, logarithmic, exponential, step function, and/or other components, each covering a similar or dissimilar span of percent powers. The example NPSC also may include, for example, one or more look-up tables, series of points, digital approximations, or other components. For the range of reactor powers that includes 100% reactor power, PDHRPS may equal HRPS100. Considerations regarding the at least one second high reactor pressure setpoint are similar to those regarding the PDHRPS. FIG. 12 illustrates a protection system 200 for a nuclear boiling water reactor according to example embodiments. The protection system 200 may include a device 202 configured to monitor reactor power; a device 204 configured to monitor reactor pressure; a device 206 configured to determine a power-dependent high reactor pressure setpoint based on the monitored reactor power; and a device 208 configured to initiate a protection system action when the monitored reactor pressure is greater than the power-dependent high reactor pressure setpoint. As discussed above, the RPS monitoring of valve closure(s) may change, for example, below about 25% reactor power, below about 30%-40% reactor power, below about 40%-60% reactor power, and/or at or near additional values of reactor power related to other plant-specific needs and design features. Changes in the NPSCs may or may not reflect one or more of these changes to the RPS monitoring of valve closure(s). RPSs include, for example, analog, digital, or analog and digital components. Digital components may allow, for example, employment of more complex NPSCs that may provide additional improvements in safety margin for a given plant. In addition to the PDHRPS and/or the at least one second high reactor pressure setpoint, one or more signals in the system that correspond to one or more values of percent power may be delayed in time before the one or more signals affect the power-dependent high reactor pressure setpoint. The delay may result in the power change during a pressurization transient not significantly changing the PDHRPS and/or the at least one second high reactor pressure setpoint during the transient. This may be particularly true during slow pressurization transients. The one or more signals may be delayed in time by greater than or equal to about 1 second. For example, the one or more signals may be delayed in time by greater than or equal to about 2, 3, 4, 5, 6, 7, 8, 9, 10, 11, 12, 13, 14, 15, 16, 17, 18, 19, 20, 25, 30, 45, 60, or more seconds. The one or more signals may be delayed in time by less than or equal to about 60 seconds. For example, the one or more signals may be delayed in time by less than or equal to about 45, 30, 25, 20, 19, 18, 17, 16, 15, 14, 13, 12, 11, 10, 9, 8, 7, 6, 5, or fewer seconds. The one or more signals may be delayed in time by greater than or equal to about 1 second and less than or equal to about 60 seconds. The one or more signals also may be delayed in time by greater than or equal to about 1, 2, 3, 4, 5, 6, 7, 8, 9, 10, or more seconds and less than or equal to about 20, 19, 18, 17, 16, 15, 14, 13, 12, 11, 10, 9, 8, 7, 6, 5, or fewer seconds. The delay may be optimized, for example, to obtain the most effective system for the most desirable combination of the operating thermal limit(s) and/or operating flexibility. In an example embodiment, the one or more signals may be delayed in time by greater than or equal to about 5 seconds and less than or equal to about 10 seconds. In another example embodiment, the one or more signals may be delayed in time by greater than or equal to about 6 seconds and less than or equal to about 8 seconds. In yet another example embodiment, the one or more signals may be delayed in time by greater than or equal to about 8 seconds and less than or equal to about 10 seconds. Apparatuses and methods to delay in time the one or more signals in the system that correspond to one or more values of percent power are known by one of ordinary skill in the art. In addition or in the alternative, one or more signals in the system that correspond to one or more values of percent power may be lagged relative to one or more signals in the system that correspond to reactor pressure. The lag may result in the power change during a pressurization transient not significantly changing the PDHRPS and/or the at least one second high reactor pressure setpoint during the transient. This may be particularly true during slow pressurization transients. The lag may be implemented, for example, by electronically filtering the measured neutron flux signal in a manner similar to that used to produce the STP signal. The measured neutron flux signal may be, for example, the Average Power Range Monitor (“APRM”) signal. The lag may be implemented, for example, by a single-time-constant filter that may approximate the loop transit time of the reactor and/or a value that may envelop a typical timing for one or more slow pressurization transients. Such loop transit times may be, for example, greater than or equal to about six seconds and less than or equal to about eight seconds. Apparatuses and methods to lag the one or more signals in the system that correspond to one or more values of percent power relative to one or more signals in the system that correspond to reactor pressure are known by one of ordinary skill in the art. While example embodiments have been particularly shown and described, it will be understood by those of ordinary skill in the art that various changes in form and details may be made in the example embodiments without departing from the spirit and scope of the present invention as defined by the following claims. |
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062263552 | description | FIG. 1 shows an embodiment of an X-ray examination apparatus. The X-ray source 2 emits an X-ray beam 15 in order to irradiate an object 16. Due to differences in the X-ray absorption within the object 16, for example a patient to be radiologically examined, an X-ray image is formed on an X-ray-sensitive surface 17 of the X-ray detector 3 which is arranged so as to face the X-ray source. The X-ray detector is provided, for example with an image intensifier pick-up chain which includes an X-ray image intensifier 18 for converting an X-ray image into a light image on an exit window 19, and a video camera 23 for picking up the light image. An entrance screen 20 acts as an X-ray-sensitive surface which converts incident X-rays into an electron beam which is imaged onto an exit window 19 by means of an electron optical system 21. The incident electrons generate the light image by way of a phosphor layer 22 of the exit window. The video camera 23 is optically coupled to the X-ray image intensifier 18 by way of an optical coupling. The optical coupling includes, for example a lens system or an optical fiber coupling 24. The video camera derives an electronic image signal from the light image and applies it to a monitor 25 in order to visualize the image information of the X-ray image. The electronic image signal can also be applied, for example to an image processing unit 26 so as to be processed further. In order to realize local attenuation of the X-ray beam so as to adjust a two-dimensional intensity profile, an X-ray filter 4 is arranged in the X-ray beam 15 between the X-ray source 2 and the object 16. The X-ray filter includes a large number of filter elements 5. Furthermore, a filter element 5 preferably includes a capillary tube whose inner side is covered with an electrically conductive layer. The capillary tubes communicate, via a first opening, with a reservoir (not shown in the Figure) containing an X-ray absorbing liquid. The construction of such an X-ray filter and the composition of the X-ray-absorbing liquid are known from the cited European patent application EP-A-740839. The X-ray absorptivity of X-ray filters of this kind can be adjusted by application, using an adjusting unit 7, of electric voltages across the inner side of the capillary tubes 5 and the X-ray-absorbing liquid. This is because the adhesion of the X-ray-absorbing liquid to the inner side of the capillary tubes is dependent on the electric voltage applied across the inner side of the capillary tubes and the X-ray-absorbing liquid. The capillary tubes are filled with a given quantity of X-ray-absorbing liquid in dependence on the electric voltage across the individual tubes 5 and the X-ray-absorbing liquid. Because the capillary tubes extend approximately parallel to the lines from the X-ray source 2 via the X-ray filter 4 to the X-ray detector 3, the X-ray absorptivity of the individual capillary tubes is dependent on the relative amount of X-ray absorbing liquid present in such a capillary tube. The electric adjusting voltages for the individual filter elements are adjusted by means of the adjusting unit 7 while taking into account the brightness values in the X-ray image and/or the adjustment of the X-ray source 2. To this end, the adjusting unit 7 is coupled to an output terminal 40 of the video camera 23 and to the power supply 11 for the X-ray source 2. FIG. 2 is a sectional view of an X-ray filter in a first embodiment of the X-ray examination apparatus according to the invention. In FIG. 2 the cross-sections of capillary tubes 5 are oriented in a longitudinal direction towards a center F.sub.2 of, for example a curved surface. A first cross-section 52 of the X-ray filter preferably comprises in a first direction a first segment of a first circle having a first radius CR.sub.1 whereas a second cross-section of the X-ray filter comprises, in a second direction which is perpendicular to the first direction, a second segment of a second circle having a second radius CR.sub.2. A spherical entrance surface of the X-ray filter 5 is obtained, for example when the first radius CR.sub.1 is chosen so as to be equal to the second radius CR.sub.2. For such a spherical surface the radii CR.sub.1 and CR.sub.2 amount to, for example 10 cm. The resolution uniformity of the X-ray filter 4 is enhanced by arranging the tubes 5 in the X-ray filter in this manner. Furthermore, in practice the cross-sections 53 of the capillary tubes 5 amount to, for example approximately 300 .mu.m and the wall thickness 50 of the capillary tubes amounts to, for example 10 .mu.m. The number of capillary tubes 5 in the X-ray filter 4 amounts to, for example 256.sup.2, said number being arranged in a square matrix of 256.times.256 tubes. The positioning of the X-ray filter 4 in the first embodiment of the X-ray examination apparatus 1 according to the invention will be described with reference to FIG. 3. FIG. 3 shows an X-ray source 2 which has a focus with a first focus radius R.sub.1 and a center F.sub.1. FIG. 3 also shows an embodiment of an X-ray filter 4 whose tubes 5 are arranged on a spherical surface as shown, for example in FIG. 2. A center 60 of the X-ray filter, being coincident with a center of the spherical surface, a cross-section 52 of which is shown in FIG. 3, contains a center of the first circle as well as a center of the second circle. FIG. 3 also shows an X-ray sensitive surface 17 of the X-ray detector 18. In order to ensure that the lines trough all the longitudinal axes of the tubes intersects the lines from the X-ray source via the X-ray filter to different points of the X-ray detector, the distance from a first point F.sub.2 of a center 60 of the spherical surface of the X-ray filter 5 to a midpoint F.sub.1 of a focus of the X-ray source 2 is preferably chosen to be at least equal to a focus radius R.sub.1 of the focus. In practice the focus radius R.sub.1 amounts to, for example 150 .mu.m and said distance thus amounts to, for example 150 .mu.m. X rays starting from the X-ray source travelling to different points of the X-ray detector then traverses a substantial part of the absorbing liquid of at least one tube of the X-ray filter so that efficient X-ray absorption is achieved. The absorption efficiency in the X-ray-absorbing liquid is thus enhanced, and hence also the dynamic range of the X-ray filter. As a result of this step the theoretically feasible dynamic range is approached A cylindrically symmetrical surface can be used as an alternative for a spherical surface of the X-ray filter 5. The center 60 of the cylindrically symmetrical surface then contains a line. FIG. 4 shows an example of the positioning of the X-ray filter 4 in a second embodiment of the X-ray examination apparatus according to the invention. The second embodiment of the X-ray examination apparatus includes an X-ray source 2 with a focus having a first focus radius R.sub.1 and a center F.sub.1, an X-ray filter 4 in which the capillary tubes 5 are arranged so as to extend parallel to one another, and an X-ray detector 18. FIG. 4 also shows an X-ray-sensitive surface 17 of the X-ray detector 18. The X-ray filter 4 comprises, for example 256.sup.2 capillary tubes. The length of such capillary tubes 5 amounts to, for example 25 mm and their cross-section is, for example 200 .mu.m. Furthermore, the entrance surface 54 of the X-ray filter extends parallel to the image plane of the X-ray-sensitive surface 17 of the X-ray detector 18. In order to enhance the dynamic range of the X-ray filter 5, an angle .alpha. enclosed by a longitudinal direction of a tube 5 of the X-ray filter 4 with respect to a second normal l.sub.2 to the X-ray-sensitive surface 17 of the X-ray detector 18 is chosen to be equal to the arc tan of the ratio of a first distance D.sub.1 between a center of an entrance surface 54 of an X-ray filter 4 and a tube 55 near an edge of the X-ray filter 5 to a second distance D.sub.2 between the center of the entrance surface 54 and a center F.sub.1 of the focus of the X-ray source 2. It is thus achieved that the lines trough all the longitudinal axes of the tubes intersects the lines from the X-ray source 2 via the X-ray filter 4 to different points of the X-ray detector 18 and that X rays starting from the X-ray source travelling to different points of the X-ray detector traverses a substantial part of the absorbing liquid of at least one tube of the X-ray filter so that efficient X-ray absorption is achieved. The distance D.sub.1 in practice amounts to, for example approximately 25 mm whereas the distance D.sub.2 amounts to approximately 100 mm. The angle .alpha. then amounts to approximately 14 degrees. In order to prevent the appearance of an undesirable pattern in the X-ray image due to a difference in absorption between the X-ray absorbing liquid and the glued joints between various walls at the side of the capillary tubes which communicates with the X-ray absorbing liquid and a first plate of the X-ray filter, the adhesive contains X-ray absorbing particles. This step will be described in detail with reference to FIG. 5. FIG. 5 is a sectional view of an X-ray filter 4 which includes capillary tubes 5. The X-ray filter also includes a first plate 70 and a second plate 73, wherebetween the capillary tubes 5 are arranged. Preferably, the first and second plates 70,73 are positioned perpendicular to the capillary tubes 5. The tubes also have walls 50 which contain, for example a synthetic material. The tubes contain an X-ray-absorbing liquid 72. Different walls 50 of different tubes 5 are connected to the first plate 70 by way of a first glued joint 71 and to the second plate 73 of the X-ray filter by way of a second glued joint 74. The first and the second glued joint 71, 74 comprise an adhesive as customarily used by those skilled in the art, for example an epoxy-type adhesive or an adhesive of the two-component type. In order to prevent the adhesive in the first glued joint 71 between the walls 50 at the sides of the tubes which communicate with a reservoir containing the X-ray-absorbing liquid and the first plate 70 from causing an undesirable pattern in the X-ray image, an X-ray absorbing material, for example molybdenum (Mo), lead (Pb) or tungsten (W) is added to the adhesive of the first glued joints 71, so that the X-ray absorption of the glued joint approximately equals that of the X-ray absorbing liquid 72. An advantage of the use of molybdenum and tungsten over lead is that their effects on the environment are less severe. |
description | This Application claims foreign priority to Australian Patent Application No. 2008243144, entitled “IMPROVEMENTS IN RADIOPHARMACEUTICAL PURIFICATION,” and having a filing date of Nov. 6, 2008. The present invention generally relates to the production of radiopharmaceuticals, and more particularly to improved methods and apparatus for the purification of the products of radiopharmaceutical synthesis. Radioactive compositions have a range of diagnostic and therapeutic applications. However, the half-life of radiopharmaceuticals is typically on the order of hours, and it is therefore not possible to retain stocks of such compounds within hospitals, clinics and/or research laboratories. Rather, it is necessary to synthesise and purify radiopharmaceuticals on-demand, for example within a hospital or clinical laboratory. Purification is often achieved via chromatographic techniques, such as high-performance liquid chromatography (HPLC), which perform temporal and spatial separation of the desired product from attendant impurities and by-products. Such methods of purification carry an associated risk of the product including impurities, for example if collection is commenced too early, or concluded too late. It is undesirable, and potentially highly detrimental, for impurities or by-products of radiopharmaceutical synthesis to be injected into the patient or subject. The operator of the synthesis and purification apparatus may therefore adopt a conservative approach, for example collecting only a portion of the desired product by commencing collection later than necessary and/or halting collection earlier than necessary. This results in a reduced volume of the desired radiopharmaceutical product, with an associated risk that an insufficient activity of the desired product will be collected. One aspect of the invention has been developed to, at least in part, provide an improved automated method and apparatus enabling mitigation of the uncertainties and risks inherent in existing approaches to radiopharmaceutical purification. More particularly, aspects of the invention may be applied in an apparatus for purifying the product of a reaction to synthesise a radiopharmaceutical compound, wherein the apparatus includes a chromatographic separating device, such as an HPLC column, a UV-absorption detector located at a UV-monitoring location, a radioactivity detector, such as a scintillation counter, located at a radioactivity monitoring location, and a collection container having a collection inlet valve located at a collection location. The separating device, the UV-absorption detector, the radiation detector, and the collection valve are interconnected via suitable tubing to permit flow of eluent therebetween. In one aspect, the invention provides a method of collecting the purified radiopharmaceutical compound which comprises the steps of: determining a first eluent propagation time between the UV-monitoring location and the radiation-monitoring location; determining a second eluent propagation time from at least one of the UV-monitoring location and the radioactivity-monitoring location to the collection location; passing the synthesis product through the chromatographic separating device to produce a corresponding eluent; monitoring UV absorption of the eluent to identify a first time period in which a substantially pure sample of the radiopharmaceutical compound is present at the UV-monitoring location; monitoring radioactivity of the eluent to identify a second time period in which a substantially high concentration of the radiopharmaceutical compound is present at the radioactivity-monitoring location; determining a third time period in which a substantially pure sample of the radiopharmaceutical compound is present at the collection location in a substantially high concentration, based upon said first time period, said second time period, said first eluent-propagation delay, and said second eluent-propagation delay; and opening the collection inlet valve during the third time period to collect the purified radiopharmaceutical compound in the collection container. It will be appreciated that, in the context of the invention, the term “substantially pure” refers to the sample having sufficient purity for the intended application, typically injection into a patient or other subject. The criteria may therefore vary, depending upon the particular radiopharmaceutical compound and synthesis process, however in any given case it would be possible to specify a suitable threshold of UV absorption, measured at the UV-monitoring location, corresponding with the presence of the desired radiopharmaceutical compound in a sufficiently pure form. Similarly, the term “substantially high concentration”, in the context of the invention, means a sufficient concentration for the intended purpose of collection. Due to the correlation between concentration and radioactivity levels, in any given case it would be possible to establish an appropriate threshold of radioactivity at the radioactivity-monitoring location corresponding with the required concentration. In another aspect of the invention, the purifying apparatus comprises: means for determining a first eluent-propagation time between the UV-monitoring location and the radiation-monitoring location; means for determining a second eluent-propagation time from at least one of the UV-monitoring location and the radiation-monitoring location to the collection location; means for monitoring the UV absorption of the eluent to identify a first time period in which a substantially pure sample of the radiopharmaceutical compound is present at the UV-monitoring location; means for monitoring radioactivity of the eluent to identify a second time period in which a substantially high concentration of the radiopharmaceutical compound is present at the radioactivity-monitoring location; means for determining a third time period in which a substantially pure sample of the radiopharmaceutical compound is present at the collection location in a substantially high concentration, based upon said first time period, said second time period, said first eluent-propagation delay, and said second eluent-propagation delay; and means for opening the collection inlet valve during the third time period to collect the purified radiopharmaceutical compound in the collection container. In preferred embodiments, the apparatus comprises: at least one microprocessor; at least one memory device operatively associated with the microprocessor; at least one peripheral interface operatively associated with the microprocessor and enabling the microprocessor to receive UV-absorption data from the UV-absorption detector and radioactivity data from the radioactivity detector, wherein the memory device comprises computer-executable instruction code stored therein which, when executed by the microprocessor, causes the microprocessor to execute the steps of: determining a first eluent-propagation time between the UV-monitoring location and the radiation-monitoring location; determining a second eluent-propagation time from at least one of the UV-monitoring location and the radiation-monitoring location to the collection location; receiving UV-absorption data from the UV-absorption detector in order to identify a first time period in which a sample of the radiopharmaceutical compound is present at the UV-monitoring location; receiving radioactivity data from the radioactivity detector so as to identify a second time period in which a substantially high concentration of the radiopharmaceutical compound is present at the radioactivity-monitoring location; determining a third time period in which a substantially pure sample of the radiopharmaceutical compound is present at the collection location in a substantially high concentration, based upon said first time period, said second time period, said first eluent-propagation delay, and said second eluent-propagation delay; and generating an indication, during said third time period, that the collection inlet valve should be opened in order to collect the purified radiopharmaceutical compound in the collection container. In some embodiments, the collection inlet valve may include an electrical control input for opening and closing the valve, and the apparatus may include means, such as a suitable peripheral interface between the microprocessor and the valve-control input, enabling the collection inlet valve to be opened during the third time period under automated control, in response to the generated indication. Alternatively, or additionally, the apparatus may include a display, and a visual indication may be generated on the display during the third time period enabling the operator of the apparatus to open the collection inlet valve under manual control. Further preferred features and advantages of the invention will be apparent to those skilled in the art from the following description of a preferred embodiment of the invention, which should not be considered to be limiting of the scope of the invention as defined in the preceding statements, or in the claims appended hereto. FIG. 1 illustrates schematically the major elements of a system for purifying the product of a radiopharmaceutical synthesis reaction, according to a preferred embodiment of the invention. In particular, the system 100 includes an input 102, via which raw synthesis products are received. Details of the synthesis process will depend upon specifics of a radiopharmaceutical product that is to be obtained, and are not relevant to the present invention. For present purposes, it is sufficient to note that the raw synthesis product includes the desired radiopharmaceutical compound, along with undesired impurities and by-products. In accordance with embodiments of the invention, the desired radiopharmaceutical compound is temporally and spatially separable from the undesired components via chromatographic techniques, such as HPLC. A six-way rotary valve 104 facilitates communication of fluids between various elements of the system 100, as described in greater detail below. In one part of the system 100, a valve 106 provides for selection between a mobile phase solution 108 and a rinse solvent 110. The valve 106 communicates the selected liquid to a pump 112, which is used to drive the separation process, as described in greater detail below. The pump 112 is connected to a port of the six-way valve 104. A sample storage loop 114, consisting of a suitable length of tubing, is connected between two further ports of the six-way valve 104. Another port of the valve 104 is connected to a waste storage container 116. The final port of the valve 104 is connected to one or more chromatographic separating columns, such as HPLC column 118. While the system 100 includes only one separating column 118, where multiple columns are provided an additional column selection valve may also be included. When the raw synthesis product is available for purification, the rotary valve 104 is configured to allow the raw product to enter the storage loop 114. Preferably, the storage loop 114 is of an appropriate length to contain the exact volume of raw product available. However, any excess is directed to the waste storage container 116. Prior to and/or during this period, the valve 106 may be used to select the rinse solvent 110, and the pump 112 operates to drive the rinse solvent through the HPLC column 118, and the rest of the purification system, in preparation for the purification process. As will be appreciated, rinsing is optional, and may be performed before and/or after purification, depending upon requirements. Prior to purification, however, the valve 106 should be operated in order to select the mobile phase solution 108, which is then allowed to fill the system prior to commencement of purification. Once the storage loop 114 is full, the six-way rotary valve 104 is operated such that the pump 112 drives the raw synthesis products into the HPLC column 118. Within the column 118, chromatographic separation of the component parts of the raw synthesis products occurs, and the resulting eluent emerges at the column output 120. The eluent passes, via suitable tubing, through a UV-absorption detector 122, a radiation detector, such as a scintillation counter 124, and via a further valve 126, to either a waste container 128, or a radiopharmaceutical product container 130. The valve 126 is a collection valve, which is operated at the appropriate time in order to direct the desired purified radiopharmaceutical product into the collection container 130, on the basis of UV absorption and radioactivity measurements acquired from the detectors 122, 124. The UV-absorption detector 122 and the radiation detector 124 provide information that may be used to identify the presence of the desired radiopharmaceutical compounds within the eluent output from the HPLC column 118. This information may therefore be used to direct the operation of the collection valve 126, as described in greater detail below. Turning firstly to the UV-absorption detector 122, this device typically comprises a UV light source and a corresponding detector, wherein the path of the eluent passes between the source and the detector. The output frequency or spectrum of the UV source may be selectable, or controllable, in accordance with the contents of the eluent. For example, by selection of an appropriate wavelength of UV light, the corresponding absorption of the desired radiopharmaceutical compound may be substantially lower than the absorption of undesired impurities and by-products in the eluent, such that the presence of the desired compound corresponds with an increase in UV energy received at the UV detector. Accordingly, as the purity of the radiopharmaceutical compound increases, an increase in the detected signal, photocurrent or corresponding voltage, will be observed. Collection of the radiopharmaceutical compound should be timed to coincide with the passage of a portion of the eluent for which the UV-absorption readings indicate a substantially pure sample of the desired radiopharmaceutical compound. Furthermore, readings obtained using the radioactivity detector 124 are correlated with the concentration of the radiopharmaceutical compound passing through the detector. At higher levels of concentration, greater radioactivity will be detected, in the form of an output current, or corresponding voltage of the detector 124. Significantly, the eluent passes from the UV detector 122 to the radiation detector 124 via a length of feed line or tubing 132, and from the radiation detector 124 to the collection valve 126 via a further length of feed line or tubing 134. Accurate determination of the appropriate time for collection of the purified radiopharmaceutical product by operation of the collection valve 126 is complicated by the fact that the propagation delay of eluent passing through the lengths of tubing 132, 134 is dependent upon the particular setup of the purification apparatus, and the fact that these delays may be significant compared to the relevant collection period. FIG. 2 is a schematic diagram illustrating a microprocessor-based apparatus 200 for monitoring and/or controlling the process of purification of the desired radiopharmaceutical product, using the system 100 illustrated in FIG. 1. The exemplary apparatus 200 includes a computer 202 having at least one microprocessor 204, which is interfaced, or otherwise associated, with a high-capacity, non-volatile memory/storage device 206, such as one or more hard-disk drives. The storage device 206 is used for permanent or semi-permanent storage of program instructions and data relating to the operation of the computer 202, and the implementation of the preferred embodiment of the present invention. The computer 202 further includes an additional storage medium 208, typically being a suitable type of volatile memory, such as random access memory, for containing executable program instructions and transient data relating to the operation of the computer 202. In particular, the memory device 208 contains a body of program instructions 212 implementing various software-implemented features of the present invention, as described in greater detail below, with reference to the remaining drawings. In general, these features include analysis and processing functions, such as for receiving data from various components of the system 100, and for controlling various other components of the system 100, as well as for interacting with an operator of the apparatus 200 via a suitable user interface. To facilitate operation of the apparatus 200, the computer 202 has associated input/output devices 214, such as a display monitor, keyboard and/or suitable pointing device (ie a mouse). A further peripheral interface 216 which may be a custom interface, or a standard interface such as an RS-232 and/or USB serial port. The peripheral interface 216 is used to connect the computer 202 to a data acquisition and control circuit 218. In an exemplary embodiment the data acquisition and control card 218 is a commercially-available National Instruments NI DAQPad-6015 Multifunction Data Acquisition and Control Card. This particular device is interfaced to a conventional PC via a USB port, and includes 16 analog inputs, eight digital input/output ports, two analog outputs, and two counter/timers. Associated driver software is available for installation on the PC, and is compatible with National Instruments LabVIEW software, as well as Measurement Studio for Visual Studio .NET, and other programming environments. However, the invention is not limited to this particular hardware and software, and in other embodiments alternative data acquisition and control hardware may be employed, including custom hardware designed and constructed using suitable analog and digital electronic components. In accordance with the exemplary apparatus 200, the data acquisition and control card 218 is interfaced to a number of devices within the system 100. An input signal 220 may be provided by the synthesiser (not shown), indicating when the synthesised raw product is available at the input 102. This signal may be acquired and used to initiate the purification process, either under automated control, or manual control. In the latter case, the presence of the signal 220 may be used to trigger a visual indication via the user interface devices 214, informing the operator that the synthesised product is available for purification. The data acquisition and control card 218 is configured to provide output signals controlling the pump 112, the six-way rotary valve 104, the selection valve 106, and the collection valve 126. In embodiments having multiple separating columns, a column selection valve may also be provided that is able to be controlled via the data acquisition and control card 218. Inputs, for example analog voltage inputs, are received from the UV-absorption detector 122 and the radiation detector 124, which are converted into digital form on board the data acquisition and control card 218, whereby they are made available via the peripheral interface 216 for processing by the microprocessor 204 under control of program instructions 212. In general, the entire system 100 may be relevantly monitored and controlled from the microprocessor 204 via the data acquisition and control card 218, and such monitoring and control is limited only by the input and output interfaces provided by the various devices utilised within the system 100. Accordingly, embodiments of the invention that are more or less sophisticated in their degree of automation, monitoring and control, as compared with the presently preferred embodiment, may readily be implemented by appropriate selection of components, and corresponding programming. The discussion will now turn to the algorithms implemented within the apparatus 200, with reference to FIGS. 3 and 4. In particular, FIG. 3 is a flowchart 300 which illustrates system configuration and propagation time calculation in accordance with the preferred embodiment of the invention, whereas FIG. 4 is a flowchart 400 illustrating the process of collection of purified products. The system 100 has a number of parameters that are relevant to the purification process. For the purpose of describing the preferred embodiment, the relevant parameters include, without limitation, the pumping rate of the pump 112, as well as the respective interior diameters and lengths of the connecting tubes 132 and 134. For ease of reference, the diameter and length of tube 132 are hereafter denoted D1 and L1, whereas the diameter and length of tube 134 are denoted D2 and L2. The pumping rate of the pump 112 is denoted by Q, and may be measured in units such as mm3/s or ml/min. Lengths and diameters are conveniently represented in millimeters. In some embodiments of the invention, the pumping rate of the pump may be electronically controllable, and therefore subject to total control via the computer 202. In other embodiments, such as the presently preferred embodiment described herein, the pumping rate of the pump 112 is fixed, or manually selectable, and the relevant pumping rate is then preferably input to the computer 202 via the user interface devices 214. Similarly, the relevant dimensions of the tubes 132, 134 may be provided by the operator via the input devices 214. For example, the internal diameter of the tubes 132, 134 will generally be a known property of the type of tubing employed, and the relevant lengths may be measured, and this data manually entered into the computer 202. Alternatively, one or more standard configurations of tubing may be provided, each having an associated standard identifier, whereby entry of the relevant identifier into the computer 202 fully specifies the relevant diameters and lengths. In still further embodiments, various components of the system 100 may be provided as preassembled installable cartridges, and markings or other features of the cartridges, that may be either manually or computer-readable, may be used to indicate the relevant dimensions of the tubes 132, 134. Accordingly, as shown in the flowchart 300, at step 302, the computer receives relevant parameters of the system 100 either via the input devices 214, or directly by interrogation and/or control of the relevant components via the data acquisition and control card 218. Subsequently, at step 304, the computer is programmed to calculate relevant propagation times of the eluent through the tubes 132, 134. In particular, the propagation time T for eluent passing through a tube of the internal diameter D, and of length L, is given by the following equation: T = π D 2 L 4 Q By way of example, for a typical flow rate of 4 ml/min, and for ID tubing having a 1/32″ inner diameter, the propagation delay is around 7.5 s/m. Using the received system parameters, a first eluent propagation time between the UV-monitoring location and the radiation-monitoring location, T1, and a second eluent propagation time between the radioactivity monitoring location and the collection location, T2, are calculated as follows: T 1 = π D 1 2 L 1 4 Q T 2 = π D 2 2 L 2 4 Q It will be appreciated that, depending upon the system setup, and specific design choices and requirements, an alternative, but equivalent, pair of propagation delays may be calculated. For example, the radiation detector 124 may be placed prior to the UV-absorption detector 122 within the path of flow of the eluent, in which case corresponding adjustments may be made in the definitions of the calculated propagation delays. Furthermore, an alternative definition of the relevant second eluent propagation time could be based upon the total propagation time from the first (ie UV-absorption) detector 122 to the collection valve 126, and its alternative delay would be equal to the sum of T1 and T2 computed above. All such readily-implemented variations fall within the scope of the present invention. Once the system parameters have been received, and relevant propagation times calculated, purification and collection of the desired radiopharmaceutical compound may commence, in accordance with the flowchart 400. In particular, at step 402 separation commences, as previously described, by operation of the six-way rotary valve to inject the raw product into the HPLC column 118. Step 404 represents a monitoring of the UV-absorption and radioactivity of the eluent output 120 via the detectors 122, 124. Processing of the data received from the detectors 122, 124 enables a decision 406 to be made, regarding whether purified compound is available. If not, then monitoring continues. However, if the desired radiopharmaceutical compound is determined to be present in a sufficiently pure form, after an appropriate delay 408, to allow the pharmaceutical compound to reach the collection valve 126, the valve is opened at step 410. Collection of the radiopharmaceutical compound within the collection container 130 then commences, while the computer 202 continues to monitor the detectors 122, 124, as indicated at step 412. Processing of the data received from the detectors 122, 124 enables a determination as to whether passage of the purified radiopharmaceutical compound is complete, as indicated via decision step 414. If not, then monitoring continues. However, when it is determined that passage of the sufficiently pure desired compound is complete, an appropriate delay is allowed to pass, at step 416, while collection is completed, and then the collection valve 126 is closed at step 418. Suitable criteria and algorithms for determining the presence of sufficiently pure radiopharmaceutical compound, and corresponding control of the collection valve 126, will now be described. Specifically, the collection valve 126 is opened, and collection continued, during a time period which depends upon a first time period during which data received from the UV detector indicates the presence of a sufficiently pure sample of the radiopharmaceutical compound, and a second time period during which data received from the radioactivity detector indicates that a sufficiently high concentration of the radiopharmaceutical compound is present. The collection period depends upon these first and second time periods, as well as the propagation delays T1 and T2. It is accordingly a requirement to define suitable criteria for determining the commencement and duration of the relevant time periods. In general terms, a first criterion or trigger may be defined in relation to the data received from the first detector in the purification system 100, while a second criterion or trigger may be defined in relation to the data received from the second detector. When both of these criteria, conveniently denoted C1 and C2, are simultaneously satisfied, then collection of the purified radiopharmaceutical compound is indicated. It will be appreciated that in the system 100, in which the UV detector 122 precedes the radioactivity detector 124, the criterion C1 is associated with the purity of the product assessed via UV-absorption at the detector 122, while the criterion C2 is associated with levels of radioactivity detected at the detector 124. For simplicity, the following analysis is based upon the embodiment 100, although it will be appreciated that the order of the detectors 122, 124 may be reversed in alternative embodiments, with corresponding changes made as required to the analysis. In order to maximize yield, and avoid the collection of insufficiently pure product, proper account must be taken of the delays in the tubes 132, 134. It is convenient to define the collection criteria as boolean functions of time C1(t), C2(t). Each of these functions is “true” at times when the relevant criteria are satisfied, and “false” otherwise. Defining a function T(C) to be the time at which the condition (C) becomes true, it is possible to define a collection start time Tstart and a collection stop time Tstop in the following manner:Tstart=T(C1(t−T1) AND C2(t))+T2 Tstop=T( C1(t−T1) AND C2(t) C1(t−T1) AND C2(t))+T2 Tstop>Tstart It will be appreciated that the above equations require the conditions C1 and C2 to become “simultaneously” satisfied for collection to commence, taking into account the propagation delay T1 between the UV-absorption detector 122 and the radiation detector 124. Additionally, the start and stop times are further delayed in accordance with the propagation time T2 through the tube 134. If desired, a single purification run may comprise multiple collection periods, each having corresponding collection start and stop criteria and times. In a very simple implementation, the criteria C1 and C2 may be defined in terms of threshold values. For example, the UV-absorption criteria C1 may be “true” whenever the UV-absorption or transmission falls below, or rises above, a specified trigger level, or equivalently whenever the detected UV photocurrent or voltage rises above a corresponding threshold level. A similar threshold might be defined comprising the criteria C2 in relation to the radioactivity detector 124. Additional refinements may include filtering, or performing a running average, of the detected signals, in order to reduce noise, and avoid repetitive switching and/or false triggering. Alternatively, or additionally, a degree of hysteresis may be built into the criteria, such that the “turn on” threshold is set higher than the “turn off” threshold, in order to further reduce false triggering and/or repetitive switching, due to system noise. As another option, operator input may be utilized to determine the criteria C1 and C2. For example, the operator may observe the detected UV-absorption data and radioactivity data in real time, for example via a graphical display, and manually indicate when appropriate detected levels have been reached, using the user interface devices 214. The computer 202 may then automatically account for the eluent propagation times, avoiding the need for the user to mentally “align” the data from the two detectors, and determine when to open and close the collection valve 126. In another implementation of this type, the computer 202 may adjust and “synchronise” (ie delay) the display of data received from the detectors 122, 124, so that displayed waveforms correspond with “notional” detected signals that would be measured at the collection valve 126, so that the user may operate the valve 126 manually while directly observing the displayed waveforms. In accordance with a further variation, delays within the tubes 132, 134 may be sufficiently large that they substantially exceed the period of time for which the pure radiopharmaceutical compound is available for collection (ie the time period for which the collection valve 126 will be open). In this case, completed data regarding the UV-absorption and the radioactivity of the portion of the eluent containing the desired radiopharmaceutical product will be available to the computer 202 before the compound has arrived at the collection valve 126 (ie prior to Tstart). In this case, the computer 202 is able to perform a detailed analysis of the captured detector data over the relevant period of time in order to determine an optimum period during which the collection valve 126 should be open. Such analysis may include, for example, identifying relevant peaks in the UV-absorption data and/or the radioactivity data, identifying corresponding periods around the peaks during which sufficiently pure radiopharmaceutical product is present in the eluent, calculating the corresponding values of Tstart, and Tstop, and operating the collection valve 126 accordingly and/or directing an operator to do so via the interface devices 214. As will be appreciated, such processing may avoid the need to rely upon simplistic criteria, such as predetermined threshold levels, and may therefore enable greater yield of higher purity product to be collected, with a high level of repeatability, and a reduced incidence of human error. In view of the foregoing technical details, FIGS. 5 to 8 show exemplary screen shots illustrating a user interface in accordance with the presently preferred embodiment of the invention. In particular, FIG. 5 shows a control screen 500, in which a schematic diagram of the system 100 is depicted. For convenience, reference numerals in FIG. 5 correspond as appropriate, with those utilized in the system 100 depicted in FIG. 1, in order to avoid duplication of corresponding description. The operator is able to use the control screen 500 in order to observe the operation of the system 100, and to set relevant system parameters, as well as entering other pertinent information. The flow rate of the pump 112 is displayed in the text box 502. In the preferred embodiment, this value is entered by the operator to correspond with the actual flow rate of the pump utilized within the system. In alternative embodiments, this text entry box 502 may be used to set the pump flow rate, for example where the pump 112 is of a type that is able to be controlled in this manner via the control interface card 218. In still further embodiments, the computer 202 may be able to read a current pump flow rate setting from the device via the control card 218, and the displayed flow rate value may therefore be automatically obtained from the current pump configuration. A text entry box 504 is provided, within which the user may enter any information related to the current purification process. This information is subsequently saved in a file, along with other data acquired in the course of purification. The lengths and/or diameters of the lengths of tubing 132, 134 may be conveniently entered by the operator using the text boxes 506, 508. In the system under control by the computer corresponding with the display 500, there is additionally included a nitrogen supply source 510, and a corresponding valve 512, enabling the tubes feeding the connection container 130 to be evacuated with the inert gas once collection is complete. FIG. 6 shows a capture screen 600, in which graphs of detected radioactivity 602 and UV-photovoltage amplitude 604 are displayed. The graphs 602, 604 are updated under control of the computer 202, and in real time, to enable the operator to observe the radioactivity and UV-absorption data during the process of purification. In the embodiment depicted, collection of the products (by opening valve 126) is performed under user control, and is initiated by operation of the “Collect Product” button 606. Additional information provided to the operator in the capture screen 600 includes a running total of collected activity 608, and collected volume 610. FIG. 7 shows an archive screen 700 of the user interface. The archive screen is used to display previously acquired data. Such previously acquired information may be useful in assisting the operator to predict the expected product retention times during a subsequent synthesis and purification process of the same radiopharmaceutical compound. The screen 700 includes the previously recorded graphs of radioactivity 702 and UV-photovoltage amplitude 704. The time period during which collection was performed in the purification run is automatically indicated by the vertical cursors 706, 708. As can be seen, the desired product has been collected during a time period during which a radioactivity peak 710 is observed, and while simultaneously the detected UV-photovoltage amplitude is low. FIG. 8 shows a further archive screen 800 of the user interface, including previously acquired data illustrating the potential challenges of determining the most appropriate collection time period. In the purification process represented by the data in the archive screen 800, the radioactivity detector 124 precedes the UV-absorption detector 122. A peak 802, representing the presence of a high concentration of the desired product at the radioactivity detector, precedes a peak 804 in UV-transmission, representing the presence of an undesired impurity in the eluent immediately following the desired product. The nominal collection period, observed at the radioactivity detector 124, is indicated by the cursors A and B (806, 808), however this must be correctly referred to the collection valve location, taking into account the eluent propagation delays between the radioactivity detector 124 and the UV-absorption detector 122 (eg 8 seconds), and between the UV-absorption detector 122 and the collection valve 126 (eg 4 seconds). Due to the very close proximity within the eluent of the desired product and the impurity, failure to properly account for these delays may result in either contamination of the product, or reduced yield. In particular, accounting for the delay between the radioactivity detector 124 and the UV-absorption detector 122, there is an overlap between the radioactivity peak 802 and the UV-transmission peak 804 that is greater than is apparent in the data shown in the archive screen 800. It would accordingly be very difficult for an unaided human operator to collect the radiopharmaceutical product reliably and efficiently in this case. However, in accordance with an embodiment of the present invention, reliable and repeatable collection is facilitated. Finally, FIG. 9 shows a report screen 900 of the user interface. Notably, the report screen 900 corresponds with the production and purification of a different type of radiopharmaceutical tracer from that represented in the archive screen 800, and accordingly the UV-photovoltage trace is quite different in form. The report screen 900 provides a summary of the most recent purification results for future reference. In particular, the summary includes graphical representations of the radioactivity and UV-photovoltage amplitude traces, in which the relevant peaks 902, 904 have been highlighted. A peak summary table 906 is displayed in the top left hand corner of the screen 900. The vertical cursors 908 and 910 in the radioactivity and UV graphs may be manipulated by the operator to identify the desired peak periods, which are subsequently shaded as shown. The table 906 summarizes the specified cursor settings, and the computed areas under the enclosed curves. The operator is able to zoom in on selected portions of the display, and to adjust and align the graphs, in order to facilitate review and analysis of the purification results, which may be useful for improving the performance of future purification runs. In the particular example shown in FIG. 9, the observed radioactivity peak extends from 10 minutes and 45 seconds into the purification run, until 12 minutes and 15 seconds. In the purification setup, the UV detector 122 precedes the radioactivity detector 124, and the corresponding period in the UV-photovoltage amplitude trace extends from 10 minutes and 37 seconds to 12 minutes and seven seconds. Collection was performed in this example between 10 minutes and 49 seconds and 12 minutes and 19 seconds. This collection period encompasses the main portion of the radioactivity peak 902, as well as the UV transmission peak 904 which is known to correspond with the desired product. It is not clear, however, whether the UV transmission peak 912 which precedes the product peak 904, represents an unacceptable impurity presence. Taking into account the delays between the UV detector 122, the radioactivity detector 124, and the collection valve 126, it is apparent that a portion of the eluent corresponding with the peak 912 has been collected in this case. Typically, an objective of 98% purity of the collected product is established. Operation of the system 100 in accordance with an embodiment of the present invention, based upon analysis of relevant prior purification runs such as that shown in the report screen 900 in order to define appropriate collection criteria C1 and C2, is desirable to achieve reliable and repeatable purification satisfying specified purity objectives. Criteria may readily be established, for example, which would exclude all of the eluent corresponding with the peak 912 from collection. While the foregoing description has covered various exemplary features of a preferred embodiment of the invention, it will be appreciated that this is not intended to be exhaustive of all possible functions provided within various embodiments of the invention. It will be understood that many variations of the present invention are possible, and the overall scope of the invention is as defined in the claims appended hereto. |
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050664514 | description | DETAILED DESCRIPTION OF THE INVENTION In the following description, like reference characters designate like or corresponding parts throughout the several views of the drawings Also in the following description, it is to be understood that such terms as "forward", "rearward", "left", "right", "upwardly", "downwardly", and the like, are words of convenience and are not to be construed as limiting terms. PRIOR ART Referring now to the drawings, and particularly to FIG. 1, there is illustrated a prior art nuclear reactor, generally designated 10. The nuclear reactor 10 includes a reactor vessel 12 containing a core barrel 14 which houses a plurality of nuclear fuel assemblies 16 defining a nuclear core 18. The fuel assemblies 16 extend and are supported in a side-by-side parallel spaced array between upper and lower core plates 20, 22 of the core barrel 14. The nuclear reactor 10 also includes a closure head 24 enclosing the top of the reactor vessel 12 and an upper support plate 26 mounted within and across the upper portion of the vessel 12 spaced above the upper core plate 20. Further, the nuclear reactor 10 includes a plurality of guide tube assemblies 28 supported by and extending between the upper core plate 20 and the upper support plate 26 in alignment with the fuel assemblies 16. Extensions 30 of the guide tube assemblies 28 are supported by and extend between the closure head 24 and the upper support plate 26. A plurality of control rod drive mechanisms 32 mounted to the closure head 24 movably support a plurality of control rod cluster assemblies 34 (see also FIG. 31 and 32) within the guide tube assemblies 28 and their respective extensions 30. The control rod drive mechanisms 32 are operable to lower the control rod cluster assemblies 34 toward the fuel assemblies 16 so as to insert control rods 36 (FIG. 31) into the fuel assemblies 16 for reducing the power output of the core 18. The reactor vessel 12 has inlets 38 and outlets 40 (only one of each being shown in FIG. 1) through which coolant flows into and from the reactor vessel 12. The inlets 38 communicate with the exterior of the core barrel 14, whereas the outlets 40 communicate with the interior of the core barrel 16. Coolant flows into the vessel 12 through the inlets 38 and then down to the bottom of the vessel 12. The coolant then flows upwardly into the reactor core 18 through the lower core plate 20, upwardly along the interior and exterior of the fuel assemblies 16, from the reactor core 18 through the upper core plate 22, and upwardly along the interior and exterior of the guide tube assemblies 28, finally exiting from the reactor vessel 12 through the outlets 40 thereof. Referring to FIGS. 2-5, each control rod drive mechanism 32, being of the magnetic jack type, supports one control rod cluster assembly 34 (FIGS. 31 and 32) from above within one guide tube assembly 28 (FIG. 33). The control rod drive mechanism 32 basically includes an elongated pressure housing 42, an elongated drive rod shaft 44, movable and stationary gripper assemblies 46, 48, and an electromagnetic actuating assembly 50. The pressure housing 42 is formed of non-magnetic material and has an elongated central passage 52 defined by an elongated central guide sleeve 54 of the housing. The housing 42 is threaded and hermetically sealed to the reactor vessel 12 and supported by the closure head 24 thereof. The drive rod shaft 44 of the mechanism 32 extends through the hollow passage 52 of the housing 42. The drive rod shaft 44 has multiple ring-like grooves 56 axially spaced apart from one another by predetermined equal distances and circumferentially formed about the shaft 44. By way of example, the grooves 56 are defined on the shaft 44 at 5/8" intervals. One control rod cluster assembly 34 (FIGS. 31 and 32) is connected to the lower end of each drive rod shaft 44, as can be seen in FIG. 1. The number of grooves 56 provided on the rod shaft 44 are sufficient so that the attached control rod cluster assembly 34 can be positioned at any location within the associated fuel assembly 16 at 5/8" intervals or "steps". The total travel of the control rod cluster assembly 34 is 144 inches or 228 steps. The movable and stationary gripper assemblies 46, 48 of the mechanism 32 are mounted at the interior of the housing 42. As seen in FIGS. 4 and 5, each gripper assembly 46, 48 includes a support tube 58 attached to the central guide sleeve 54 and a set of latch and linkage arrangements 60. Each set contains three such arrangements 60 angularly positioned 120 degrees apart about the drive rod shaft 44. Referring to FIGS. 6-11, each latch and linkage arrangement 60 is composed of a latch 62 and a link 64. An upper pivot pin 66 pivotally mounts the latch 62 adjacent its upper end to its respective support tube 58 such that the latch is positioned through longitudinal slots defined in the support tube 58 and central guide sleeve 54 to permit pivotal movement of the latch 62 relative to the drive rod shaft 44. The lower pivot pin 68 pivotally connects the latch 62 adjacent its partially bifurcated lower end to an inner end of the link 64. The partial bifurcation of the lower end of the latch 62 defines a recess 62A for receiving the inner end of the link 64 and contains aligned holes 62B for receiving the lower pin 68 which connects the latch 62 with the link 64. A flat key 70 is also installed in slots in the support tube 58 and guide sleeve 54 for preventing the support tube 58 from rotating relative to the housing 42 and guide sleeve 54. The latch 62 also has an inwardly projecting tooth 72 defined on its lower end. The tooth 72 has an outer surface 72A with an arcuate shaped profile for mating with and engaging a portion of the drive rod shaft groove 56. In FIG. 6, the latch 62 and link 64 of the arrangement 60 are shown extended toward the drive rod shaft 44 with the latch tooth 72 engaged in one of the circumferential grooves 56 on the shaft 44. FIG. 7 shows the latch 62 and link 64 of the arrangement 60 retracted from the drive rod shaft 44 with the latch tooth 72 disengaged from the one circumferential groove 56 on the shaft 44. As seen in FIGS. 2, 3 and 12-17, the electromagnetic actuating assembly 50 of the mechanism 32 has components mounted respectively at the exterior and interior of the pressure housing 42. In particular, the actuating assembly 50 includes three independent annular electromagnetic lift coils C mounted about the exterior of the housing 42, namely, an upper coil 74, a middle coil 76 and a lower coil 78. Annular flux rings (not shown) are mounted about the housing 42 radially inward from the coils. The actuating assembly 50 also includes annular upper and lower poles 80, 82 fixedly mounted about the central guide sleeve 54 of the housing 42 in axially spaced relation to one another, and annular upper, middle and lower armatures 84, 86, 88 slidably mounted about the central guide sleeve 54. Upper, middle and lower armature return coil springs 90, 92, 94 encircle the guide sleeve 54. The upper return spring 90 is disposed between the upper fixed pole 80 and upper movable armature 84. The middle return spring 90 is disposed between the upper movable armature 84 and middle movable armature 86. The lower return spring 94 is disposed between the lower fixed pole 82 and lower movable armature 88. A load transfer return spring 96 is disposed between the lower movable armature 88 and the support tube 58 of the stationary gripper assembly 48. The links 64 of the latch and linkage arrangements 60 of the movable and stationary gripper assemblies 46, 48 are pivotally connected at their outer ends by pivot pins 98 to the respective middle and lower movable armatures 86, 88. Thus, the control rod drive mechanism 32 is a three-coil, electromagnetic jack which raises and lowers the control rod cluster assembly 34 via the drive rod shaft 44. The three coils 74, 76, 78, mounted outside the pressure housing 42, actuate the movable armatures 84, 86, 88 contained within the housing 42. The movable armatures 84, 86, 88 operate the latches 62 of the movable and stationary gripper assemblies 46, 48 which grip the grooved drive rod shaft 44. The latches 62 of the lower stationary gripper assembly 48 are used to hold the drive rod shaft 44 in a desired stationary position. The latches 62 of the upper movable gripper assembly 46, which are raised and lowered by the upper movable armature 84, are used to raise and lower the drive rod shaft 44. Each step of the mechanism 32 moves the drive rod shaft 5/8" (1.58 cm). More particularly, the latches 62 of the movable and stationary gripper assemblies 46, 48 are actuated and the drive rod shaft 44 is moved vertically by the coordinated operation of the coils 74, 76, 78, poles 80, 82 and armatures 84, 86, 88. When the coils 74, 76, 78 are energized, a magnetic flux field is created which passes through the nonmagnetic pressure housing 42 and couples with the fixed poles 80, 82. Force sufficient to provide vertical motion of the movable armatures 84, 86, 88 is obtained by the solenoid principle. Referring to FIGS. 12-17, there is illustrated a sequence of steps in raising the drive rod shaft 44 of the prior art mechanism 32 in stepping fashion relative to the pressure housing 42. During normal steady state operation of the reactor 10, the drive rod shaft 44 is held in a stationary position, as shown in FIG. 12. In this mode, called the "hold" mode, only the lower coil 78 is energized. When energized, the lower coil 78 raises the lower armature 88 causing inward pivoting of the lower set of latches 62 of the stationary gripper assembly 48 to engage the drive rod shaft 44. If the current to the lower coil 78 is interrupted, either by choice or by electrical malfunction, the lower armature 88 is released and the latches 62 of the stationary gripper assembly 48 are pivoted to disengage from the drive rod shaft 44, permitting the drive rod shaft to drop and thereby insert the control rod cluster assembly 34 into the nuclear core 18. This same action applies during all phases of the operation of the mechanism 32 and provides for quick shutdown of the reactor 10. The repositioning, or stepping, of the drive rod shaft 44 is most easily understood by description of the sequence of events which occur during a rod withdrawal step, as shown in FIGS. 13-17. The withdrawal step begins by increasing the current in the lower coil 78. The middle coil 76 is then energized causing the middle armature 86 to raise and the upper set of latches 62 of the movable gripper assembly 46 to cam radially into a position where they can engage the drive rod shaft 44. At this point, however, they are located slightly less than one-sixteenth of an inch below a drive rod groove 56. When the current in the lower coil 78 is then lowered, the stationary gripper assembly 48 drops approximately one-sixteenth of an inch, transferring the load of the drive rod shaft 44 from the latches 62 of the stationary gripper assembly 48 to the latches of the movable gripper assembly 46, as seen in FIG. 13. Continued reduction of lower coil current causes the latches 62 of the stationary gripper assembly 48 to cam outward from the drive rod shaft groove 56 to the position seen in FIG. 14. This action is called the load transfer function and assures that neither set of latches 62 is carrying the full weight of the drive rod shaft 44 during camming in or camming out operations. As seen in FIG. 14, the drive rod shaft 44 is now being held by the latches 62 of the movable gripper assembly 46 only. The upper, or lift, coil 74 is now energized. This results in raising the upper armature 84 and the movable gripper assembly 46 connected therewith through a step of 5/8 of an inch, raising the drive rod shaft 44 and attached control rod cluster assembly 34 through the one step, from the position of FIG. 14 to that of FIG. 15. The lower coil 78 is then energized, raising the lower armature 88 as seen in FIG. 16. The latches 62 of the stationary gripper assembly 48 then pull up approximately one-sixteenth of an inch, picking up the drive rod shaft load from the latches 62 of the movable gripper assembly 46. The middle coil 76 is then de-energized and the latches 62 of the movable gripper assembly 46 cam out as seen in FIG. 17. Finally, as also seen in FIG. 17, the lift coil 74 is de-energized and the movable gripper assembly 46 is lowered to its normal position Referring to FIGS. 18 and 19, there is illustrated in detail the prior art control rod cluster assembly 32 which is raised and lowered in the prior art guide tube assembly 28 by carrying out the above-described stepping action of the drive rod shaft 44 of the control rod drive mechanism 32. The control rod cluster assembly 32 includes a plurality of the control rods 36 and a spider 100 having radially extending flukes 102 connected to the upper ends of the control rods 36. The control rods 36 extend downwardly from the spider 100 and generally parallel to one another. The control rods 36 are arranged in a pattern matched to that of guide thimbles (not shown) in the one of the fuel assemblies 16 above which the cluster assembly 32 is aligned. The spider 100 also has a central cylindrical member 104 by which it is coupled to the lower end of the drive rod shaft 44 of the control rod drive mechanism 32. FIG. 20 shows the control rod cluster assembly 32 mounted in the prior art guide tube assembly 28. The guide tube assembly 28 includes an outer tubular housing 106 and a plurality of longitudinally slotted guide tubes 108 mounted internally of the housing by brace plates 110. The slotted guide tubes 108 are arranged in a pattern matched to that of the control rods 36 of the cluster assembly 32 so that the control rods 36 are slidably raised and lowered within the guide tubes 108 by operation of the control rod drive mechanism 32. FIG. 21 illustrates an end 36A of one control rod 36 positioned in one guide tube 108. It is in this region that a wear interface 110 develops between the walls of the control rod 36 and guide tube 108 due to rubbing therebetween caused by hydraulic forces imposed on the control rod 36 by coolant flow upward through the guide tube 108. FIG. 22 diagrammatically depicts the cluster assembly 34 with its control rods 36 positioned in the guide tubes 108 of the guide tube assembly 28. In FIG. 22, the initial or starting position of the control rods 36 is shown at the start of the fuel cycle of the nuclear reactor 10. In FIG. 23, the position of the control rods 36 is shown after repositioning of the cluster assembly 34 three steps at one time at the end of the fuel cycle, which typically is a twelve month period, in preparation for the next fuel cycle. The control rods 36 are repositioned by operation of the control rod drive mechanism 32, as described earlier, so as to move the drive rod shaft 44 successively through three steps. Although the three-step repositioning scheme of the prior art avoids the problems caused by random misstepping of the control rod drive mechanism 32, the drawback of the three-step repositioning scheme is that the wear is still poorly distributed throughout the total available clad thickness of the control rod. As an example, assume that a particular nuclear plant has a control rod wear rate that uses 60% of the permissible wear thickness in one fuel cycle. With the three-step repositioning scheme, the control rods would be relocated every fuel cycle so that the permissible minimum level of wear thickness is not exceeded. This effectively "wastes" 40% of the wear thickness at that elevation of the control rods. IMPROVEMENTS FIGS. 24-27 diagrammatically depict the sequence of stages in the method of repositioning the cluster assembly 34 relative to the guide tube assembly 28 in accordance with the present invention by a single step at each of a plurality of times during each fuel cycle. In the example illustrated, the control rod drive mechanism 32 is operated to move the drive rod shaft 44 and thus the cluster assembly 34 a single step at three separate times. In the case of a twelve month fuel cycle, the single-step repositioning of the cluster assembly 34 might be carried out every month, or twelve times, during a one year fuel cycle, or carried out three times, at four, eight and twelve months, during the twelve month fuel cycle. As stated above, the single-step repositioning of the cluster assembly 34 three times during a single fuel cycle is shown respectively in FIGS. 25, 26 and 27. By increasing the frequency of the repositioning to more than once per fuel cycle, the wear exposure time due to a misstep is decreased, and the consequences of a misstep are reduced proportionately. The more frequent repositioning better utilizes the control rod clad thickness and/or the wall thickness of the guide tube available for wear. For example, if the control rod cluster assembly is repositioned three times per fuel cycle, the wear would be 20% of the allowable wear per repositioning. Eventually, each wear location could be used five times. With a wear of 20% per repositioning, 100% of the permissible wear thickness would be utilized. Referring to FIGS. 28-36, there is illustrated one embodiment of an improved latch and linkage arrangement, generally designated 112, in accordance with the invention of the cross-referenced patent application. As mentioned in the background section supra, the prior art latches 62 each have either one or a pair of teeth 72 which engage within either a single groove or a pair of adjacent of grooves 56 in the drive rod shaft 44. Because the latches 62 pivotally move in arcuate paths toward and away from the drive rod shaft 44, the teeth 72 are placed at the end of the latch 62 opposite from the pivotal axis of the latch. This latch teeth placement results in the generation of a moment load through the latch which over time tends to cause cracking at the root of the teeth and eventual failure of the latch. The improved latch and linkage arrangement 112 employs a parallel linkage which permits a latch 114 to move along a curvilinear path toward and away from a series of circumferential grooves on the drive rod shaft of the control rod drive mechanism. Because the mounting geometry of the improved arrangement produces uniform movement of the latch 114 toward and away from the drive rod shaft at every point along the length of the latch, a plurality of teeth 116 can be deployed over the entire length of the latch 114 which minimizes the bending moment load through the body 118 of the latch and increases the wear capability of the latch. The elongated body 118 of the latch 114 has a pair of opposite ends 118A, 118B, a pair of opposite side edges 118C, 118D and a pair of opposite faces 118E, 118F. The plurality of latch teeth 116 are defined along the one side edge 118C, extending in a row between the opposite ends 118A, 118B of the body. A slot 120 is defined into one end 118A of the latch body 118 and a pair of recesses 122A, 122B are defined in the opposite faces 118E, 118F of the body. A pair of longitudinally-spaced transversely-extending upper and lower holes 124A, 124B are formed through the latch body 118 opening within the recesses 122A, 122B and the upper hole 124A intersecting with the slot 120. The improved latch and linkage arrangement 112 also includes a plurality of links 126, 128 and 130. Each link has a hole 132 through it at each of its opposite ends. A first link 126 is pivotally connected to the latch body 118 by a first pivot pin 134 through the upper hole 124A and pivotally connected to the respective one of the middle and lower armatures 86, 88 of the actuating assembly 50 which is the same as in the prior art. These armatures apply the motive force longitudinally of the latch body 118 between its opposite ends 118A, 118B via the first link 126. Second and third pairs of links 128, 130 are pivotally connected to the latch body 118 and to the respective support tube 58 by second and third sets of pivot pins 136, 138 in a parallel relation to one another so as to define a parallelogram therewith such that when the motive force is applied to the latch body 118 via the first link 126, the latch body 118 and thereby the latch teeth 116 undergo curvilinear movement toward and away from the engaging or latching position with the plurality of drive rod shaft grooves 56. Except for the composition of the latch and linkage arrangement 112, the movable and stationary gripper assemblies 46, 48 of the mechanism 32 are the same as described in the prior art above. Thus, there are three sets of the arrangements 112 angularly positioned 120 degrees apart about the drive rod shaft 44. In FIGS. 28 and 29, the arrangement is seen in respective engaged and disengaged positions relative to a plurality of the circumferential grooves 56 on the drive rod shaft 44. FIGS. 37 and 38 show a second embodiment of the improved latch and linkage arrangement. The second embodiment is the same as the first embodiment, except as follows. Instead of the recesses 122A, 122B in the opposite faces 118E, 118F of the latch body 118, the latch body 140 in the second embodiment has the slot 120 extending longitudinally between the opposite ends 140A, 140B of the latch body. Also, a separate hole 124C is provided for the first link 126 and only one each of second and third links 128, 130 are employed. FIGS. 39 and 40 illustrate a third embodiment of the improved latch and linkage arrangement. The third embodiment is the same as the second embodiment, except as follows. The first link 142 has a bifurcated end by which it is pivotally mounted to the latch body 144. It is thought that the present invention and many of its attendant advantages will be understood from the foregoing description and it will be apparent that various changes may be made in the form, construction and arrangement thereof without departing from the spirit and scope of the invention or sacrificing all of its material advantages, the form hereinbefore described being merely a preferred or exemplary embodiment thereof. |
abstract | IR emission devices comprising an array of polaritonic IR emitters arranged on a substrate, where the emitters are coupled to a heater configured to provide heat to one or more of the emitters. When the emitters are heated, they produce an infrared emission that can be polarized and whose spectral emission range, emission wavelength, and/or emission linewidth can be tuned by the polaritonic material used to form the elements of the array and/or by the size and/or shape of the emitters. The IR emission can be modulated by the induction of a strain into a ferroelectric, a change in the crystalline phase of a phase change material and/or by quickly applying and dissipating heat applied to the polaritonic nanostructure. The IR emission can be designed to be hidden in the thermal background so that it can be observed only under the appropriate filtering and/or demodulation conditions. |
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abstract | A digital reordering unit, an ultrasonic front-end device and operating method thereof are provided. The ultrasonic front-end device may be connected between a probe and a detector of the ultrasonic system and controlled by a primary controller of the ultrasonic system; the ultrasonic front-end device having an ultrasonic transmission part and an ultrasonic reception part, wherein the ultrasonic transmission part includes a transmission beamformer and M transmission driving units, and has M transmission channels; the ultrasonic reception part includes M high-voltage isolation circuits, RC amplifiers, RC ADCs and a beamformer electrically connected in said order and has RC reception channels, where RC=[N,2N,3N . . . p*N], N being an integer larger than or equal to 1, being characterized in that, M low-voltage analog switches and a network of resistors are serially connected between the M high-voltage isolation circuits and the RC amplifiers. |
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claims | 1. A generation system having a natural circulation reactor comprising:a reactor pressure vessel of the natural circulation reactor having a core with a chimney disposed above the core, the core and the chimney being disposed in the reactor pressure vessel;a feed water pipe connected to the reactor pressure vessel and configured to supply feed water which flows therein to the reactor pressure vessel;a feed water temperature detection apparatus disposed at the feed water pipe and configured to detect a temperature of the feed water in the feed water pipe;a power controller section which outputs a power adjustment demand signal so as to control reactor power in accordance with the detected temperature of the feed water in the feed water pipe which is detected by the feed water temperature detection apparatus; anda control rod driving section which controls the reactor power by inserting and withdrawing a control rod with respect to the core in accordance with the power adjustment demand signal from the power controller section. 2. The generation system to claim 1, wherein the feed water temperature detection apparatus has a pair of ultrasonic sensors disposed at a position outside of the reactor pressure vessel so as to enable detection of the temperature of the feed water in the feed water pipe in proximity thereto. 3. The generation system according to claim 1, wherein the feed water temperature detection apparatus has a plurality of pairs of ultrasonic sensors disposed at a position outside of the reactor pressure vessel so as to enable detection of the temperature of the feed water in the feed water pipe in proximity thereto. 4. The generation system according to claim 3, wherein the plurality of the pairs of the ultrasonic sensors enable detection of a temperature distribution of the feed water in the feed water pipe in proximity thereto. 5. The generation system according to claim 3, wherein the plurality of the pairs of the ultrasonic sensors are disposed on an outer peripheral portion of said feed water pipe so as to be parallel to each other. 6. The generation system according to claim 5, wherein the plurality of the pairs of the ultrasonic sensors are disposed so as to have a prescribed interval between them. 7. The generation system according to claim 3, wherein the plurality of the pairs of the ultrasonic sensors are disposed on an outer peripheral portion of the feed water pipe so as to cross each other. 8. The generation system of claim 1, wherein the plurality of the pairs of the ultrasonic sensors are disposed at a position outside of and in the vicinity of the reactor pressure vessel so as to enable detection of the temperature of the feed water in the feed water pipe in proximity thereto. 9. The generation system according to claim 1, wherein the feed water pipe is arranged externally of the reactor pressure vessel. 10. The generation system according to claim 9, wherein the feed water temperature detection apparatus is disposed with respect to the feed water pipe at a position externally of said reactor pressure vessel so as to enable detection of the temperature of the feed water in the feed water pipe in proximity thereto. 11. A generation system according to claim 1, wherein the feed water pipe is connected to a condenser so as to enable supply of the feed water which flows within the feed water pipe from the condenser to the reactor pressure vessel. 12. A generation system according to claim 1, wherein the feed water temperature detection apparatus is disposed at a position of the feed water pipe externally of the reactor pressure vessel and adjacent thereto so as to detect the temperature of the feed water in the feed water pipe immediately before the feed water flows into the reactor pressure vessel. 13. A generation system according to claim 1, wherein the feed water pipe is connected to the reactor pressure vessel so as to supply feed water to an interior of the reactor pressure vessel without extending within the reactor pressure vessel to an inlet of the core within the reactor pressure vessel. 14. A generation system according to claim 1, wherein the feed water pipe supplies feed water to an interior of the reactor pressure vessel so that the feed water is mixed with water within the reactor pressure vessel and the temperature of the feed water which flows in the feed water pipe to the reactor pressure vessel is different from a temperature of the feed water which is mixed with the water within the reactor pressure vessel at least at an inlet of the core. |
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abstract | An electron exit window foil for use with a high performance electron beam generator operating in a corrosive environment is provided. The electron exit window foil comprises a sandwich structure having a film of Ti, a first layer of a material having a higher thermal conductivity than Ti, and a flexible second layer of a material being able to protect said film from said corrosive environment, wherein the second layer is facing the corrosive environment. |
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summary | ||
abstract | The invention concerns a method for repairing guide rails (27) of an assembly (10) radially maintaining a core supporting plate (5), which consists in measuring the spacing between the spans of the lateral branches of the guide rail (27) to be repaired, cutting and removing at least one lateral branch of said guide rail (27) to be repaired, measuring the dimensions of said lateral branch and machining with identical dimensions at least one substitution lateral branch, mounting and fixing said substitution lateral branch and measuring the spacing between the spans of the lateral branches of the repaired guide rail. |
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description | This invention relates to electromagnetic wave-absorbing compositions having a good electromagnetic wave absorbing ability and a high breakdown voltage. With the ever-increasing utilization of electromagnetic waves in broadcasting, mobile communication, radar, cellular phones, wireless local area networks (LAN) and other systems, more electromagnetic waves are scattering in the ambient atmosphere. This situation frequently gives rise to the problems of electromagnetic wave disturbance and electronic equipment malfunction. With the advance toward a higher density and higher integration of central processor units (CPU), micro processor units (MPU), large scale integrated circuits (LSI) other components used in electronic equipment such as personal computers and mobile phones as well as higher density mounting of electronic components on printed circuit boards, there arise problems that electromagnetic waves are radiated in the interior of equipment and reflected thereby so that the interior is full of electromagnetic waves. Electromagnetic interference can occur with the electromagnetic wave emitted by the equipment itself. In the prior art, an artisan with specialized knowledge and experience of noise suppression must be engaged in taking a countermeasure against disturbances by electromagnetic interference. It is a time-consuming task to find an effective countermeasure. Another drawback is that a guarded component requires an extra space for mounting. To solve these problems, engineers are interested in electromagnetic absorbers which absorb electromagnetic waves for thereby reducing reflected and transmitted waves. The drive toward higher density and higher integration of electronic equipment components such as CPU, MPU and LSI entails increased amounts of heat release. Ineffective cooling will allow thermal runaway, causing malfunction. One typical means for effectively radiating heat to the exterior is to place silicone grease or silicone rubber filled with heat conductive powder between electronic components (e.g., CPU, MPU and LSI) and heat sinks to reduce contact thermal resistance. This means, however, cannot avoid the problem of electromagnetic interference within the equipment interior. Therefore, members having electromagnetic wave absorbing and heat transfer abilities are needed for high density and highly integrated components (e.g., CPU, MPU and LSI) within electronic equipment. Depending on the necessary situation, the state-of-the-art makes a choice among three types: (1) sheets of a base polymer with a magnetic powder dispersed therein having only an electromagnetic wave absorbing ability, (2) sheets of a base polymer with a heat conductive powder (like alumina) dispersed therein having only a heat transfer ability, and (3) sheets filled with both the powders having both electromagnetic wave absorbing and heat transfer abilities. In these years, the signal processing speed of personal computers and other electronic equipment is drastically increasing. Many devices have an operating frequency of several hundred MHz to several GHz. Then electromagnetic noises generating in electronic equipment often have frequencies in the GHz band. To suppress such electromagnetic noises, the use of sheets having a spinel type cubic ferrite powder (typically manganese zinc base ferrite or nickel zinc base ferrite) uniformly dispersed in a base polymer may be useful. These ferrite sheets are effective mainly in the MHz band, but less in the GHz band. Then, sheets having uniformly dispersed in a base polymer a metal base soft magnetic powder which is more effective in the GHz band become the mainstream shield used at present. Since soft magnetic metals are generally electrically conductive, sheets having such powder dispersed in a base polymer have a low breakdown voltage. Then, when a sheet is mounted within an electronic equipment, care must be taken so as to avoid short-circuits between individual parts with which the sheet can come in contact. On use, the sheet having both electromagnetic wave absorbing and heat transfer abilities is often sandwiched between a device and a heat sink. The sheet cannot be used if electrical connection between a device and a heat sink is a problem. In this situation, an electrically insulating sheet having only a heat transfer ability is sandwiched between a device and a heat sink for dissipating the heat from the device. At the same time, a sheet having only an electromagnetic wave absorbing ability is placed at a nearby area where no electrical problem occurs, for suppressing electromagnetic noise. The use of two types of sheet is cumbersome. Most of electromagnetic noise-generating sites within electronic equipment are high speed operating devices such as CPU, MPU and LSI, whereas pins or legs for connecting the device to the pattern of a printed circuit board and the printed circuit pattern itself can act as antennas to generate electromagnetic noises. In the latter case, it is preferred that an electromagnetic wave-absorbing material be directly mounted at the noise generating site. However, a material having uniformly dispersed in a base polymer a metal base soft magnetic powder which is fully effective to noise in the GHz band cannot be used because it causes short-circuits. Basically, in the material having a metal base soft magnetic powder uniformly dispersed in a base polymer, the electroconductive soft magnetic metal particles are electrically insulated from each other since the base polymer is electrically insulating. To enhance the electromagnetic wave absorbing ability, the material must be heavily loaded with soft magnetic metal powder. As a consequence, metal particles are spaced only a close distance and can even be brought into contact, resulting in the composition having a lower breakdown voltage. JP-A 11-45804 describes an electromagnetic wave-absorbing material in which a soft magnetic metal powder is covered on the surface with an insulating coating of a silane coupling agent. JP-A 2001-308584 describes an electromagnetic wave-absorbing material in which a soft magnetic metal powder is covered on the surface with an insulating coating of a long-chain alkyl silane. The coatings of molecules having such organic radicals are difficult to provide an electromagnetic wave-absorbing composition with a satisfactory breakdown voltage. Therefore, an object of the present invention is to provide an electromagnetic wave-absorbing composition having an enhanced electromagnetic wave-absorbing ability and a satisfactory breakdown voltage. Another object of the present invention is to provide an electromagnetic wave-absorbing composition having both enhanced electromagnetic wave-absorbing and heat transfer abilities as well as a satisfactory breakdown voltage. It has been found that when magnetic powder particles coated with electrically insulating inorganic fines are dispersed in a base polymer, there is obtained an electromagnetic wave-absorbing composition having an enhanced electromagnetic wave-absorbing ability and a satisfactory breakdown voltage. It has also been found that by adding a heat conductive powder to the above composition, the electromagnetic wave-absorbing composition is endowed with both enhanced electromagnetic wave-absorbing and heat transfer abilities as well as a satisfactory breakdown voltage. In a first aspect, the invention provides an electromagnetic wave-absorbing composition in which a magnetic powder coated with electrically insulating inorganic fines is dispersed in a base polymer. In a second aspect, the invention provides an electromagnetic wave-absorbing composition in which a magnetic powder coated with electrically insulating inorganic fines and a heat conductive powder are dispersed in a base polymer. The electromagnetic wave-absorbing composition of the invention is arrived at by uniformly dispersing a magnetic powder coated with electrically insulating inorganic fines and optionally, a heat conductive powder in a base polymer. The magnetic powder with which the electromagnetic wave-absorbing composition is filled is not critical as long as it can absorb electromagnetic waves. Among others, soft magnetic metal materials are useful. Since soft magnetic metal materials are electrically conductive, a coating of electrically insulating inorganic fines thereon is effective for significantly improving the breakdown voltage of the composition. From the standpoints of constant supply and cost, metal materials containing iron element are preferred as the soft magnetic metal powder. Examples include carbonyl iron, electrolytic iron, Fe—Cr base alloys, Fe—Si base alloys, Fe—Ni base alloys, Fe—Al base alloys, Fe—Co base alloys, Fe—Al—Si base alloys, Fe—Cr—Si base alloys, and Fe—Si—Ni base alloys, but are not limited thereto. Those metal materials containing at least 15% by weight of iron element are preferred from the economical standpoint. The magnetic powder may be used alone or as a combination of two or more. The powder particles may be either tabular or granular in shape or both. Preferably the magnetic powder (particles) has an average particle size of 0.1 to 100 μm, more preferably 1 to 50 μm. Particles with an average particle size of less than 0.1 μm have too large a specific surface area, which may prohibit heavy loading. If particles have an average particle size of more than 100 μm, they may develop minute asperities on the surface of the cured electromagnetic wave-absorbing composition, to increase the contact thermal resistance, which is undesirable when heat transfer is required. The magnetic powder particles are coated with electrically insulating inorganic fines, which are selected from oxides such as alumina, silica, titanium oxide and ferrite, nitrides such as silicon nitride, boron nitride and aluminum nitride, and carbides such as silicon carbide, but not limited thereto. The magnetic powder coated with electrically insulating inorganic fines is sometimes referred to as “coated magnetic powder.” The coating weight (or buildup) of electrically insulating inorganic fines on the magnetic powder is preferably 0.5 to 20 parts by weight, more preferably 1 to 10 parts by weight per 100 parts by weight of the magnetic powder. Any desired method may be employed in coating the magnetic powder particles with electrically insulating inorganic fines. For instance, an RF thermal plasma method is useful. Another method involves the steps of dispersing submicron fines of alumina or the like in an organic solvent such as alcohol to form a dispersion, immersing magnetic powder particles in the dispersion, uniformly agitating the dispersion, and evaporating off the organic solvent, whereby the magnetic powder particles having submicron fines borne thereon are left behind. Since electrically insulating inorganic fines have a strong cohesion force to magnetic particles, the radio frequency (RF) thermal plasma method is advantageously used. Referring to FIG. 1, the RF thermal plasma method is described. A mixture of argon gas and hydrogen gas from an argon gas source 1 and a hydrogen gas source 2 is fed into a chamber 7. Art RF power supply 6 conducts an RF current at a frequency of 0.5 to 40 MHz, typically 4 MHz to a copper coil 5 wrapped around the chamber 7 for generating an RF thermal plasma arc 4 within the chamber 7. To the RF thermal plasma arc 4, electrically insulating inorganic fines to become an inorganic insulating coating are fed from its source 3 using argon gas as a carrier gas. Then the electrically insulating inorganic fines are once gasified in the plasma and then agglomerate and deposit on surfaces of magnetic particles which are pneumatically fed from a magnetic powder source 8. The magnetic particles coated with the electrically insulating inorganic fines are recovered in a reservoir 9 under suction of a vacuum pump 10. In the electromagnetic wave-absorbing composition, the magnetic powder coated with electrically insulating inorganic fines is preferably contained in an amount of 5 to 80% by volume, more preferably 20 to 70% by volume of the entire composition. Less than 5 vol % of the coated magnetic powder may fail to achieve a satisfactory electromagnetic wave absorbing ability whereas loading of more than 80 vol % of the coated magnetic powder may render the composition brittle. When the electromagnetic wave-absorbing composition is used at a site where heat transfer is required, it is recommended to further incorporate a heat conductive powder in the composition in addition to the coated magnetic powder, for endowing the composition with an enhanced heat transfer ability. The heat conductive powder is selected from metals such as copper and aluminum, metal oxides such as alumina, silica, magnesia, red iron oxide, beryllia, and titania, metal nitrides such as aluminum nitride, silicon nitride and boron nitride, and silicon carbide, but not limited thereto. Of these, electrically non-conductive ones are preferred. Preferably the heat conductive powder (particles) has an average particle size of 0.1 to 100 μm, more preferably 1 to 50 μm. Particles with an average particle size of less than 0.1 μm have too large a specific surface area, which may prohibit heavy loading. If particles have an average particle size of more than 100 μm, they may develop minute asperities on the surface of the cured electromagnetic wave-absorbing composition, to increase the contact thermal resistance, which is undesirable when heat transfer is required. The heat conductive powder is used herein for achieving closest packing with the coated magnetic powder and for increasing a thermal conductivity. In the electromagnetic wave-absorbing composition, the amount of the heat conductive powder loaded is preferably 5 to 80% by volume of the entire composition. Additionally, the amount of the heat conductive powder and the coated magnetic powder combined is preferably 10 to 90% by volume, especially 30 to 80% by volume of the entire composition. If the amount of the heat conductive powder and the coated magnetic powder combined is less than 10 vol %, the composition may not have a satisfactory thermal conductivity. If the same amount is more than 90 vol %, the composition may become brittle. In the embodiment having the heat conductive powder compounded, a sheet formed of the electromagnetic wave-absorbing composition should preferably have a thermal conductivity of at least 1 W/mK, especially at least 2 W/mK. In the electromagnetic wave-absorbing composition of the invention, the base polymer may be a thermosetting resin, thermoplastic resin, rubber or the like, but is not limited thereto. A suitable base polymer for the intended application may be selected from such materials. Illustrative examples of the base polymer include organopolysiloxane, acrylic resins, chlorinated polyethylene, polyethylene, polypropylene, ethylene-propylene copolymers, polyvinyl chloride, fluoro rubber, and urethane resins. In the application where heat transfer is required, in order to reduce the contact thermal resistance between the electromagnetic wave-absorbing composition and a heat generating object and/or a heat sink, the electromagnetic wave-absorbing composition of the invention should preferably be flexible enough to conform to minute asperities on the surface of the member. In this event, it is preferred to use as the base polymer an organopolysiloxane which is easy to adjust the hardness of the electromagnetic wave-absorbing composition and heat resistant. Suitable compositions using an organopolysiloxane as the base polymer include, but are not limited to, unvulcanized patty silicone resin compositions, silicone gel compositions comprising a curable organopolysiloxane as the base polymer, silicone rubber compositions of the addition reaction type, and silicone rubber compositions of the peroxide crosslinking type. It is noted that the composition comprising a curable organopolysiloxane as the base polymer, in the cured state, preferably has a rubber hardness of up to 80, especially up to 70 in Asker C hardness. In the unvulcanized patty silicone, silicone rubber, and silicone gel compositions described above, the base polymer may be any well-known organopolysiloxane. Typically the organopolysiloxane used herein has the average compositional formula (1) below.R1aSiO(4−a)/2 (1) In formula (1), R1, which may be the same or different, stands for substituted or unsubstituted monovalent hydrocarbon radicals, preferably having 1 to 10 carbon atoms, more preferably 1 to 8 carbon atoms, for example, unsubstituted monovalent hydrocarbon radicals including alkyl radicals such as methyl, ethyl, isopropyl, butyl, isobutyl, tert-butyl, hexyl and octyl; cycloalkyl radicals such as cyclohexyl; alkenyl radicals such as vinyl and allyl; aryl radicals such as phenyl and tolyl; aralkyl radicals such as benzyl, phenylethyl and phenylpropyl; and substituted monovalent hydrocarbon radicals including the foregoing radicals in which some or all of the hydrogen atoms attached to carbon atoms are substituted with halogen atoms, cyano and other radicals, for example, halogenated alkyl radicals and cyano-substituted alkyl radicals such as chloromethyl, bromoethyl and cyanoethyl. Of these, methyl, phenyl, vinyl and trifluoropropyl radicals are preferable. More preferably methyl accounts for at least 50 mol %, especially at least 80 mol % of the R1 radicals. The subscript “a” is a positive number from 1.98 to 2.02. Preferably the organopolysiloxane has at least two alkenyl radicals per molecule, especially with the alkenyl radicals accounting for 0.001 to 5 mol % of the R1 radicals. The organopolysiloxane of formula (1) may have any molecular structure and is preferably blocked at ends of its molecular chain with triorganosilyl radicals or the like, especially diorganovinylsilyl radicals such as dimethylvinylsilyl. In most cases, the organopolysiloxane is preferably a linear one. A mixture of two or more different molecular structures is acceptable. The organopolysiloxane preferably has an average degree of polymerization of about 100 to 100,000, especially about 100 to 2,000, and a viscosity of about 100 to 100,000,000 centistokes at 25° C., especially about 100 to 100,000 centistokes at 25° C. When the above silicone rubber composition is formulated to the addition reaction curing type, the organopolysiloxane is one having at least two alkenyl radicals such as vinyl radicals per molecule, and the curing agent is a combination of an organohydrogenpolysiloxane and an addition reaction catalyst. The organohydrogenpolysiloxane is preferably of the following average compositional formula (2):R2bHcSiO(4−b−c)/2 (2)wherein R2 is a substituted or unsubstituted monovalent hydrocarbon radical of 1 to 10 carbon atoms, the subscript “b” is a number from 0 to 3, especially from 0.7 to 2.1, and “c” is a number from more than 0 to 3, especially from 0.001 to 1, satisfying 0<b+c≦3, especially 0.8≦b+c≦3.0. This organohydrogenpolysiloxane is liquid at room temperature. In formula (2), R2 stands for substituted or unsubstituted monovalent hydrocarbon radicals of 1 to 10 carbon atoms, especially 1 to 8 carbon atoms, examples of which are the same as exemplified above for R1, preferably those free of aliphatic unsaturation, and include alkyl, aryl, aralkyl and substituted alkyl radicals, such as methyl, ethyl, propyl, phenyl, and 3,3,3-trifluoropropyl among others. The molecular structure may be straight, branched, cyclic or three-dimensional network. The silicon atom-bonded hydrogen atoms (i.e., SiH radicals) may be positioned at ends or midway of the molecular chain or both. The molecular weight is not critical although the viscosity is preferably in the range of 1 to 1,000 centistokes at 25° C., especially 3 to 500 centistokes at 25° C. Illustrative, non-limiting, examples of the organohydrogenpolysiloxane include 1,1,3,3-tetramethyldisiloxane, methylhydrogen cyclic polysiloxane, methylhydrogensiloxane/dimethylsiloxane cyclic copolymers, both end trimethylsiloxy-blocked methylhydrogenpolysiloxane, both end trimethylsiloxy-blocked dimethylsiloxane/methylhydrogensiloxane copolymers, both end dimethylhydrogensiloxy-blocked dimethylpolysiloxane, both end dimethylhydrogensiloxy-blocked dimethylsiloxane/methylhydrogensiloxane copolymers, both end trimethylsiloxy-blocked methylhydrogensiloxane/diphenylsiloxane copolymers, both end trimethylsiloxy-blocked methylhydrogensiloxane/diphenylsiloxane/dimethylsiloxane copolymers, copolymers comprising (CH3)2HSiO1/2 units and SiO4/2 units, copolymers comprising (CH3)2HSiO1/2 units, (CH3)3SiO1/2 units and SiO4/2 units, and copolymers comprising (CH3)2HSiO1/2 units, SiO4/2 units and (C6H5)3SiO1/2 units. The organohydrogenpolysiloxane is preferably blended in the base polymer in such amounts that the ratio of the number of silicon atom-bonded hydrogen atoms (i.e., SiH radicals) on the organohydrogenpolysiloxane to the number of silicon atom-bonded alkenyl radicals on the base polymer may range from 0.1:1 to 3:1, more preferably from 0.2:1 to 2:1. The addition reaction catalyst used herein is typically a platinum group metal catalyst. Use may be made of platinum group metals in elemental form, and compounds and complexes containing platinum group metals as the catalytic metal. Illustrative examples include platinum catalysts such as platinum black, platinic chloride, chloroplatinic acid, reaction products of chloroplatinic acid with monohydric alcohols, complexes of chloroplatinic acid with olefins, and platinum bisacetoacetate; palladium catalysts such as tetrakis(triphenylphosphine)palladium and dichlorobis(triphenylphosphine)palladium; and rhodium catalysts such as chlorotris(triphenylphosphine)rhodium and tetrakis(triphenylphosphine)rhodium. The addition reaction catalyst may be used in a catalytic amount, which is often about 0.1 to 1,000 ppm, more preferably about 1 to 200 ppm of platinum group metal, based on the weight of the alkenyl radical-containing organopolysiloxane. Less than 0.1 ppm of the catalyst may be insufficient for the composition to cure whereas more than 1,000 ppm of the catalyst is often uneconomical. In the other embodiment wherein the silicone rubber composition is of the peroxide curing type, organic peroxides are used as the curing agent. The organic peroxide curing is useful when the organopolysiloxane as the base polymer is a gum having a degree of polymerization of at least 3,000. The organic peroxides used may be conventional well-known ones, for example, benzoyl peroxide, 2,4-dichlorobenzoyl peroxide, p-methylbenzoyl peroxide, o-methylbenzoyl peroxide, 2,4-dicumyl peroxide, 2,5-dimethyl-bis(2,5-t-butylperoxy)hexane, di-t-butyl peroxide, t-butyl perbenzoate, 1,1-bis(t-butylperoxy)-3,3,5-trimethylcyclohexane, and 1,6-bis(t-butylperoxycarboxy)hexane. An appropriate amount of the organic peroxide blended is about 0.01 to 10 parts by weight per 100 parts by weight of the organopolysiloxane as the base polymer. In addition to the above components, the electromagnetic wave-absorbing composition of the invention may further include well-known components. Also a wetter is preferably used for the purpose of improving the wetting (or dispersibility) of the coated magnetic powder and heat conductive powder with the base polymer for thereby increasing the amount of the powders loaded. Suitable wetter include silanes and low molecular weight siloxanes having hydrolyzable radicals such as hydroxyl and alkoxy radicals which are usually incorporated in conventional silicone rubber compositions, with a hydrolyzable radical-containing methylpolysiloxane having trifunctionality at one end being especially preferred. Any desired conventional methods may be employed in producing, molding and curing the electromagnetic wave-absorbing composition of the invention. The shape of the electromagnetic wave-absorbing composition of the invention is not critical. The composition may take the form of a sheet, a molded part of arbitrary shape, a variable shape material, or a material which has a variable shape upon application and then cures into a fixed shape, among which a choice may be made for the intended application. When molded into a sheet of 1 mm thick, the electromagnetic wave-absorbing composition of the invention preferably has a breakdown voltage of at least 50 V, especially at least 100 V. A breakdown voltage of lower than 50 V has the increased risk of short-circuiting within electronic equipment, which may restrict the site where the composition is applicable, failing to achieve the requisite electromagnetic noise suppressing effects. By virtue of the high breakdown voltage, the electromagnetic wave-absorbing composition of the invention can be applied to any adequate site within electronic equipment without a need to pay substantial attention to short-circuits on printed circuit boards and other devices. Such desired application of the composition ensures to suppress electromagnetic noise within electronic equipment more than ever, and to restrain the leakage of electromagnetic waves to the exterior. When the electromagnetic wave-absorbing composition of the invention endowed with a heat transfer ability is disposed between a device and a heat sink within electronic equipment, the composition ensures to suppress electromagnetic noise and to dissipate the heat generated in the device to the exterior of the equipment. Examples of the invention are given below by way of illustration and not by way of limitation. An electromagnetic wave-absorbing composition within the scope of the invention was prepared. First, spherical particles of soft magnetic Fe—Cr—Si alloy (DEPS1 by Daido Steel Co., Ltd.) were coated with alumina fines as electrically insulating inorganic fines by an RF thermal plasma method, obtaining a magnetic powder coated with alumina fines. Analysis confirmed that a coating layer of alumina fines was formed to a buildup of 5% by weight based on the weight of the coated powder. A liquid silicone rubber composition of the addition reaction type was prepared using a vinyl-containing dimethylpolysiloxane blocked at either end with a dimethylvinylsiloxy radical and having a viscosity of 30 Pa's at room temperature as the base oil. An organopolysiloxane containing silicon atom-bonded alkoxy radicals as the surface treating agent for fillers was added to the base oil in an amount of 1 part by weight per 100 parts by weight of the fillers combined. Then the alumina fine-coated magnetic powder and an alumina powder as the heat conductive filler were added to the base oil. The mixture was agitated and mixed at room temperature. With agitation and mixing continued, the mixture was heat treated at 120° C. for 1 hour, obtaining a base compound for the electromagnetic wave-absorbing composition. Next, an organohydrogenpolysiloxane having at least two silicon atom-bonded hydrogen atoms in a molecule, a platinum group metal catalyst, and an acetylene alcohol reaction regulator were added to the base compound. The proportion of the respective components was adjusted such that the final composition contained 1,000 parts by weight of the alumina fine-coated magnetic powder and 400 parts by weight of the alumina powder as the heat conductive filler per 100 parts by weight of the silicone component. The composition was press molded and heat cured at 120° C. for 10 minutes to form a 1 mm (thick) sheet of the inventive electromagnetic wave-absorbing composition having both electromagnetic wave absorbing and heat transfer abilities. An electromagnetic wave-absorbing composition was molded into a sheet of 1 mm thick as in Example 1 except that the magnetic powder was coated with titanium oxide as the electrically insulating inorganic fines. An electromagnetic wave-absorbing composition was molded into a sheet of 1 mm thick as in Example 1 except that the magnetic powder was coated with aluminum nitride as the electrically insulating inorganic fines. An electromagnetic wave-absorbing composition within the scope of the invention was prepared. First, tabular particles of soft magnetic Fe—Si—Cr—Ni alloy (JEM powder by Mitsubishi Materials Corp.) were coated with alumina fines as electrically insulating inorganic fines by an RF thermal plasma method, obtaining a magnetic powder coated with alumina fines. Analysis confirmed that a coating layer of alumina fines was formed to a buildup of 6% by weight based on the weight of the coated powder. Using a liquid silicone rubber composition of the addition reaction type as in Example 1, an electromagnetic wave-absorbing composition was prepared. The final composition contained 400 parts by weight of the alumina fine-coated magnetic powder per 100 parts by weight of the silicone component. A sheet of 1 mm thick was molded therefrom. An electromagnetic wave-absorbing composition was molded into a sheet of 1 mm thick as in Example 4 except that an acrylic rubber RV-2520 (Nisshin Chemical Co., Ltd.) was used as the base polymer. An electromagnetic wave-absorbing composition within the scope of the invention was prepared. First, spherical particles of soft magnetic Fe—Cr—Si alloy (DEPS1 by Daido Steel Co., Ltd.) were immersed in a dispersion of 15 wt % submicron alumina fines in toluene. The dispersion was intimately milled and then heated to evaporate off the toluene, yielding magnetic particles (powder) coated on surfaces with submicron alumina fines. Analysis confirmed that a coating layer of alumina fines was formed to a buildup of 7% by weight based on the weight of the coated powder. Otherwise as in Example 1, an electromagnetic wave-absorbing composition was molded into a sheet of 1 mm thick. An electromagnetic wave-absorbing composition was molded into a sheet of 1 mm thick as in Example 1 except that spherical particles of soft magnetic Fe—Cr—Si alloy (DEPS1 by Daido Steel Co., Ltd.) were used without coating of electrically insulating fines. Spherical particles of soft magnetic Fe—Cr—Si alloy (DEPS1 by Daido Steel Co., Ltd.) were added to a dispersion of 5% by weight a long-chain alkyl-containing silane (KBM-3103 by Shin-Etsu Chemical Co., Ltd.) in hexane. The dispersion was agitated, mixed, and then heated to evaporate off the hexane, yielding magnetic particles (powder) coated on surfaces with KBM-3103. Analysis confirmed that a coating layer was formed to a buildup of 1.5% by weight based on the weight of the coated powder. Otherwise as in Example 1, an electromagnetic wave-absorbing composition was molded into a sheet of 1 mm thick. The sheets obtained in Examples 1 to 6 and Comparative Examples 1 to 2 were examined for breakdown voltage, attenuation of radiated electromagnetic waves (electromagnetic wave absorbing ability) and thermal conductivity by the following tests. The results are shown in Table 1. Breakdown Voltage: Measured according to JIS C-2110. Attenuation of radiated electromagnetic waves (EM) The attenuation of radiated electromagnetic waves was determined using a system as shown in FIG. 2. Disposed in an electromagnetic dark chamber 11 is a dipole antenna 13 around which a sheet of 1 mm thick molded from an electromagnetic wave-absorbing composition is wrapped. A receiver antenna 15 is located in the chamber and spaced 3 m from the dipole antenna 13. This setting complies with the 3-in testing according to the Federal Communications Commission (FCC). Disposed in a shielded chamber 12 are a signal generator 14 which is connected to the dipole antenna 13 and an electromagnetic emission (EMI) receiver or spectral analyzer 16 which is connected to the receiver antenna 15. The signal generator 14 is operated to generate from the antenna 13 electromagnetic waves at a frequency of 1 GHz, which are received by the receiver antenna 15 and measured by the spectral analyzer 16. The difference between this measurement and the quantity of electromagnetic waves generated when the sheet is omitted is the attenuation. Thermal Conductivity: Measured according to ASTM E1530. TABLE 1BreakdownAttenuation ofThermalvoltageradiated EMconductivity(V)(dB)(W/mK)Example 116503.53.5Example 217203.43.3Example 314803.83.7Example 410208.51.0Example 513608.70.8Example 61803.83.3Comparative Example 1204.03.4Comparative Example 2403.83.5 By virtue of a high breakdown voltage, the electromagnetic wave-absorbing composition of the invention can be applied to any adequate site within electronic equipment without a need to pay substantial attention to short-circuits on printed circuit boards and other devices. When the electromagnetic wave-absorbing composition endowed further with a heat transfer ability is disposed between a device and a heat sink within electronic equipment, the composition ensures to suppress electromagnetic noise and allows the heat generated in the device to dissipate to the exterior of the equipment. The use of the electromagnetic wave-absorbing composition within electronic equipment permits the countermeasures to electromagnetic noise and heat release to be taken at any desired site without caution and in a more effective and simple manner than before. Japanese Patent Application No. 2002-149341 is incorporated herein by reference. Although some preferred embodiments have been described, many modifications and variations may be made thereto in light of the above teachings. It is therefore to be understood that the invention may be practiced otherwise than as specifically described without departing from the scope of the appended claims. |
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048246077 | claims | 1. A process for denitrating dilute aqueous nitric acid and nitrate salt containing waste solutions, comprising mixing the waste solution with ethyl alcohol at room temperature, and heating the mixture to at least 75.degree. C. and up to 100.degree. C. at normal pressure. 2. A process according to claim 1, wherein the mixture is heated to a temperature of about 95.degree. to 100.degree. C. 3. A process according to claim 1, wherein the waste solution is an actinide containing waste solution. 4. A process according to claim 1, wherein the mixture is heated for about 1.0 to 4 hours. 5. A process according to claim 1, wherein the ethyl alcohol is mixed with the waste solution at a mole ratio of 0.5 to 1.5 ethyl alcohol to 1.0 nitric acid. 6. A process according to claim 5, wherein the mixture is heated to a temperature of about 95.degree. to 100.degree. C. for about 1.0 to 4 hours. 7. Process according to claim 1, wherein the process is continued for a sufficient length of time to decompose at least 73% of the originally existing HNO.sub.3. 8. Process according to claim 1, wherein the process is continued for a sufficient length of time to decompose at least 94.5% of the originally existing HNO.sub.3. 9. Process according to claim 1, wherein the process is continued for a sufficient length of time to decompose from 73 to 98% of the originally existing HNO.sub.3. 10. Process according to claim 1, wherein a mole ratio of 0.5 to 1.0 moles ethyl alcohol to 1 mole nitric acid is employed. 11. Process according to claim 1, wherein the concentration of nitric acid in the starting waste solutions is between 0.5 and 5 moles per liter of waste. |
description | This application is a division of U.S. application Ser. No. 15/970,316 filed on May 3, 2018, now U.S. Pat. No. 10,490,311, which is a continuation of U.S. patent application Ser. No. 13/213,389, filed on Aug. 19, 2011, now abandoned, the entire disclosures of which are incorporated by reference herein. The following relates to the nuclear reactor arts, electrical power generation arts, nuclear reactor control arts, nuclear electrical power generation control arts, and related arts. Nuclear reactors employ a reactor core comprising a critical mass of fissile material, such as a material containing uranium oxide (UO2) that is enriched in the fissile 235U isotope. The fuel rod may take various structural configurations, for example including fissile material as pellets embedded in a ceramic matrix or so forth. To promote safety, it is conventional to assemble the core as rods containing the fissile material. A set of rods is preassembled to form a fuel assembly. Preferably, the mass of fissile material in the fuel assembly remains below critical mass. The fuel assemblies are shipped to the reactor site, and are installed in a grid in the reactor pressure vessel to form the reactor core. To prevent a premature chain reaction, suitable neutron absorbing material is provided during installation, for example by inserting neutron-absorbing control rods into the fuel assemblies before they are brought together in the pressure vessel, and by omitting the neutron moderator (e.g., water ambient) if employed. With reference to FIGS. 1 and 2, an illustrative example of such an assembly is shown. FIG. 1 shows an illustrative fuel assembly 10 including a set of fuel rods 12 secured together with a controlled spacing by mid-spacer grid elements 14 and by end-spacer grid elements 16, 18. In the illustrative example, the fuel rods 12 form a 17×17 array. The fuel assembly 10 is typically substantially elongated, and is shown in part in FIG. 1 with an indicated gap G. The fuel assembly 10 also suitably includes other elements, such as control rod guide tubes or thimbles 20 through which neutron-absorbing control rods may pass. One or more of these or similar tubes or thimbles may also serve as instrumentation conduits for in-core sensors. Upper and lower nozzle plates 22, 24 may be provided to facilitate coupling of control rods, instrumentation bundles, or so forth into or out of the fuel assembly 10. The illustrative upper and lower nozzle plates 22, 24 include respective upper and lower alignment pins 26, 28 at the corners of the respective nozzle plates 22, 24 for facilitating alignment of the fuel assemblies during installation in the reactor core. FIG. 2 shows the assembled reactor core 30, including a closely packed grid of fuel assemblies 10 disposed in a core former 32. In FIG. 2, a control rod assembly (CRA) is fully inserted into each fuel assembly 10. In the view of FIG. 2, only an upper support element 34 of the CRA is visible extending above each corresponding fuel assembly 10. The upper support element of each CRA may in be a conventional spider or (as in FIG. 2) a larger element (see “Terminal Elements for Coupling Connecting Rods and Control Rods in Control Rod Assemblies for a Nuclear Reactor”, U.S. Ser. No. 12/862,124 filed Aug. 24, 2010, which is incorporated herein by reference in its entirety, for some illustrative examples). The illustrative reactor core 30 includes sixty-nine (69) fuel assemblies, although in general more or fewer fuel assemblies may be included. The reactor core has a designed lifetime, typically in a range of a year to a few years. The core lifetime is controlled by the reduction in fissile material caused by operation of the nuclear chain reaction. To continue operation, a refueling operation must be performed, in which the spent fuel assemblies are removed and replaced by new fuel assemblies. Typically, this entails shutting down the reactor, opening the pressure vessel and removing any components in order to gain overhead access to the fuel assemblies, and removing the fuel assemblies with the assistance of a crane. To enable coupling with the fuel assembly, each fuel assembly is typically fitted with a box structure with leaf springs mounted on top of the box, or a plate-and-post structure with preloaded helical coil springs mounted between the posts. The fuel assembly is lifted by a grappling mechanism that engages the fixed top plate of the box structure or the movable top plate of the plate-and-post structure via hooks that swing laterally under the top plate in four orthogonal directions. In box designs, the hooks swing outward to engage the top plate of the box, while in plate-and-post designs the hooks swing inward to engage the top plate. These refueling approaches have substantial disadvantages. The swinging motion of the grappling hooks calls for a large working space proximate to the top of each fuel assembly. However, this working space is constrained by the presence of closely adjacent neighboring fuel assemblies in the array disposed in the core former. Moreover, if the CRA is left fully inserted into the fuel assembly during refueling (which is desirable to maintain suppression of the neutron population in the fuel assembly during the refueling process), then either the spider must be removed entirely (a process entailing individually detaching each of the numerous control rods from the spider), or the spider must be of sufficiently low profile to enable the grappling hooks to operate above the spider. Disclosed herein are improvements that provide various benefits that will become apparent to the skilled artisan upon reading the following. In one aspect of the disclosure, a method comprises performing refueling of a nuclear reactor. The refueling includes removing a fuel assembly from a reactor core of the nuclear reactor. The removal method includes: connecting a lifting tool of a crane with a top of the fuel assembly, the lifting tool comprising an assembly of downwardly extending elements, the connecting including locking lower ends of the downwardly extending elements with respective mating features located at a top and periphery of the fuel assembly; moving the fuel assembly connected with the lifting tool into a spent fuel pool using the crane; and releasing the lifting tool from the top of the fuel assembly, the releasing including unlocking the lower ends of the downwardly extending elements from the respective peripherally located mating features at the top and periphery of the fuel assembly. In another aspect of the disclosure, a method comprises performing refueling of a nuclear reactor. The refueling includes removing a fuel assembly having a control rod assembly (CRA) inserted in the fuel assembly from a reactor core of the nuclear reactor. The removal method includes: lowering a lifting tool of a crane onto a top of the fuel assembly, the lowered lifting tool including a plurality of downwardly extending elements that surround and vertically overlap a portion of the CRA extending above the top of the fuel assembly; locking the downwardly extending elements of the lowered lifting tool with corresponding mating features at the top of the fuel assembly in order to connect the lifting tool with the fuel assembly; moving the fuel assembly connected with the lifting tool into a spent fuel pool using the crane; and disconnecting the lifting tool from the top of the fuel assembly in the spent fuel pool by unlocking the downwardly extending elements from the corresponding mating features at the top of the fuel assembly. In another aspect of the disclosure, an apparatus comprises a lifting tool including an upper end configured for attachment with a crane, and a plurality of downwardly extending elements surrounding an open central region disposed below the upper end, lower ends of the downwardly extending elements being configured to mate with mating features at the top of a fuel assembly of a nuclear reactor core. In another aspect of the disclosure, an apparatus comprises: a nuclear fuel assembly including mating features at a top of the nuclear fuel assembly; and a lifting tool including an upper end configured for attachment with a crane and a plurality of downwardly extending elements surrounding an open central region disposed below the upper end, lower ends of the downwardly extending elements being configured to mate with the mating features at the top of the nuclear fuel assembly. In another aspect of the disclosure, an apparatus comprises: a nuclear fuel assembly including mating features at a top of the nuclear fuel assembly; a control rod assembly (CRA) inserted in the nuclear fuel assembly with an upper end of the CRA extending out of the top of the nuclear fuel assembly; and a lifting tool including an upper end configured for attachment with a crane and a plurality of downwardly extending elements surrounding an open central region disposed below the upper end, lower ends of the downwardly extending elements being configured to mate with the mating features at the top of the nuclear fuel assembly. The open central region of the lifting tool that is surrounded by the plurality of downwardly extending elements is configured to receive the upper end of the CRA when the lower ends of the downwardly extending elements mate with the mating features at the top of the nuclear fuel assembly. With reference to FIGS. 3-5, an illustrative nuclear reactor is shown. FIG. 3 shows the nuclear reactor 40 in conjunction with a diagrammatically indicated spent fuel pool 42 and a diagrammatically indicated crane 44. FIG. 4 shows an exploded view of the pressure vessel of the nuclear reactor of FIG. 3. The pressure vessel includes a lower vessel portion 50, an upper vessel portion 52, and a skirt or support structure 54. In the illustrative arrangement, the pressure vessel is mounted vertically (as shown) with at least part of the lower vessel portion 50 disposed below ground level. The bottom of the skirt or support structure 54 is at ground level and supports the pressure vessel and/or biases the pressure vessel against tipping. In the illustrative example of FIG. 3, the spent fuel pool 42 is a below-ground pool containing water and optional additives such as, by way of illustrative example, boric acid (a soluble neutron poison). FIG. 5 shows an exploded perspective view of the lower vessel portion 50 including selected internal components. The lower vessel 50 contains the nuclear reactor core comprising the core former 32 and an array of fuel assemblies 10 (only one of which is shown by way of example in FIG. 5). The reactor core is disposed in and supported by the core former 32 which is in turn disposed in and supported by a core basket 56, which may include radiation shielding, optional emergency coolant tubing (not shown), or so forth. The illustrative nuclear reactor includes upper internals 58 which include wholly internal control rod drive mechanism (CRDM) units. In the illustrative example, the upper internals 58 are supported by a mid-flange 60 that also forms a structural joint of the pressure vessel (being disposed at the junction between the lower and upper vessel portions 50, 52). Alignment between the fuel assemblies 10 and the upper internals 58 is suitably provided by the upper alignment pins 26 at the corners of the upper nozzle plates 22 of the fuel assemblies 10. These pins 26 are designed to accommodate the differential thermal expansion between the fuel assembly 10 and the reactor internals 58 and the fuel assembly growth due to irradiation without losing engagement. The illustrative nuclear reactor is a thermal nuclear reactor employing light water (H2O) as a primary coolant that also serves as a neutron moderator that thermalizes neutrons to enhance the nuclear reaction rate. Alternatively, deuterium dioxide (D2O) is contemplated as the coolant/moderator. The primary coolant optionally contains selected additives, such as optional boric acid which, if added, acts as a neutron poison to slow the reaction rate. The pressure vessel suitably includes a cylindrical central riser or other internal compartments or structures (details not shown) to guide circulation of the primary coolant in the pressure vessel. The primary coolant circulation may be natural circulation caused by the heating of the primary coolant in the vicinity of the reactor core, or may be assisted or generated by illustrative primary coolant pumps 62 also mounted via the mid-flange 60. Although not illustrated, in some embodiments the nuclear reactor is intended to generate steam. Toward this end, primary coolant heated by the reactor core flows through a primary loop that is in thermal communication with a secondary coolant loop through which secondary coolant flows. Heat transfer from the primary loop to the secondary loop heats the secondary coolant and converts it to steam. The thermally coupled primary/secondary coolant loops thus define a steam generator. In some embodiments, the steam generator is external to the pressure vessel, while in other embodiments the steam generator is internal to the pressure vessel, for example mounted in the upper pressure vessel portion 52 in some contemplated embodiments. The steam may for example, be employed to drive a turbine of a generator of an electrical power plant, thus generating electrical power from the nuclear reaction. The illustrative nuclear reactor is of a type generally known as a pressurized water nuclear reactor (PWR), in which the primary coolant (water) is maintained in a superheated state during normal operation. This is suitably accomplished by maintaining a steam bubble located at the top of the upper vessel portion 52 at a desired pressure during normal reactor operation. Alternatively, the nuclear reactor could be configured as a boiling water reactor (BWR) in which the primary coolant is maintained in a boiling state. The illustrative nuclear reactor 40 and other components, e.g. spent fuel pool 42 and diagrammatically represented crane 44, is shown as an example. Numerous variations are contemplated. For example, the pressure vessel can have other portioning, such as having a removable top or “cap” section, and can have access manways provided at various points for maintenance or so forth. In some embodiments the entire pressure vessel may be located underground. Similarly, while the illustrative spent fuel pool 42 is below-ground and surrounds the lower vessel portion 50, more generally the spent fuel pool can be located anywhere within “reach” of the crane 44, and may in some embodiments be above-ground (or, conversely, may be buried deep underground with suitable access from above). The reactor 40 and auxiliary components 42, 44 are typically housed in a concrete or steel containment structure, which is also not shown. The crane 44 is diagrammatically shown, and may in general have any suitable configuration that provides the desired horizontal and vertical travel, lifting capacity, and so forth while fitting within the containment structure. Some suitable crane configurations include an overhead crane configuration, a gantry crane configuration, a tower or hammerhead crane configuration, or so forth. With continuing reference to FIGS. 3-5 and with further reference to FIGS. 6-9, reactivity control is suitably achieved using a control rod assembly (CRA) 70 associated with each fuel assembly 10. FIG. 6 shows an illustrative fuel assembly with the fuel rods omitted, denoted by reference number 10′. By omitting the fuel rods for illustrative purposes, the diagrammatic element 10′ reveals that the control rod guide tubes or thimbles 20 through which neutron-absorbing control rods may pass extend through the entire (vertical) height of the fuel assembly. Corresponding control rods 72 of the CRA 70 are shown in the fully withdrawn position in FIG. 6 (that is, fully withdrawn out of the guide tubes or thimbles 20). The CRA 70 also includes upper support element 74 that secures the bundle of control rods 72 together in a pattern matching that of the guide tubes or thimbles 20. The upper support element 74 may be a conventional spider; in the illustrative example, however, the upper support element 74 is a larger element intended to provide various benefits such as a longer (vertical) length over which to secure the upper ends of the control rods 72, and optionally increased mass for the CRA 70. The illustrative upper support element 74 is shown in isolation in FIG. 8, and in side sectional view in FIG. 9. The illustrative upper support element 74 is further described in “Terminal Elements for Coupling Connecting Rods and Control Rods in Control Rod Assemblies For a Nuclear Reactor”, U.S. Ser. No. 12/862,124 filed Aug. 24, 2010, which is incorporated herein by reference in its entirety. FIG. 7 shows the CRA 70 fully inserted into the fuel assembly 10. It will be noted in FIG. 7 that a portion of the CRA 70, including at least the upper support element 74, extends above the top of the fuel assembly 10 in the fully inserted position. With continuing reference to FIGS. 6-9, the CRA 70 is inserted into the fuel assembly 10 (as per FIG. 7), or withdrawn from the fuel assembly 10 (as per FIG. 6) in order to control the reaction rate of reactivity of the reactor core. The control rods 72 comprise a neutron-absorbing material—accordingly, as the control rods 72 are inserted further into the fuel assembly 10 the reaction rate is reduced. In the fully inserted position (FIG. 6) the reaction is typically extinguished entirely. A connecting rod 76 is employed in order to raise or lower the CRA 70. As illustrated in FIGS. 6, 7, and 9, the lower end of the connecting rod 76 is connected with the upper support element 74 of the CRA 70. The opposite upper end of the connecting rod 76 is not illustrated, but is connected with a suitable control rod drive mechanism (CRDM) unit. In the illustrative embodiment (see FIG. 5) the CRDMs are wholly internal and are part of the upper internals 58 contained within the pressure vessel. Alternatively, the CRDMs may be mounted externally above the pressure vessel (as is typical in a PWR) or externally below the pressure vessel (as is typical in a BWR), with the connecting rods passing through suitable vessel penetrations to connect with the corresponding CRA. With returning reference to FIGS. 3-5, the reactor core has a sufficient quantity of fissile material to support reactor operation for a designed operational time period, which is typically of order one to a few years, although shorter or longer designed periods are also contemplated. Thereafter, the nuclear reactor 40 is refueled and then restarted. Toward this end, the crane 44 includes or is operatively connected with lifting tool 80 that is designed to connect with one of the fuel assemblies. During refueling, the crane 44 operating in conjunction with the lifting tool 80 transfers spent fuel assemblies out of the lower vessel 50 and deposits the spent fuel assemblies in the spent fuel pool 42. By way of diagrammatic illustration, FIG. 3 shows several spent fuel assemblies 10spent which have been transferred into the spent fuel pool 42. (It should be noted that while the illustrative spent fuel pool 42 is below-ground and surrounds the lower vessel portion 50, more generally the spent fuel pool can be located anywhere within “reach” of the crane 44, and may in some embodiments be above-ground.) The crane 44 operating in conjunction with the lifting tool 80 also transfers (i.e., loads) new fuel assemblies into the lower vessel 50, and more particularly into the core former 32. With reference to FIGS. 10-16, the refueling process is described. In an operation S1, the reactor is shut down preparatory to the refueling. The shutdown S1 includes inserting each CRA 70 into its corresponding fuel assembly 10, producing the inserted configuration shown in FIG. 7. A suitable time delay is allowed in order for the reactor to cool down to a sufficiently low temperature to allow opening of the pressure vessel. Some primary coolant may also be removed from the pressure vessel in order to reduce the water level. In an operation S2 (see also FIGS. 3-5), the upper vessel portion 52 is removed (for example, using the crane 44). The effect of the operation S2 is to provide access to the (now spent) fuel assemblies 10 disposed in the core former 32. In an operation S3, for each fuel assembly 10 the connecting rod 76 is detached from the corresponding CRA 70 so as to leave the combination of the fuel assembly 10 and the inserted CRA 70, as shown in FIG. 11. With brief reference to FIG. 9, a suitable approach for performing the removal S3 of the connecting rod 76 is described. In this embodiment, the lower end 76L of the connecting rod 76 terminates in a bayonet or (illustrated) J-lock coupling that is designed to lock into a mating receptacle 76M (see FIG. 8) of the upper support element 74 of the CRA 70. The perspective sectional view of FIG. 9 shows the lower end 76L of the connecting rod 76 in the locked position biased by a spring SS against a retaining feature RR inside the mating receptacle 76M of the CRA upper support element 74. Thus, by pressing the connecting rod 76 downward against the bias of the spring SS and rotating the connecting rod 76 to disengage from the retaining feature RR, the connecting rod 76 is released from the CRA upper support element 74. More generally, a bayonet, J-lock, or other “quick-release” type rotatable coupling can be employed to enable the operation S3 to be quickly performed, with the “groove” and “pin” or other retaining combination being variously disposed (e.g., with the groove on the connecting rod and the pin or pins on the CRA receptacle, or vice versa). Some further illustrative description is set forth in “Terminal Elements for Coupling Connecting Rods and Control Rods in Control Rod Assemblies for a Nuclear Reactor”, U.S. Ser. No. 12/862,124 filed Aug. 24, 2010, which is incorporated herein by reference in its entirety. Although a quick-release approach is advantageous, it is also contemplated to employ a different approach for performing the operation S3—for example, the connecting rod may be permanently connected with the CRA (for example, by a weld or the like), and the operation S3 may entail cutting the connecting rod at a point at or near its junction with the CRA. With continuing reference to FIG. 10, after completion of the operation S3 the resulting unit includes the fuel assembly 10 with the CRA 70 inserted, with a top portion of the CRA 70 including the upper support element 74 extending above the top of the fuel assembly 10. This is illustrated in FIG. 11. In an operation S4 (see also FIG. 12), the lifting tool 80 is lowered onto the top of the fuel assembly 10. As seen in FIG. 12, the lifting tool 80 includes an upper end 81 configured for attachment with the crane 44. In the illustrative lifting tool 80, the upper end 81 includes a loop for attachment with the cable or arm of the crane 44. The lifting tool 80 also includes a plurality of downwardly extending elements 82, namely four downwardly extending rods or bars 82 in the illustrative example, that surround and vertically overlap the portion of the CRA 70 extending above the top of the fuel assembly 10 (e.g., the upper support element 74). The illustrative downwardly extending elements 82 are vertical rods or bars that are aligned such that lower ends 82L of the downwardly extending elements 82 of the lowered lifting tool 80 align with respective peripherally located mating features at a top and periphery of the fuel assembly 10. In the illustrative embodiment, the upper alignment pins 26 of the fuel assembly 10 located at the corners of the upper nozzle plate 22 also serve as the mating features 26 (namely lifting pins 26 in the illustrative example) at a top and periphery of the fuel assembly 10. However, other mating features are also contemplated. For example, the mating features can be protrusions, openings, or recesses at a top and periphery of the fuel assembly. The mating features (e.g., lifting pins 26) are designed to be weight-bearing such that the entire fuel assembly 10 can be raised upward by lifting on the mating features. In the case of the illustrative fuel assembly 10, this is accomplished by constructing the upper and lower nozzle plates 22, 24, the control rod guide tubes or thimbles 20, and the spacer grid elements 14, 16, 18 as a welded assembly of steel or another suitable structural material (best seen as the structure 10′ in FIG. 6). The lifting pins 26 at a top and periphery of the fuel assembly 10 are secured to the upper nozzle plate 22 by welding, a threaded connection, a combination thereof, or another suitably load-bearing connection. With continuing reference to FIG. 10 and with further reference to FIGS. 14, 15, 15A, 16, and 16A, in an operation S5 the lowered lifting tool 80 is connected with the top of the fuel assembly 10. The connection operation S5 includes locking the lower ends 82L of the downwardly extending elements 80 with the respective peripherally located mating features, e.g. lifting pins 26, at the top and periphery of the fuel assembly 10. In the illustrative approach (see FIGS. 14, 15, 15A, 16, and 16A), the locking operation is performed by rotating at least the lower ends 82L of the downwardly extending elements 80 to lock the lower ends disposed over (as illustrated) or inside the respective lifting pins 26 with the respective lifting pins 26. Toward this end, the lower ends 82L and the respective lifting pins 26 define a lockable bayonet coupling. FIG. 14 shows an enlarged view of one of the lower ends 82L aligned with and being lowered over the respective lifting pin 26. In this view a groove 86 in the lifting pin 26 is visible, as well as a narrowed portion 88 of the lifting pin 26. These features 86, 88 are designed to cooperate with a recess 90 in the lower end 82L with a narrowed region 92 to form a rotationally engaging lock. FIG. 15 shows an enlarged view of the lower end 82L fully lowered over the lifting pin 26. FIG. 15A shows the view of FIG. 15 with partial cutaway of the lower end 82L to reveal internal components of the (unlocked) locking configuration. FIG. 16 shows an enlarged view of the lower end 82L after a rotation of about 90°. This rotation causes the narrowed region 92 to move into the groove 86 to form the lock. FIG. 16A shows the view of FIG. 16 with partial cutaway of the lower end 82L to reveal internal components of the (locked) locking configuration. In other embodiments, other rotationally locking “quick-release” configurations can be employed. For example, in another contemplated embodiment the J-lock coupling shown in FIG. 9 for coupling the connecting rod 76 with the CRA upper support element 74 can be used in coupling the lower end of the downwardly extending rod or bar with a mating recess at the top and periphery of the fuel assembly. Another rotationally locking quick-release configuration contemplated for use in the lower ends of the downwardly extending elements of the lifting tool are threaded connections. In this embodiment, the lower ends have threads that mate with threaded holes located at the top periphery of the nuclear fuel assembly. The locking in this case is a frictional lock obtained by rotating the lower ends to thread into the threaded holes until a designed torque is reached. With reference to FIG. 17, in any embodiment employing a rotational lock, the downwardly extending elements, or at least their lower ends, should include motorized rotation capability. In an illustrative example shown in FIG. 17, each downwardly extending rod or bar 82 includes a diagrammatically indicated motor 94 providing the motorized rotation of the lower end 82L. Although FIG. 17 illustrates a separate motor 94 for each downwardly extending rod or bar 82, in other embodiments it is contemplated to employ a single motor that drives rotation of all lower ends via a suitable drive train (e.g., geared rotating shafts or the like). It is also noted that since the lifting tool 80 is not disposed inside the pressure vessel except when the reactor is shut down, the lifting tool 80 (including the motors 94) does not need to be rated for operation at the operating temperature of the nuclear reactor. The motors 94 should be robust against immersion in the primary coolant and in the fluid of the spent fuel pool 42 (see FIG. 3), for example by being hermetically sealed. While various embodiments of rotational locks (e.g., bayonet or J-lock couplings) are disclosed herein, other types of locks, including non-rotational locks, are also contemplated. For example, in another contemplated embodiment the locks may employ motorized clamps that clamp onto respective mating features at the top of the fuel assembly. With returning reference to FIG. 10, in an operation S6 the fuel assembly 10 connected with the lifting tool 80 is moved into the spent fuel pool 42 using the crane 44. In an operation S7 the lifting tool 80 is released from the top of the fuel assembly. The release operation S7 includes unlocking the lower ends 82L of the downwardly extending elements 82 from the respective peripherally located mating features (e.g. lifting pins 26) at the top and periphery of the fuel assembly 10. In the illustrative embodiment, this entails rotating the lower ends 82L in the opposite direction to that used in the locking operation and then lifting the unlocked lifting tool 80 upward away from the spent fuel assembly now residing in the spent fuel pool 42. Other unlocking operations may be employed depending upon the nature and configuration of the locking coupling. Since the reactor core typically includes a number of fuel assemblies 10 (see the example of FIG. 2 in which the reactor core 30 includes sixty-nine fuel assemblies). Accordingly, after the release operation S7, an operation S8 is performed in which the next fuel assembly to be unloaded is selected, and the process repeats beginning at operation S4. Once all fuel assemblies have been unloaded, an operation S9 is performed in which the lifting tool 80 is parked in a storage location. Alternatively, if new fuel is to be loaded into the reactor, operations analogous to operations S4, S5, S6, S7, S8 are performed to pick up new fuel assemblies from a loading dock or other source location and place the new fuel assemblies into the core former 32, followed by performing control rod reattachment (analogous to operation S3), replacement of the upper vessel portion 52 (analogous to operation S2), and restarting the reactor (analogous to operation S1, and optionally further including performing various integrity or safety checks prior to the restart). Note that these analogous loading operations are not shown in FIG. 10. Additionally, the reloading may include performing other maintenance such as replacing the connecting rods or other internal reactor components, various inspection and/or cleanup operations, or so forth. An advantage of the lifting tool 80 is that it accommodates a CRA inserted into the fuel assembly 10 that extends substantially above the top of the fuel assembly 10. Because no swing action is required to engage the lifting mechanism; the fuel assembly can be lifted even when most or all of the inboard volume above the fuel assembly is occupied by the upper portion 74 of the inserted CRA. The peripherally arranged downwardly extending elements 80 accommodate the exposed portion of the CRA by surrounding the exposed upper end of the inserted CRA (e.g., the upper support element 74) when the fuel assembly 10 is connected with the lifting tool. The downwardly extending elements 82 surround an open central region disposed below the upper end 81 of the lifting tool 80, such that the open central region can accommodate the upward extension of the inserted CRA out of the top of the fuel assembly 10. In this way, the CRA vertically overlaps the lifting tool 80 when the fuel assembly 10 is connected with the lifting tool 80 (see FIG. 13). In some embodiments the overlap is at least one-half of the vertical height of the lifting tool 80. In some embodiments the overlap between the CRA and the lifting tool 80 is at least one-half of the vertical height of the downwardly extending elements 82 of the lifting tool 80. With reference to FIGS. 18-20, it is to be appreciated that the fuel assemblies, CRA, and lifting tool can have various geometries. FIG. 18 shows the illustrative geometry of the fuel assembly 10, which has a rectangular cross-section when viewed from above as per FIG. 18, with the CRA including the upper support element 74 inserted in illustrative FIG. 18. FIG. 19 illustrates a hexagonal fuel assembly 22H having six sides, with the same CRA including the same upper support element 74 inserted. In this embodiment there are six mating features 26H located at a top and periphery of the fuel assembly. The illustrative six mating features 26H are the same as the lifting pins 26 of the fuel assembly 10. The corresponding lifting tool (not shown) suitably includes six downwardly extending elements, e.g. six downwardly extending rods or bars, arranged in a hexagonal pattern to mate with the respective six lifting pins 26H. Finally, as a further example, FIG. 20 illustrates a triangular fuel assembly 22T having three sides, with a conventional spider 74T with six branches serving as the upper support element of the CRA. In this embodiment there are three mating features 26T, which in this embodiment are embodied as recesses or openings 26T. The corresponding lifting tool (not shown) suitably includes three downwardly extending elements, e.g. three downwardly extending rods or bars, arranged in an equilateral triangular pattern to mate with the respective three openings 26T. In general, the geometry of the fuel assembly preferably promotes a closely packed arrangement. The preferred embodiments have been illustrated and described. Obviously, modifications and alterations will occur to others upon reading and understanding the preceding detailed description. It is intended that the invention be construed as including all such modifications and alterations insofar as they come within the scope of the appended claims or the equivalents thereof. |
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053316789 | abstract | A grid strip (10) for a fuel rod support grid of a nuclear fuel assembly is cold-formed from a flat annealed plate having length, height, and width dimensions. A plurality of slots (12,14) extend along the height dimension at regular intervals along the length dimension, thereby defining successive cell walls (16,18,20) between successive slots along the length dimension. Each cell wall has upper, central, and lower regions (28,30,32) along the height dimension, each region including a substantially flat base area (34,36,38) and fuel rod support structure (22,24,26) projecting integrally from the base area along the width dimension of the strip. The support structure in each of the upper and lower regions includes a relatively stiff, arched stop (22,26) which projects in a first direction and the support structure in the central region includes a relatively soft, arched spring (24) which projects in a second direction opposite the first direction. The spring includes spaced apart pedestals 40,42 formed in the base area of the central region and projecting in the second direction, and a resilient beam 44 extending between and rigidly supported by the pedestals, so as to project in the second direction beyond the projection of the pedestals. |
abstract | Systems for enhancing preignition conditions of a fusion reaction are disclosed. A first system includes a target chamber for receiving a fusion fuel, and energy driving means oriented to direct plasma confinement structure onto to the fusion fuel to facilitate ignition of a controlled fusion reaction of said fusion fuel. A plurality of electron sources provides electron beams of a predetermined energy and one of fluence and quantity, directed onto and illuminating, a fusion fuel-derived plasma for controlling the ratio of ion temperature and electron temperature of the plasma. A second system comprises a central target chamber for receiving a spherical pellet of fusion target material and at least first and second pluralities of energy drivers oriented to supply temporally-staged X-ray pulses to the fusion target material in a 3-dimensionally symmetric manner about said pellet. A third system combines aspects of the first and second systems. |
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047327286 | summary | FIELD OF INVENTION The present invention relates to detecting a beam of neutrino or antineutrino particles and more particularly to a method of and apparatus for detecting such a beam by utilizing a single crystal containing coherent elastic scatterers for the particles. BACKGROUND ART In my copending application, Ser. No. 295,002, filed Aug. 21, 1981, now U.S. Pat. No. 4,576,777, issued Mar. 18, 1966 and entitled Energy Detection Method and Apparatus, there is disclosed a method of and apparatus for detecting a beam of neutrino or antineutrino particles wherein the beam is incident on a crystal arranged so it has coherent inelastic scatterers for the particles. Individual atoms in the crystal absorb energy from the particles to scatter the particles and produce stimulated coherent radiant energy fields. The coherent fields produced by the individual atoms are detected to provide an indication of the presence of the neutrino or antineutrino particles in the beam. In the invention of the copending application, the inelastic scattering and energy exchange between the neutrino or antineutrino particles and the scatterers occur because the crystal is placed in an electromagnetic field which induces nuclear spin changes in the crystals. In the present invention, there is elastic scattering because the crystal is not subjected to an electromagnetic field and no nuclear spin state change is induced in the crystal by the neutrino or antineutrino particles of the beam incident thereon. By employing elastic, rather than inelastic scattering, the present invention has a scattering cross-section that is 10.sup.5 times the cross-section of the inelastic scatterers. The elastic scatterers do not absorb significant energy from the neutrino or antineutrino particles, in contrast to the energy absorption by the inelastic scatterers. It is an object of the present invention to provide a new and improved apparatus for and method of detecting a beam of neutrino or antineutrino particles that has greater sensitivity than the prior art neutrino or antineutrino detection methods and apparatus. A further object of the invention is to provide a new and improved apparatus for and method of detecting a beam of neutrino or antineutrino particles by utilizing crystals containing coherent scatterers for the particles. An additional object of the invention is to provide a new and improved apparatus for and method of detecting a beam of neutrino or antineutrino particles wherein a crystal containing coherent scatterers for the particles includes scatterers that recoil as a single entity, and thereby provide greater sensitivity than the prior art method and apparatus. THE INVENTION In accordance with the present invention there is provided a new and improved apparatus for and method of detecting a beam of neutrino or antineutrino particles wherein a single crystal arranged to contain coherent elastic scatterers for the particles is located in a path of the beam so the beam particles are incident on the elastic scatterers. The scatterers respond to the particles of the beam incident thereon by transferring momentum from the particles to mechanical momentum in the crystal. The mechanical momentum transferred to the crystal from the momentum of the particles is detected. The scatterers in the crystal for the particles are sufficiently stiff as to recoil as a single entity. Such crystals are selected from the group including sapphire, silicon and diamond, the same crystals which I used in the prior art devices; however in the present invention these crystals are elastic scatterers because they are not in a significant electromagnetic field. Detection is preferably provided by detecting the motion imparted by the momentum transfer from the crystal to a mass on which the crystal is mounted. Mounted on the mass in opposed relation to the crystal is a structure or means for substantially balancing gravitational effects imparted to the mass by the crystal. The balancing means is arranged so that there is no momentum transfer from the beam particles to it so that the mass moves in response to the net momentum transferred to it by the beam particles and is unresponsive to the gravitational and other effects imparted to the crystal and the balancing means. In one embodiment, the mass is a torsion balance having a member that turns in response to the net momentum transferred to the crystal elastic scatterers. The extent this member turns provides an indication of the intensity of the neutrino or antineutrino particles in the beam. The balancing means is a mass mounted on the member in opposed relation to the crystal. In a second embodiment, the crystal containing the coherent elastic scatterers is mounted on a tine of a tuning fork. A second tine of the tuning fork includes the balancing means, whereby the motion of the tuning fork, as detected by a piezoelectric crystal mounted on an arm connecting the two tines together, provides an indication of the presence and intensity of the neutrino or antineutrino particles in the beam. It is to be understood, however, that the crystal and balancing means can be mounted on other types of mechanical resonators. Preferably, the neutrino or antineutrino particles in the beam incident on the elastic scatterers of the crystal mounted on the mechanical resonator are amplitude modulated at the resonant frequency of the mechanical resonator or a harmonic thereof. Such modulation substantially increases the sensitivity of the detecting structure and process for the neutrino and antineutrino particles. The amplitude modulation is preferably provided by chopping the beam incident on the crystal containing the elastic coherent scatterers mounted on the mechanical resonator. Chopping of the neutrino beam is preferably provided by mounting a plurality of scatterers for the particles in the beam on a structure that is turned by a motor. The theory of the present invention is described in detail in my papers entitled "Gravitons, Neutrinos and Antineutrinos", Volume 14, No. 12, December, 1984, pp. 85-1209 and "Method for Observation of Neutrinos and Antineutrinos," Physical Review C, Volume 31, No. 4, April, 1985, pp. 1468-1475; these papers are incorporated herein by reference. As discussed in these papers, when the wave length of incident particles is small in comparison with the dimensions of a macroscopic volume of scatterers, the total cross-section of the volume interacting with the energy is proportional to the square of the number of scatterers, if the crystal is very stiff. For inelastic scattering, the change in energy of the scatterers is determined by the spin state change in an applied magnetic field. For elastic scattering, the energy change is equal to the square of the momentum change divided by two times the mass of the entire body performing the scattering. Hence, for elastic scattering, there is a very low energy change because the value of the mass of the entire body is very large. For inelastic scattering, however, the energy change is relatively large, because a large number of spins change their states. By employing the elastic scatterers of the present invention, in contrast to the inelastic scatterers of the prior art, the momentum change imparted to the scattering crystal is not masked by energy changes imparted to the crystal. Hence, detecting the momentum change imparted to the elastic scatterers by the neutrino or antineutrino particles provides a highly sensitive measure of the intensity of the neutrino beam incident on the crystal, even though there is only a weak interaction between the beam and the scatterers of the crystal. In the elastic scattering process employed in the present invention, no significant fraction of the neutrino energy is converted into heat or change of internal energy state of the scatterers. The scattering employed in the present invention corresponds to projecting a large number of light elastic spheres, such as tennis balls, against a wall. The detection process involved in the present invention is analagous to measuring the small forces exerted by the wall in scattering the spheres. The coherent elastic scattering single crystals employed in the present invention have very stiff properties. As stiffness increases, there is a greater total cross-section for a given amount of matter intercepting the particles of the neutrino or antineutrino beam. A measure of crystal stiffness is the Debye temperature of the crystal. The above and still further objects, features and advantages of the present invention will become apparent upon consideration of the following detailed description of several specific embodiments thereof, especially when taken in conjunction with the accompanying drawings. |
description | The present application claims priority under 35 U.S.C. 119(e) to U.S. Provisional Patent Application No. 60/574,555 for TECHNIQUE FOR DIGITALLY REMOVING X-RAY SCATTER IN A RADIOGRAPH USING A MODULATING GRID filed on May 25, 2004, the entire disclosure of which is incorporated herein by reference for all purposes. The present invention relates to techniques for improving radiograph images. Since the discovery of x-rays and the early days of radiography (i.e., the exposure of x-ray shadowgrams on film), scatter has been recognized as a major contributing factor to image quality degradation. Scatter is the reemission of x-rays (called secondary rays) caused by the absorption of primary rays as illustrated in FIG. 1A. When the reemission of a secondary ray is in a different direction than the direction of the absorbed primary ray, the secondary ray blurs the shadowgram of the object. Without scatter, the shadowgram sharpness is determined by the size of the x-ray source (called the focal spot), and the distances between the source, the object, and the image. Suitable detectors (e.g., films, intensifying screens, storage-phosphor plates, or solid-state detectors) are placed in the image plane to capture the shadowgram. When scatter is present, the shadowgram sharpness is reduced regardless of the resolution of the detector. Scatter is more pronounced in low-Z materials (such as soft tissue) than in high-Z materials (such as bone). Scatter is also more pronounced when a large x-ray cone beam is used rather than a small one. This is why radiographs of extremities (e.g., hands and legs) are much sharper than radiographs of the pelvis (i.e., a large area with significant amount of soft tissue). A number of techniques have been suggested to suppress (or at least reduce) x-ray scatter. Some complex techniques involve sweeping a pencil beam or a fan beam of x-rays along with a secondary collimator (to block scatter). A simpler and commonly used technique invented by Gustave Bucky in 1913 involves using a cone beam of x-rays and placing a collimating grid 102 between the object (patient 104) and the detector 106 as shown in FIG. 1B. The purpose of the collimating grid is to absorb secondary rays and transmit primary rays. Unfortunately, the collimating grid also absorbs a significant percentage of primary rays, and therefore the dose to the patient has to be increased accordingly so as to maintain the correct exposure level on the detector. The dose increase varies with the construction of the grid and its efficiency at reducing scatter, i.e., the greater the scatter reduction, the greater the dose increase. A “grid ratio” of 8:1 to 16:1 is not uncommon. This translates into a dose increase to the patient (i.e., the Bucky factor) of 4 to 6 times the dose necessary when no grid is used. In reality, not using a grid for certain exams (e.g., the pelvis) is not even an option since unblocked scatter would degrade the image quality below any acceptable level. In view of the foregoing, it is clear that techniques for reducing the scatter in a radiograph are desirable. According to the present invention, methods and apparatus are provided for generating an x-ray image corresponding to an object. According to a specific embodiment, the object is interposed between a detector and an x-ray source. A grid is interposed between the x-ray source and the object. The grid includes a plurality of elements defining interstices. A first area of the grid corresponding to the elements is larger than a second area of the grid corresponding to the interstices. The grid is exposed to primary x-ray energy generated by the x-ray source, thereby exposing the object to a first portion of the primary x-ray energy via the interstices of the grid. A second portion of the primary x-ray energy is received with first areas of the detector corresponding to the interstices of the grid. Secondary x-ray energy is received with the first areas of the detector and with second areas of the detector corresponding to the elements of the grid. The secondary x-ray energy results from interaction of the first portion of the primary x-ray energy with the object. X-ray image data are generated for the object by altering first data corresponding to the first areas of the detector with reference to second data corresponding to the second areas of the detector. According to another specific embodiment of the invention, an apparatus operable to capture an x-ray image corresponding to an object is provided. An x-ray source is operable to generate primary x-ray energy. A detector apparatus is provided for holding a detector operable to capture the x-ray image at an image plane. A grid comprising a plurality of elements defining interstices is provided. A first area of the grid corresponding to the elements is larger than a second area of the grid corresponding to the interstices. The grid is positioned between the x-ray source and the detector apparatus such that the x-ray apparatus is operable to receive the object between the grid and the detector apparatus. The grid is operable to transmit a first portion of the primary x-ray energy via the interstices. A further understanding of the nature and advantages of the present invention may be realized by reference to the remaining portions of the specification and the drawings. Reference will now be made in detail to specific embodiments of the invention including the best modes contemplated by the inventors for carrying out the invention. Examples of these specific embodiments are illustrated in the accompanying drawings. While the invention is described in conjunction with these specific embodiments, it will be understood that it is not intended to limit the invention to the described embodiments. On the contrary, it is intended to cover alternatives, modifications, and equivalents as may be included within the spirit and scope of the invention as defined by the appended claims. In the following description, specific details are set forth in order to provide a thorough understanding of the present invention. The present invention may be practiced without some or all of these specific details. In addition, well known features may not have been described in detail to avoid unnecessarily obscuring the invention. As described above, a conventional collimating grid is placed between the patient and the detector so as to absorb secondary rays and transmit primary rays. Following a counterintuitive notion, the present invention instead places a grid 202 between the object (e.g., patient 204) and the x-ray source 206 as shown in FIG. 2. At first, it may seem that the grid would not be effective since it would not stop secondary rays from reaching the detector. In fact, the purpose of the grid in the present invention is no longer to absorb secondary rays and transmit primary rays, but rather to create a spatially modulated x-ray exposure on the detector 208. That is, the grid of the present invention is not a collimating grid but a spatially modulating grid, which creates a lattice of x-ray pencil beams over the patient (e.g., a checkerboard pattern 302 as shown in FIG. 3). If no scatter is present, only the lattice of (primary x-ray) pencil beams reaches the detector and the gaps between the beams are not exposed. When scatter is present as shown in FIG. 4, some areas of the detector 402 are exposed to primary and secondary x-rays (i.e., where the lattice of pencil beams hit the detector), whereas some areas of the detector 402 are only exposed to secondary x-rays (i.e., the gaps between the pencil beams). As shown, the spatially modulating grid 404 does not stop scatter radiation from reaching the detector. Rather, according to the present invention, the presence of the spatially modulating grid makes it possible to digitally remove the scatter from the radiograph. This is possible because the presence of the spatially modulating grid results in the capture of two interlaced sub-images within a single radiograph. One of the interlaced sub-images (created where the x-ray pencil beams hit the detector) is referred to herein as the “shadow+scatter” image. This image is similar to a radiograph taken with a conventional cone beam x-ray source without a collimating grid (i.e., both primary and secondary rays reach the detector). The other interlaced sub-image (created in the gaps between x-rays pencil beams where primary rays are prevented from reaching the detector by the grid) is referred to herein as the “scatter only” image since only secondary rays reach the detector. These interlaced sub-images can be extracted and subsequently processed provided the detector has sufficient spatial resolution. Although it is not clear from the figures, it should be noted that, in some cases, scatter received by a specific area of the detector may be significant as compared to the desired image or shadow information corresponding that area (e.g., up to 3 times). If the pitch of the lattice of x-rays pencil beams (called p) is small (e.g., 100 μm to 300 μm) the “scatter only” image is a very close approximation of the scatter contribution in the “shadow+scatter” image, since both images are only offset by p/2. Because of this close approximation, it is then possible to remove scatter from the radiograph by digitally subtracting the “scatter only” image from the “shadow+scatter” image. According to a specific embodiment, the digital image subtraction is performed by subtracting each pixel of the “scatter only” image from the corresponding pixel of the “shadow+scatter” image. Digital image subtraction is a common feature of most image processing software programs (such as ImagePro from Media Cybernetics) and is also referred to as image background subtraction. For accurate scatter removal, adequate sampling of the scatter image is important. Previous attempts have been made to remove scatter but such methods have not provided for sufficient sampling of the scatter image (as compared to the final image). When the scatter image is strongly undersampled compared to the final image, scatter removal is not accurate and the resulting image is not diagnostically acceptable. Therefore, according to specific embodiments of the invention, the number of samples for the “scatter only” image is actually higher (in some cases significantly higher) than for the “shadow+scatter” image. This may be understood with reference to grid configurations in which each interstice of the grid is surrounded by an absorbing region, i.e., small transmitting regions (i.e., the interstices) surrounded by large absorbing regions (i.e., the elements of the grid). According to a specific embodiment, the fill factor of the grid is around 16% (i.e., 16% of the grid is transmitting while 84% is absorbing). Previous techniques in which the scatter image is undersampled compared to the final image have relied on a 2D interpolation to create a complete scatter image. By contrast, according to specific embodiments of the invention, the “scatter only” image is computed using a 2D intrapolation (as opposed to interpolation). More specifically, the scatter value for a given shadow+scatter pixel (corresponding to a particular interstice) is computed by averaging the surrounding pixels (corresponding to the absorbing area of the grid). In a particular embodiment, four pixels corresponding to a particular shadow+scatter pixel are used for this intrapolation. In this embodiment (illustrated in FIG. 5), four “scatter only” pixels 502 surrounding “shadow+scatter” pixel 504 correspond to areas of the grid with maximum absorption (i.e., the overlap regions of vertical and horizontal wires). Even if the x-ray detector does not have sufficient spatial resolution to perfectly resolve the “scatter only” image from the “shadow+scatter” image, i.e., there is some crosstalk between the two images, sufficient information can still be extracted to perform the scatter removal. Crosstalk between pixels in the x-ray detector is equivalent to adding “scatter only” signal to the “shadow+scatter” image and adding “shadow+scatter” signal to the “scatter only” image. Proper calibration can alleviate this issue. Additionally, image processing software that can analyze image frequency content (e.g., using Fast Fourier Transforms (FFT) algorithms) can be useful in extracting the “shadow+scatter” image and the “scatter only” image from the raw image. As will be understood by those of skill in the art, the foregoing digital processing may be performed by a wide variety of computing platforms, an example of which is described below with reference to FIG. 9. It should be understood that even though this technique requires a digital operation, the acquisition of the radiograph itself does not have to be done digitally. That is, a conventional x-ray detector can be used and the captured image subsequently digitized for the digital subtraction of the “scatter only” image. Thus, although a digital detector may be more appropriate for some applications, other detectors (e.g., conventional film, storage-phosphor, etc.) may be employed with embodiments of the invention as long as they have sufficient spatial resolution to over-sample the image plane and capture both interlaced sub-images. As will be understood, any suitable technique for digitizing radiograph images may be employed. According to some implementations of the present invention, the detector has twice the linear resolution (i.e., four times the number of pixels) than the resolution of the scatter-corrected image. This implies that the shadow of the spatially modulating grid be perfectly registered with the pixel lattice on the detector. Because it is practically difficult to achieve such registration, specific embodiments of the invention employ detectors having a significantly greater resolution than the resolution of the scatter-corrected image. According to one embodiment, the detector has four times the linear resolution of the scatter-corrected image (i.e., sixteen times the number of pixels). For example, in order to obtain a scatter-corrected pelvis radiograph at 2.8 lp/mm (176 μm pixel), a suitable detector would feature a 44 μm pixel. Most conventional digital detectors are limited to a minimum pixel size of 100 μm (commonly 127 μm or 143 μm) because of fundamental manufacturing process limitation (micro-lithography on glass). Such detectors are therefore inadequate for implementing this scatter-removal technique. By contrast, new generations of digital detectors based on a novel technique for stimulating and reading storage-phosphor plates are capable of achieving a 44 μm pixel geometry and are therefore ideal for implementing the scatter-removal technique of the present invention. A detailed description of this technique is provided in U.S. Pat. No. 6,800,870, issued Oct. 5, 2004, the entire disclosure of which is incorporated herein by reference for all purposes. An important benefit of this technique is the fact that use of the spatially modulating grid, unlike the traditional collimating grid, eliminates the requirement for any dose exposure increase to the patient. The detector captures all the remnant rays through the patient (primary as well as secondary rays). The spatially modulating grid does absorb a significant portion of the incoming x-rays, but since it is placed between the source and the patient, it does not necessitate a dose increase to the patient. And because the purpose of the grid of the present invention is to spatially modulate the x-ray exposure (as opposed to collimating the x-rays), the construction of the spatially modulating grid can be very different than collimating grids. Whereas a collimating grid must stop x-rays at a certain incidence and transmit others at a different incidence, the spatially modulating grid is configured to create a lattice of x-ray pencil beams regardless of the angle of incidence of the incoming x-ray cone beam. According to one embodiment of the invention, in order to minimize the effect of the angle of incident x-rays onto the spatially modulating grid, the grid is made of a crisscross web of high-Z wires having circular cross-sections. Unlike the rectangular cross-section elements used in collimating grids, circular cross-section elements cast the same shadow regardless of the angle of incident x-rays. Another effect of the angle of incident x-rays, α, onto the grid is the apparent spacing of the elements. That is, as the angle of incidence increases the apparent spacing decreases. According to a specific embodiment, a 114 μm circular cross-section Tungsten wire is used to build the grid. The grid 600 (of FIGS. 6A-C) is made of two substantially parallel sets of wires placed perpendicular to each other. As shown in FIGS. 6A-C, various implementations of the grid may be constructed with or without a weave, e.g., the plain or twill weaves of FIGS. 6A and 6B, or the “no weave” configuration of FIG. 6C. Each set of wires covers an approximately 17″ by 17″ area and contains approximately 2,700 17″ long sections of wire (602 or 604) laid next to each other with an inter-space of 46 μm. A 46 μm inter-space provides a 160 μm wire pitch, which results in a 176 μm pitch once projected on the detector with a 110% magnification. The magnification is caused by the relative distances between the source, the grid and the detector. At normal incidence, each pencil beam projects an unobstructed spot of approximately 50 μm×50 μm on the detector. The actual spot is larger since the edge of the wires do not absorb nearly as much as the center of the wires. For angles of close to normal incidence, the pencil beam pattern does not change significantly. This means that the angle of the grid with respect to the source is not nearly as critical as with a collimating grid where cutoff can occur. As illustrated by FIG. 7 (which shows horizontal grid wires 702 in cross-section), it would take a 25° angle of incidence, α, to reduce the unobstructed spot to zero. The x-ray throughput of the specific implementation of the spatially modulating grid described above is given by its fill factor: (46/114)2=16%. This means that 6 times more incident x-rays are required to produce the same exposure as without the grid. This value is similar to the Bucky factor of a conventional collimating grid. Therefore, conventional x-ray generators can be used with this technique without requiring unacceptably long exposure times. The magnification and focal spot blur effects mentioned above can be minimized by placing the spatially modulating grid as close as possible to the detector (i.e., as far as possible from the source) while still providing enough space for the patient. In order to obtain a close to ideal “scatter only” image, it is desirable for the gaps between the pencil beam spots to receive the least number of primary x-rays possible. The locations where the primary x-rays are least transmitted by the grid are where the two orthogonal wires cross. At the precise location of the intersection, the x-ray cross-section is twice the diameter of the wire. So, for example, in the case of a 114 μm wire, the maximum x-ray cross-section is 228 μm. Alternatively, in the case of a 150 μm wire, the maximum x-ray cross-section would be 300 μm. FIG. 8 shows the transmission through 300 μm of Tungsten for different beam energies (140 kVp, 120 kVp and 80 kVp). Even though some primary x-rays do get transmitted through the 300 μm of Tungsten, they do not contribute significantly to the image since their energy is mostly between 60 keV and 70 keV, a range in which, for example, a storage phosphor detector is not very sensitive. As mentioned above, a wide variety of computing platforms or data processing systems may be employed to implement various aspects of the invention, e.g., the processing of image data to remove scatter, the digitization of captured images, etc. FIGS. 9A and 9B illustrate a computer system 900 suitable for implementing embodiments of the present invention. FIG. 9A shows one possible physical form of the computer system. Of course, the computer system may have many physical forms ranging from an integrated circuit, a printed circuit board and a small handheld device up to a huge super computer. Computer system 900 includes a monitor 902, a display 904, a housing 906, a disk drive 908, a keyboard 910 and a mouse 912. Disk 914 is a computer-readable medium used to transfer data to and from computer system 900. FIG. 9B is a block diagram of an exemplary architecture for computer system 900. Attached to system bus 920 are a wide variety of subsystems. Processor(s) 922 (also referred to as central processing units, or CPUs) are coupled to storage devices including memory 924. Memory 924 includes random access memory (RAM) and read-only memory (ROM). As is well known in the art, ROM acts to transfer data and instructions uni-directionally to the CPU and RAM is used typically to transfer data and instructions in a bi-directional manner. Both of these types of memories may include any suitable of the computer-readable media described below. A fixed disk 926 is also coupled bi-directionally to CPU 922; it provides additional data storage capacity and may also include any of the computer-readable media described below. Fixed disk 926 may be used to store programs, data and the like and is typically a secondary storage medium (such as a hard disk) that is slower than primary storage. It will be appreciated that the information retained within fixed disk 926, may, in appropriate cases, be incorporated in standard fashion as virtual memory in memory 924. Removable disk 914 may take the form of any of the computer-readable media described below. CPU 922 is also coupled to a variety of input/output devices such as display 904, keyboard 910, mouse 912 and speakers 930. In general, an input/output device may be any of: video displays, track balls, mice, keyboards, microphones, touch-sensitive displays, transducer card readers, magnetic or paper tape readers, tablets, styluses, voice or handwriting recognizers, biometrics readers, or other computers. CPU 922 optionally may be coupled to another computer or telecommunications network using network interface 940. With such a network interface, it is contemplated that the CPU might receive information from the network, or might output information to the network in the course of performing the above-described method steps. Furthermore, method embodiments of the present invention may execute solely upon CPU 922 or may execute over a network such as the Internet in conjunction with one or more remote CPUs that each share a portion of the processing. In addition, embodiments of the present invention further relate to computer storage products with a computer-readable medium that have computer code thereon for performing various computer-implemented operations. The media and computer code may be those specially designed and constructed for the purposes of the present invention, or they may be of the kind well known and available to those having skill in the computer software arts. Examples of computer-readable media include, but are not limited to: magnetic media such as hard disks, floppy disks, and magnetic tape; optical media such as CD-ROMs and holographic devices; magneto-optical media such as floptical disks; and hardware devices that are specially configured to store and execute program code, such as application-specific integrated circuits (ASICs), programmable logic devices (PLDs) and ROM and RAM devices. Examples of computer code include machine code, such as produced by a compiler, and files containing higher level code that are executed by a computer using an interpreter. While the invention has been particularly shown and described with reference to specific embodiments thereof, it will be understood by those skilled in the art that changes in the form and details of the disclosed embodiments may be made without departing from the spirit or scope of the invention. In addition, although various advantages, aspects, and objects of the present invention have been discussed herein with reference to various embodiments, it will be understood that the scope of the invention should not be limited by reference to such advantages, aspects, and objects. Rather, the scope of the invention should be determined with reference to the appended claims. |
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041586018 | abstract | Automatic apparatus for loading a predetermined amount of nuclear fuel pellets into a nuclear fuel element to be used in a nuclear reactor. The apparatus consists of a vibratory bed capable of supporting corrugated trays containing rows of nuclear fuel pellets and arranged in alignment with the open ends of several nuclear fuel elements. A sweep mechanism is arranged above the trays and serves to sweep the rows of fuel pellets onto the vibratory bed and into the fuel element. A length detecting system, in conjunction with a pellet stopping mechanism, is also provided to assure that a predetermined amount of nuclear fuel pellets are loaded into each fuel element. |
claims | 1. A charged particle beam apparatus comprising:a charged particle source configured to discharge a charged particle beam;charged particle optics that adjusts the charged particle beam on the basis of an input voltage value;scanning means for moving an irradiation point of the charged particle beam with respect to a sample;observing means for obtaining an image of a surface of the sample irradiated with the charged particle beam; anda control unit that sets an adjusted input voltage value to the charged particle optics, the control unit comprising spot pattern forming means for sequentially forming spot patterns at different positions on the sample surface in accordance with different voltage values input to the charged particle optics, and spot pattern analyzing means for analyzing the spot patterns on the image and selecting the adjusted input voltage value corresponding to a spot pattern having the smallest outer diameter and/or the most circular shape of the spot patterns. 2. The charged particle beam apparatus according to claim 1; wherein the spot pattern forming means of the control unit moves the irradiation point of the charged particle beam with respect to the sample by the scanning means by an amount obtained by multiplying the amount of change of the input voltage value by a predetermined coefficient every time when the charged particle beam is irradiated on the sample after setting the input voltage value to the different value. 3. The charged particle beam apparatus according to claim 1; wherein the charged particle optics includes an electrostatic lens configured to focus the charged particle beam to cause the sample to be irradiated therewith by applying a voltage to an electrode of the electrostatic lens, and the control unit causes the spot pattern forming means to form the spot patterns by setting voltage values to be applied to the electrode of the electrostatic lens to different values as the input voltage values and causes the spot pattern analyzing means to select and set the adjusted input voltage value to thereby adjust the focal length of the electrostatic lens. 4. The charged particle beam apparatus according to claim 1; further including a stigmator which includes a multi-pole having a pair of opposed positive poles and a pair of negative poles opposed substantially orthogonally to the direction of arrangement of the positive poles and which corrects the cross-sectional shape of the charged particle beam into a substantially circular shape by applying a voltage between each pair of positive and negative poles of the multi-pole, and wherein the control unit causes the spot pattern forming means to set the voltage value to be applied between the positive poles and the negative poles of the stigmator to different values as input voltage values to form spot patterns, and causes the spot pattern analyzing means to select the spot pattern having the smallest spot characteristic value as the ratio of the short diameter with respect to the long diameter of the outer diameters of the spot pattern in the orthogonal two directions, and set the voltage value to the voltage value corresponding to the spot pattern having the smallest value, so that the beam diameter ratio in the orthogonal two directions of the charged particle beam is adjusted. 5. The charged particle beam apparatus according to claim 4; wherein the spot pattern forming means of the control unit brings the charged particle beam into an over focus state with respect to the sample and forms a first plurality of spot patterns on the surface of the sample with different voltage values, and then brings the charged particle beam into an under focus state with respect to the sample and forms a second plurality of spot patterns on the surface of the sample with the same voltage values as in the case of the over focus state, respectively, and the spot pattern analyzing means of the control unit performs pattern matching between the spot patterns formed in the over focus state and in the under focus state with the same voltage values and selects the set of the spot patterns whose ratio of matching of pattern matching elements is the highest. 6. The charged particle beam apparatus according to claim 5; wherein the spot pattern forming means of the control unit matches the arrangement of the first and second pluralities of spot patterns to the corresponding voltage values between the state of the over focus and in the state of under focus to form the spot patterns. 7. The charged particle beam apparatus according to claim 4; wherein the stigmator includes two of the multi-poles including a first multi-pole and a second multi-pole, the voltage value includes a set of a first voltage value to be applied between the positive pole and the negative pole of the first multi-pole and a second voltage value to be applied between the positive pole and the negative pole of the second multi-pole, and the spot pattern forming means of the control unit combines the first voltage value and the second voltage value in different manners, moves the charged particle beam in a first direction by an amount obtained by multiplying an amount of change of the first voltage value by a predetermined coefficient and in a second direction intersecting the first direction by an amount obtained by multiplying an amount of change of the second voltage value by the coefficient relatively with respect to the sample by the scanning means, and causes the same to be applied a plurality of times. 8. The charged particle beam apparatus according to claim 1; wherein the observing means comprises a scanning electron microscope (SEM) column that scans the sample surface with an electron beam to obtain a SEM image of the sample surface having the spot patterns. 9. The charged particle beam apparatus according to claim 8; further comprising a rare gas ion beam column that irradiates the sample with a rare gas ion beam. 10. A method of adjusting charged particle optics of a charged particle beam apparatus which is configured to irradiate a sample with a charged particle beam and in which the charged particle optics is configured to adjust and set a beam characteristic value of the charged particle beam on the basis of the value of an input voltage applied to the charged particle optics, the method comprising: a spot pattern forming step for forming a plurality of discrete spot patterns at different positions on a surface of a sample by setting the input voltage of the charged particle optics to different value, each corresponding to a different one of the spot patterns, and irradiating the sample with the charged particle beam at different positions, each corresponding to a different one of the set input voltage values; an image forming step for forming an image of the sample surface that has been spot patterned by the charged particle beam; a spot pattern analyzing step for selecting from the image of spot patterns the spot pattern having the smallest value from spot characteristic values which indicate the shapes of the respective spot patterns; and an input voltage value setting step for setting the input voltage value of the charged particle optics to a value equal to the input value corresponding to the charged particle beam irradiated when the spot pattern selected in the spot pattern analyzing step is formed. 11. The method of adjusting charged particle optics according to claim 10; wherein the spot pattern forming step moves an irradiation point of the charged particle beam with respect to the sample by an amount obtained by multiplying an amount of change of the input voltage value by a predetermined coefficient every time when the charged particle beam is irradiated on the sample after setting the input voltage value to the different value. 12. The method of adjusting charged particle optics according to claim 10; wherein the charged particle beam apparatus includes an electrostatic lens configured to focus the charged particle beam to cause the sample to be irradiated therewith by applying a voltage to an electrode of the electrostatic lens, the spot pattern forming step forms the spot patterns by setting voltage values to be applied to the electrode of the electrostatic lens to different values as the input voltage values, and the spot pattern analyzing step selects the spot pattern having the smallest spot characteristic value as the outer diameter of the spot pattern, and sets the input voltage value by the input voltage value setting step, so that the focal length of the electrostatic lens as the beam characteristic value is adjusted. 13. The method of adjusting charged particle optics according to claim 10; wherein the charged particle beam apparatus includes a stigmator which includes a multi-pole having a pair of opposed positive poles and a pair of negative poles opposed substantially orthogonally to the direction of arrangement of the positive poles and which corrects the cross-sectional shape of the charged particle beam into a substantially circular shape by applying a voltage between each pair of positive and negative poles of the multi-pole, the voltage value to be applied between the positive poles and the negative poles of the stigmator is set to different values as input voltage values to form the spot patterns in the spot pattern forming step, the spot pattern having the smallest spot characteristic value is selected as the ratio of the short diameter with respect to the long diameter of the outer diameters of the spot pattern in the orthogonal two directions in the spot pattern analyzing step, and the input voltage value is set in the input voltage value setting step, so that the beam diameter ratio in the orthogonal two directions of the charged particle beam as the beam characteristic value is adjusted. 14. The method of adjusting charged particle optics according to claim 13; wherein the spot pattern forming step includes a first step of bringing the charged particle beam into an over focus state with respect to the sample and forming a first plurality of spot patterns on the surface of the sample with different voltage values, and a second step of bringing the charged particle beam into an under focus state with respect to the sample and forming a second plurality of spot patterns of the surface of the sample with the same voltage values as in the first step, respectively, and the spot pattern analyzing step performs pattern matching between the spot patterns formed in the first step and in the second step with the same voltage values and selects the set of the spot patterns whose ratio of matching of pattern matching elements is the highest. 15. The method of adjusting charged particle optics according to claim 14; wherein the first step and the second step of the spot pattern forming step match the arrangement of the first and second pluralities of spot patterns to the corresponding input voltage values to form the spot patterns. 16. The method of adjusting charged particle optics according to claim 13; wherein the stigmator of the charged particle beam apparatus includes two of the multi-poles including a first multi-pole and a second multi-pole, the voltage value includes a first voltage value to be applied between the positive pole and the negative pole of the first multi-pole and a second voltage value to be applied between the positive pole and the negative pole of the second multi-pole and, in the spot pattern forming step, the first voltage value and the second voltage value are combined in different manners, and the charged particle beam is moved in a first direction by an amount obtained by multiplying an amount of change of the first voltage value by a predetermined coefficient and in a second direction intersecting the first direction by an amount obtained by multiplying an amount of change of the second voltage value by the coefficient relatively with respect to the sample, and applied by a plurality of times. 17. The method of adjusting charged particle optics according to claim 10; wherein the spot pattern analyzing step compares the smallest spot characteristic value and a preset spot reference value and, if the spot characteristic value is larger than the spot reference value, the procedure goes again to the spot pattern forming step, and the spot pattern forming step changes the input voltage value by an amount of change smaller than the amount of change when the input voltage value of the charged particle optics is changed in the previous spot pattern forming step to form the spot pattern. 18. The method of adjusting charged particle optics according to claim 10; wherein in the spot pattern analyzing step, the input voltage value corresponding to the spot pattern having the smallest spot characteristic value is compared with the input voltage values corresponding to the other spot patterns and, if the input voltage value corresponding to the spot pattern having the smallest spot characteristic value is the smallest or the largest, the procedure goes to the spot pattern forming step again, and in the spot pattern forming step, the input voltage value is changed within a range including values smaller than the input voltage value if the input voltage value corresponding to the spot pattern having the smallest spot characteristic value is the smallest, and within a range including values larger than the input voltage value if the input voltage value is the largest to form the spot pattern again. 19. The method of adjusting charged particle optics according to claim 10; wherein the spot pattern analyzing step creates binary data obtained by binarizing the image of the sample on which the spot pattern is formed, and selects the spot pattern having the smallest spot characteristic value from the binary data. 20. The method of adjusting charged particle optics according to claim 10; further comprising an adjustment preparation step for arranging a standard sample and a target sample as the sample, and a processing and observation preparation step for adjusting the position of the target sample with respect to the irradiation point of the charged particle beam, wherein the spot pattern forming step, the spot pattern analyzing step, and the input voltage value setting step for the standard sample are performed after the adjustment preparation step, and the processing and observation preparation step is carried out after the input voltage value setting step. 21. The method of adjusting charged particle optics according to claim 10; wherein the image forming step is performed using a scanning electron microscope (SEM) column to obtain a SEM image of the sample surface that has been spot patterned by the charged particle beam. |
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054266793 | abstract | A strainer device for filtering water to an emergency cooling system in a nuclear power plant containment area having a lower part which forms a pool in which the strainer device is mounted. The device includes a strainer housing with separate strainer walls between which a flexible strainer means operates. The unit is arranged to be kept in a first position when water is drawn through one of the strainer walls, in which position it interrupts the connection between the other strainer wall and a suction conduit. By inherent flexibility, a shield means is automatically switched over to a second position which opens the connection between the second strainer wall and the suction conduit, as soon as the first strainer wall is clogged by impurities to such an extent that a low-pressure arises in the space between the first strainer wall and the shield means unit, which low-pressure brings about the switching-over by its suction effect. |
claims | 1. Container intended for transport/storage of radioactive materials, comprising a container body and at least one handling device assembled on said container body, each handling device being provided with a main part capable of cooperating with a gripping mechanism and projecting from the container body, and a base fixed to the main part and located in a base housing delimited by a base housing wall formed on the container body, the container also comprising firstly a plurality of attachment screws for each handling device distributed around the main part and attaching the base onto the container body, and secondly sealing means inserted between the base of the handling device and the container body, characterised in that the sealing means for each handling device comprise a sealing plate located in a plate housing delimited jointly by a plate housing wall provided on the base of the handling device and by a portion of the base housing wall, said sealing plate being installed removably in the plate housing so as to surround the main part of said handling device and to cover each attachment screw, said sealing means also including an external seal inserted between a peripheral wall external to said sealing plate and the portion of the base housing wall partially delimiting the plate housing, and an internal seal inserted between a peripheral wall internal to said sealing plate and the plate housing wall. 2. Container according to claim 1, characterised in that the external peripheral wall of the sealing plate comprises an external edge in contact with the external seal, and in that the internal peripheral wall of the sealing plate comprises an internal edge in contact with the internal seal. 3. Container according to claim 2, characterised in that the external edge has an external groove extending all along said external edge and inside which the external seal is located, and in that the internal edge has an internal groove extending all along said internal edge and inside which the internal seal is located. 4. Container according to claim 3, characterised in that for each handling device, the plate housing wall provided on the base of the handling device comprises a shoulder, the internal seal housed in the groove of the internal edge of the sealing plate bearing in contact with an inside surface of said shoulder in order to maintain said sealing plate in the plate housing, the internal seal being compressed between the groove of the internal edge and a part of the maximum diameter of the shoulder, to enable assembly/disassembly of said sealing plate. 5. Container according to claim 4, characterised in that at least one access orifice provided in the main part of the handling device is capable of holding pressurisation/vacuum creation means that can generate a pressure/vacuum inside a space partially delimited by the inside surface of the sealing plate and surrounding the main part of the handling device, through a channels network in order to cause assembly/disassembly of the sealing plate. 6. Container according to claim 2, characterised in that the internal edge of the sealing plate and the plate housing wall provided on the base of the handling device, each have a threaded portion cooperating with each other. 7. Container according to claim 1, characterised in that each handling device is provided with a channels network for making a sealing test of the sealing means, the channels network communicating at least with an access orifice provided in the main part of the handling device so as to open up on the outside of said main part, each access orifice being closed off using a removable plug. 8. Container according to claim 7, characterised in that for each handling device, the sealing plate has an inside surface partially delimiting a space surrounding the main part of the handling device and partly filled in by the heads of the attachment screws, the channels network being arranged so as to enable communication between said space and at least one access orifice. 9. Container according to claim 8, characterised in that at least one access orifice provided in the main part of the handling device is capable of holding pressurisationl/vacuum creation means that can generate a pressure/vacuum inside the space partially delimited by the inside surface of the sealing plate and surrounding the main part of the handling device, through the channels network in order to cause assembly/disassembly of the sealing plate. 10. Container according to claim 1, characterised in that the sealing plate for each handling device is in the shape of a ring and in that the external and internal seals are each in the shape of an annular seal. 11. Container according to claim 10, characterised in that the sealing plate for each handling device is installed screwed in the plate housing. 12. Container according to claim 11, characterised in that an internal edge-of the sealing plate and the plate housing wall provided on the base of the handling device, each have a threaded portion cooperating with each other. 13. Container according to claim 1, characterised in that the sealing plate for each handling device, is in the shape of a frame, and in that each of the external and internal seals is also in the shape of a frame. 14. Container according to claim 1, characterised in that the sealing plate for each handling device is installed clipped in the plate housing. 15. Container according to claim 14, characterised in that for each handling device, the plate housing wall provided on the base of the handling device comprises a shoulder, the internal seal housed in a groove of an internal edge of the sealing plate bearing in contact with an inside surface of said shoulder in order to maintain said sealing plate in the plate housing, the internal seal being compressed between the groove of the internal edge and a part of the maximum diameter of the shoulder, to enable assembly/disassembly of said sealing plate. 16. Container according to claim 1, characterised in that the sealing plate for each handling device is made of stainless steel. 17. Container according to claim 1, characterised in that for each handling device, each of the external and internal seals is made from an elastomer material. |
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051990571 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS The configuration of an image formation-type X-ray microscopic apparatus in accordance with the present invention is described below with reference to the embodiment shown in the drawings. In FIG. 1, the laser beams emitted from a pulse laser 1 are condensed on a disk- or tape-shaped thin film target 5 by a condensing lens 3 through a vacuum holding window 4 so as to cause the target 5 to generate X-rays having necessary intensity and wavelength. The light emission of the pulse layer 1 is controlled by a pulse control unit 2 with a desired pulse separation (maximum, 30 Hz). The X-rays generated from the X-ray thin film target 5 are condensed on a sample 13 in a sample container 12 by a rotary elliptical multilayer film reflecting mirror 9. A sample image, which is enlarged 100 times (resolving power, 100 nm) to 500 times (resolving power 20 nm), is formed on a two-dimensional X-ray imaging element 15 by using as an image-formation optical system a phase zone plate PZP 14. For example, it is effective to use as the two-dimensional X-ray imaging element 15 a solid-state image sensor such as a back irradiation-type FT-CCD. As shown in the drawing, in this arrangement, the X-ray thin film target 5, which is excited by the pulse laser, is disposed at the first focal point of the rotary elliptical multilayer film reflecting mirror 9, and the sample 13 is disposed at the second focal point thereof. The X-rays are monochromatized by the multilayer film so that the X-rays, which are emitted by excitation by the pulse laser, are applied to the sample, and photon counting imaging is performed by the two-dimensional X-ray imaging element 15. In order to observe the sample 13, which is horizontally held, the X-rays for irradiation and observation are arranged in the vertical direction, and the laser beams for excitation of the X-rays are arranged in the horizontal direction. Specifically, the angle between the target 5 and the excitation laser beams is about 35.degree., and the angle of incident of the X-rays on the rotary elliptical multilayer film reflecting mirror 9 is about 65.degree.. A maintenance apparatus 6 for replacing the X-ray thin film target 5 and removing scattered substances and the like, a diaphragm 7 and a shielding window for neutral particles and plasma are provided so that the X-rays are applied to the sample in a predetermined direction. The shielding window 8 also serves as a filter for adjusting the intensity n.sub.max of the X-rays to a value of less than 200. A water cooling apparatus is provided on the rear side of the rotary elliptical multilayer film reflecting mirror 9 serving as a condenser for the purpose of preventing temperature rising and deterioration, which are caused by the absorption of the X-rays. A field diaphragm 11 is provided just ahead of the sample container 12, the aperture diameter thereof being changed to an appropriate value in correspondence with the observation magnification. However, even if the magnification is the same, the aperture diameter is changed to an appropriate value for the purpose of preventing the occurrence of flare and improving contrast. The image information output from the two-dimensional X-ray imaging element 15 is processed by an image processing part 16 and then output to an image output part 17 such as a display, a printer or the like. The image processing part 16 has the function to adjusting the maximum number n.sub.max of the detected photons to a value of less than 200 in linkage with the intensity adjusting filter 8 provided on the X-ray source side. Of the above-described components, the components from the rotary elliptical multilayer film reflecting mirror 9 serving as a condenser to the light-receiving surface of the two-dimensional X-ray imaging element 15 are received in a vacuum container 18 for holing them under vacuum. The pressure in the vacuum container 18 is kept at about 10.sup.-2 Pa at which absorption of the X-rays is negligible. Since the maintenance apparatus 6 or the like must be provided around the X-ray thin film target 5 because scattered substances are generated around the target 5, it is necessary to isolate the X-ray source part by another vacuum container 19. A detailed description will now be given of the optimum specifications of the X-ray source and the optical system for realizing photon counting imaging for a moment of 1 pulse in a level of msec or less when the maximum number n.sub.max of the detected photons is 100 in the above-described arrangement. If the efficiency of utilization of the X-ray source and the efficiency of the optical system are determined, the final pulse intensity (specification of the X-ray source) of the X-ray source is determined. When a laser excitation type of X-ray source is used, the specifications of laser are determined. Various viewpoints are in turn described below. (vii) Diameter and Efficiency of Utilization of Nondirectional Light Source FIG. 4 shows the basic configuration of the image-formation X-ray microscope. In the drawing, the X-rays emitted from the target 5 are condensed on the sample 13 in the sample container 12 by the condenser 9, and an enlarged image is formed on the two-dimensional X-ray imaging element 15 through an objective optical system 14. In this embodiment, the efficiency and the transmittance are defined as follows: NA.sub.c : Numerical aperture on incident side of condenser NA.sub.o : Numerical aperture on incident side of object .phi..sub.s : Diameter of X-ray source .phi..sub.c : Diameter of X-ray source which can be utilized by condenser .phi..sub.o : Diameter of real visual field of sample (diaphragm diameter) .eta..sub.w : Spectral utilization efficiency of X-ray source .eta..sub.s : Spatial utilization efficiency of X-ray source .eta..sub.c : Efficiency of condenser (condensation efficiency) .eta..sub.o : Efficiency of object T.sub.s : Transmittance of sample T.sub.c : Transmittance of window of sample container The utilization efficiency of the X-ray source is the product of the spatial utilization efficiency and the spectral utilization efficiency. It is necessary to design the X-ray microscope by selecting the type of the X-ray source and the mode and type of the optical system so that the necessary performance is satisfied, and the efficiency is the highest. (1) The efficiency .eta..sub.s of spatial utilization of the X-ray source by the optical system is expressed by the following equation: ##EQU13## wherein .phi..sub.c .ltoreq..phi..sub.s and NA.sub.c .ltoreq.1. From the sine condition and the objective visual field and numerical aperture, the following equation is established: EQU NA.sub.c .multidot..phi..sub.c =NA.sub.o .multidot..phi..sub.o(23) The use of this equation permits the determination of the actual efficiency .eta..sub.s. As described below, in the present invention, the value of the equation (23) is 1.125 .mu.m regardless of the magnification. As shown by the above equation, the high the utilization efficiency, the smaller the diameter of the light source as far as a condenser having a large numerical aperture can be formed. The optimum diameter of the light source and the numerical aperture of the condenser are the following: ##EQU14## Since the numerical aperture must be increased if the diameter of the light source is smaller than the above value, the condenser cannot be easily designed and manufactured, and, when a laser plasma X-ray source is used, the efficiency of X-ray generation is decreased. Conversely, if the diameter of the light source is large, the utilization efficiency is deteriorated, and the light source cannot be used for imaging with one pulse because of its shortage of intensity. (2) From the spectral width (.DELTA..lambda./.lambda.=1/200) obtained by the multilayer film reflecting mirror and the spectral characteristics of the X-ray source, the spectral utilization efficiency of the X-ray source is the following: EQU .eta..sub.w .perspectiveto.10.sup.-1 .multidot..DELTA..lambda./.lambda..perspectiveto.5.times.10.sup.-4 (viii) Efficiency of Optical System On the other hand, when the wavelength used is 2.5 nm, the realizable efficiency of the optical system is described below. If the multilayer film comprises nickel Ni and vanadium V, the reflection efficiency .eta..sub.c of the rotary elliptical multilayer film reflecting mirror serving as a condenser is the following: EQU .eta..sub.c .perspectiveto.0.3 If the sample container is composed of Si.sub.3 N.sub.4, the transmittance T.sub.c of the window of the sample container is the following: EQU T.sub.c .perspectiveto.0.63 Since the efficiency of the object .eta..sub.o is equal to the efficiency .eta..sub.1 of the first diffracted light of the phase zone plate, the object efficiency .eta..sub.o is the following: EQU .eta..sub.o .perspectiveto.0.3 (ix) X-Ray Pulse Intensity and Output The pulse intensity and output of the X-ray source are calculated by using the above-described utilization efficiency of the light source, the efficiency of the optical system, the transmittance of the sample and the sensitivity of the detector (photon counting). From (vii) and (viii), the overall efficiency .eta. of the optical system of the X-ray microscope is as follows: ##EQU15## If the number of the photons of the X-ray source corresponding to a single pixel of the detector is n.sub.s, the number n.sub.I of the photons which reach a single pixel of the detector is expressed by the following equation: EQU n.sub.I =T.sub.s .multidot..eta..multidot.n.sub.s (25) If the maximum transmittance of the sample is T.sub.smax =T.sub.w =0.273, and if the maximum number of the photons detected is n.sub.max =100, the necessary photon number n.sub.s for the X-ray source is the following: ##EQU16## If the energy of the X-ray photons .epsilon..sub.p is .epsilon..sub.p =h.gamma.=7.9.times.10.sup.-17 J [.lambda.=2.5 nm], and if the pixel number N.sub.I in the circumscribed circle of the imaging element, which obtained by the pixel number in accordance with the NTSC system is the following: EQU N.sub.I .perspectiveto..pi..times.(450).sup.2 =6.4.times.10.sup.5, the X-ray pulse intensity P.sub.x required for obtaining an X-ray image on a sheet and the X-ray output I.sub.x required for obtaining 30 images per second are the following: ##EQU17## (x) Laser Pulse Intensity and Output It is preferable to generate the X-rays by using laser excitation plasma, as described below. If the efficiency .eta..sub.x of generation of the X-rays by using a laser is .eta..sub.x =0.1, from the equation, P.sub.x =.eta..sub.x .multidot.P.sub.L, the laser pulse intensity P.sub.L and the laser output I.sub.L are the following: EQU P.sub.L =500 [mJ/image] EQU I.sub.L =15 W The above-described X-ray source and pulse laser can be manufactured. (xi) Specification and Type of X-Ray Source Although the optimum specifications of the X-ray source are determined as described above, a laser plasma X-ray source, which uses a pulser laser, is optimum as an X-ray source which conforms to the specifications. Other types of plasma X-ray sources and electron beam excitation X-ray sources cannot be brought into practical use because the utilization efficiency is deteriorated owing to the sizes of the X-ray sources, which are as large as 0.1 to 1 mm .phi., as described in (vii) (1). On the basis of the above results, a nondirectional X-ray source must satisfy the following conditions: a) Diameter of X-ray source: .about.10 .mu.m.phi. b) X-ray spectrum: 2.3 nm to 4.4 nm c) X-ray pulse width <1 .mu.s d) X-ray pulse intensity >50 mJ e) X-ray repetition frequency .about.30 Hz Although a laser excitation plasma X-ray source is suitable for satisfying these conditions, it is effective that higher harmonics of a slab laser or a pulse laser such as an excimer laser is used as the laser. Specifically, a target material having a thickness of several .mu.m is selected, and the above-described conditions are satisfied by optimization with respect to the items (f) to (j) described below. f) Laser/X-ray conversion efficiency >0.1 g) Laser wavelength .about.250 nm h) Laser condensing diameter .about.10 .mu.m.phi. i) Laser pulse width <1 .mu.s j) Laser pulse intensity >500 mJ (xii) Condenser System and Optimization Thereof From the efficiency hc and the numerical aperture of the condenser and the degree of monochromatization, which are described above, the following conditions are determined: a) In order to reduce the size of the X-ray source, it is necessary that the reflectance (efficiency) hc is about 30%. b) In order to effectively utilize the X-ray source, it is necessary that the numerical aperture MAc is about 0.125. c) The degree of monochromatization (spectral width) for photon counting imaging with one pulse is the following: EQU .lambda./.DELTA..lambda.=200 Further, the illumination system must satisfy the following conditions (d), (e) and (f): (d) In order to perform effective illumination by using a minimum optical element, a critical illumination method is used in which a light source has the following magnification M: M=real visual field/diameter of X-ray source (e) In order to effectively condense X-rays and remove the irregularity of illumination, it is necessary to effectively correct aberration. f) In order to reduce the size of the illumination system, it is preferable that the vertical distance between the X-ray source and the sample is about 200 to 800 mm. g) Since extremely strong X-rays with intensity, which is about 2000 times that of the X-rays applied to the sample, is applied to the condenser, as described above in (vii) (2), the condenser must be cooled for protecting the optical element. In order to satisfy the conditions, it is effective to use as the condenser a rotary elliptical multilayer film mirror having a multilayer film composed of nickel and vanadium with a number of periods N.sub.c of about 200. The above conditions cannot be easily satisfied by any other optical elements such as a zone plate and total reflecting mirrors. For example, although the annular zone number and the numerical aperture of a zone plate must be increased, such a zone plate cannot be easily manufactured. Specifically, when the distance L.sub.c between the light source and the sample is 100 mm, NA.sub.c is 0.125 and the magnification M of the light source is 2 (resolving power, 20 nm; combination with an object with magnification of .times.500), from the equations below, the annular zone number of the zone plate is 312500, the minimum line width is 6.7 nm and the zone radius is 4.17 mm. Such a zone plate cannot be easily produced by the present technique. In addition, the zone plate cannot be easily cooled because of its structure. In a case of a total reflecting mirror, since another spectral element is required for monochromatizing X-rays the spatial utilization efficiency of the light source is deteriorated owing to the shielding of the aperture portion. In a case of a spherical multilayer film mirror, since it is difficult to correct aberration by using a single surface, uniform illumination cannot be effectively made. The use of two-surface reflection for correcting aberration causes a deterioration in efficiency, which caused by low reflectance and shielding of the aperture portion. For example, in a case of a Schwarzschild type which uses two reflecting spherical surfaces, even if a multilayer film having reflectance of 30% is used, the condensing efficiency is about 5% at most if the shielding of the central portion is considered. From the above-described viewpoints, a rotary elliptical multilayer film reflecting mirror, which permits good aberration correction and effective condensation by one reflection and which has a simplest shape, is optimum as the condenser. In addition, with respect to the processing precision of a rotary elliptical surface, the shape precision is about 200 nm and the surface roughness is about 0.2 nm, apart from the surface used in an image-formation system. Hence such a rotary elliptical surface can be easily produced. In regard to cooling of the element, the rear side of the rotary elliptical multilayer film mirror can be effectively cooled with water. FIG. 5 shows the spectral characteristics of the reflected X-rays in such a multilayer film mirror. In FIG. 5, the wavelength .lambda. (or reflection angle .theta.) is shown on the abscissa, and the X-ray intensity is shown on the ordinate. FIG. 6 is a schematic sectional view which shows the structure of the multilayer film. In the drawing, the angle of incidence of an X-ray upon the multilayer film is considered as an angle .theta. measured on the basis of the nominal line. The width .DELTA..lambda. of the spectrum obtained by the multilayer film is determined as described below. The coherence distance (.lambda..sup.2 /.DELTA..lambda.) and the number Nc of periods of the multilayer film and the thickness d.sub.T of one period of the multilayer film have the following relation: EQU 2N.sub.c .multidot.d.sub.T cos .theta.=.lambda..sup.2 /.DELTA..lambda.(28) From the conditions for coherence between the adjacent periods, the following Bragg's condition is established: EQU 2d.sub.T cos .theta.=.lambda. (29) From the equations (28) and (29), the relation expressed by the above-described equation (21) is obtained. As shown in FIG. 7, in the rotary elliptical multilayer film mirror, the incident angles of the X-rays generated from the target 5 are different in the respective reflection regions of the rotary elliptical multilayer film mirror 9, and the variation in incident angle exceeds the allowable width of reflection angle described below. In a multilayer film having the uniform structure, therefore, reflectance is significantly decreased in some region. It is therefore necessary to change the thickness ratio of in each region of the multilayer film in correspondence with the change in incident angle. However, the width .alpha. of the incident angle at the same reflection point is within an allowable width of reflection angle. Specifically, as shown in FIG. 7, the incident angle of the X-ray at the central portion of the rotary elliptical reflecting mirror 9 is .PSI., and the incident angles at both ends of the reflecting mirror 9 are respectively .PSI.+.DELTA..PSI. and .PSI.-.DELTA..PSI.. If an angle width, which can maintain predetermined reflectance as a multilayer film if the incident angle is changed, i.e., the allowable incident angle width, is d.theta., from the equations (21) and (29), the allowable angle width for effective reflection of the multilayer film mirror and the number of period N.sub.c of the multilayer film have the following relationship: EQU tan .theta..multidot.d.theta..perspectiveto.1/N.sub.c (30) On the other hand, if the sine condition is satisfied in the illumination system, the following equation is established: EQU .DELTA..PSI..perspectiveto.{sin.sup.-1 (NAc)=sin.sup.-1 (NA.sub.c /M)}/2 In addition, since NAc<<1, EQU .DELTA..PSI..perspectiveto.NA.sub.c (M-1)/(2.multidot.M) It is therefore preferable to divide the reflection region of the rotary elliptical body into bands by the number obtained by the following equation: ##EQU18## It is also preferable to form the multilayer film comprising regions each having an optimized thickness ratio. For example, when .theta.=65.degree. and M=2, .DELTA..PSI./d.theta..perspectiveto.13. It is therefore preferable that the reflection region of the multilayer film of the rotary elliptical body is divided into 26 band regions. As described above, .theta.=65.degree. is an angle which is suitable for observing the sample while horizontally maintaining it. For the condenser optical system in accordance with the present invention, a rotary elliptical multilayer film reflecting mirror such as that described above is most practical. However, an arrangement based on a combination of a concave reflecting mirror and a diffraction grating is also effective. In this case, a wavelength selecting function is achieved by the diffraction grating, while in the case of a rotary elliptical multilayer film reflecting mirror, a multilayer film is used to reflect X-rays from an X-ray source selectively in a desired wavelength range at a high efficiency. Examples of such an arrangement will be described later in (xiv) "Other Arrangements of the Condenser System." (xiii) System and Optimization of Objective Optical System A phase zone plate, which has high efficiency, high resolving power and high magnification and which can be formed into a small size, is used in the objective optical system. The conditions necessary for the objective optical system are the following items (a) to (d): a) The efficiency is increased for reducing damage to the sample. EQU .eta..sub.o .perspectiveto.30% b) Since the resolving power of an optical microscope, which is capable of observing organisms, is 200 nm, the resolving power .delta. is 20 to 100 nm which is superior to the above value. c) In order to enable the attainment of high resolving power by using the imaging element, the magnification .beta. is determined as a value which is obtained by dividing the pixel dimension (10 nm) of the imaging element by desired resolving power. EQU .beta.=100 to 500 (high magnification) d) In order to reduce the overall size of the apparatus, the distance L.sub.o between the object and the image is about 400 mm or less. Each of the above conditions can be satisfied by the phase zone plate, as in the design examples below. However, it is difficult to realize an objective optical system having the same specifications by using a mirror such as a total reflecting mirror or a multilayer film reflecting mirror for the reasons described below. It is difficult to reduce the size to a level, in which the distance between the object and the image is 200 nm, by using a mirror with high magnification. It is also difficult to manufacture a mirror with high magnification because the shape accuracy is as restrict as the order of wavelength (several nm). In addition, when an objective multilayer film mirror is formed, since two surfaces having low reflectance must be used for aberration correction, the aperture is partly shielded. Thus, the efficiency is low, and the optical adjustment is difficult. As shown in FIG. 8, a zone plate is generally a zonal optical element and, as shown in FIG. 9, it is classified into a Fresnel zone plate FZP, a phase zone plate PZP and a saw tooth zone plate BZP according to their sectional shapes. Although BZP among these zone plates exhibits the highest efficiency, since BZP used for the X-ray region cannot be easily manufactured, PZP, which is next to BZP with respect to efficiency, is generally used. It is necessary to make an attempt to improve the efficiency of such a phase zone plate and reduce the occurrence of flare. It is also necessary to coordinate the irradiation spectral width and the chromatic aberration, secure the effective visual field diameter and keep the focal surface constant (so-called "parfocal") even if the magnification is changed. Each of the above viewpoints is described below with reference to typical examples of design of the phase zone plate. (1) Efficiency of Phase Zone Plate FIG. 10 is a drawing which shows the diffracted light of the zone plate. As shown in the drawing, many orders of diffracted light are present. The focal distance and diffraction efficiency of each of the orders of diffracted light of the phase zone plate, which has a negligible thickness, are generally expressed by the following equations: ##EQU19## wherein f.sub.m : Focal distance of m-order diffracted light f: Focal distance of first diffracted light which is used for forming an image .eta..sub.m : Diffraction efficiency of m-order diffracted light (.eta..sub.o .ident..eta..sub.1) .eta..sub.o : Diffraction efficiency of zero-order diffracted light (direct light) .eta..sub.ab : Absorption efficiency of phase zone plate T.sub.z : Amplitude transmittance of phase zone plate .chi.: Phase difference between adjacent (dark and light) zones If the amplitude transmittance is T.sub.z =1 and the phase difference is .chi.=.pi., the efficiency is ideally 40%. However, in fact, since the material of the zone plate absorbs light, the highest efficiency with a wavelength of 2.5 nm is about 30%. However, if the sectional shape is a step-like shape or a saw tooth-like shape, the efficiency can be further increased (for example, Japanese Patent Laid-Open No. 1-142604 by the same applicant as this application). Such a high-efficiency zone plate is also significantly effective for reducing flare, as described below. (2) Flare of Phase Zone Plate Although the first diffracted light having the highest diffraction efficiency is generally used for forming an image, the other orders of diffracted light are made flare and thus they must be removed. FIG. 11 is a drawing which shows the state wherein flare is generated by the m-order diffracted light. In the drawing, the solid lines show the image formed by the first diffracted light, and the broken lines show the image and flare formed by the m-order diffracted light. A small image having a height of h.sub.m is formed near the zone plate by the m-order diffracted light, and this image is effectively enlarged to a height of h.sub.mf to form a faded image on the surface of the image (height h.sub.I) formed by the first diffracted light. The relative intensity .zeta.m of the faded image formed by the m-order diffracted light on the surface of image formed by the first diffracted light is expressed by the following equation: EQU .zeta.m=(h.sub.I /h.sub.mf).sup.2 If the magnification by the first diffracted light of the zone plate is .beta., the radius of the phase zone plate (the radius of the outermost annular zone ring) is r.sub.N and the field diaphragm diameter is .phi..sub.o, the following equations are established: EQU h.sub.I =.beta..phi..sub.o /2 EQU h.sub.mf .perspectiveto.{(m-1)r.sub.N .+-..phi..sub.o /2} (wherein the plus sign denotes a case where m is 1 or more, and the minus sign denotes a case where m is 1 or negative). The relative intensity .zeta.m is therefore expressed by the following equation: EQU .zeta.m.perspectiveto.1/{(m-1)2r.sub.N /.phi..sub.o .+-.1}.sup.2(36) Since the flare coefficient .eta..sub.f of the optical system in the equation (10) is the sum of the products of the diffraction efficiency of diffracted light and the relative intensity of the geometrical faded image, the flare coefficient is the following: ##EQU20## In an optical system which uses a usual Fresnel zone plate of T.sub.z =0 or an actual phase zone plate of T.sub.z .noteq.1, the illumination method can be designed so that the zero-order diffracted light in the first term of the above equation is cut off. Only the flare produced by the higher-order diffracted light in the second term and the subsequent terms is therefore a problem. As seen from the equation (37), since the flare coefficient .eta..sub.fm of the m-order diffracted light is inversely proportional to the fourth power of the diffraction order, it is sufficient that zero-, -1-, .+-.3- and .+-.5-order diffracted light are considered as the main diffracted light for flare. If the field diaphragm diameter is one fifth of the zone plate diameter (.phi.o=2rN/5), the flare is reduced to about 1/100. In the microscope of this embodiment in which the maximum number of the photons detected is about 100, the photon number of flare can be restricted to about 1 or less by diaphragming the field so as to be negligible. From the above viewpoint, a field diaphragm (the real field is determined by the diaphragm diameter) which is determined by the diagonal length of the imaging element and a field diaphragm having a diameter of about one fifth of the zone plate diameter are prepared for each object so that a clear image without flare can be obtained as occasion demands. This is a condition for obtaining a clear image without flare even in a state wherein flare is remarkably produced, as in a case (T.sub.sm /T.sub.smax .perspectiveto.1) where black points are slightly produced in a light field. It is unnecessary to so severely restrict the visual field in practical use. As a result of further detailed investigation, the allowable number of flare n.sub.f is 1 or less per pixel on an average, as described above. Since this is expressed by the equation, n.sub.f <<n.sub.max, from the equations (10) and (11), the following equation is obtained: EQU n.sub.f .perspectiveto..eta..sub.f .multidot.(T.sub.sm /T.sub.smax).multidot.n.sub.max <1 Since the thickness of a usual living sample is 10 .mu.m, and the average protein thickness is t.sub.pm =1.5 .mu.m, from the equation (7), the average transmittance T.sub.sm =0.035. In addition, since T.sub.smax =T.sub.w =0.273, the following equation is obtained: EQU T.sub.sm /T.sub.smax =0.128 The allowable value Of the flare coefficient .eta..sub.f is the following: EQU .eta..sub.f <1/(0.128.multidot.n.sub.max) When the maximum number n.sub.max of the photons detected is 100, .eta..sub.f <0.08. From the equation (37), the field diaphragm diameter .phi..sub.o having the above flare coefficient is determined by the following equation: EQU .phi..sub.o .perspectiveto.7r.sub.N /5 In this way, when a usual sample is observed, the occurrence of flare can be made negligible by limiting the aperture diameter of the field diaphragm to 70% of the diameter of the zone plate. As described above, in order to form a clear image with negligible flare, the visual field may be limited in accordance with the sample used. It is sufficient for a usual sample that the field is reduced to 70% of the diameter of the zone plate. Even if the visual field is further reduced, it just becomes narrow. The real field diameter, which is determined by the diagonal length of the imaging element, a diaphragm for ordinary samples, which has an aperture of 70% of the zone plate diameter, and a diaphragm for limit samples, which has an aperture of 20% of the zone plate diameter, are provided for practical use so that practical observation can be made by properly using them. As disclosed in the above-described Japanese Patent Appln. Laid-Open No. 1-142604, when a high-efficiency zone plate producing no minus-order diffracted light is used, no diaphragm is inserted, but a shield of a size, which is substantially the same as the +3-order image or +5-order image, is disposed at the position of the image so that flare can be significantly reduced. (3) Relation Between Various Quantities of Zone Plate Equations necessary for designing a general optical system are given below. In regard to the resolving power .delta. and the focal depth D.sub.f, if the numerical aperture is NA.sub.o, and the wavelength is .lambda., the following equations are established: EQU .delta.=.lambda./{2(NA.sub.o)} (38) EQU D.sub.f =.+-..lambda./{2(NA.sub.o).sup.2 } (39) If the focal distance is f, the object-image distance is L.sub.o, and the magnification is .beta., the following equation is established: EQU f=.beta..multidot.L.sub.o /(.beta.+1).sup.2 (40) On the other hand, in the zone plate, the focal distance is the focal distance of the first diffracted light and has the following relation to the annular zone radius r.sub.k of the zone plate: EQU r.sub.k.sup.2 .perspectiveto.k.lambda.f (41) wherein the annular zone number; k=1, 2, 3, . . . , N.sub.o (light and dark zones are numbered 2) If the width of the outermost annular zone as the minimum zone width of the zone plate .DELTA.r.sub.N is expressed by the following equation: EQU .DELTA.r.sub.N =r.sub.N -r.sub.N-1, the various quantities of a general optical system and the various quantities of the zone plate have the following relations: ##EQU21## Since the magnification .beta., the resolving power .delta. and the wavelength .lambda. used are determined as specifications, if N.sub.o or L.sub.o is determined in accordance with the equation (45), all the quantities of the zone plate excluding the efficiency are determined by the above equations. The factors which determine N.sub.o or L.sub.o is the above-described flare and the allowable spectral width and effective field diameter below. (4) Coordination of Irradiation Spectral Width and Chromatic Aberration The chromatic aberration dZ' is generally expressed by the following equation: EQU dZ'=(1+.beta.).sup.2 df (46) If the chromatic aberration is smaller than the focal depth, there is no problem in practical use. Since the focal depth on the image surface is .beta..sup.2 D.sub.f, the following equation may be satisfied: EQU dZ'.ltoreq..beta..sup.2 D.sub.f (47) This is a condition for determining the allowable chromatic aberration. In regard to the zone plate, from the equation (41), the following equation is established: EQU df=-f.multidot.d.lambda./.lambda. From this equation and the equations (40), (45) and (46), the chromatic aberration of the zone plate is expressed by the following equation: EQU dZ'={N.sub.o (.beta./NA.sub.o).sup.2 }d.lambda. (48) From the equation, .DELTA..lambda.=2d.lambda., and the equations (39), (47) and (48), the following equation is established: EQU .lambda./.DELTA..lambda..gtoreq.N.sub.o (49) This equation (49) is a condition for the allowable chromatic aberration of the zone plate. It is necessary that the spectral width .DELTA..lambda. and the annular zone number N.sub.o of the zone plate satisfy the above relation. In the configuration of this embodiment, since the spectral width is determined by the multilayer film mirror so that .DELTA./.DELTA..lambda.=N.sub.c =200, as described above, it is necessary that N.sub.o .ltoreq.200. (5) Effective Field of Zone Plate FIG. 12 is a drawing which shows the state where an image surface is curved by the zone plate. The image formed by the zone plate is generally curved as shown by the solid line in the drawing. Since the numerical aperture NA.sub.o of the zone plate in the present invention is relatively as small as <0.1, the quality of the image formed can be evaluated by using the tertiary aberration coefficient of the zone plate. Since the curvature aberration coefficient is zero in accordance with the tertiary aberration coefficient of the zone plate, astigmatism and the curvature of the image are dominating factors which deteriorate the image. From the tertiary aberration coefficient, the radius RM of curvature of a meridional image surface is expressed by the following equation; EQU R.sub.M =f/3 (50) The effective visual field is determined by this value and the focal depth .beta..sup.2 D.sub.f. Namely, the region which, allows the curvature of the image surface to be within the focal depth of the image surface, is the effective field. FIG. 7 schematically shows the optical path in a relation of image formation between the zone plate and an object. An image of a sample 13 having a height of .phi..sub.o /2 is enlarged to a height of .beta..phi..sub.o /2 by a zone plate 14. Since a curved image surface I is within the range of the focal depth .beta..sup.2 D.sub.f on the image surface, the image can be detected as a clear image within the height range to .beta..phi..sub.o /2. This relation is expressed by the following equation: EQU (.beta..phi..sub.o /2).sup.2 /(2R.sub.M).ltoreq..beta..sup.2 D.sub.f(51) If the half view angle of the zone plate is .omega., the following equation is established: EQU .omega.=.beta..phi..sub.o /{2(1+.beta.)f} From the equations (45) and (51), therefore, the following equation is obtained: ##EQU22## Since the effective field .phi..sub.I is the following: EQU .phi..sub.I .perspectiveto.2.omega.L.sub.o, therefore, if L.sub.o in the equation (45) is substituted into this equation, the following equation is obtained: EQU .phi..sub.I .perspectiveto.2.sqroot.N.sub.o /3.multidot.{(.lambda..multidot..beta.)/NA.sub.o.sup.2 } (53) As shown in the plan view of FIG. 13, it is necessary that the effective image size is greater than the diagonal length of the imaging element 15. Since the effective diagonal length of the imaging element is about 9 mm, as described below, N.sub.o may be determined so that .phi..sub.I .gtoreq.9 mm on the basis of the equation (53). In FIG. 13, DI denotes the picture profile in the NTSC system, and C denotes the circumscribed circle thereof. (6) Design Example of Zone Plate In the above-described investigation, the condition No .ltoreq.200, which is determined from the viewpoint of coordination of the irradiation spectral width and the chromatic aberration described in (4), is incompatible with the condition, which is determined by the equation (53) assuming that .phi..sub.I .gtoreq.9 mm in practical use from the viewpoint of the effective field of the zone plate described in (5). It is therefore necessary to design in serious consideration of one of the viewpoints described in (4) and (5) to form a compromise between both viewpoints. Tables 1, 2 and 3 show the specifications of the image-forming X-ray microscopic apparatus which is designed on the basis of the above viewpoints in accordance with the present invention. Table 1 shows the specifications of an example which is designed in serious consideration of the coordination of the irradiation spectral width and the chromatic aberration described above in (4). Although this example has little chromatic aberration, the example significantly produces flare and a narrow effective visual field, which are caused by the zone plate having a small diameter, in accordance with the equation (37). The example shown in Table 2 is designed in serious consideration of the effective field of the zone plate described above in (5). In this case, although the effective field is wide, the chromatic aberration is large. Table 3 shows an example designed in serious consideration of flare. In this example, flare is little, and the effective field is wide. However, since the chromatic aberration is extremely large, it is necessary to further monochromatize X-rays. During actual design, since the degree of freedom for design includes only one of the annular zone number N.sub.o of the zone plate and the object-image distance L.sub.o, it is necessary to think one of the chromatic aberration, the effective field and flare important. Although it is therefore necessary to design in correspondence with the purpose of use, it it thought that the example shown in Table 2 is generally practical. In any one of the embodiments below, the magnification of the zone plate as an object is trivariant, and the focal surface is not changed even when the magnification is changed (parfocal). Since .beta.>>1, the focal distance f of the first diffracted light is substantially the same as the working distance. r.sub.I denotes the minimum radius of the annular zones. The flare coefficient .eta..sub.f is a value obtained when T.sub.z =1, .chi.=.pi. or when an illumination method for cutting off zero-order diffracted light is employed, and .phi..sub.o is the real field. TABLE 1 ______________________________________ Object-image distance L.sub.o = 64 mm Zone plate radius r.sub.N = 8 .mu.m Flare Magnifi- Focal Annular Minimum Effective co- cation distance zone radius field .phi..sub.I efficient .beta. f mm number N.sub.o r.sub.I mm mm .eta..sub.f ______________________________________ 100 0.628 40 1.25 11.7 0.605 200 0.317 80 0.89 8.26 0.409 500 0.128 200 0.59 5.23 0.152 ______________________________________ TABLE 2 ______________________________________ Object-image distance L.sub.o = 160 mm Zone plate radius r.sub.N = 20 .mu.m Flare Magnifi- Focal Annular Minimum Effective co- cation distance zone radius field .phi..sub.I efficient .beta. f mm number N.sub.o r.sub.I mm mm .eta..sub.f ______________________________________ 100 1.578 100 1.98 18.5 0.334 200 0.792 200 1.41 13.1 0.152 500 0.319 500 0.89 8.26 0.039 ______________________________________ TABLE 3 ______________________________________ Object-image distance L.sub.o = 400 mm Zone plate radius r.sub.N = 50 .mu.m Flare Magnifi- Focal Annular Minimum Effective co- cation distance zone radius field .phi..sub.I efficient .beta. f mm number N.sub.o r.sub.I mm mm .eta..sub.f ______________________________________ 100 3.92 250 3.13 29.5 0.113 200 1.98 500 2.22 20.8 0.039 500 0.797 1250 1.41 12.7 0.008 ______________________________________ The above-described numerical examples have the common specifications shown below in Table 4. TABLE 4 ______________________________________ Magnifi- Numerical Resolving Focal Real cation aperture power depth field .beta. NA.sub.o .delta. nm D.sub.f nm .phi..sub.o mm ______________________________________ 100 0.0125 100 8000 90 200 0.025 50 2000 45 500 0.0625 20 300 18 ______________________________________ (The resolving power equals to the minimum zone width since .beta. >> 1) (xiv) Other Arrangements of the Condenser System As mentioned above, it is also effective to arrange the condenser system of the present invention by using a combination of a concave reflecting mirror and a diffraction grating in place of the rotary elliptical multilayer film mirror. An example of this arrangement will be described below. One of the simplest arrangements based on a combination of a concave reflecting mirror and a diffraction grating is such that a toroidal concave diffraction grating is used and is integrally combined with a concave reflecting mirror. In this case, the construction is generally the same as that shown in FIG. 1. That is, the same construction can be used except that the rotary elliptical surface multilayer film mirror 9 shown in FIG. 1 is used as a toroidal concave diffraction grating. In this arrangement, X-rays from the x-ray thin film target 5 caused by the pulse laser 1 are condensed by the toroidal concave diffraction grating 9, and a sample 13 is disposed at the focal point of this condensing. The X-rays are monochromatized by the diffraction grating 9. The sample is thereby irradiated with the one-pulse X-rays excited and emitted by the pulse laser, and photon count imaging is effected by the two-dimensional X-ray imaging element 15. If the radius of curvature of the toroidal concave diffraction grating is Rv, the angle of incidence from the X-ray source upon the center of the diffraction grating (main incidence angle) is .alpha., the distance therebetween is r, the angle of diffraction of X-rays from the center of the diffraction grating to the the sample is b, and the distance therebetween is r', then it is desirable that the following equations (17) and (18) are satisfied. ##EQU23## If the X-ray source and the sample are on the Rowland circle of the diffraction grating, high-order spherical aberrations can be corrected. It is therefore advantageous to set the X-ray source and the sample in this manner. In this case, r and r' are EQU r=Rh cos .alpha., r'=Rh cos .beta.. (19) The radius of curvature can be calculated from EQU Rv/Rh=cos .alpha..multidot.cos .beta.. (20) Examples of the values of this system specified when the wavelength of X-rays is 3.7 nm and .alpha. is 86.degree. are shown below. ______________________________________ Diffraction grating ______________________________________ Slit distance d: 0.5 .mu.m Main curvature radius Rh: 1.5 m Sub curvature radius Rv: 14.6 mm Effective size: 100 .times. 20 mm Wavelength dispersion: 0.33 .ANG./100 .mu.m ______________________________________ The disposition of the X-ray source and the sample EQU r=104.6 mm, .alpha.=86.degree. EQU r'=209.9 mm, .beta.=81.96.degree. It is known that the efficiency of the diffraction grating is about 5% in this wavelength range. In this example, NA changes with respect to meridional and sagittal light beams but is 0.04 on average; the numerical aperture also satisfies the condition. If the size of the X-ray source is 45 .mu.m, the field of view on the sample surface is 45.times.90 .mu.m and the wavelength width is 0.15 .ANG., so that .lambda./.DELTA..lambda.=240. A rotary elliptical surface diffraction grating may be used instead of the toroidal surface diffraction grating. If the diffraction grating slit distance is the same as the above-described embodiment, i.e., 0.5 .mu.m, suitable spectral illumination light aberration-corrected by the same disposition of the X-ray source, the diffraction grating and the sample as that described above can be obtained by using a diffraction grating in which the lengths of the three axes (ra, rb, rc) of the rotary elliptical surface are EQU ra=1.5 m, rb=1.5 m, and rc=148 mm. The surface of the diffraction grating is coated with a material such as gold, platinum or aluminum oxide having a high reflectivity. Aluminum oxide is suitable in a range of wavelengths longer than 3 nm. The reflectivity of the diffraction grating changes with the incidence angle, and the diffraction efficiency also changes with it. To maximize the quantity of effective light, an incidence angle is selected at which the product of NA (numerical aperture) and the reflectivity is maximized. If the width of the diffraction grating is constant, NA on the plane containing the X-ray source, the center of the diffraction grating and the sample is proportional to cos .alpha.. .eta.c(%).multidot.cos .alpha. is thereby obtained as shown in FIG. 5. The incidence angle may be set to about 86.degree. to maximize the quality of light. More specifically, it is preferable to set the angle .alpha. of main incidence of X-rays upon the diffraction grating used as a spectroscopic condenser in the range of EQU 84.degree..ltoreq..alpha..ltoreq.88.degree.. If the angle .alpha. is smaller than the lower limit of this range, the reflectivity of the diffraction grating is reduced so that the loss of the quantity of light is excessively large. If the upper limit is exceeded, the aperture of the diffraction grating and the reflecting mirror is excessively large and restrictions on the optical disposition are increased. From the viewpoint of the above, a toroidal surface diffraction grating which enables suitable aberration correction by one-time reflection and, hence, efficient spectral diffraction and condensing is suitable as a condenser. It is also very effective with respect to cooling of the element, because it is possible to cool a diffraction grating from the reverse surface The provision of a multilayer film on a diffraction grating is also effective in improving the reflectivity of the diffraction grating. FIG. 14 schematically shows the construction of the second embodiment of the present invention in which a diffraction grating is used for the condenser system. In this embodiment, a spectroscopic condenser is constituted of a concave diffraction grating 9a and a toroidal surface reflecting mirror 9b in place of the toroidal surface diffraction grating 9 used alone as shown in FIG. 1. Light from a target 5 which is an X-ray source is reflected by the toroidal surface reflecting mirror 9b, and sagittal beam of it is condensed in the vicinity of an intermediate aperture 9c of a shielding plate and is then made incident upon the concave diffraction grating 9a. X-rays thereby separated are condensed on a sample 13 in sample container 12. On the other hand, a meridional light beam orthogonal to this light beam is reflected by the toroidal surface reflecting mirror 9b and travels via the concave diffraction grating 9a to be incident upon the sample 13. In other respects, the construction is the same as that of the first embodiment shown in FIG. 1. In FIG. 14, the components having the same functions as those of the arrangement shown in FIG. 1 are indicated by the same reference characters. By such a combination of a toroidal surface reflecting mirror and a concave diffraction grating, an image from the X-ray source can be formed on the sample surface. There are several possible variations of such a combination. However, one in which the intermediate aperture 9c and the sample are disposed so as to be located on the Rowland circle of the concave diffraction grating 9a is good in terms of aberration correction. Examples of the values of this system specified therefor are shown below. ______________________________________ Concave diffraction grating Slit distance d: 0.5 .mu.m Curvature radius R: 1.5 m Toroidal surface reflecting mirror Main curvature radius Rh: 2.5 m Sub curvature radius Rv: 18.0 mm The distance between the X-ray source and 174.4 mm the toroidal surface reflecting mirror: The angle of incidence of X-rays upon the 86.degree. toroidal surface reflecting mirror: The distance between the toroidal surface 174.4 mm reflecting mirror and the intermediate aperture: The distance r between the intermediate 104.6 mm aperture and the concave diffraction grating: The angle .alpha. of incidence upon the concave 86.degree. diffraction grating: The distance r' between the concave 209.9 mm diffraction grating and the sample: The diffraction angle .beta.: 81.96.degree. ______________________________________ The number of reflecting surfaces of this embodiment is greater than that in the first embodiment by one, and the quantity of effective light is therefore reduced to 1/2 to 1/3. However, it is possible to provide an X-ray source strong enough to cancel this reduction. Since the toroidal reflection mirror may effect condensing alone, the allowance of the contour accuracy is increased in comparison with the first embodiment and the facility with which the reflecting mirror is worked is thereby improved. Also, the diffraction grating which must be formed with high accuracy can be provided as a concave grating, it can therefore be easily worked. Ordinarily, the intermediate aperture corresponds to the incidence slits of the spectroscopic section and is necessary for determining the spectrum width. In this embodiment, however, the light beam size at the position of the intermediate aperture is determined by the size of the X-ray source. The intermediate aperture may therefore be removed but it is effective in previously adjusting the spectroscopic condenser or removing scattered light components. FIG. 15 schematically shows the third embodiment of the present invention in which a diffraction grating is used for the condenser system. In this embodiment, the spectroscopic condenser is constituted of a flat plane diffraction grating 9d and two out-of-axis paraboloidal mirrors 9e and 9f. A target 5 which is an X-ray source is placed at the focal point of the first paraboloidal mirror 9f. X-rays from the target 5 are reflected by the first paraboloidal mirror 9f and thereby travel as a parallel light beam to be incident upon the flat plane diffraction grating 9d. The diffracted light from the plane diffraction grating 9d is reflected by the second paraboloidal mirror 9e so that separated X-rays are condensed on a sample 13 placed at the focal point of this mirror. In other respects, the construction is the same as that of the first embodiment shown in FIG. 1. In FIG. 15, the components having the same functions as those of the arrangement shown in FIG. 1 are indicated by the same reference characters. In this embodiment, the magnification of the spectroscopic condenser is changed through a wide range by selecting the combination of the focal distances of the two paraboloidal mirrors 9e and 9f. Examples of the values of this system specified when the magnification is 1 are shown below. ______________________________________ Assuming that the paraboloid is expressed by y.sup.2 + z.sup.2 = 4 px, a general formula: p = 0.1827 The distance between the X-ray source and the 150.0 mm first paraboloidal mirror: The angle of incidence upon the first 86.0.degree. paraboloidal mirror: The distance between the first paraboloidal 200.0 mm mirror and the diffraction grating: The flat plane diffraction grating slit 0.5 .mu.m distance: The Angle of incidence upon the flat plane 86.0.degree. grating: The diffraction angle: 81.96.degree. The distance between the diffraction grating 200 mm and the second paraboloidal mirror: The angle of incidence upon the second 86.degree. paraboloidal mirror: The distance between the second paraboloidal 150 mm mirror and the sample: ______________________________________ With respect to these values, the diffraction grating has a wavelength dispersion in the plane of FIG. 5. That is, the slits of the diffraction grating are perpendicular to the plane of projection of FIG. 15. It is also possible to arrange the system by turning the disposition of the diffraction grating by 90.degree. so that the slits of the diffraction grating are in the plane of the diagram. In this case, the definition of the angles of incidence and diffraction at the flat plane grating is complicated. However, the diffraction grating can be slightly rotated on an axis contained in the plane of the diagram so that the main light beam travels as if it is reflected by normal reflection in the plane of the diagram by the diffraction grating. If the angle of this incidence in the plane of the diagram is 86.degree., a spectroscopic condenser can be made while other values are the same as those of the third embodiment listed above. In this spectroscopic condenser, the use of the diffraction grating is different from the use in the above embodiment. However, it is known that by this use the diffraction efficiency is made closer to the reflectivity. In this case, a diffraction efficiency higher than that in the first to third embodiments, which is 8%, can be realized and the reduction in the quantity of light due to the increase in the number of reflecting surfaces can be sufficiently compensated. (xv) Combination of X-ray Microscope and Optical Microscope If the X-ray microscope of the present invention is used in combination with an optical microscope, a narrow field of view of the X-ray microscope can easily be confirmed and optical adjustment of the X-ray microscope can be performed in a simple manner. In combining microscopes of these kinds, it is effective to use a co-axial construction of an illumination optical system and an objective optical system of an optical microscope for confirming the field of view for observation while constructing the overall system based on the image formation type X-ray microscope described first with reference to FIG. 1. That is, a reflection area for reflecting illumination light for the optical microscope is provided around a concave aspherical surface multilayer film mirror condenser provided as a light condensing means in the X-ray illumination system to enable the concave aspherical surface mirror for X-rays to be also used as a condenser for the optical microscope, and an imaging optical path is formed in a through aperture of an objective lens of the optical microscope on the optical axis thereof by a zone plate provided as an X-ray objective. In this arrangement, X-rays can be reflected and condensed by a central portion of the concave aspherical reflecting mirror in the illumination system while the illumination light for the optical microscope is reflected by the central portion and the peripheral portion; the concave aspherical reflecting mirror can be used for both the illumination systems for the X-ray microscope and the optical microscope. It is possible to efficiently supply illumination light for the optical microscope requiring a numerical aperture (NA) greater than the illumination numerical aperture of the X-ray microscope. Thus, the overall system has a simple construction and the optical axis adjustment of the X-ray microscope can be performed easily and precisely with the optical microscope. FIG. 16 is a diagram schematically showing the construction of an image formation type soft X-ray microscope designed in accordance with this conception. Laser light generated by a pulse laser 1 is condensed on a disk- or tape-like thin film target 5 through a vacuum maintaining window 4 by a condenser lens 3 so that X-rays having an intensity and a wavelength required are generated. The light emission of the pulse laser 1 is controlled by a pulse control unit 2 with a desired pulse separation (0.01 to several Hz). X-rays from the X-ray thin film target 5 are condensed on a test piece 13 in a test piece container 12 by a rotary elliptical multilayer film reflecting mirror 9. A test piece image is enlarged and formed on a two-dimensional X-ray imaging element 15 by using a zone plate (ZP) 14 for an imaging optical system. It is effective to use a solid-state image sensor, e.g., a back irradiation-type FT-CCD as the two-dimensional X-ray imaging element 15. As illustrated, in this arrangement, the X-ray thin film target 5, which is excited by the pulse laser, is disposed at the first focal point of the rotary elliptical multilayer film reflecting mirror 9, and the test piece 13 is disposed at the second focal point thereof. X-rays are monochromatized by the multilayer film reflecting mirror 9. The test piece is irradiated with one-pulse X-rays emitted by excitation by the pulse laser, and photon counting imaging is performed by the two-dimensional X-ray imaging element 15. To observe the test piece 13 while horizontally maintaining the same, the X-rays for irradiation and observation are made to travel in a vertical direction, and the laser beam for excitation of X-rays is made to travel in a horizontal direction. Specifically, the angle between the target 5 and the excitation laser beam is set to about 35.degree., and the angle of incidence of X-rays upon the rotary elliptical multilayer film reflecting mirror 9 is set to about 65.degree.. A means 6 for interchanging the X-ray thin film target 5 and removing scattered waste materials and the like and a diaphragm 7 are arranged to make X-rays travel in a predetermined direction. At the rear of the rotary elliptical multilayer film reflecting mirror 9 serving as a condenser, a water cooling unit 10 is provided to prevent temperature rising and deterioration of the mirror caused by absorption of X-rays. A field diaphragm 11 is provided in front of and close to the test piece container 12. It is changed to select an appropriate aperture size according to the observation magnification. However, it may be changed to select a different aperture as desired for the purpose of preventing flare or improving the contrast, while the same magnification is set. Image information output from the two-dimensional X-ray imaging element 15 is processed by an image processing unit 16, and is then output to an image output unit 17 such as a display, a printer or the like. These components, the components from the rotary elliptical multilayer film concave reflecting mirror 9 serving as a condenser to the light-receiving surface of the two-dimensional X-ray imaging element 15, are accommodated in a vacuum container 18 to be maintained under vacuum. The pressure in the vacuum container 18 is kept at about 10.sup.-2 Pa at which the X-ray absorption effect is negligible. Since the removing means 6 or the like must be provided around the X-ray thin film target 5 to cope with the generation of scattered substances therearound, it is necessary to isolate the X-ray source section by placing the same in another vacuum container 19. Visible light from an illumination light source 22 for the optical microscope is condensed on the X-ray target 5 through a condenser lens 21 and a dichroic mirror 20 provided as a light separating device. The point to which the light is thereby condensed coincides with the point to which laser light from the pulse laser 1 is condensed. Light reflected by the target 5 is condensed on the test piece 13 by the rotary elliptical multilayer film concave reflecting mirror 9, as in the case of X-rays from the target 5. As shown in FIG. 17 in section and in FIG. 18 in plan, a reflection area 9a used exclusively for visible light is formed on a peripheral portion of the rotary elliptical multilayer film concave reflecting mirror 9, while a multilayer film reflection area 9b used to reflect X-rays is formed on a central portion of this mirror. The visible light reflection area may be provided by aluminum deposition. The X-ray reflecting multilayer film is capable of reflecting visible light at a substantially high rate, the arrangement may alternatively be such that the whole surface of the reflecting mirror 9 is formed as an X-ray multilayer film mirror, a central portion thereof is used for an area for reflecting X-rays, and the whole surface is used as a visible light reflection area. In either case, X-rays are reflected by the central portion of the concave reflecting mirror 9 while visible light is reflected by the central and peripheral portions It is thereby possible to sufficiently provide illumination with a large NA (aperture value) necessary for an objective lens 23 along with illumination with a comparatively small NA of the zone plate 14 serving as an X-ray objective while maintaining the coaxial state of these optical elements. The zone plate 14 serving as an X-ray objective and the objective lens 23 of the optical microscope are arranged coaxially with each other, and the zone plate 14 optical path of the X-ray microscope is formed in an aperture formed through the objective 23 of the optical microscope along the optical axis thereof. Details of the zone plate 14 and the objective 23 of the optical microscope are illustrated in FIGS. 19 and 20. The body tube of the objective 23 of the optical microscope has a portion 104 projecting on the test piece side, and the zone plate 14 is supported in this portion by a crisscross support member 107 such as that shown in the plan view of FIG. 20. As shown in FIG. 19, the X-ray imaging optical path of the zone plate 14 is formed in the aperture formed through the optical microscope objective 23 along the optical axis thereof. Although the optical path of the optical microscope is slightly obstructed by the support member 107 as shown in FIG. 20, a sufficient quantity of observation light can be obtained through openings 108. Strictly, it is necessary to make a beam of light 102 from the zone plate 14 slightly convergent in order that the image of the test piece is enlarged and imaged on the two-dimensional X-ray imaging element 15 by the zone plate 14. The optical focal point of the zone plate 14 is slightly deviated to the emergent light side from the optical focal point of the objective 23. Therefore a beam of light 101 from the objective 23 is formed generally parallel, as shown in FIG. 19. Thus, the zone plate 14 and the objective 23 is integrally arranged so that the object point of the image formed by the zone plate 14 and the object point of the image formed by the objective 23 coincide with each other. The objective 23 may therefore be focused with respect to the test piece by being moved on the axis to automatically complete the focusing of the X-ray microscope with the zone plate 14. Generally, the numerical aperture of the zone plate 14 is very small, and it is therefore difficult to focus the X-ray microscope. However, this integral arrangement with the objective 23 of the optical microscope makes it easy to focus the X-ray microscope. As shown in FIG. 16, the beam of parallel light from the optical microscope objective 23 is reflected by a reflecting mirror 24 with an aperture obliquely disposed, and is led to out of the vacuum container 18 through an window glass 2 to form a spatial image 28 via an optical path flexing mirror 27. The spatial image 28 can be observed through an ocular lens 29 at a predetermined magnification. In this embodiment, the X-ray image and the visible light image are separated by the oblique reflecting mirror 24 having an aperture. However, if a detector having a characteristic such as to be sensitive to visible light well as to X-rays, e.g., a CCD, is used as the two-dimensional X-ray imaging element 15, it is possible to construct the observation systems of the X-ray microscope and the optical microscope completely coaxially. Preferably, for the rotary elliptical multilayer film concave reflecting mirror 9, a very small absorptive member, such as a very small piece of aluminum, is provided at a central position (on the optical axis) to prevent occurrence of noise in the imaging element 15 due to incidence of X-rays upon the optical axis of the zone plate 14. (xvi) OTHERS (1) Magnification Change In order to minimize damage to the sample, the X-rays must be applied only to a necessary region. In addition, in order to effectively utilize the X-ray source, it is necessary to combine a multilayer film mirror condenser, which is exclusively used for illumination within a necessary region and with a necessary numeral aperture for each resolving power and magnification of the objective optical system. When the magnification of the objective optical system is changed to each of the magnifications shown in the tables, it is necessary to change the rotary elliptical multilayer film mirror serving as a condenser to an appropriate one in correspondence with the objective optical system changed. As shown in FIG. 21, in order to the rotary elliptical mirror, a rotary elliptical multilayer film mirror 9a having magnification, which is different from that of the rotary elliptical multilayer film mirror 9, is disposed at a different position in the vertical direction, and the laser condensing lens 3 and the target 5 are integrally moved in the horizontal direction. At this time, as a matter of course, the target 5 and the sample 13 are positioned at the first focal point and the second focal point, respectively, of the rotary elliptical multilayer film mirror. (2) Window Material of Sample Container It is necessary to use as the window material of the container for receiving the sample a material such as Si.sub.3 N.sub.4 film or the like, which has high X-ray transmittance. In this case, it is preferable that the thickness of the sample container is 50 nm, and the window diameter is about 100 .mu.m in view of the relation to the effective visual field. A container having the structure disclosed in Japanese Patent Appln. Laid-Open No. 63-298200 by the same applicant as this application may be used. (3) Two-dimensional X-ray Imaging Element In the two-dimensional imaging element, which is capable of counting photons, it is preferable that the wavelength region with detection sensitivity is 2.3 to 4.4 nm, the quantum efficiency and aperture efficiency are substantially 100%, the pixel dimension is 10 mm, the pixel number in accordance with the TSC system is 700.times.525=367500, the diagonal length is 9 mm and the image number per second is 30. The system for such an element is, for example, back irradiation FT-CCD or the like. (4) Monitor Observation by Optical Microscope Since constant observation of a sample by using an X-ray microscope accelerates damage to the sample, as shown in FIG. 22, it is effective that the sample is usually observed by an optical microscope 20, and, if required, the sample is transferred onto an X-ray microscope so that the biological cells only in a necessary portion is observed by the X-ray microscope. In this case, the system of the optical microscope is preferably designed so that it is possible to perform microscopic methods for a phase difference, differential interference, polarization, fluorescent light, a dark field and the like. In the above-described image-forming soft X-ray microscope in accordance with the present invention, X-rays are monochromatized by the multilayer film mirror, the X-rays of one pulse emitted from the pulse X-ray source are condensed on the sample by the elliptical multilayer film reflecting mirror serving as a condenser, an enlarged image is formed by the phase zone plate serving as an objective optical system, and photon counting imaging is photoelectrically made by the two-dimensional X-ray imaging element. It is therefore possible to achieve high efficiency and high resolving power. This permits the minimization of damage to a living sample and dynamical observation of the living sample with high resolving power of about 20 nm, without breaking it. In addition, although a sample is observed by a general electron microscope in a state where the sample is fixed after it has been instantaneously frozen, sliced and stained. However, the present invention permits dynamical observation of living cells, without fixing and breaking the cells. Although dynamical observation can be made by an optical microscope without fixing and breaking a sample, the resolving power is 200 nm at most. However, the X-ray microscope of the present invention permits observation with extremely high resolving power which reaches 20 nm. |
039873064 | claims | 1. The method of treating a material by irradiation with U.V. radiation comprising the steps of providing a discharge chamber, at least a portion which is transparent to U.V. radiation, and a pair of spaced apart electrodes therein, supplying water to the region between the electrodes, applying a high voltage between the electrodes whereupon the water undergoes electrical breakdown and an electrical discharge occurs between the electrodes, the electrodes and the supply of water being so arranged that at least a part of the electrical discharge path is submerged in the water to produce thereby a hot submerged plasma of ionized water which is unable to expand freely and which emits U.V. radiation, and disposing the material to be treated in a treatment region adjacent said radiation-transparent portion where it is irradiated by the U.V. radiation transmitted from the discharge chamber. 2. The method of claim 1, including the steps of transporting the material continuously through the treatment region, and repetitively applying the high voltage to effect electrical discharge at a predetermined rate to obtain a required treatment of the material. 3. The method of claim 2 wherein the material to be treated is a liquid, and wherein the transporting step comprises feeding the liquid through a treatment chamber which constitutes the treatment region. 4. The method of claim 2 wherein the electrodes are mounted in and insulated from the discharge chamber and wherein the high voltage is repetitively applied between the electrodes, and inluding the step of continuously sensing the electrical characteristics of the electrical discharges and continuously and automatically adjusting the spacing of the electrodes in response to the sensed characteristics so as to maintain predetermined electrical characteristics during the repetitive production of U.V. radiation. 5. The method of claim 2 wherein one of the electrodes is constituted by the discharge chamber, the other electrode is immersed in water within the discharge chamber, the water is arranged to have a circulatory motion within the discharge chamber, and the high voltage is repetitively applied between the electrodes, and including the step of continuously sensing the electrical characteristics of the electrical discharges and continuously and automatically controlling the amount of this circulatory motion whereby the spacing of said other electrode and the overlying water surface is adjusted in response to the sensed characteristics so as to maintain predetermined electrical characteristics during the repetitive production of U.V. radiation. 6. An apparatus for treating material by irradiation with U.V. radiation comprising a pair of spaced apart electrodes, a discharge chamber containing the region between the electrodes, at least a portion of the chamber being transparent to U.V. radiation, means arranged to supply water to the region between the electrodes such that in use upon application of a high voltage between the electrodes the water will undergo electrical breakdown and the resulting electrical discharge between the electrodes has at least a part of its path submerged in the water whereby the hot submerged plasma of ionized water thus produced is unable to expand freely and emits U.V radiation, means arranged to apply between the electrodes a high voltage of such magnitude as to cause electrical breakdown of the water and thereby producing U.V. radiation from said hot submerged plasma, a treatment region adjacent the U.V. transparent portion of the chamber, means arranged to transport the material to be treated continuously through the treatment region, and wherein the voltage applying means is arranged to effect continously electrical discharges at a predetermined rate to obtain thereby a required treatment of the material. 7. The apparatus of claim 6 wherein the treatment region is in the form of a chamber having inlet and outlet passages. 8. The apparatus of claim 7 wherein the treatment chamber is encircled by the discharge chamber and is formed by a hollow tube of U.V. transparent material which constitutes the radiation-transparent portion of the discharge chamber. 9. The apparatus of claim 8 wherein the discharge chamber has an elliptical cross-sectional configuration, the treatment chamber is arranged at one focus, and the electrodes are arranged such that the discharge region is at the other focus, and wherein the discharge chamber, other than the radiation-transparent portion thereof, has an internal surface arranged to be reflective for at least some of the generated U.V. radiation. 10. The apparatus of claim 8 wherein the discharge chamber has a cylindrical configuration, the treatment chamber is constituted by a plurality of the hollow tubes arranged symmetrically around the axis of the discharge chamber, and wherein the discharge chamber, other than the radiation-transparent portion thereof, has an internal surface arranged to be reflective for at least some of the generated U.V. radiation. 11. The apparatus of claim 8 wherein the treatment chamber encircles the discharge chamber and is in the form of an annular space between inner and outer sleeves, the inner sleeve being formed of U.V. transparent material and constituting the radiation-transparent portion of the discharge chamber. 12. The apparatus of claim 8 wherein the U.V. transparent portion of the discharge chamber is in the form of a plate, and wherein the treatment chamber is arranged such as to provide thin film flow across the outer surface of the plate of liquid-form material to be treated when said material passes through the treatment chamber. 13. The apparatus of claim 12 and including another plate spaced from the transparent portion and shaped such as to provide the treatment chamber in the form of a thin disc-like space, wherein the inlet and outlet passages are arranged such that the thin film flow is substantially radial within the treatment chamber. 14. The apparatus of claim 13 wherein the inlet passage passes through said other plate and is aligned with the axis of the treatment chamber, and the outlet passage communicates with the peripheral regions of the treatment chamber via an annular collecting chamber therearound. 15. An apparatus for treating material by irradiation with U.V. radiation comprising a discharge chamber, two electrodes mounted in and insulated from the discharge chamber such as to form a discharge region between the electrodes, at least one of the electrodes being mounted for adjustment of the electrode spacing, means arranged to apply repetitively a high voltage between the electrodes, means arranged to supply water to the discharge region such that upon application of the high voltage the water undergoes an electrical breakdown and the electrical discharge between the electrodes is along a submerged path in the water whereby the hot plasma of ionized water produced by the discharge is not able to expand freely and emits U.V. radiation means arranged to continuously sense the electrical characteristics of the electrical discharges and to continuously and automatically adjust the spacing of the electrodes in response to the sensed characteristics so as to maintain predetermined electrical characteristics during the repetitive production of U.V. radiation, at least a portion of the discharge chamber being transparent to permit the U.V. radiation to be transmitted from the discharge chamber, a treatment region adjacent the radiation-transparent portion of the discharge chamber, and means arranged to transport material to be treated through the treatment region. 16. An apparatus for treating material by irradiation with U.V. radiation comprising a discharge chamber which constitutes an electrode, another electrode disposed in the bottom of the discharge, means arranged to supply water into the discharge chamber such as to cover said another electrode, means arranged to apply repetitively a high voltage between the discharge chamber and said another electrode, whereby the water repetitively undergoes electrical breakdown, the electrical discharge path being from said another electrode through the water to the surface and thence along the surface to the discharge chamber and whereby the hot submerged plasma of ionized water produced by the discharge is not able to expand freely and emits U.V. radiation, means arranged to sense continuously the electrical characteristics of the electrical discharges and to continuously and automatically control the water supplying means such as to alter the height of the water surface above said other electrode in response to the sensed characteristics so as to maintain predetermined electrical characteristics during the repetitive production of U.V. radiation, at least a portion of the discharge chamber being transparent to permit the U.V. radiation to be transmitted from the discharge chamber, a treatment region adjacent the radiation-transparent portion of the discharge chamber, and means arranged to transport material to be treated through the treatment region. 17. The apparatus of claim 16 wherein the water supplying means comprise liquid inlet and outlet passages in the discharge chamber arranged such as to effect a circulatory motion in the water in the discharge chamber, and wherein the sensing and controlling means is arranged to control the rate at which the water supplying means supplies the water, thereby to control the amount of said circulatory motion and thus the height of the water above said other electrode. 18. The apparatus of claim 17 wherein said other electrode is in the form of a rod positioned such that in use it is under the region where the surface of the circulating water has maximum depression. 19. The apparatus of claim 15 wherein said transparent portion is formed of quartz doped with a selected metallic ion such as to obtain selective absorption of an unwanted portion of the generated radiation. 20. The apparatus of claim 19 wherein the metallic ion is tungsten. 21. The apparatus of claim 16 wherein said transparent portion is formed of quartz doped with a selected metallic ion such as to obtain selective absorption of an unwanted portion of the generated radiation. 22. The apparatus of claim 21 wherein the metallic ion is tungsten. |
047770082 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to an isolating device for a water chamber in a steam generator for a nuclear reactor. 2. Description of the Prior Art In a pressurized water type nuclear reactor, as shown in FIG. 8, a nuclear reactor vessel (01), a water chamber (02a) in a steam generator (02) and a coolant pump (03) are mutually connected with one another by means of coolant pipes (04) to form a coolant circulating loop. And in an atomic power plant, for the purpose of inspection for confirming safety of the component instruments and the systems as well as replacement of the fuels, operation of the nuclear reactor is interrupted nearly once a year. Upon the inspection and fuel replacement, tasks relating to the nuclear reactor such as opening of a nuclear reactor vessel, extraction of fuels, inspection of a reactor core, charging of fuels, reassembly of a nuclear reactor, etc. are carried out in shielding water after the nuclear reactor vessel (01), a nuclear reactor cavity (05) and the like have been filled with shielding water. Then, inspection of thin tubes in the steam generator to be effected at that time is necessitated to be carried out under the condition that the water has been extracted, but the tasks relating to the nuclear reactor and the inspection of the thin tubes in the steam generator cannot be performed simultaneously, because the shielding water is filled during the work relating to the nuclear reactor. Accordingly, the time required for the inspection would become long, and this would limit improvement of the rate of operation of the atomic power plant. The present invention has been worked out in view of the above-mentioned background in the art. SUMMARY OF THE INVENTION It is therefore one object of the present invention to provide a novel isolating device for a water chamber in a steam generator for a nuclear reactor, which can reliably isolate the water chamber in the steam generator from nuclear reactor coolant pipes, a nuclear reactor vessel, a nuclear reactor cavity and the like and thereby makes it possible to carry out inspection of thin tubes in the steam generator simultaneously with the performance of other tasks relating to the nuclear reactor. According to one feature of the present invention, there is provided an isolating device for a water chamber in a steam generator for a nuclear reactor which steam generator includes, at its bottom, a water chamber communicating with a nuclear reactor vessel via coolant pipings and having a nozzle section formed between the water chamber and the coolant piping, which isolating device comprises a support arm having its outer end mounted to a manhole section of the water chamber and its inner end butted against an inner surface of the nozzle section, a plug pivotably attached to a tip end portion of the support arm and positioned within the nozzle section, and an angle adjusting screw interposed between the plug and the support arm, the plug being constructed in a foldable manner. The above-mentioned and other features and objects of the present invention will become more apparent by reference to the following description of preferred embodiments of the invention taken in conjunction with the accompanying drawings. |
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053295714 | description | In FIGS. 1-3, 1 designates a fuel channel of substantially square cross section. The fuel channel surrounds with no mentionable play an upper, square portion of a bottom part 2 with a circular, downwardly facing inlet opening 3 for cooling water and moderator water. The bottom part 2 supports, in addition to the fuel channel 1, a supporting plate 4. At its lowest part the fuel channel 1 has a relatively thick wall portion which is fixed to the bottom part 2 and the supporting plate 4 by means of a plurality of horizontal bolts, indicated by means of dash-dotted lines 5. The fuel channel 1 is divided, by means of a hollow support member 7 of cruciform cross section, into four vertical tubular parts 6, hereinafter referred to as casings, with at least substantially square cross section. The support member 7 is welded to the four walls 1a, 1b, 1c and 1d of the fuel channel 1 and has four hollow wings 8. The cruciform space formed from the support member is designated 32 and is connected at its lowest part to an inlet tube 9 for moderator water. Each casing 6 comprises a bundle 25 containing twenty-five fuel rods 10. The rods are arranged in a symmetrical lattice in five rows each containing five rods. Each rod is included in two rows perpendicular to each other. Each bundle is arranged with a bottom-tie plate 11, a top-tie plate 12 and a plurality of spacers 13. A fuel rod bundle 25 with a bottom-tie plate 11, a top-tie plate 12, a spacer 13 and a casing 6 forms a unit which in this application is referred to as fuel assembly and is designated 40a, whereas the device illustrated in FIGS. 1-3 comprising four such fuel assemblies is referred to as a composed fuel assembly and is designated 40. In the composed fuel assembly 40 each fuel assembly 40a is arranged with interspaces 8 with respect to adjacent fuel assemblies 40a in that the cruciform space 32 extends through the composed fuel assembly in a vertical direction. The four bottom-tie plates 11 are supported in the composed fuel assembly by the supporting plate 4 and are each partially inserted into a respective square hole 14 in this plate. In each fuel assembly at least one of the fuel rods is designed with relatively long, threaded end plugs 33 and 34 of solid cladding material, the lower end plug 33 being passed through the bottom-tie plate 11 and provided with a nut 15, the upper end plug 34 being passed through the top-tie plate 12 and provided with a nut 16. In the embodiment shown, the centre rod 26 is designed in this way. This rod also serves as a spacer holder rod. An upper end portion of the fuel channel 1 surrounds a cruciform lifting plate 17 with four horizontal arms 18, 19, 20 and 21, which extend from a common central portion. At the outer end each arm has an arrow-head-like portion 22 which, in respective corners of the fuel channel 1, makes contact with the inner wall surface of the fuel channel 1. A lifting handle 23 is fixed to the arms 20 and 21. The lifting plate 17 and the handle 23 together form a lifting member of steel cast in one piece. The lifting plate 17 is fixed to the support member 7 by inserting each of four vertical bars 28 into a respective wing 8 of the support member 7 and welding them thereto, At the top each bar 28 has a vertical, bolt-like portion 29 which is passed with a play through a corresponding hole in the mid-portion of the lifting plate 17 and provided with a nut 30. As will be clear from the figures, the fuel channel 1 is provided with indentations 31, intermittently arranged in the longitudinal direction, to which the support member 7 is welded. FIG. 4 shows a small part of a reactor core. The section comprises nine whole composed fuel assemblies 40 of the kind illustrated in FIGS. 1-3. Of the fuel assemblies only one is shown in detail, the other ones only as empty squares. The spaces between the fuel rods 10 within each fuel assembly 40a is traversed by water as is the cruciform space 32 in the composed fuel assembly 40. The interspaces in the form of gaps 37a and 37b between the fuel assemblies 40 are also traversed by water. Those gaps 37a into which control rods 38 can be inserted are wider than those gaps 37b into which no control rods can be inserted. The control rods 38 have blades 38a, 38b, 38c and 38d which form a right-angled cross. The fuel assembly 40a schematically shown in FIG. 5 has, in the exemplified case, six spacers 13a-f. Each fuel rod 10, of which only one is shown in the figure, contains between the two end plugs 33 and 34 (FIG. 1) a large number of pellets, stacked one above the other, of uranium dioxide enriched with U 235. At the top there is a plenum 10a filled with helium. This plenum without nuclear fuel material is thus not included in the active length of the fuel rod. On two sides the fuel assembly is provided with outlet holes 41 for releasing water. The outlet holes are located at a level above the centre of the active length of the fuel rods and below the two uppermost spacers 13a and 13b. FIG. 5 does not show the front side of the casing 6 but only the positions of the outlet holes thereon. According to the embodiment of the present invention illustrated in FIG. 6, four fuel assemblies 40a of the kind shown in FIG. 5 are arranged to form a composed fuel assembly 40 of the kind described with reference to FIGS. 1-3. The outlet holes 41 for water for release of water from the four fuel assemblies 40a to the space 32 in the composed fuel assembly 40 are arranged on those sides of the casing 6 facing the space 32. Outlet holes could also be arranged in those sides of the casing 6 facing the water gaps 37a and 37b (FIG. 4). The water gaps 37a and 37b shown in FIG. 4 are in FIG. 6, as well as in FIGS. 7 and 8, marked by dashed lines. In the exemplified case, 40% of the total number of fuel assemblies in the core are arranged with outlet holes for water of the kind described. The fuel assembly 42 illustrated in FIG. 7 corresponds to the composed fuel assembly 40 shown in FIGS. 3 and 6 and each one of the composed fuel assemblies 40 in the reactor core according to FIG. 4 can be replaced by such a fuel assembly 42. The fuel assembly 42 comprises 8.times.8 fuel rods and has no cruciform internal space traversed by water. According to the present invention, outlet holes 43 for water are arranged on all side of the casing 6 which are facing the water gaps 37a and 37b. As in the case illustrated in FIGS. 5 and 6, the outlet holes are arranged at a level above the centre of the active length of the fuel rods and below the two uppermost spacers. The fuel assembly 44 illustrated in FIG. 8, like the fuel assembly illustrated in FIG. 7, corresponds to the composed fuel assembly 40 shown in FIGS. 3 and 6 and can replace each one of the composed fuel assemblies 40 in the reactor core according to FIG. 4. The fuel assembly 44 is provided with an internally arranged vertical channel 45, through which water is conducted in a vertical direction from below and upwards through the assembly. The wall of the channel is designated 46. The fuel assembly 44 comprises 9.times.9 fuel rods, with 3.times.3 removed for the arrangement of the channel 45 that is 72 fuel rods in all. According to the present invention, outlet holes 47 for water are arranged in the walls of the vertical channel for releasing water from the space between the fuel rods to the channel. The outlet holes 47 are located at the same level as the outlet holes 41 and 43 in the examples illustrated in FIGS. 6 and 7. Outlet holes could also be arranged in those sides of the casing facing the water gaps 37a and 37b. In the cases exemplified in FIGS. 5, 6, 7 and 8, the total area of the outlet holes in each one of the fuel assemblies constitutes 30% of the flow area in the respective fuel assembly (40a, 42, 44). |
061887412 | abstract | An integral stub tube which simplifies the reactor pressure vessel fabrication process, provides a structural transition between the penetration and head, and facilitates future remote inspection of the attachment weld is described. In one embodiment, a stub tube is machined into the bottom head dome. Specifically, a penetration is formed in the bottom head dome by a bore having a stub tube portion. The stub tube portion has a cylindrical shape and a length of the stub tube portion is selected to provide a transition between a penetration housing and an adjacent portion of bottom head dome. The penetration housing extends through the penetration, in the dome. A weld attaches the stub tube portion to the penetration housing. |
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claims | 1. A storage product for storing radioactive isotopes, the storage product including:(a) at least one radioactive isotope; (b) a metal alloy including a chemically active metal fraction in an amount effective for chemically reducing organic feed materials, tungsten in an amount effective for producing a close-packed crystalline structure throughout the metal alloy upon solidification of the metal alloy from a molten state, and for each type of expected radioactive emission associated with a radioactive isotope in the storage product, at least one corresponding radiation absorbing metal, each corresponding radiation absorbing metal being capable of absorbing the respective type of expected radioactive emission; and (c) wherein the metal alloy is solidified from the molten state to form a storage ingot with each radioactive isotope, the tungsten, the chemically active metal fraction, and each radiation absorbing metal being substantially evenly distributed within the ingot. 2. The storage product of claim 1 including a radiation absorbing encapsulant encapsulating the ingot, the encapsulant material including at least one radiation absorbing material for each type of expected radioactive emission within the ingot and further including tungsten in an amount effective for producing a close-packed crystalline structure in the encapsulant. 3. The storage product of claim 1 including no less than approximately one atom of tungsten for every twenty-seven atoms of other elements in the storage product. 4. The storage product of claim 1 wherein each chemically active metal in the chemically active metal fraction is selected from the group consisting of magnesium, aluminum, lithium, zinc, calcium, and copper. 5. The storage product of claim 1 wherein the cumulative total amount of die chemically active metal fraction is no less than approximately forty percent by weight of the total metal alloy. 6. A storage product for storing a fast neutron emitting isotope, the storage product including:(a) a metal alloy including a chemically active metal fraction in an amount effective for chemically reducing organic feed materials, the metal alloy also encompassing a quantity of a fast neutron emitting isotope; (b) a transmutation target fraction forming part of the metal alloy, the transmutation target fraction made up of a transmutation target material for absorbing fast neutrons emitted by the fast neutron emitting isotope; (c) a transmutation emission absorbing fraction forming part of the metal alloy, the transmutation emission absorbing fraction made up of a transmutation emission absorbing material for absorbing emissions resulting from the absorption of a respective fast neutron by the transmutation target material; and (d) tungsten in an amount effective for producing a close-packed crystalline structure upon solidification of the metal alloy; and (e) wherein the metal alloy is solidified from a molten state to form a storage ingot with the fast neutron emitting isotope, the chemically active metal fraction, the transmutation target fraction, and the transmutation emission absorbing fraction being substantially evenly distributed within the ingot. 7. The storage product of claim 6 including no less than approximately one atom of tungsten for every 27 atoms of other elements in the storage product. 8. A storage product for storing radioactive isotopes including:(a) at least one radioactive isotope tungsten in an amount effective for producing a close-packed crystalline structure throughout the storage product upon solidification of the storage product from a molten state, and for each type of expected radioactive emission in the storage product, no less than approximately seven hundred and twenty-seven (727) atoms of a corresponding radiation absorbing metal, each corresponding radiation absorbing metal capable of absorbing the respective type of expected radioactive emission in the storage product; (b) a chemically active metal fraction in an amount effective for chemically reducing organic feed material; and (c) wherein the storage product is solidified from the molten slate to form a storage ingot with each radioactive isotope, the tungsten, the chemically active metal fraction, and each radiation absorbing metal being substantially evenly distributed within the ingot. 9. The storage product of claim 8 including no less than approximately one atom of tungsten for every 27 atoms of other elements in the storage product. |
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047175281 | claims | 1. A system for controlling a nuclear reactor comprising a nuclear core comprising a plurality of fissile fuel elements having openings therebetween, a plurality of control rods containing a neutron absorbing material arranged throughout said core and fitting within said openings between said fuel elements, said control rods being movably insertable and withdrawable from said openings by drive means comprising linear motion apparatus having lift, movable gripper and stationary gripper solenoid coils, said control rods being arranged in groups of a predetermined number of control rods which predetermined number move together, means for directing the movement of said groups of control rods to control the core radial power distribution and the core axial power distribution and means for sharing the power circuitry among said control rod groups, said power circuitry being connected to the drive mechanism of each group of control rods, said means for sharing the power circuitry comprising a plurality of storage enclosures, with one enclosure containing the power circuitry for the lift and movable gripper coils for all of said groups of control rods in combination with separate enclosures which contain the power circuitry for the stationary gripper coils for each respective group of control rods. 2. The control system of claim 1, wherein the power circuitry for the lift and movable gripper coils shared by all the groups of control rods is distributed to each of said other enclosures by means of a bus bar arrangement. 3. The control system of claim 1, wherein said means for directing the movement of said control rod groups comprises a control rod movement strategy processor comprising means for receiving signals for reactivity changes from a reactor power control system and converting said signals into motion commands for said rod groups by selecting from the groups of rods which will satisfy the reactivity changes that combination of groups of rods which minimizes power distribution peaking factors. 4. The control system of claim 3, wherein said means for directing the movement of said control rod groups further comprises a partial trip processor comprising means for receiving signals for immediate power reductions from said reactor power control system and converting said signals into commands to drop selected rod groups into said core by deenergizing the drive mechanisms of said control rod groups, said partial trip processor and said rod motion strategy processor being functionally connected by means for exchanging and acting upon each other's receiving and command signals. 5. The control system of claim 4, wherein said means for directing the movement of said control rod groups further comprises a power distribution calculator comprising means for determining local power density at a number of local regions throughout the core by considering at each region the effective reactivity the neutron leakage to and from adjacent regions and the amount of absorber material in each region, said power distribution calculator being functionally connected to said partial trip processor and said rod motion strategy processor. 6. The control system of claim 5, wherein said means for directing the movement of said control rod groups further comprises a rod group worth calculator comprising means to determine the instantaneous differential and integral reactivity worth of each control rod group by using the calculated power distribution and the then current control rod positions, said rod group worth calculator being functionally connected to said partial trip processor, said rod motion strategy processor and said power distribution calculator. 7. The control system of claim 6, wherein said means for directing the movement of said control rod groups further comprises a fuel burnup distribution control processor comprising means for maintaining uniform core burnup by maintaining a history of the power distribution of fuel elements with control rods relative to the neighboring elements and generates commands to remove control rods from elements which have been at a lower power level for a relatively extended period of time, said fuel burnup distribution control processor being functionally connected to said partial trip processor, said rod motion strategy processor, said power distribution calculator and said rod group worth calculator. 8. The control system of claim 7, wherein said means for directing the movement of said control rod groups further comprises a communications processor comprising means for functionally connecting said partial trip processor, said rod/motion strategy processor, said burnup distribution control processor, said power distribution calculator, and said rod group worth calculator by a shared memory bus arrangement moving data from one module to another. 9. A system for controlling a nuclear reactor comprising a nuclear core comprising a plurality of fissile fuel elements having openings therebetween, a plurality of control rods containing a neutron absorbing material arranged throughout said core and fitting within said openings between said fuel elements, said control rods being movably insertable and withdrawable from said openings, said control rods being arranged in groups of a predetermined number of control rods which predetermined number move together, means for directing the movement of said groups of control rods to control the core radial power distribution and the core axial power distribution, and means for sharing the power circuitry among said control rod groups comprising a plurality of storage enclosures with one enclosure containing the power circuitry for the lift and movable gripper coils for all of said groups of control rods in combination with separate enclosures which contain the power circuitry for the stationary gripper coils for each respective group of control rods, said power circuitry being connected to the drive mechanism of each group of control rods, said means for directing the movement of said control rod groups comprising: a control rod movement strategy processor comprising means for receiving signals for reactivity changes from a reactor power control system and converting said signals into motion commands for said rod groups by selecting from the groups of rods which will satisfy the reactivity changes that combination of groups of rods which minimizes power distribution peaking factors, a partial trip processor comprising means for receiving signals for immediate power reductions from said reactor power control system and converting said signals into commands to drop selected rod groups into said core by de-energizing the drive mechanisms of said control rod groups, a power distribution calculator comprising means for determining local power density at a number of local regions throughout the core by considering at each region: the effective reactivity, the neutron leakage to and from adjacent regions, and the amount of absorber material in each region, a rod group worth calculator comprising means to determine the instantaneous differential and integral reactivity worth of each control rod group by using the calculated power distribution and the then current control rod positions, a fuel burnup distribution control processor comprising means for maintaining uniform core burnup by maintaining a history of the power distribution of fuel elements with control rods relative to the neighboring elements and generates commands to remove control rods from elements which have been at a lower power level for a relatively extended period of time, and a communications procesor comprising means for functionally connecting said partial trip processor, said rod motion strategy processor, said burnup distribution control processor, said power distribution calculator, and said rod group worth calculator by a shared memory bus arrangement moving data from one module to another. |
claims | 1. A method for high spatial resolution imaging of one or more sources of x-ray and gamma-ray radiation comprising:a) locating the sources of radiation proximally to a point;b) supplying a plurality of elements, each element comprisinga first collimator, defined by a pair of first plates, wherein the first collimator is adapted to direct radiation emanating from locations proximal to the point;a diffracting crystal adapted to receive and diffract the radiation directed by the first collimator;a second collimator defined by a pair of second plates, wherein the second collimator is adapted to direct radiation diffracted by the diffracting crystal; anda detector adapted to detect the radiation directed by the second collimator;c) supplying a means for analyzing said detected radiation to collect data as to the type and location of the source of the radiation; andd) supplying a means for converting the data to an image. 2. The method as recited in claim 1 wherein said image is produced by an array of elements such that the first plates are parallel to a first line that joins one said crystal to said point. 3. The method as recited in claim 1 further comprising choosing said diffracting crystals from ‘high grade’ commercially available diffracting crystals and mechanically bending said crystals in an apparatus. 4. The method as recited in claim 1 wherein said method further comprises providing means to impart linear and rotational motion to the sources. 5. The method as recited in claim 1 further comprising positioning said diffracting crystals so as to employ Laue diffraction in diffracting radiation from said sources. 6. The method as recited in claim 1 further comprising positioning said diffracting crystals so as to employ Bragg diffraction in diffracting radiation from said sources. 7. The method as recited in claim 6 further comprising positioning said diffracting crystals so that said crystals act as focusing crystals and said elements have a spatial resolution of 20 arcsec FWHM. 8. The method as recited in claim 1 wherein said image is produced by a plurality of arrays each array comprising one or more of said elements such that the first plates in each array are parallel to a first line first line that joins one said crystal to said point, with said first line being different for each array and with no two said first lines parallel to each other. 9. The method as recited in claim 8 wherein said first plates in each array are perpendicular to a second line with said second line being different for each array, and with no two said second lines parallel to each other. 10. The method as recited in claim 1 wherein said method further comprises providing means to impart linear and rotational motion to the elements. 11. A device for high spatial resolution imaging of one or more sources of x-ray and gamma-ray radiation comprising:a) a means for locating the sources of radiation proximally to a point;b) a plurality of elements, each element comprisinga first collimator, defined by a pair of first plates, wherein the first collimator is adapted to direct radiation emanating from locations proximal to the point;a diffracting crystal adapted to receive and diffract the radiation directed by the first collimator;a second collimator defined by a pair of second plates, wherein the second collimator is adapted to direct radiation diffracted by the diffracting crystal; anda detector adapted to detect the radiation directed by the second collimator;c) means for analyzing said detected radiation to collect data as to the type and location of the source of the radiation; andd) a means for converting the data to an image. 12. The device as recited in claim 11 further comprising means to impart linear and rotational motion to the sources. 13. The device as recited in claim 11 further comprising means to impart linear and rotational motion to the elements. 14. The device as recited in claim 11 wherein the elements form a plurality of arrays. 15. The device as recited in claim 11 wherein said image is produced by a plurality of arrays, each array comprising one or more of said elements such that the first plates in each array are parallel to a first line first line that joins one said crystal to said point, with said first line being different for each array and with no two said first lines parallel to each other. 16. The device as recited in claim 15 wherein said first plates in each array are parallel to a second line that is perpendicular to the array's first line, with no two said second lines parallel to each other. 17. The device as recited in claim 16 wherein no more than two of said first lines of said arrays are co-planar. 18. The device as recited in claim 11 wherein said diffracting crystals are so positioned as to employ Laue diffraction in diffracting radiation from said sources. 19. The device as recited in claim 11 wherein said diffracting crystals are so positioned as to employ Bragg diffraction in diffracting radiation from said sources. 20. The device as recited in claim 11 wherein said crystals act as focusing crystals with said elements having a spatial resolution 20 arcsec FWHM. |
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abstract | A plant with piping mounted on branch pipe, wherein said piping which introduces gas has a nozzle portion as a joint portion to a vessel and a branch portion connected with a branch pipe; and wherein an enlarged passage portion is formed at least at one of said branch portion and said nozzle portion, and a passage sectional area of said enlarged passage portion is larger than that of said piping other than said enlarged passage portion. Since the flow velocity of the gas flowing inside slows down at the enlarged passage portion, the occurrence of acoustic resonance at the branch portion or the nozzle portion can be suppressed. Accordingly, the fluctuation pressure of the gas flowing in the piping can be reduced even more. |
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claims | 1. A detector apparatus for electromagnetic radiation originating from a source, the detector apparatus comprising:a housing;at least one multilayer disposed within the housing, the at least one multilayer defining a leading edge and a trailing edge, wherein the at least one multilayer is adapted to interact with a plurality of high-energy photons within the X-ray portion of the electromagnetic spectrum, impingent from the leading edge, to permit passage of photons of at least one selected from a plurality of different energies;a first securement bracket disposed within the housing, wherein the at least one multilayer is pivotally fixed to the first securement bracket at a position adjacent to the leading edge;a second securement bracket disposed within the housing, wherein the at least one multilayer is adjustably supported by the second securement bracket at a position adjacent to the trailing edge;at least one detector disposed adjacent to the trailing edge of the at least one multilayer, wherein the detector is adapted to detect the impingent high-energy photons after interaction with the at least one multilayer; andat least one adjustment mechanism operatively connected to the second securement bracket, wherein the at least one adjustment mechanism adjusts the position of the second securement bracket to alter an angular position of the at least one multilayer wherein the at least one multilayer is oriented along a radial line extending outwardly from the source. 2. The detector apparatus of claim 1, wherein the at least one multilayer comprises a plurality of multilayers. 3. The detector apparatus of claim 1, wherein the at least one multilayer comprises a linearly gradient surface. 4. The detector apparatus of claim 1, wherein the at least one multilayer comprises a radially gradient surface. 5. The detector apparatus of claim 4 wherein the at least one multilayer has a body defining a surface with a grading ratio, wherein the grading ratio follows the equation:d/d0=R/R0wherein d is a d-spacing of the multilayer at a distance R from a center point, andwherein d0 is the d-spacing at R0, which is a minimum value for the at least one multilayer. 6. The detector apparatus of claim 1, further comprising:a collimating portion defined within the housing adjacent to the at least one multilayer to collimate the plurality of impingent high-energy photons after interaction with the at least one multilayer; anda plurality of collimating plates disposed in the collimating portion to facilitate collimation of the plurality of impingent high-energy photons. 7. The detector apparatus of claim 1, wherein the at least one adjustment mechanism comprises a motor operative on the second securement bracket. 8. The detector apparatus of claim 1, further comprising:a processor connected to the at least one detector to receive signals generated by the detector and generate a control signal to operate the at least one adjustment mechanism. 9. The detector apparatus of claim 8, wherein the processor is adapted to generate a spectrum associated with the plurality of high-energy photons. 10. The detector apparatus of claim 1, further comprising:a screen disposed adjacent to the leading edge of the at least one multilayer, wherein the screen is adapted to adjust the vertical acceptance of the impingent high-energy photons. 11. The detector apparatus of claim 10, further comprising:a second adjustment mechanism operatively connected to the screen, wherein the second adjustment mechanism adjusts the position of the screen. 12. A detection system, comprising:a detector apparatus includinga housing;at least one multilayer disposed within the housing, the at least one multilayer defining a leading edge and a trailing edge, wherein the at least one multilayer is adapted to interact with a plurality of high-energy photons within the X-ray portion of the electromagnetic spectrum, impingent from the leading edge, to permit passage of photons of at least one selected from a plurality of different energies,a first securement bracket disposed within the housing, wherein the at least one multilayer is secured to the first securement bracket at a position adjacent to the leading edge,a second securement bracket disposed within the housing, wherein the at least one multilayer is secured to the second securement bracket at a position adjacent to the trailing edge,at least one detector disposed adjacent to the trailing edge of the at least one multilayer, wherein the detector is adapted to detect the impingent high-energy photons after interaction with the at least one multilayer, andat least one adjustment mechanism operatively connected to the second securement bracket, wherein the at least one adjustment mechanism adjusts the position of the second securement bracket to alter an angular position of the at least one multilayer;a first communication link connected to the at least one detector to transmit electrical signals from the at least one detector;a processor connected to the first communication link to receive the electrical signals from the at least one detector and generate control signals for the at least one adjustment mechanism; anda second communication link connected between the processor and the at least one adjustment mechanism. 13. The detection system of claim 12, further comprising:a signal conditioning module disposed between the at least one detector and the processor to amplify the electrical signals from the at least one detector before being transmitted to the processor. 14. The detection system of claim 12, wherein the processor is adapted to generate a spectrum associated with the plurality of high-energy photons. 15. A modular detector apparatus for electromagnetic radiation originating from a source, the modular detector apparatus comprising:a first detector module, the first detector module comprisinga first housing,at least a first multilayer disposed within the first housing, the first multilayer defining a first leading edge and a first trailing edge, wherein the first multilayer is adapted to interact with a plurality of high-energy photons within the X-ray portion of the electromagnetic spectrum, impingent from the first leading edge, to permit passage of photons of at least one selected from a plurality of different energies,a first securement bracket disposed within the first housing, wherein the first multilayer is pivotally fixed to the first securement bracket at a position adjacent to the first leading edge,a second securement bracket disposed within the first housing, wherein the first multilayer is adjustably supported by the second securement bracket at a position adjacent to the first trailing edge,a first screen disposed adjacent to the first leading edge of the first multilayer, wherein the first screen is adapted to adjust the vertical acceptance of the impingent high-energy photons,at least a first detector disposed adjacent to the first trailing edge of the first multilayer, wherein the first detector is adapted to detect the impingent high-energy photons after interaction with the first multilayer;a second detector module, the second detector module comprisinga second housing,at least a second multilayer disposed within the second housing, the second multilayer defining a second leading edge and a second trailing edge, wherein the second multilayer is adapted to interact with a plurality of high-energy photons within the X-ray portion of the electromagnetic spectrum, impingent from the second leading edge, to permit passage of photons of at least one selected from a plurality of different energies,a third securement bracket disposed within the second housing, wherein the second multilayer is pivotally fixed to the third securement bracket at a position adjacent to the second leading edge,a fourth securement bracket disposed within the second housing, wherein the second multilayer is adjustably supported by the fourth securement bracket at a position adjacent to the second trailing edge,a second screen disposed adjacent to the second leading edge of the second multilayer, wherein the second screen is adapted to adjust the vertical acceptance of the impingent high-energy photons,at least a second detector disposed adjacent to the fourth trailing edge of the second multilayer, wherein the second detector is adapted to detect the impingent high-energy photons after interaction with the second multilayer;a screen adjustment mechanism operatively connected to one of the first screen and the second screen;a screen connector configured to couple the first screen and second screen providing, in cooperation with the screen adjustment mechanism, synchronized movement of the first screen and the second screen;an adjustment mechanism connected to one of the second securement bracket and the fourth securement bracket; anda ring connector configured to couple the second securement bracket and the fourth securement bracket providing, in cooperation with the adjustment mechanism, synchronized movement of the second securement bracket and the fourth securement bracket allowing adjustment of the position of the second securement bracket and the fourth securement bracket to alter an angular position of the first multilayer and the second multilayer wherein the first multilayer is oriented along a first radial line extending outwardly from the source and second multilayer is oriented along a second radial line extending outwardly from the source. 16. The modular detector apparatus of claim 15 wherein one of the first multilayer and the second multilayer has a body defining a surface with a grading ratio, wherein the grading ratio follows the equation:d/d0=R/R0wherein d is a d-spacing of the multilayer at a distance R from a center point, andwherein d0 is the d-spacing at R0, which is a minimum value for the at least one multilayer. 17. A method for detecting at least a portion of a spectrum of electromagnetic radiation comprising:positioning at least one detector at a predetermined distance from an electromagnetic radiation source, wherein the electromagnetic radiation source emits at least X-rays and the at least one detector detects at least the X-rays;positioning at least one graded multilayer between the at least one detector and the electromagnetic radiation source, wherein the at least one graded multilayer interacts with the X-rays to permit X-rays with first predetermined energies to pass to the at least one detector and wherein the at least one multilayer has a body defining a surface with a grading ratio, wherein the grading ratio follows the equation:d/d0=R/R0wherein d is a d-spacing of the multilayer at a distance R from a center point, andwherein d0 is the d-spacing at R0, which is a minimum value for the at least one multilayer;detecting a first quantity of X-rays with the first predetermined energies by the detector;adjusting the position of the at least one graded multilayer to permit X-rays with second predetermined energies, different from the first predetermined energies, to pass to the at least one detector; anddetecting a second quantity of X-rays with the second predetermined energies by the detector. 18. The method of claim 17, further comprising:iteratively repeating the adjusting of the position of the at least one graded multilayer and the detecting of a quantity of X-rays until a spectrum spanning a predetermined range of predetermined X-ray energies is collected. 19. The method of claim 17, wherein the detector is at least one selected from a group comprising a proportional counter, a scintillation counter, an ionization chamber, and a solid-state detector. 20. A method for detecting at least a portion of a spectrum of electromagnetic radiation comprising:positioning at least one detector at a predetermined distance from an electromagnetic radiation source, wherein the electromagnetic radiation source emits at least X-rays and the at least one detector detects at least the X-rays;positioning at least one graded multilayer between the at least one detector and the electromagnetic radiation source, wherein the at least one graded multilayer interacts with the X-rays to permit X-rays with first predetermined energies to pass to the at least one detector;mounting a first edge of the at least one graded multilayer on a first radius line;mounting a second edge of the at least one graded multilayer on a second radius line;mounting a first edge of at least a second graded multilayer on the first radius line;mounting a second edge of at least the second graded multilayer on the second radius line;wherein the first radius line and the second radius line are concentric, with a common center point at the electromagnetic radiation source;positioning a plurality of adjustable beam entrance slits between the electromagnetic radiation source and the at least one graded multilayer and the second graded multilayer, wherein the plurality of adjustable beam entrance slits are adapted to control the vertical and horizontal acceptance angles relative to the center point;adjusting an orientation of the at least one graded multilayer and the second graded multilayer to maintain an angular relationship between the multilayers; andwherein adjusting the orientation comprises moving the second edges of the multilayers while maintaining the first edges in a stationary position;detecting a first quantity of X-rays with the first predetermined energies by the detector;adjusting the position of the at least one graded multilayer to permit X-rays with second predetermined energies, different from the first predetermined energies, to pass to the at least one detector; anddetecting a second quantity of X-rays with the second predetermined energies by the detector. 21. The method of claim 20, further comprising:iteratively repeating the adjusting of the position of the at least one graded multilayer and the detecting of a quantity of X-rays until a spectrum spanning a predetermined range of predetermined X-ray energies is collected. 22. The method of claim 20, wherein the detector is at least one selected from a group comprising a proportional counter, a scintillation counter, an ionization chamber, and a solid-state detector. |
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061987875 | summary | BACKGROUND OF THE INVENTION The present invention relates to a method of handling a nuclear reactor and an apparatus used in the handling method and in particular to a handling technique of an internal structure of a nuclear reactor with which a taking-out working of the internal structure of the nuclear reactor to an outside of a nuclear reactor building or containment is provided. In the nuclear reactor building or containment of an atomic power plant station, a nuclear reactor pressure vessel is received and in this nuclear reactor pressure vessel a shroud for surrounding a reactor core of a nuclear reactor and an upper portion lattice plate and a reactor supporting plate etc. exist as an internal structure. In the nuclear reactor internal structure of the atomic power plant station, a damage state of the nuclear reactor internal structure is carried out to inspect and mend during every regular inspection time, however from the aspects of a conservation performance, an economic performance and a prevention preservation in a midway of durable years the nuclear reactor internal structure is exchanged over and renewed with a new nuclear reactor internal structure. In a renewal working or processing of the nuclear reactor internal structure according to the prior art, in a nuclear reactor well pool, an upper portion lattice plate, an upper portion shroud and a reactor core supporting plate etc. are cut roughly and taken out in advance. In a nuclear reactor building using a previous-established ceiling crane, the roughly cut-down upper portion lattice plate, the roughly cut-down upper portion shroud and the roughly cut-down reactor core supporting plate etc. are temporally placed in a dry separator (herein, it is called as "DS"). Further, in DS pool, a decomposition taking-out working and a fine cutting working for the roughly cut-down upper portion lattice plate, the roughly cut-down upper portion shroud and the roughly cut-down reactor core supporting plate etc. are performed with a remote control in the water. Accordingly, the upper portion lattice plate, the upper portion shroud and the reactor core supporting plate etc. are converted to waste materials. Accordingly, so as to isolate the nuclear reactor well pool and DS pool, a DS pool gate is newly established. Further, the fine cut waste materials (the small chips of the nuclear reactor internal structure) are received in plural receptacles such as container or casks and on occasion the fine cut waste materials are transferred and kept on the premises to a sight bunker or a solid waste material reservoir. Accordingly, in accordance with the working for cutting roughly in the nuclear reactor well pool, the working for cutting finely in DS pool, the supplementary working for increasing the storage containers and the supplementary material amount increase, since the above stated workings extend over a long period, there is problems from the aspects of the cost and radiation exposure etc. Further, a long period for a plant stop period in a total becomes a big burden economically to an electric power company from a reduction in an operation of rate of an equipment. On the other hand, when the nuclear reactor pressure vessel is exchanged over, as described in Japanese application patent laid-open publication No. Hei 8-285981 and Japanese application patent laid-open publication No. Hei 8-262190, a technique has known in which the nuclear reactor pressure vessel is drawn in in a transfer receptacle or container which has waited on a roof of a nuclear reactor building by a crane and the nuclear reactor pressure vessel with the transfer container is transferred to a desired point of the nuclear reactor building. In the above stated prior art example, the nuclear reactor pressure vessel with the transfer container is moved at an upper portion to the roof by performing an uncover to the nuclear reactor pressure vessel and by lifting the nuclear reactor pressure vessel with the transfer container in the nuclear reactor building. Accordingly, an exposure reduction countermeasure against to working men in the nuclear reactor building who relates to radiation from the nuclear reactor pressure vessel is insufficient. Taking into consideration about the exposure reduction countermeasure against to the working men in the nuclear reactor building, as disclosed in Japanese application patent laid-open publication No. Hei 9-145882, a techniques has known in which plural divided shield bodies are assembled so as to surround the nuclear reactor pressure vessel in the nuclear reactor building and the reduction of the radiation and the diffusion of the radioactivity are shielded by the shield bodies and as a result the nuclear reactor pressure vessel is taken out from the roof of the nuclear reactor building. However, in the above stated prior art method, since the assembling working of the shield body in the nuclear reactor building equipment is accompanied with, such an assembling working requires the labor time and accordingly it takes the time for carrying out the shield bodies to an outside of the nuclear reactor building. To avoid the labor time for requiring the assembly of the shield bodies in the nuclear reactor building, a technique described in Japanese application patent laid-open publication No. Sho 62-285100 has known. The contents described in this patent publication will be shown as follows. Namely, a cask for storing the radioactive waste material is lifted by a crane which is provided at an outside of the nuclear reactor building and the cask is passed through a provisional opening which is provided with an opening and closing state on the roof of the nuclear reactor building and further the cask is lifted in the nuclear reactor building. After that, the cask is separated once from the crane and the cask is placed on a floor of the nuclear reactor building and further the cask is moved in a horizontal direction at a position in which the radioactive waste material to be subjected to the carry-out exists. Next, the radioactive waste material (in concretely, an upper portion cover of the nuclear reactor pressure vessel) is lifted in the cask using a lifting machine in the cask and after that a bottom portion of the cask is closed according to an opening and closing door. After that, the cask is moved toward a horizontal direction at just an under portion of the provisional opening which has provided with the opening and closing state on the roof of the nuclear reactor building equipment. The cask is connected together with the crane and this cask is lifted up by the crane and further the cask is passed through the provisional opening of the cask and is carried out to the outside of the nuclear reactor building. In the prior art technique described in Japanese application patent laid-open publication No. Sho 62-285100, it takes no labor time for requiring the assembly of the shield bodies in the nuclear reactor building, however it is effective to the reduction of radiation exposure and the diffusion of the radioactivity. However, in this prior art technique, the cask is separated once on the floor of the nuclear reactor building equipment from the crane and the cask is placed and after that the cask is moved toward the horizontal direction. Accordingly, from the lift-in of the cask in the nuclear reactor building equipment until the lift-out of the cask, however since the working of the horizontal movement of the cask and the working of the connection and the separation of the cask and the crane are accompanied with, it takes the labor time yet. In a case of the requirement of the above stated labor time, since the exchange-over of the internal structure is delayed, and further since also a re-operation time period of the atomic power plant station after the exchange-over of the internal structure is delayed, an operation efficiency of the atomic power station becomes worse. Further, in a case of the decomposition of the atomic power plant station, there is a problem about a long-pending of the decomposition working. SUMMARY OF THE INVENTION An object of the present invention is to provide to a method of handing a nuclear reactor and an apparatus used in this handling method wherein it can be compatible with a speedy handling being accompanied with a carry-out for an internal structure of a nuclear reactor from a nuclear reactor building and a handling by reducing an exposure amount which is received by the working men from the radioactive internal structure. For attainment the above stated object according to the present invention, it employs a method of handling an internal structure of a nuclear reactor, comprising the steps of lifting in a container in a nuclear reactor building from an outer side of the nuclear reactor building through an opening which is opened at an upper portion of the nuclear reactor building, maintaining the container under a condition in which the container is lifted at an upper portion of the internal structure of a nuclear reactor pressure vessel, storing the internal structure in the container which is maintained under the lifted condition, and lifting out the container which stores the internal structure toward an outside of the nuclear reactor building through the opening. Further, it employs a method for exchanging over an internal structure of a nuclear reactor, comprising the steps of, taking out the internal structure through the opening which is opened at the upper portion of the nuclear reactor building in accordance with the above stated handing method of the internal structure, lifting up a new internal structure in the nuclear reactor building from the outside of the nuclear reactor building through an opening which is opened at an upper portion of the nuclear reactor building, installing the new internal structure in the nuclear reactor pressure vessel of the nuclear reactor building, and restoring the opening which is opened at the upper portion of the nuclear reactor building. As an apparatus to realize the above stated methods, according to the present invention, it employs a handling apparatus of an internal structure of a nuclear reactor, comprising an opening opened at a portion of a nuclear reactor building which is positioned at an upper portion of a nuclear reactor well pool, a cask for storing the internal structure which has taken off from a nuclear reactor pressure vessel in the nuclear reactor building, a crane for lifting out and lifting in the cask from an outer side of the nuclear reactor building through the opening, and a hoisting device lifted up together with the cask according to the crane and for drawing in the internal structure in the cask. |
039883973 | summary | For high temperature power reactors a uranium-thorium fuel cycle is used with the object of recovering the uranium 233 bred from the thorium as well as the thorium which was not exhausted to gain it in the reprocessing plant and to use it again as a fuel after refabrication. At the same time the recovered uranium 233 replaces a part of the otherwise necessary uranium 235. The atom ratio of uranium to thorium lies between 1:5 to 1:20 in the reactor core in the case of known fuel element types. The arrangement in the fuel element is selected in such a way that the coated fuel particles are available from pure uranium 235 or uranium 233 as well as the coated particles from thorium which contain only a few percent by weight of uranium. Since the valuable uranium 233 is recovered practically only from particles containing thorium, one must have already strived during the head-end process for reprocessing to separate the coated particles containing uranium from those containing thorium, in order to obtain an uranium 233 as pure as possible, i.e., an uranium free of other uranium isotopes. This object can be achieved only in an unsatisfactory manner in the known block fuel element concepts. Since the coated particles containing uranium and thorium in the case of the concepts which have been known hitherto, are used as a mixture, a simple separation is not possible particularly because they are connected together in the fuel particles by the graphite matrix. Suggestions of producing the coated particles containing uranium and thorium in variable sizes, in order to be able to separate them mechanically, have not led to any satisfactory result either, because such block fuel elements must be mechanically comminuted. In the certainly necessary mechanical comminution of the blocks, and in the burning of the graphite in the head-end step, a high percentage of the particles is destroyed so that even in the case of a variable size of the particles no satisfactory separation of the uranium containing particles from the thorium containing particles is possible. The mechanical comminution of the block fuel elements and the subsequent separation of the graphite from the coated particles therefrom make the reprocessing considerably more difficult. The present invention by-passes the above mentioned difficulties. In published German Offenlegungsschrift 1,902,994, Sept. 24, 1970, and Hrovat U.S. application Ser. No. 218,244, filed Jan. 17, 1972 now abandoned, (corresponding to published German application P 21 04 431.5) entitled "Process For The Production Of Block Fuel Elements For Gas Cooled High Temperature Power Reactor", there is described a monolithic block fuel element with pressed in cooling channels. The entire disclosures of said German applications and said U.S. application Ser. No. 218,244 are hereby incorporated by reference. As described in those applications the block fuel element is a compact prism consisting only of a homogeneous graphite matrix. The coated particles are embedded in the graphite matrix and constitute the fuel zones whereby the fuel material in the coated particles either can be a homogeneous mixture of uranium oxide and thorium oxide (or carbide), same in each particle, or otherwise each fuel zone contains a mixture of uranium particles as feed material and thorium particles as breed materials. At the same time it is essential that these zones be connected with the remaining graphite matrix without transition and that they produce the actual fuel element structure. The fuel zones themselves are the integrated components of the block element and they contribute substantially to the strength of the fuel element. From this the possibility results of making the fuel zones substantially larger and designing them with more flexibility. The high mechanical integrity of the pressed block element achieved thereby permits attachment of the uranium containing coated particles separate from the thorium containing particles in the form of fuel columns located side by side in a so-called heterogeneous arrangement (according to FIGS. 1 - 3). In the previously known fuel element types made of prefabricated and drilled electrographite blocks, such an arrangement is not possible, since such a block has been too greatly weakened in its strength as a result of the large number of bores. Therefore the object of the invention is the arrangement of feed and breed zones in a block of a fuel element, in which these zones contribute to the mechanical strength of said block, arranging said zones in such a way that they can be selectively drilled out of this block. Thereby the thorium containing breed particles and the uranium containing feed particles can be taken separately from the structural graphite. As a result, the burning of the graphite, which raises a difficult exhaust gas problem, is essentially avoided. Therefore, the thorium containing particles can be supplied free of feed particles for the recovery of uranium 233. The feed zones and breed zones can be separated independently from the structural graphite prior to the processing operation. In the reprocessing, uranium 233 is produced from the breed particles and uranium 235 and fission products are produced from the feed particles. The feed and breed arrangement can be selected differently, dependent upon the power density of the core and the useful life of the fuel element, as well as on the conversion rate striven for. |
summary | ||
claims | 1. A method of operating a reactor (1) of a nuclear plant in which the reactor (1) comprises a reactor vessel (6) enclosing a core having a plurality of fuel elements (7) and a number of control rods (8),wherein each fuel element (7) includes a plurality of elongated fuel rods (9), which each has an upper end (9′) and a lower end (9″) and includes a cladding (10) and nuclear fuel in the form of fuel pellets (11) enclosed in an inner space (12) formed by the cladding,wherein the fuel pellets (11) are arranged in the inner space to leave a free volume in the inner space, wherein the free volume comprises an upper plenum (12′), containing no nuclear fuel and provided in the proximity of the upper end of the fuel rod, a lower plenum (12″), containing no nuclear fuel and provided in the proximity of the lower end of the fuel rod, and a pellet-cladding gap between the fuel pellets (11) and the cladding (10),wherein a reactor coolant, during operation of the reactor, is re-circulated as a coolant flow through the core in contact with the fuel rods (9) and is added to the reactor via a feed-water conduit (4) as feed-water having a normal feed-water temperature providing a sub-cooling of the reactor coolant, andwherein each of the control rods (8) is displaceable a control rod distance to be inserted into and extracted from a respective position between respective fuel elements in the core, the method including the following steps of operation:operating the reactor at a normal power and a normal sub-cooling during a normal state,monitoring the reactor for detecting a defect on the cladding of any of the fuel rods,upon detecting a defective fuel rod having said defect on the cladding thereof, changing the operation of the reactor to a particular state that causes an increase of the free volume at least in the defective fuel rod in which the defect is detected,operating the reactor at the particular state during a limited time period, andoperating, after said time period, the reactor at substantially the normal state. 2. A method according to claim 1, comprising providing the lower plenum (12″) to have a longitudinal length along the elongated fuel rod (9) and the upper plenum (12′) to have a longitudinal length along the elongated fuel rod (9), wherein the longitudinal length of the lower plenum (12″) is provided to be shorter than the longitudinal length of the upper plenum (12′). 3. A method according to claim 2 , wherein the longitudinal length of the lower plenum is provided to be less than 30% of the total longitudinal length of the upper plenum (12′) and the lower plenum (12″). 4. A method according to claim 1, wherein changing the operation of the reactor to the particular state comprises at least one of the following steps of operation:operating the reactor at a reduced power in relation to the normal power during the normal state, andoperating the reactor at an increased sub-cooling of the reactor coolant in relation to the normal sub-cooling during the normal state in order to achieve a larger temperature gradient over the fuel rod. 5. A method according to claim 4, wherein changing the operation of the reactor to the particular state comprises reducing the coolant flow of the reactor coolant through the core. 6. A method according to claim 4, wherein the added reactor coolant is preheated outside the reactor during the normal state by means of a preheating arrangement (14), and wherein said increased sub-cooling of the reactor coolant is obtained by reducing the preheating of the added reactor coolant. 7. A method according to claim 4, wherein said reduced power is obtained by displacing at least some of the control rods (8) into the core at least a part of the control rod distance. 8. A method according to claim 7, wherein substantially all control rods (8) are at least periodically displaced at least a part of the control rod distance during the particular state. 9. A method according to claim 4, wherein said reduced power is obtained by displacing successively different groups of the control rods (8) at least a part of the control rod distance, wherein each such group defines a respective specific part of the core. 10. A method according to claim 4, wherein the reactor is operated at the reduced power during the whole time period of the particular state. 11. A method according to claim 4, wherein substantially all control rods are displaced at least a part of the control rod distance during the whole time period of the particular state. 12. A method according to claim 1, wherein the particular state is initiated at least within 72 hours after the detection of a defect. 13. A method according to claim 1, wherein the particular state is initiated at least within 48 hours after the detection of a defect. 14. A method according to claim 1, wherein the particular state is initiated at least within 24 hours after the detection of a defect. 15. A method according to claim 1, wherein the particular state is initiated immediately after the detection of a defect. 16. A method according to claim 1, wherein the particular state involves that at least some of the control rods are alternately displaced in the core for obtaining an alternating increase and decrease of the power. 17. A method according to claim 1, wherein said monitoring includes continuous monitoring during the operation of the reactor. 18. A method according to claim 1, wherein the monitoring includes sensing of a radioactive activity in a gas flow from the reactor. 19. A method according to claim 1, wherein the fuel rod is provided to comprise a hydrogen absorbing element (21′) in at least one of the upper plenum (12′) and the lower plenum (12″). 20. A method according to claim 19, wherein the hydrogen absorbing element (21′, 21″) is provided to comprise a hydrogen absorbing body having a surface coated with a layer of a substance that is non-oxidizing and permeable to hydrogen. 21. A method according to claim 20, wherein the absorbing body is enclosed in an imaginary body having a substantially convex outer surface, and wherein a surface area of the absorbing body is greater than a surface area of the outer surface of the imaginary body. 22. A method according to claim 20, wherein the substance is provided to comprise at least one metal in the group consisting of palladium, rhodium, rhenium and alloys comprising one or more of these metals. 23. A method according to claim 20, wherein the absorbing body is provided to comprise at least one metal in the group consisting of zirconium, titanium, nickel and alloys comprising one or more of these metals. 24. A method according to claim 1, wherein the fuel rod (9) is provided to comprise a distance element (20′) provided in the upper plenum (12′) and/or a distance element (20″) provided in the lower plenum (12″). 25. A method according to claim 24, wherein at least one of the distance elements (20′, 20″) is provided to form the absorbing element. 26. A method according to claim 24, wherein at least one of the distance elements is provided to be deformable for permitting swelling of the fuel pellets. |
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abstract | Disclosed is an X-ray source, including: a cathode including a shielding channel through which an X-ray passes; emitters formed on an upper surface of the cathode, and arranged around the shielding channel; an anode positioned so as to face the cathode, and including an anode target in which an E-beam is focused; and a gate electrode positioned between the cathode and the anode, and including gate holes at positions corresponding to those of the emitters. |
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summary | ||
abstract | An inspection apparatus may include an installation platform and a scan head. The scan head may be configured to engage in an index movement, a theta movement, a wrist fold movement, and a scan movement. During the scan movement, a transducer of the scan head travels a circumferential path so as to allow an inspection of a surface that is opposite of a surface on which the installation platform is mounted. The inspection apparatus may be used to inspect a reactor component in a nuclear reactor. |
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claims | 1. A nuclear fission fuel mixture, comprising:at least one naturally fissioning actinide; anda ternary metallic additive comprising:a metal first component as a first percentage of the additive total weight;a metal second component as a second percentage of the additive total weight;a metal third component as a third percentage of the additive total weight, wherein the first percentage, second percentage, and third percentage sum to about one hundred percent;wherein the first percentage, second percentage, and third percentage are selected in a solid phase region of an isothermal ternary phase diagram of the ternary metallic additive taken at a temperature between about 450 Celsius degrees and about 700 Celsius degrees. 2. The nuclear fission fuel mixture of claim 1, wherein the first component is molybdenum, the second component is titanium, and the third component is zirconium. 3. The nuclear fission fuel mixture of claim 2, wherein the ternary metallic additive comprises:molybdenum in a range from about 18 percent to about 55 percent of the additive total weight;titanium in a range from about 38 percent to about 59 percent of the additive total weight; andzirconium in a range from about 2 percent to about 27 percent of the additive total weight. 4. The nuclear fission fuel mixture of claim 2, wherein the ternary metallic additive comprises:molybdenum in a range from about 45 percent to about 55 percent of the additive total weight;titanium in a range from about 38 percent to about 48 percent of the additive total weight; andzirconium in a range from about 2 percent to about 12 percent of the additive total weight. 5. The nuclear fission fuel mixture of claim 2, wherein the ternary metallic additive comprises:molybdenum in a range from about 18 percent to about 28 percent of the additive total weight;titanium in a range from about 48 percent to about 58 percent of the additive total weight; andzirconium in a range from about 17 percent to about 27 percent of the additive total weight. 6. The nuclear fission fuel mixture of claim 2, further comprising niobium, wherein the nuclear fission fuel mixture contains less niobium by weight than any one of molybdenum, titanium, and zirconium. 7. The nuclear fission fuel mixture of claim 1, wherein the naturally fissioning actinide is uranium. 8. The nuclear fission fuel mixture of claim 7, further comprising plutonium. 9. The nuclear fission fuel mixture of claim 1, wherein the first component is molybdenum and the second component is tungsten. 10. The nuclear fission fuel mixture of claim 1, wherein the first component is molybdenum and the second component is tantalum. 11. A nuclear fission fuel mixture for use in a fission reactor in which the nuclear fission fuel mixture remains solid under all anticipated operating conditions, the nuclear fission fuel mixture comprising:at least one naturally fissioning actinide as a first percentage of the total weight of the nuclear fission fuel mixture;molybdenum as a second percentage of the total weight of the nuclear fission fuel mixture; andone or more metals other than molybdenum as a third percentage of the total weight of the nuclear fission fuel mixture,wherein the first percentage, second percentage, and third percentage are selected such that the nuclear fission fuel mixture exhibits a solidus temperature above the likely to exceed the fuel operating temperature of a sodium cooled fast reactor. 12. The nuclear fission fuel mixture of claim 11, wherein the one or more metals comprises at least one of titanium, zirconium, tungsten, tantalum, niobium and palladium. 13. The nuclear fission fuel mixture of claim 11, wherein the at least one naturally fissioning actinide comprises at least one of uranium and plutonium. 14. The nuclear fission fuel mixture of claim 11, wherein the nuclear fission fuel mixture comprises:uranium in a range from about 85 percent to about 99 percent of the total weight of the nuclear fission fuel mixture;molybdenum in a range from about 1 percent to about 13 percent of the total weight of the nuclear fission fuel mixture; andtitanium in a range from about 1 percent to about 4 percent of the total weight of the nuclear fission fuel mixture. 15. The nuclear fission fuel mixture of claim 11, wherein:the first percentage, second percentage, and third percentage sum to about one hundred percent; andthe first percentage is in a range of about 80 percent to about 97 percent of the total weight of the nuclear fission fuel mixture. 16. The nuclear fission fuel mixture of claim 11, wherein:the first percentage, second percentage, and third percentage sum to about one hundred percent;the first percentage, second percentage, and third percentage are selected in a solid phase region of an isothermal ternary phase diagram of the nuclear fission fuel mixture taken at a temperature below the upper temperature limit. 17. The nuclear fission fuel mixture of claim 11, wherein:the first percentage, second percentage, and third percentage are selected in a triangular region of a uranium-molybdenum-titanium ternary diagram having three corners selected from a set of four points consisting of:uranium as about 98 percent and titanium as about 2 percent of the total weight of the nuclear fission fuel mixture;uranium as about 90 percent and molybdenum as about 10 percent of the total weight of the nuclear fission fuel mixture;uranium as about 87 percent, molybdenum as about 11 percent, and titanium as about 2 percent of the total weight of the nuclear fission fuel mixture; anduranium as about 90 percent, molybdenum as about 9 percent, and titanium as about 1 percent of the total weight of the nuclear fission fuel mixture. 18. A method of making nuclear fission fuel, comprising:providing at least one naturally fissioning actinide;providing molybdenum;providing a metal other than molybdenum; andpreparing a total weight of a fuel mixture by mixing the at least one naturally fissioning actinide, the molybdenum, and the metal, the fuel mixture comprising:the at least one naturally fissioning actinide as a first percentage of the total weight;the molybdenum as a second percentage of the total weight; andthe metal as a third percentage of the total weight,wherein the first percentage, second percentage, and third percentage are selected in a body-centered cubic solid phase region of a phase diagram of the fuel mixture. 19. The method of claim 18, wherein:the first percentage, second percentage, and third percentage are selected in a triangular region of a uranium-molybdenum-titanium ternary diagram having three corners selected from a set of four points consisting of:uranium as about 98 percent and titanium as about 2 percent of the total weight of the nuclear fission fuel mixture;uranium as about 90 percent and molybdenum as about 10 percent of the total weight of the nuclear fission fuel mixture;uranium as about 87 percent, molybdenum as about 11 percent, and titanium as about 2 percent of the total weight of the nuclear fission fuel mixture; anduranium as about 90 percent, molybdenum as about 9 percent, and titanium as about 1 percent of the total weight of the nuclear fission fuel mixture. 20. The method of claim 18, wherein the fuel mixture comprises:the at least one naturally fissioning actinide; anda ternary metallic additive comprising:molybdenum in a range from about 18 percent to about 55 percent of the weight of the additive;titanium in a range from about 38 percent to about 59 percent of the weight of the additive; andzirconium in a range from about 2 percent to about 27 percent of the weight of the additive. |
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043448720 | description | A vacuum evaporator 1, especially a wiping blade evaporator, possibly a two-step evaporator, is charged at 2 with the solution of fission products, particularly a nitric solution. The solution is concentrated in the evaporator with recovery of acid condensate at 3, which condensate can be returned for fuel dissolving. The concentrate is mixed in the mixing device 4 with vitrifiers (added at 5) and a nitrate decomposing agent, particularly urea (added at 6), and passes either directly or after intermediate drying at 7 into the vitrifying oven 8. The vitrification of the mass along with simultaneous decomposition of nitrates takes place in the oven 8. The vitrified material leaves the oven at 9, while the practically nitrous-free waste gases are liberated at 10. The method proceeds as follows: (1) Vacuum Concentration The separation of the nitric acid from the fission product solution is preferably effected in a wiping blade evaporator, System-Sambay, at about 20 mm Hg pressure (corresponding to 35.degree. to 40.degree. C. vapor temperature). During concentration to 1/10 of the starting volume, a good 40% of the total nitrate quantity is separated out as nitric acid. Preferably, the concentrate residue obtained is rediluted with water (approximately in the ratio of 1:1), and is again concentrated as before, resulting in an increase of the nitric acid separation to a good 70%. Ruthenium was detectable in the acid condensate only in quantities below 1 ppm (decontamination factor: 10.sup.3). Up to a solid content of 25%, the concentrate flows off freely and without forming a crust. The nitrate content is about 9 M/l, of which about 1/3 is difficult to decompose nitrate. The preceding concentration can be utilized to reduce the volume of fresh solutions of fission products which are first to be temporarily stored. As a result, the space which is required and the corrosiveness are reduced. The degree of concentration depends upon the maximum allowable decay energy of the fission products per volume and the chemical characteristics of the solution. However, after the addition of urea and the corresponding vitrifiers, the concentrate can also be supplied directly to a subsequent solidification process, particularly a drying and vitrifying process. To adapt to reliable and practiced technologies, a drying by means of a roller or drum dryer can be provided. The addition of urea can take place prior to the drying or directly before the fusing. A fluid dosing of waste-vitrifier-urea-suspension into the smelting or vitrifying oven is also possible. After being concentrated, the separated-off acid can be returned to the dissolving process. (2) Vitrification, particularly with the addition of urea The fission product concentrate obtained pursuant to step (1) is, in conformity with the respective formula, mixed with vitrifiers as well as with a quantity of urea which, according to the total nitrate content and the ratio of nitrate to free acid, is about 20 to 300% of the nitrate content. Determinative is the nitrous fumes content of the melting off gas, which should be below 3% by volume NO.sub.x. The thus obtained mixtures can readily be dehydrated, for example, with a roller dryer, and yield a less powdery, more granular product, which can be easily and freely melted in a melting device to a clean glass mass. EXAMPLE 12 l fission product solution (with about 1.5 M/l nitrate and 1.6% fission product oxide) with 17.9 Mol nitrate were concentrated in a wiping blade evaporator at 40 mm Hg pressure and a corresponding vapor temperature of 44.degree. C. 1.083 l concentrate (with 9.35 M/l nitrate and 11.8% fission product oxide) as well as 10.4 l distillate (with an acid content of 0.7 M/l) was obtained. The degree of concentration (12 l:1.083 l) was about 11. The nitrate balance is computed as follows: From the starting amount of 17.9 Moles nitrate 10.1 Moles nitrate remained in the concentrate corresponding to a reduction of about 44%; 7.28 Moles nitrate are found in the distillate. The concentrate was rediluted with water (1:1), and was again evaporated under the same conditions, yielding 520 ml concentrate (with 9.82 M/l nitrate and 24.6% fission product oxide) as well as 1460 ml distillate (with an acid content of 2.6 M/l) at a degree of concentration (altogether) of 23. The nitrate balance of this two-step treatment is as follows: ______________________________________ Starting solution Concentrate 2 Distillate 1 and 2 ______________________________________ 17.9 M nitrate 5.1 M nitrate 7.28 + 3.80 M = 11.1 M nitrate ______________________________________ i.e., a reduction of 71.5% of the starting nitrate quantity. 100 g fission product concentrate (with 9.36 M/l nitrate and 20% fission product oxide) was, with the addition of water, mixed with 38 g silicic acid (SiO.sub.2), 30 g borax, 11 g boron oxide, and 15 g calcium oxide and, in the ratio nitrate:urea=3:1, was mixed with 15 g urea. The mixture was dried and continuously dosed into a melting crucible which was at 1100.degree. C. The collected waste gas contained 3% by volume NO.sub.x, 10% CO.sub.2 and 10.8% CO, as well as unknown quantities of N.sub.2 and N.sub.2 O from the reaction between nitrate and urea and was nearly colorless. The fused glass was yellowish-gray, ceramic-like, and "homogeneous" (i.e. in itself uniform). The advantages of the method of the present invention, which were in part already expanded upon above, comprise a reduction of the method steps, an extensive recovery of acid, a lower usage of reduction agents, and a simplification of the waste gas treatment. The present invention is, of course, in no way restricted to the specific disclosure of the specification, examples, and drawing, but also encompasses any modifications within the scope of the appended claims. |
claims | 1. An apparatus for measuring the spin orientation and magnetization of charged particle beams, comprised of a cavity resonator with a bore for the passage of the beam to be measured; anda conductive ring positioned coaxially within the bore of said cavity resonator for the passage of the beam to be measured, whereby the interaction of the beam passing through the conductive ring couples to a resonance in the cavity resonator; andan antenna coupled to the cavity's resonance for measurement of its amplitude and phase. |
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abstract | A method of testing an electronics module (11) for an underwater well installation, comprises the steps of: providing a test equipment (7) comprising a processor (8) and a Local Area Network (LAN) switch (9), such that the processor (8) may communicate with the switch (9); providing an electronics module (11) comprising a data acquisition means (12) and a second LAN switch (10), such that the data acquisition means (12) may communicate with the second switch (10); passing test data from the processor (8) to the data acquisition means (12) via the first and second LAN switches (9, 10); and monitoring the response of the electronics module (11) in response to the test data. |
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summary | ||
041994043 | abstract | A fuel-pellet composition for use in fast breeder reactors. Uranium carbide particles are mixed with a powder of uranium-plutonium carbides having a stable microstructure. The resulting mixture is formed into fuel pellets. The pellets thus produced exhibit a relatively low propensity to swell while maintaining a high density. |
055966187 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS Preferred embodiments of the present invention will be explained in conjunction with the drawings. FIG. 1 is a perspective view of a first embodiment. The exposure apparatus of this embodiment comprises a cylindrical mirror 2 for expanding a sheet-like SR-X-ray beam L (synchrotron radiation light) (illumination light), emitted from an emission point 1 of a charged particle accumulation ring, in the direction of the thickness thereof (Y-axis direction); an unshown exposure chamber having a beryllium window 3 for transmitting the SR-X-ray beam; an exposure amount controlling device 4 disposed within the exposure chamber; a mask 5 held by an unshown holding device; and a wafer stage 6 for holding a wafer W. The wafer stage 6 is movable along a plane perpendicular to the optical axis of the SR-X-ray beam L, and it serves to hold the wafer W at a position spaced by about 30 microns from the mask 5 and in parallel to the mask 5. Mounted on the wafer stage 6 is an X-ray detector 6b. The X-ray beam intensity distribution within the exposure angle can be measured by this detector by scanningly moving the wafer stage 6. The wafer W can be attracted to an attracting chuck 6a mounted on the wafer stage 6. In operation, the wafer stage 6 is moved along a plane perpendicular to the optical axis of the SR-X-ray beam L impinging on the wafer W, in the Y-axis direction and in a direction orthogonal thereto (X-axis direction), so as to move stepwise each exposure zone of the wafer W to the plane of irradiation of the SR-X-ray beam L. Any positional deviation between the pattern of the mask 5 and each exposure zone of the wafer W can be detected by alignment optical systems 8a-8d, and the wafer W can be finally positioned by actuating the aforementioned driving mechanism of the wafer stage 6 or an unshown fine-motion adjusting mechanism. The SR-X-ray beam L, being expanded by the cylindrical mirror 2, has a non-uniform intensity distribution similar to a Gaussian distribution, with respect to the Y-axis direction (predetermined direction). It is directed through the beryllium window 3 and then into the exposure chamber. It goes via the exposure amount controlling device 4 and the mask 5, and it impinges on the wafer W. By this, the pattern of the mask 5 is transferred and printed on the wafer. The exposure amount controlling device 4 comprises a Y-aperture stage 4a being movable and adjustable in the Y-axis direction; an X-aperture stage 4b which is an aperture stage being reciprocally movable in the X-axis direction along the Y-aperture stage 4a; an unshown scanning mechanism comprising driving means for scanningly moving the X-aperture stage 4b in the X-axis direction; an aperture movable member (protruded member) 4c which is supported and is movable and adjustable in the X-axis direction along the X-aperture stage 4b; and an X-ray detector array (detecting means) 4d disposed along the Y-axis direction upon the X-aperture stage 4b. The X-aperture stage 4b has an aperture 9 of substantially rectangular shape for passing the SR-X-ray beam L. The Y-aperture stage 4a has a large aperture (not shown) of a size not blocking the path of the SR-X-ray beam regardless of the movement of the same to any position on the Y-axis direction. Formed at an end edge of the aperture movable member 4c in the X-axis direction is a protruded portion 4e, protruding toward the aperture 9. It has a shape which is determined on the basis of the X-ray intensity distribution I.sub.0 (y) of the SR-X-ray beam measured beforehand. It serves to change the aperture width D (FIG. 2) of the aperture 9 in the X-axis direction, in accordance with a predetermined pattern and in the Y-axis direction. For exposure of each exposure zone of the wafer W, the X-aperture stage 4b is scanningly moved in the X-axis direction at a predetermined scan speed V to adjust the exposure time and to avoid exposure non-uniformness due to the X-ray intensity distribution I.sub.0 (y) of the SR-X-ray beam. This will be explained in more detail, below. On the surface of the wafer W, there is a relation such as below among the exposure amount E(y) at a certain position in the Y-axis direction, the X-ray intensity I(y) of the SR-X-ray beam and the exposure time T(y): EQU E(y)=I(y).multidot.T(y) (1) Also, the following relation lies among the exposure time T(y), the aperture width D(y) of the aperture 9 of the X-aperture stage 4b in the X-axis direction and the scan speed V of the X-aperture stage 4b: EQU T(y)=D(y)/V (2) From equations (1) and (2), it follows that: EQU E(y)=I(y).multidot.D(y)/V (3) For example, if the X-ray intensity distribution I.sub.0 (y) of the SR-X-ray beam in the Y-axis direction such as represented by a curve R.sub.0 in FIG. 3 wherein the intensity is highest at the central portion in the Y-axis direction, about 1.10 times higher as compared with the lowest level at the opposite end portions thereof, then the shape and the amount of protrusion of the protruded portion 4e of the aperture movable member 4c are set so that the aperture width D(y) of the aperture 9 of the X-aperture stage 4b in the X-axis direction changes in accordance with a curve S.sub.0 (FIG. 4) wherein the width is smallest at the central portion with respect to the Y-axis direction, about 0.91 times of the largest width at the opposite end portions. Then, for each exposure cycle, the X-aperture stage 4b is scanningly moved at the predetermined speed V. By this, from equation (3), the exposure amount E(y) upon the wafer W surface becomes uniform at any position along the Y-axis direction, and occurrence of exposure non-uniformness is prevented. Since the scan speed V of the X-aperture stage 4b is substantially constant, there is no necessity of a large acceleration and deceleration of the driving motor of the X-aperture stage 4b. Thus, there is substantially no possibility of transfer error due to vibration attributable to such acceleration and deceleration of the driving motor. The X-ray detector array 4d serves to measure the X-ray intensity distribution I(y) of the SR-X-ray beam L in the Y-axis direction each time the X-ray aperture stage 4b is scanned. If the path or X-ray intensity of the SR-Xray beam L changes during repetition of exposure cycle, such a change is fed back so as to change the Y-axis position of the X-aperture stage 4b, the X-axis position of the aperture movable member 4c on the X-aperture stage 4b and/or the scan speed V of the X-aperture stage 4b on the basis of outputs of the X-ray detector array 4d. By this, it is possible to prevent a change in exposure amount or non-uniform exposure, over the whole wafer W surface. If the change in intensity or intensity distribution of the SR-X-ray beam L is slow as compared with the exposure time, the X-ray detector 6b may be scanned in the Y-axis direction to measure the X-ray intensity distribution beforehand. On that occasion, the X-ray detector array 4d may be omitted. Now, the manner of making the exposure amount uniform will be explained. If the path of the SR-X-ray beam L shifts in the Y-axis direction, a deviation .increment.y in the Y-axis direction (hereinafter "y-offset") of the peak position of the output of the X-ray detector array 4d may be detected and the Y-aperture stage 4a may be moved in the Y-axis direction by .increment.y. By this, the aperture 9 of the X-aperture stage 4b is moved in the same direction by the same amount. If, as shown in FIG. 5, the X-ray intensity I of the SR-X-ray beam L decreases generally to about a half, for example, the scan speed of the X-aperture stage 4b may be controlled and decreased to a half. This effectively avoids a change in exposure amount. If, as shown in FIG. 6, the X-ray intensity of the SR-X-ray beam L decreases generally to about a half and, additionally, the X-ray intensity distribution I(y) changes locally such that the highest level in the central portion becomes 0.60 while the lowest level at the opposite end portions becomes 0.50, then the scan speed of the X-aperture stage 4b may be decreased while the aperture movable member 4c may be moved in the X-axis direction along the X-aperture stage 4b, to avoid a change in exposure amount over the whole wafer surface and to minimize non-uniformness of exposure. The movement amount of the aperture movable member on the X-aperture stage 4b can be calculated as follows: From equation (3), the condition for attaining the same exposure amount with the central portion of and with the opposite end portions of the SR-X-ray beam impinging on the wafer W surface, is: EQU 0.60.multidot.Dm=0.05.multidot.(Dm+De) (4) where Dm is the aperture width of the central portion in the Y-axis direction of the aperture 9, and De is the amount of protrusion in the X-axis direction of the central portion, in the Y-axis direction, of the aperture movable member 4c. From FIG. 4. De=0.09. Substituting this into equation (4) to calculate Dm, if follows that: EQU Dm=0.45 Namely, it is seen that, from the state of FIG. 4, the aperture movable member 4c may be moved in the X-axis direction until, as shown in FIG. 7, the aperture width Dm at the central portion in the Y-axis direction of the aperture 9 becomes equal to 0.45. Here, the scan speed of the X-aperture stage 4b is decreased to 0.27 times. FIG. 8 shows the sequence of measurement of the X-ray intensity distribution I(y) of the SR-X-ray beam L for each exposure cycle as well as adjustment of the exposure amount based on the output of the measurement. Initially, before the start of the exposure process, the X-aperture stage 4b is scanned in the X-axis direction and the X-ray intensity distribution I(y) of the SR-X-ray beam L is measured by using the X-ray detector array 4d (Step 1). The measured intensity distribution is compared with the X-ray intensity distribution I.sub.0 (y), having been measured beforehand. Then, a deviation in the Y-axis direction between the peaks of them (i.e., y-offset) and the movement amount of the aperture movable member on the X-aperture stage 4b in the X-axis direction as well as the scan speed of the X-aperture stage 4b for the subsequent exposure cycle are calculated (Step 2). On the basis of the thus calculated y-offset and movement amount of the aperture movable member 4c, the Y-aperture stage 4a and the aperture movable member 4c are moved, respectively (Step 3). Then, the subsequent exposure cycle starts, and the X-aperture stage 4b is scanned. Here, the scan speed is controlled to the level as calculated at Step 2 (Step 4). With the scan of the X-aperture stage 4b, a change in the X-ray intensity distribution I(y) of the SR-X-ray beam L is measured by the X-ray detector array 4d, and Step 2 to Step 4 are repeated. It is to be noted that, when the X-ray intensity distribution I(y) of the SR-X-ray beam L changes locally, the aperture movable member 4c may be pivotally moved about a predetermined axis such as shown in FIG. 9, in place of moving the same in the X-axis direction, to thereby reduce non-uniformness of exposure. Alternatively, in place of using the aperture movable member 4c being movable along the X-axis direction, two pivotal plates 25 and 26 (FIG. 10) being pivotally mounted to each other by a pivot pin 24 may be provided, such that the angle of rotation of them may be changed in accordance with the change in the X-ray intensity distribution I(y) of the SR-X-ray beam L. As a further alternative, in place of using X-aperture stage 4b, a belt-like shutter 34 (FIG. 11A) of a sheet-like member having an aperture 39 similar to the aperture 9 may be used. The shutter 34 may be scanned in the X-axis direction by rotating a pair of rolls 35 and 36 around which end edges of the shutter 34 may be wound. As best shown in Figure 11B, the shutter 34 may be provided with second and third apertures 40 and 41 of different aperture shapes. When the X-ray intensity distribution I(y) of the SR-X-ray beam L changes locally, an appropriate one of these apertures 40 and 41 may be selected and used. As a still further alternative, in place of using the aperture movable member 4c being movable in the X-axis direction, an arrangement shown in FIG. 12 may be used: which comprises a belt-like elastic or resilient member 54 being able to protrude into an end edge portion of the aperture 9 of the X-aperture stage, in the X-axis direction; a plurality of pushing rods 56-58 for changing the amount of protrusion of the resilient member 54; and actuators 56a-58a for actuating the pushing rods individually. When the X-ray intensity distribution I(y) of the SR-X-ray beam L changes locally, the pushing rods 56-58 may be actuated correspondingly to deform the elastic member 54 appropriately to prevent non-uniformness of exposure. Further, as shown in FIG. 13, a second elastic or resilient member 60 separate from the resilient member 54 may be added and disposed so as to be opposed to the resilient member 54. By deforming the resilient member 60 with its pushing rods 61-66 and their actuators 61a-66a, higher precision in the correction of exposure amount is assured. Next, an embodiment of a semiconductor device manufacturing method using the exposure apparatus described hereinbefore will be explained. FIG. 14 is a flow chart of the sequence of manufacturing a semiconductor device such as a semiconductor chip (e.g. IC or LSI), a liquid crystal panel or a CCD, for example. Step 11 is a design process for designing the circuit of a semiconductor device. Step 12 is a process for manufacturing a mask on the basis of the circuit pattern design. Step 13 is a process for manufacturing a wafer by using a material such as silicon. Step 14 is a wafer process which is called a pre-process wherein, by using the so prepared mask and wafer, circuits are practically formed on the wafer through lithography. Step 15 subsequent to this is an assembling step which is called a post-process wherein the wafer processed by Step 14 is formed into semiconductor chips. This step includes assembling (dicing and bonding) and packaging (chip sealing). Step 16 is an inspection step wherein operability check, durability check and so on of the semiconductor devices produced by Step 15 are carried out. With these processes, semiconductor devices are finished and they are shipped (Step 17). FIG. 15 is a flow chart showing details of the wafer process. Step 21 is an oxidation process for oxidizing the surface of a wafer. Step 22 is a CVD process for forming an insulating film on the wafer surface. Step 23 is an electrode forming process for forming electrodes on the wafer by vapor deposition. Step 24 is an ion implanting process for implanting ions to the wafer. Step 25 is a resist process for applying a resist (photosensitive material) to the wafer. Step 26 is an exposure process for printing, by exposure, the circuit pattern of the mask on the wafer through the exposure apparatus described above. Step 27 is a developing process for developing the exposed wafer. Step 28 is an etching process for removing portions other than the developed resist image. Step 29 is a resist separation process for separating the resist material remaining on the wafer after being subjected to the etching process. By repeating these processes, circuit patterns are superposedly formed on the wafer. While the invention has been described with reference to the structures disclosed herein, it is not confined to the details set forth and this application is intended to cover such modifications or changes as may come within the purposes of the improvements or the scope of the following claims. |
056299656 | claims | 1. A diving bell type control rod having a control element including a cladding tube to be immersed longitudinally in a liquid sodium coolant, both ends of said cladding tube being provided with upper and lower end plugs; a pellet chamber disposed in said cladding tube for loading a plurality of pellets of boron carbide; an intermediate plug disposed above said pellet chamber; an upper chamber formed above said intermediate plug; a vent tube so disposed as to penetrate through said intermediate plug and to allow said pellet chamber to communicate with said upper chamber; and a vent hole so formed as to penetrate through said cladding tube located at said upper chamber; characterized in that said vent hole comprises an upper vent hole and a lower vent hole formed in upper and lower two stages in such a manner as to penetrate through said cladding tube located at said upper chamber, a sodium inflow port is so formed as to open to the upper surface of said intermediate plug, and a sodium introduction tube is so disposed as to extend from said sodium inflow port to a position below the lower end surface of said vent tube while penetrating through said intermediate plug. |
claims | 1. A lifting lug for a nuclear-waste container, the lifting lug comprising: a tubular body having an inwardly open large-diameter inner end formed with a flange adapted to be bolted to the container, an outwardly open small-diameter outer end, and a step between the ends; and a transverse wall tightly fitted to the tubular body between the inner and outer ends and bearing outwardly on the step, the wall being a separate piece from the body. 2. The waste-container lifting lug defined in claim 1 wherein the wall is press-fitted to the tubular body. claim 1 3. The waste-container lifting lug defined in claim 1 wherein the wall is mounted at a load point of the lifting lug. claim 1 4. The waste-container lifting lug defined in claim 1 wherein the wall is formed of a disk and a ring. claim 1 5. The waste-container lifting lug defined in claim 1 wherein the body is centered on an axis, the wall being a body of revolution also centered on the axis. claim 1 |
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046363360 | abstract | A process is described for reducing the volume of a liquid waste containing an organic amine chelating agent in which a finely atomized spray of the liquid waste is contacted with a gas stream having a temperature in excess of the thermal decomposition temperature of the chelating agent. The proportions of the hot gas stream and liquid waste are controlled to rapidly evaporate water from the liquid waste and cool the gas to a temperature below the decomposition temperature of the chelating agent in a time of less than about 6 seconds to produce a dry, flowable powder product including the chelating agent. |
claims | 1. A cooling apparatus (101) for a sample in an ion beam etching process, comprising:a sample stage (102) for receiving the sample;a coolant receptacle (120) containing a coolant;a cooling finger (105) connectable to the coolant receptacle (120), the cooling finger (105) including a conduit (130, 131) through which the coolant can flow when the cooling finger is connected to the coolant receptacle (120);at least one thermal conduction element (106a, 106b) connecting the sample stage (102) to the coolant in the cooling finger (105) in heat-transmitting fashion such that heat is transferred from the sample stage (102) to the coolant, wherein the thermal conduction element is fabricated of a metal;an evaporator block (109) for evaporating the coolant flowing out of the cooling finger (105), wherein the conduit (130, 131) of the cooling finger (105) opens into the evaporator block (109);a heating element (119) for heating at least the evaporator block (109); anda temperature regulation device for regulating the temperature of the evaporator block (109), wherein a signal corresponding to the temperature of the evaporator block is delivered to the temperature regulation device and the temperature regulation device regulates the temperature of the evaporator block (109) to a definable value, wherein the temperature regulation device is operable to regulate the temperature of the sample stage (102) and the temperature of the evaporator block (109) with the heating element (119) to a temperature value at which condensation of atmospheric moisture from ambient air onto the sample is prevented after completion of the ion beam etching operation. 2. The cooling apparatus according to claim 1, further comprising a means (121) for controlling the coolant flow. 3. The cooling apparatus according to claim 2, wherein the means for controlling the coolant flow includes a coolant pump (121). 4. The cooling apparatus according claim 1, wherein the cooling finger (105) has an outer wall and an inner wall spaced from the outer wall. 5. The cooling apparatus according claim 1, further comprising:a mask holder device (103);a mask (104) held by the mask holder device (103), the sample being positionable relative to the mask; andat least one further thermal conduction element (106c) extending from the cooling finger (105) to the mask holder device (103). 6. The cooling apparatus according to claim 1, further comprising:a temperature measuring means (118) for measuring the temperature of the evaporator block (109). 7. The cooling apparatus according claim 1, further comprising at least one of:a temperature measuring means (122) for measuring the temperature of the sample stage (102); anda heating element (116) for heating the sample stage (102). 8. The cooling apparatus according to claim 7, wherein a signal corresponding to the temperature of the sample stage is delivered to the temperature regulation device and the temperature regulation device regulates the temperature of the sample stage (102) to a second definable value. 9. The cooling apparatus according to claim 8, further comprising a heating element (116) associated with the sample stage (102), wherein the temperature regulation device regulates the temperature of the sample stage (102) to the definable value via the heating element (116). 10. The cooling apparatus according to claim 1, further comprising a coolant pump (121) for controlling the coolant flow, wherein the temperature regulation device regulates the temperature of the evaporator block (109) to the definable value by adjusting a pump rotation speed of the coolant pump (121). 11. The cooling apparatus according to claim 8, further comprising a coolant pump (121) for controlling the coolant flow, wherein the temperature regulation device regulates the temperature of the sample stage (102) to the second definable value by adjusting a pump rotation speed of the coolant pump (121). 12. A method for adjusting temperature of a sample in an ion beam etching operation, comprising the steps of:(a) mounting a sample on a coolable sample stage (102) of an ion beam etching apparatus, the sample stage (102) having a cooling apparatus associated therewith comprising a coolant receptacle (120) containing a coolant, a cooling finger (105) connectable to the coolant receptacle (120) and including a conduit (130, 131) through which the coolant can flow when the cooling finger is connected to the coolant receptacle (120), and at least one thermal conduction element (106a, 106b) fabricated of a metal connecting the sample stage (102) to the coolant in the cooling finger (105) in heat-transmitting fashion such that heat is transferred from the sample stage (102) to the coolant;(b) aligning the sample on the sample stage (102); and(c) cooling the sample by directing the coolant through the conduit (130, 131) of the cooling finger (105) to establish a desired temperature for the ion beam etching operation, wherein the coolant is a liquefied gas and wherein the conduit (130, 131) of the cooling finger (105) opens into an evaporator block (109) for evaporating the coolant flowing out of the cooling finger (105), and wherein a flow rate of the coolant is adjusted so that the coolant flowing through the evaporator block (109) emerges from the evaporator block (109) in gaseous form. 13. The method according to claim 12, further comprising the step of:(d) terminating the cooling of the sample after termination of the ion beam etching operation. 14. The method according to claim 13, further comprising the step of:(e) heating the sample to a temperature at which condensation of atmospheric moisture from ambient air onto the sample is prevented after step (d). 15. The method according to claim 14, wherein the sample is aligned relative to a mask (104), the mask (104) being cooled and heated along with the cooling of the sample in step (c) and the heating of the sample in step (e). 16. The method according to claim 12, wherein the desired temperature is held constant during the ion beam etching operation. 17. The method according to claim 12, wherein the coolant is selected from the group of coolants consisting of liquid nitrogen and liquid air. 18. A cooling apparatus (101) for a sample in an ion beam etching process, comprising:a sample stage (102) for receiving the sample;a coolant receptacle (120) containing a coolant;a cooling finger (105) connectable to the coolant receptacle (120), the cooling finger (105) including a conduit (130, 131) through which the coolant can flow when the cooling finger is connected to the coolant receptacle (120);at least one thermal conduction element (106a, 106b) connecting the sample stage (102) to the coolant in the cooling finger (105) in heat-transmitting fashion such that heat is transferred from the sample stage (102) to the coolant;a mask holder device (103);a mask (104) held by the mask holder device (103), the sample being positionable relative to the mask; andat least one further thermal conduction element (106c) extending from the cooling finger (105) to the mask holder device (103), wherein the thermal conduction elements are fabricated of a metal that is in direct contact with the cooling finger. |
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043228534 | claims | 1. In a nuclear reactor incorporating a plurality of columns of tubular fuel elements disposed in horizontal tubes in a mass of graphite wherein water flows through the tubes to cool the fuel elements, the improvement comprising at least one control column disposed in a horizontal tube including fewer fuel elements than in a normal column of fuel elements and tubular control elements disposed at both ends of said control column, and means for varying the horizontal displacement of the control column comprising a winch at the upstream end of the control column and a cable extending through the fuel and control elements and attached to the element at the downstream end of the column. 2. The improvement of claim 1 wherein the number of fuel elements in the control column is approximately one-half the number in a normal fuel column and the number of poison elements is sufficient so that the control column is longer than a normal fuel column. 3. The improvement according to claim 2 and including means for limiting the movement of the control column. 4. The improvement according to claim 3 wherein said means for limiting the movement of the control column includes a micropotentiometer having a shaft locked to the shaft of the winch, a housing having a resistance wire integral therewith, a wiper arm which is attached to the shaft of the micropotentiometer and which traverses the resistance wire, stops on said resistance wire which limit the movement of the wiper arm, a bar connected to the center of the end of the housing, a pair of springs engaging opposite sides of the bar, and limit switches which are operated by said bar when said springs are depressed. |
abstract | The invention relates to a hybrid phase plate for use in a TEM. The phase plate according to the invention resembles a Boersch phase plate in which a Zernike phase plate is mounted. As a result the phase plate according to the invention resembles a Boersch phase plate for electrons scattered to such an extent that they pass outside the central structure (15) and resembles a Zernike phase plate for scattered electrons passing through the bore of the central structure. Comparing the phase plate of the invention with a Zernike phase plate is has the advantage that for electrons that are scattered over a large angle, no electrons are absorbed or scattered by a foil, resulting in a better high resolution performance of the TEM. Comparing the phase plate of the invention with a Boersch phase plate the demands for miniaturization of the central structure are less severe. |
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051204916 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention is generally related to nuclear reactor fuel assemblies and in particular to thimble tubes in the fuel assembly guide tubes. 2. General Background In commercial nuclear reactors, guide tubes provided in the fuel assembly have thimble tubes therein. These fuel assembly guide tubes and thimble tubes therein are spaced throughout the fuel assembly and are adapted to receive either control rods or neutron detectors. The neutron detectors are important to efficient reactor operation as they are inserted into the thimble tubes in the reactor and are used to provide flux maps of the core. The guide tubes and thimble tubes that receive the neutron detectors are part of the reactor coolant system pressure boundary and extend from a sealed table outside of the reactor vessel to the top of the core. Such instrumentation tubes normally extend through the bottom of the reactor vessel so as not to interfere with the reactor vessel cover and control rod drive mechanism. The thimble tubes are sealed at the end inside the guide tube to prevent loss of primary coolant through the thimble tube. In the reactor coolant system, coolant flows through flow holes in the lower core plate and flow holes in the lower part of the guide tubes into and through the guide tubes and exits at the top of the guide tubes. Due to their entry point into the reactor, these thimble tubes generally receive little or no lateral support along their length and are subject to flow induced vibration by the coolant during normal reactor operations. The vibrations tend to cause the thimble tubes to contact the fuel assembly guide tubes, which has resulted in through-wall wear or thinning of the thimble tubes and cracking of the fuel assembly guide tubes. Thinning of the thimble tube wall can lead to loss of integrity as a system pressure boundary and loss of reactor coolant and pressure. Cracking of the fuel assembly guide tube can lead to disruption of proper coolant flow and/or damage to the thimble tube. The addition of flow limiters has been unsuccessful as this has been reported to aggravate the problem. Plugging or repositioning the guide tubes is also not an ideal solution as this changes coolant flow or changes the positioning of neutron detectors in the core. Related patents which applicants are aware of include the following. U.S. Pat. No. 4,229,256 discloses the use of a thimble tube having a corrugated section at its lower end to distribute the control rod deceleration forces over an extended distance in a "scram" situation. U.S. Pat. No. 4,070,241 discloses the use of a removable radial shielding assembly for closing interassembly gaps in the reactor core assembly load plane. A flexible shielding assembly is provided with a loose fitting elongated insert in an axial opening that is bounded by upper and lower end walls. The insert is constructed of a material having a higher coefficient of thermal expansion than the shielding assembly and causes bowing thereof from pressure against the upper and lower end walls. U.S. Pat. No. 4,318,776 discloses thimble tubes that extend through the wall of the pressure vessel of a boiling water reactor and are sealed from the interior of the pressure vessel to allow exchanging of detectors in the thimble tubes during reactor operation. U.S. Pat. Nos. 3,664,924; 4,077,843; 4,295,935; 4,474,730; and 4,839,136 disclose the use of bimetallic spacer grids for nuclear fuel assemblies. As it can be seen, the known art does not address the problem of damage to guide tubes and thimble tubes by flow induced vibration of thimble tubes. SUMMARY OF THE INVENTION The present invention solves the aforementioned problem in a straightforward manner. What is provided is a thimble tube constructed from materials having different coefficients of thermal expansion. This causes the thimble tube, which is straight at room temperature, to warp into a wavy or spiral shape along its length when exposed to normal reactor operating temperatures. This results in multiple or continuous support points of the thimble tube within the guide tube. The tendency of the thimble tube to respond to fluid flow is thus reduced and wear of the thimble tube and cracking of the fuel assembly guide tube is inhibited. |
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claims | 1. A method for treating a nitric aqueous liquid effluent containing nitrates of metals or metalloids, comprising the steps of:adding a dilution adjuvant to the effluent; anda step for calcination of the effluent in order to convert the nitrates of metals or metalloids into oxides of said metals or metalloids, at least one compound selected from the nitrates of metals or metalloids and the other compounds of the effluent leading upon calcination to a tacky oxide,wherein the dilution adjuvant comprises at least one nitrate of metal or metalloid leading upon calcination to a non-tacky oxide, the dilution adjuvant being added to the effluent prior to the calcination step in order to give a mixture of effluent and dilution adjuvant, in which the mixture meets the two following inequations (1) (2): mass of sodium nitrate of the mixture expressed in terms of oxide Na 2 O mass of all the compounds of the mixture expressed in terms of oxides ≤ 0.3 ( 1 ) mass of all the compounds of the mixture leading upon their calcination to tacky oxides , expressed in terms of oxides mass of all the compounds of the mixture expressed in terms of oxides ≤ 0.35 . ( 2 ) 2. The method according to claim 1, wherein the dilution adjuvant comprises aluminium nitrate and optionally at least one other nitrate selected from iron nitrate and rare earth nitrates. 3. The method according to claim 2, wherein the dilution adjuvant comprises aluminium nitrate and optionally at least one other nitrate selected from iron nitrate, lanthanum nitrate, cerium nitrate, praseodymium nitrate and neodymium nitrate. 4. The method according to claim 1, wherein said at least one compound leading upon calcination to one or more tacky oxide(s), is (are) selected from sodium nitrate, phosphomolybdic acid, boron nitrate and mixtures thereof. 5. The method according to claim 1, wherein the content of nitrate(s) and of other compound(s) leading upon calcination to a tacky oxide, expressed as oxide, based on the total mass of the salts contained in the effluent, expressed as oxide, is greater than 35% by mass. 6. The method according to claim 5, wherein the effluent has a sodium nitrate content expressed as sodium oxide Na2O, based on the total mass of the salts contained in the effluent, expressed as oxide, greater than 30% by mass. 7. The method according to claim 1, wherein the calcination step is carried out at a temperature leading to a calcinate temperature at the outlet of an oven of about 400° C. 8. The method according to claim 1, wherein the calcination is carried out in a heated rotating tube. 9. The method according to claim 1, wherein after the calcination step, a vitrification step is carried out wherein the calcinate from the calcination step is melted with a glass frit whereby a confinement, containment glass is obtained. 10. The method according to claim 1, wherein the dilution adjuvant consists essentially of at least one nitrate of metal or metalloid leading upon calcination to a non-tacky oxide. |
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040070859 | claims | 1. An elongated insert for a nuclear reactor fuel element whereby the fuel element may be individually identified, the elongated insert bearing a first array of markings at positions spaced along the length of said insert to produce a series of signals on presentation of the elongated insert to a detector responsive to each marking, each marking of said first array being located at a position spaced lengthwise of said insert from other markings of said first array, and a second array of markings at positions spaced along the length of said insert to vary the series of signals produced by the first array of markings on presentation to the detector, the marking of said second array being located at only some of the positions of said first array of markings. 2. An insert as claimed in claim 1 in the form of a metal bar wherein all the markings are holes extending through the bar and the further markings are at right angles to the array at spaced positions. 3. An insert as claimed in claim 1 in the form of a metal bar wherein all the markings are grooves and the further markings are provided by deepened grooves. 4. A nuclear reactor fuel element comprising a nuclear fuel material within a protective sheath and including within the sheath an elongated insert for identifying said fuel element, wherein the insert bears a first array of markings at positions spaced along the length of said insert to produce a series of signals on presentation of the elongated insert to a detector responsive to each marking, each marking of said first array being located at a position spaced lengthwise of said insert from other markings of said first array, and a second array of markings at positions spaced along the length of said insert to vary the series of signals produced by the first array of markings on presentation to the detector, the marking of said second array being located at only some of the positions of said first array of markings. |
abstract | A method of cooling a nuclear reactor core is disclosed. The method includes contacting the nuclear reactor core with an aqueous solution comprising at least one of polyhedral boron hydride anions or carborane anions. Nuclear reactors are also disclosed. The nuclear reactor has a neutron moderator that is an aqueous solution comprising at least one of polyhedral boron hydride anions or carborane anions, or the nuclear reactor has an emergency core cooling system including a vessel containing a volume of an aqueous solution comprising at least one of polyhedral boron hydride anions or carborane anions. The nuclear reactor can also have both an aqueous solution comprising at least one of polyhedral boron hydride anions or carborane anions as a neutron moderator and an emergency core cooling system that includes an aqueous solution comprising at least one of polyhedral boron hydride anions or carborane anions. |
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abstract | The present invention is related to an apparatus and method for irradiating product packages (1). One measures the effective dimension of a product package (1), one processes a product package (1) having everywhere an effective thickness below a predefined threshold with an e-beam source (17), and other packages with either a gas sterilisant device or an X-ray or gamma source. Packages for treating with the X-ray or gamma source are grouped in layers and stacked, in order to optimise the throughput of the apparatus. |
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abstract | The present invention relates to a method of using an atomic force microscope comprising exciting natural lower and higher vibration modes of a microlever (M) placed on a sample, and analyzing the variation of one variable of a first output signal (Ai cos(ωit−φi)) representative of the response of M to the excitation of the lower mode, with respect to the variation of a parameter influenced by one variable of a second output signal (Aj cos(ωjt−φj)) representative of the response of M to the excitation of the higher mode, and/or analyzing the variation of one variable of a second output signal (Aj cos(ωjt−φj)) representative of the response of M to the excitation of the higher mode, with respect to the variation of a parameter influenced by one variable of a first output signal (Ai cos(ωit−φi)) representative of the response of M to the excitation of the lower mode. |
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043449115 | description | DETAILED DESCRIPTION OF THE BEST MODE CONTEMPLATED The present invention provides for the protection of the first or inner wall of an inertial confinement fusion chamber from neutrons, x-rays, plasma, charged particles, debris, etc. generated by the implosion of a fusion target within the chamber. This protection is provided by the use of a fluidized wall or waterfall within the chamber in the form of liquid lithium or of solid pellets of lithium-ceramic. The waterfall contains sufficient neutron moderating material to degrade the fusion neutron spectrum to the point where neutron damage levels in structural materials are sufficiently low to allow use of smaller blanket structures which could last the lifetime of the fusion chamber (.about.30 years). Thus, the fluidized wall approach of this invention has the capability of coping with the problems of 14 MeV neutron damage and cyclical stresses. The principle feature of the invention is a thick continuously recyclable first-wall of fluidized material such as a lithium waterfall. On each shot the waterfall is in effect disassembled; between each shot it is reestablished. The lithium (liquid or pellets) is continuously pumped to the top of the evacuated fusion chamber through a reservoir region which separates the first-wall from the pressure vessel, and the waterfall is spaced from the surface of the first-wall to eliminate any shock effects produced by energy from the exploding target stricking the lithium wall. A small fraction of the lithium in the waterfall and reservoir regions is circulated to a heat exchanger which in turn transfers heat to a steam generation cycle. The lithium waterfall thus serves as the primary coolant, neutron moderator, and the fertile material for tritium breeding. Two of the major reactor vessel problems affecting the technical feasibility of a laser fusion power plant are, as pointed out above: 1. The effects of high-energy neutrons and cyclical stresses on the blanket structure. 2. The effects of x-rays and debris from the fusion microexplosion on the first-wall of the fusion chamber. The reactor concepts that have been developed (and are being assessed in terms of factors including reactor size, power density, first-wall and blanket lifetimes, duty cycle, fabrication costs, stored energy requirements, and recirculating power fractions) include: 1. A liquid-lithium-cooled stainless steel manifold. 2. A gas-cooled graphite manifold. 3. Fluidized wall of this invention, such as a liquid lithium "waterfall", and a ceramic-lithium pellet "waterfall". For a more complete description of these first two reactor concepts, though briefly described hereinafter, reference is made to report UCRL-79654 by W. R. Meier et al entitled "Reactor Concepts for Laser Fusion", dated July 1977, prepared for submission to the proceedings of the American Institute of Chemical Engineers, Nov. 13-17, 1977, New York, N.Y. Prior to describing the fluidized wall (liquid or pellet) some background information on fusion reactor technology and the design parameters and constraints that are common to all of the reactor systems will be set forth. This discussion of reactor technology deals primarily with the functions that a blanket system is required to perform and the problems associated with performing these functions in the hostile environment created by the fusion microexplosion. All of the reactor concepts here considered are based on an inertially confined deuterium-tritium fusion reaction. We have selected thermonuclear yields ranging from 400 to 4000 MJ and pulse repetition rates from 1 to 10 Hz for the basis of discussion. Selected combinations of these parameters result in reactor systems that produce 400 to 4000 MW of thermal power and 120 to 1500 MW of electrical power with net efficiencies ranging from 30 to 40%. The variance in net efficiency results from the different thermal efficiencies and recirculating power requirements of the various concepts. The selected parameter space for fusion neutron flux at the first wall ranges from 1 to 10 MW/m.sup.2, resulting in first wall radii ranging from 1.5 to 15 m. The effects of neutrons, x-rays, and debris from the thermonuclear microexplosion represent the primary technical concerns that must be dealt within laser-fusion reactor concepts. All of our reactor concepts considered here employ large focal length optics to mitigate the damaging effects to the final focusing elements. At the focal length of 10 m the final optics would survive the microexplosions, but may have to be replaced at relatively short intervals. At longer focal lengths (.about.100) the damaging effects may be greatly reduced, thus assuring the survival of the final focusing elements for intervals that are long enough not to affect adversely the plant capacity factor. High-energy neutrons also damage and activate most structural materials. The large amounts of radioactive waste thus generated represent a maintenance and disposal problem that is common to all types of DT fusion systems. X-ray and debris damage to first-wall materials is a problem primarily associated with inertial confinement fusion systems. Several different approaches to the first-wall problem have been discussed in the literature, including use of a dry wall, wetted wall, and magnetically protected wall as indicated above. These approaches differ primarily in the way in which the inner surface of the first wall interacts with the x-rays and microexplosion debris. In the dry wall approach a sacrificial metal or ceramic liner is placed between the fusion chamber and the blanket. The wetted wall concepts feature a thin layer of liquid metal that covers the metal wall and protects it from the blistering and structural ablation that would otherwise occur. The magnetic protection concept uses a solenoid to divert the pellet debris away from the sides of a cylindrical blanket and into conical collectors at the top and bottom. The fluidized wall approach of this invention is a promising approach which has been developed. In this approach, the first structural wall is shielded from x-rays, neutrons, plasma and shock effects by a thick falling region of lithium in liquid or solid pellet form. The fall (fluidized wall) will contain enough moderating material to degrade the fusion neutron spectrum to the point where neutron damage levels in structural materials are low enough to allow us to consider smaller blanket structures which could last for the useful lifetime of the plant (.about.30 years). The blanket system must perform several functions while coping with the hostile environment created by the fusion microexplosion. It must: 1. Convert the fusion energy into thermal energy. 2. Provide for efficient removal of the thermal energy. 3. Breed enough tritium to replace that which was burned in the fusion reaction. 4. Maintain the required vacuum in the fusion chamber. Sixty-five to seventy-five percent of the fusion energy is in the form of high-energy neutrons. Therefore, a neutron-moderating material is required to convert kinetic energy to thermal energy. In general, elements with low atomic numbers and high scattering cross sections are effective moderators; water, hydrides, beryllium, and graphite are common examples. Although somewhat less effective, lithium can also be considered as a neutron-moderating material. Because there is no significant natural supply of tritium, as pointed out above, a DT fusion reactor must breed its own tritium. Several neutron reactions produce tritium, but the only tritium-producing reactions with high enough cross sections to be useful are those involving lithium: EQU .sup.6 Li+n.fwdarw..sup.4 He+T, EQU .sup.7 Li+n.fwdarw..sup.4 He+T+n. The .sup.7 Li reaction has a threshold of approximately 4 MeV and a much lower cross section than the .sup.6 Li reaction; nevertheless, it is very important because it produces a T atom without depleting the neutron population. If the neutrons are moderated before reaching the fertile lithium, the .sup.7 Li reaction is not utilized (since it requires a high-energy neutron) and any lost neutrons would result in a tritium-breeding ratio less than 1.0. In such cases, the blanket may also require some sort of neutron multiplier to maintain an adequate breeding ratio. Beryllium and lead with high (n, 2n) and low capture cross sections are examples of good neutron multipliers. However, beryllium is an example of a limited resource material whose use could significantly reduce fusion's potential as a long-range source of energy. The vacuum requirements in the fusion chamber are primarily determined by considerations of laser beam propagation and damage to the injected fuel pellet. If the DT fuel can be incorporated into the pellet in a noncryogenic or insulated form, and hence be less subject to heat damage, laser beam propagation will be the primary factor determining the vacuum requirements in the fusion chamber. Our computer calculation results indicated that beam defocusing and attenuation of 1 .mu.m light by cascade breakdown and/or thermal blooming can be reduced to acceptable levels with fusion chamber pressures of 0.1 torr or less. The pumping requirements needed to maintain this vacuum will depend on the material vaporized and the type of pump used. For a lithium wetted first-wall concept, the 0.1-torr vacuum can be maintained under the worst conditions with a pump that requires about 2% of the gross electrical power and approximately 10% of the surface area (the worst conditions resulting when all the debris and x-ray energy is used to vaporize lithium). In the above-mentioned liquid-lithium-cooled stainless steel manifold approach, the microexplosion is surrounded by a cylindrical annulus of stainless steel into which vertical coolant channels have been drilled to form a manifold. Liquid lithium flows down through these channels and is recirculated to the top through a bulk coolant region, which separates the annular manifold from an outer pressure vessel. Liquid lithium serves as the primary coolant, as a neutron moderator, and as the fertile material. The stainless steel manifold concept is compatible with either the dry or wetted first-wall approach. In the dry-wall approach a graphite liner would be supported by the stainless steel and cooled by liquid lithium; while in the wet-wall approach a thin (3 mm) film of liquid lithium on the inner surface of the manifold would be used. However, tritium-breeding considerations limit the thickness of a structural wall of solid stainless steel to 10 cm or less, and which would require internal cooling. Computer calculations indicate cyclical stresses and neutron damage will limit the lifetime of the liquid manifold to a few full-power years. The above-mentioned gas-cooled graphite manifold approach is similar to the stainless steel manifold concept except that the vertical coolant channels are drilled into an array of graphite blocks that make up the fusion chamber. The vacuum vessel is an outer shell of reinforced, prestressed concrete. High-pressure helium gas is pumped through the coolant channels, some or all of which are filled with pellets of a lithium-bearing ceramic. Tritium is removed from these channels by the gas coolant as it diffuses out of the lithium compound in which it is bred. This approach exhibits low activation and low tritium inventories. Moreover, the possibility of an accident occurring that could release radioactivity to the environment is greatly reduced because the lithium is present in a solid less reactive form (Li.sub.2 O, LiAl.sub.2 O.sub.3). The graphite moderates the neutrons below activation energy levels. It also moderates the neutrons to energies below the threshold for the .sup.7 Li tritium-producing reaction. This makes it advantageous to enrich the lithium in .sup.6 Li, thereby reducing the required lithium and tritium inventories. However, without the tritium-breeding contribution from .sup.7 Li, a neutron multiplier, such as beryllium, may be required to maintain a tritium-breeding ratio greater than 1. The use of beryllium to multiply neutrons and enhance tritium-breeding presents a problem in terms of beryllium's toxicity and relative scarcity. The use of a gas coolant will allow high operating temperatures and result in high thermal conversion efficiencies. Large amounts of pumping power will be required for cooling the system and purging the tritium from the pellet-filled channels. Finally, the structural integrity of the graphite chamber in the microexplosion environment may be inadequate. An embodiment of the fluidized wall approach of this invention is illustrated in FIG. 1 wherein the blanket is formed with the implosion chamber by recirculating lithium as indicated by the arrows, and as mentioned above, a portion of the heated lithium is passed through a heat exchanger arrangement for operation of an associated steam cycle system, as known in the art. Note that the lithium blanket is spaced from the wall surface of the chamber. While FIG. 1 illustrates a liquid lithium "waterfall" embodiment of the invention, described in greater detail hereinafter, a ceramic-lithium pellet "waterfall" embodiment also constitutes part of the invention fluidized wall approach to solving the above-identified two major reactor vessel problems. It has been demonstrated in the nuclear fission industry that once scientific feasibility has been achieved, the materials development program paces the demonstration of technical and economic feasibility. Fluidized wall concepts will be less dependent on materials development because radiation damage is significantly reduced. The analysis of the liquid lithium system is further facilitated by the availability of data on the properties of liquid lithium and the existence of liquid-metal experimental facilities built in support of the liquid-metal fast breeder reactor (LMFBR) program. Referring now to the embodiment of the invention illustrated in FIGS. 2 and 3 and basically comprises a cylindrical pressure vessel or fusion chamber 10 having therein an annular stainless steel liner 11 forming therebetween a space, annulus, or cavity 12. A fuel pellet injector and vortex generator, not shown but similar to FIG. 1, is attached to the top of vessel 10 while a laser transport tube and vacuum pump ports 13 and 14 are mounted in the upper and lower ends of vessel 10. A mechanical transport mechanism 15, driven by means not shown, located at the lower end of vessel 10 forces ceramic-lithium balls or pellets 16 upward through space 12 and another mechanical transport mechanism 17, driven by means not shown, directs the balls 16 into vessel chamber 18 forming a blanket 19 of falling ceramic-lithium balls along the cylindrical wall surfaces of the vessel, these falling balls are collected at the bottom of chamber 18 and are recirculated by mechanism 15 as indicated at 16', the arrows indicating the course of circulation of the balls 16. The principal feature of the ceramic-lithium pellet waterfall (FIGS. 2 and 3) is a thick layer of falling solid ceramic-lithium pellets that shields the first structural wall from the microexplosion. The pellets are continuously recirculated to the top of the vacuum chamber through a reservoir region between the first wall and the pressure vessel. The pellets are either transported through heat exchangers, as in FIG. 1, or cooled by the flow of high-pressure helium gas in the reservoir region. Tritium is bred in the ceramic lithium compound and recovered as it diffuses out. Preliminary calculations indicate a tritium breeding ratio greater than 1 can easily be achieved. The thick region of falling pellets will moderate and absorb neutrons before they reach the first structural wall, and this will result in a significant reduction in the degree of first-wall damage and possibly the amount of radioactive waste produced by neutron activation. The use of lithium in a ceramic form is an important feature of this concept in that it eliminates the corrosive problems of liquid lithium and significantly reduces the associated chemical hazard. Major questions, such as tritium diffusion from the pellets and structural integrity of the ceramic compound, cannot be answered satisfactorily with existing data. More information may be forthcoming from the University of Wisconsin study, which uses Li.sub.2 O as a blanket and heat-transport material (see R. W. Conn et al, "Studies of the Technological Problems of Laser Driven Fusion Reactors, Annual Report-I", University of Wisconsin, Report UWFDM-190, December 1976). A means of most efficiently transporting the pellets, particularly into and out of the vacuum chamber, is another area of required development, but within the current state of the art. The liquid lithium "waterfall" concept (FIGS. 1, 4 and 5) has emerged as an extremely promising reactor concept for a laser-fusion power plant. It features a thick continuous fall of liquid lithium that protects the first structural wall, allowing it to last for the useful life of the plant. Besides moderating neutrons the fall (waterfall) also absorbs the photons (x-rays and reflected laser light), pellet debris (alpha particles, unburnt fuel, and other pellet material) and any shock wave emitted by the microexplosion. By keeping the fall off of the chamber wall this shock wave is not directly transmitted to the structural wall. The majority of the fusion energy is thus deposited in the liquid lithium, which also serves as the primary coolant and fertile material for tritium breeding. The embodiment of the liquid lithium "waterfall" illustrated in FIGS. 4 and 5 is generally similar to that of the ceramic-lithium ball embodiment of FIGS. 2 and 3 and are given similar reference numerals. The mechanical transport mechanism at the bottom of the vessel in FIG. 2 is replaced with a plurality of recirculation pumps 20 (only one shown), with the upper transport mechanism omitted. Also, the space or cavity 12 is extended to the top of vessel 10 as indicated at 12' by another stainless steel liner 21 supported by members 22 which is separated from liner 11 to form a first annular opening 23, and spaced from fuel pellet injection and vortex generating mechanism, not shown, but attached to laser transport tube 13', to form a second annular opening 24. As illustrated by the arrows, the liquid lithium, indicated at 25 in a reservoir 26 (see FIG. 5), is recirculated by pump 20 upward through spaces 11 and 11' which forms a "waterfall" or blanket 27 of falling liquid lithium along the wall of the vessel and a thin sheet 28 of lithium from the top of the vessel 10, as described hereinafter. It should be pointed out that fluids other than liquid lithium could be used to perform the neutron moderating function of the fall (waterfall). The primary constraints on the fall material are that the substance must: 1. Have a reasonably low melting point (less than about 200.degree. C.) so the fluid state can be effectively maintained. PA0 2. Have a low enough vapor pressure at the selected operating temperature (>400.degree. C. but as high as possible) to permit an adequate vacuum condition to be maintained. PA0 3. Have neutronic characteristics that permit an adequate tritium breeding ratio to be achieved. PA0 S=sticking coefficient PA0 T=vapor temperature, .degree.K. PA0 V=vacuum chamber volume, m.sup.3 PA0 t=time, sec Tritium breeding considerations preclude the use of a neutron absorber and require that lithium be incorporated in the reactor system in a suitable manner. One possibility is to use lead, which effectively degrades the high-energy neutron spectrum through (n, 2n) and inelastic scatterings, as the primary constituent of the fall. A few volume percent of .sup.6 Li in the Pb fall would be enough to maintain a tritium breeding ratio greater than 1. Alternatively, a small vol. % of lead could be added to the lithium fall to moderate more efficiently the neutron spectrum. The use of Pb-Li alloys would allow a lower system tritium inventory to be maintained. On the negative side, recirculation pumping power would increase and grain boundary corrosion of steels may present compatibility problems. The various aspects of such a system are being investigated. For the present, however, the discussion will be confined to a natural liquid lithium fall. Laser fusion reactors have a flexibility of geometry that is not available in magnetic confinement reactors. While a point source of energy is more effectively utilized in spherical geometry, we have selected a cylindrical geometry for several reasons. A vortex generator (see FIG. 1) injects a sheet of lithium (similar to sheet 28 in FIG. 4) to protect the top of the reactor (primarily from x-rays and debris). This sheet is thinner than the waterfall (see FIG. 4) and does not provide the same degree of protection from neutron damage. Therefore, it is advantageous to have the top of the chamber farther from the microexplosion than are the sidewalls. The spherical end cap on the cylindrical chamber effectively accomplishes this. At the bottom of the vacuum vessel the lithium is in direct contact with the chamber walls. Thus, shock wave can be transmitted directly to the structural components at this point. By moving the bottom region farther away and decreasing the surface area of the lithium pool at the bottom, the magnitude of this effect is reduced. Also, as will be discussed, the fall must be injected downward thus forming a cylindrical sheet. The tips of the laser beam tubes (see FIG. 1) must penetrate the waterfall and be directly exposed to the microexplosion at a distance equal to about 1/2 the chamber wall radius. Fortunately, these high damage areas represent only a minute fraction of the total surface area (a few hundredths of a percent) and sophisticated measures and/or special materials could be used to protect them. Alternatively, sacrificial tube ends could be used and remotely fed inward as the tips slowly vaporize. As previously stated, the fall will contain enough lithium to significantly degrade the neutron spectrum. Neutron damage levels in structural materials can thus be reduced by more than an order of magnitude which will permit structural members to survive for the life of the plant at first wall loadings greater than 1 MW/m.sup.2. (Wall loading will be consistently quoted here as the neutron energy flux at the first structural wall as if the fusion neutrons were only attenuated geometrically and not by any other material such as Li). The primary neutron damage mechanisms are atomic displacements and gas production (primarily helium). Displacement damage is expressed as displacements per atom (dpa) and gas production is expressed as atom-parts-per-million (appm). The damage limits for 316-SS (stainless steel) at an operating temperature of 500.degree. C. are estimated to be 150 dpa and 500 appm helium. For an unprotected first wall of 316-SS, the displacement damage rate is .about.10 dpa per full power year, and the helium production rate is .about.220 appm per full power year at a neutronic wall loading of 1 MW/m.sup.2. The damage limits for He production would thus be reached in only 2.3 years at this wall loading. As seen in FIG. 6, the allowable first-wall fluence increases exponentially with lithium thickness. Note that 40 cm of lithium is required to reduce helium production to the point where the first structural wall could last for 30 years at 1 MW/m.sup.2 (at 70% capacity factor). Displacement damage is less restrictive. We have selected a minimum thickness of 50 cm for our reference design but are considering even larger thicknesses. We have also taken advantage of the fact that the emitted 14 MeV neutrons are attenuated by compressed DT fuel; an advantageous effect unique to inertial confinement fusion. A compressed density-radius product (.rho.R) equal to 3.0 was assumed. In FIG. 6, 50 cm of Li plus the compressed target is roughly equivalent to 63 cm of lithium. This gives an allowable fluence of 60 MW-YR/m.sup.2 or about 3.0 MW/m.sup.2 for the 30-year plant life at 70% capacity factor. In the process of attenuating neutrons and interacting with the microexplosion plasma the lithium fall absorbs a large fraction of the total nuclear energy deposited in the reactor. FIG. 7 presents the cumulative energy deposition through the lithium fall, and blanket region as a function of the fraction of the total energy deposited. Again a compressed pellet .rho.R of 3.0 has been assumed resulting in a neutron energy deposition of .about.2 MeV in the pellet itself. This along with the 3.5 MeV alpha energy accounts for 32% of the total energy deposited. All of this energy, whether in the form of x-rays, alpha particles, or other energetic debris will be deposited essentially at the surface of the lithium fall. The compressed target also has an advantage in terms of neutron energy multiplication. The high-energy fusion neutrons undergo (n, 2n) reactions with both D and T resulting in an increase of about 10% in the neutron population through a mean-free-path of DT. The multiplied lower energy neutron spectrum results in a larger number of exoergic .sup.6 Li(n,.alpha.)T reactions and a smaller number of endoergic .sup.7 Li(n, n'.alpha.)T reactions than a 14 MeV neutron spectrum would. A total of 17.0 MeV of energy (per fusion reactions) is deposited in the system as a result of neutron interaction with the DT and blanket materials. This deposit represents a neutron energy multiplication factor of 1.2 compared to about 1.1 for the 14 MeV source. The total energy deposited in the reactor is 20.5 MeV per fusion reaction. As indicated, 80% of this energy is deposited in the 50-cm-thick lithium fall. Because lithium is the primary coolant, the system does not have to rely on conduction of heat through structural materials to remove the reactor energy. Lithium is in fact an excellent coolant with a specific heat capacity equal to that of water and three times better than that of sodium. In addition, its low density of 0.5 g/cm.sup.3 is advantageous in terms of pumping power considerations. A major advantage of absorbing all the plasma energy and much of the neutron energy is that the cyclical thermal stresses in the structural walls are essentially eliminated. An alternative scheme proposed for the liquid lithium waterfall would be to make the fall thick enough (>80 cm) to absorb over 90% of the total energy (see H. I. Avii et al, "The Effects of a Liquid ISSEC on Radiation Damage Parameters in Laser Fusion First Walls", University of Wisconsin, Report UWFDM, April 1977). In this way the recirculating reservoir region would not be required for tritium breeding or energy removal and the structural first wall could be independently cooled at a lower temperature. A decrease in the wall temperature would significantly relax radiation damage limits for dpa and appm He. The first wall could then be operated at a higher wall loading and thus increase the power density of the reactor system. Naturally, the higher the power density the smaller the reactor vessel size will be for a given power system, thereby reducing the capital cost of the reactor. As on might suspect, the liquid lithium waterfall concept has excellent tritium breeding characteristics. With no structural material between the fusion neutrons and the lithium fall, the design takes full advantage of the high-energy .sup.7 Li(n,n'T) reaction. Table I shows the distribution of tritium breeding from .sup.6 Li and .sup.7 Li reactions in a 50-cm lithium fall and recirculating lithium blanket region. A one-dimensional spherical model was used in these Monte Carlo calculations. TABLE I ______________________________________ TRITIUM BREEDING PERFORMANCE ______________________________________ Lithium fall T.sub.6 = 0.44 T.sub.7 = 0.56 T.sub.FALL = 1.00 Recirculating lithium T.sub.6 = 0.62 region T.sub.7 = 0.11 T.sub.RECIRC = 0.73 T.sub.TOTAL = 1.73 ______________________________________ .cndot. 50 cm natural lithium fall .cndot.100 cm natural lithium recirculating region .cndot. Target .rho.R = 3.0 gm/cm.sup.2 A compressed target .rho.R of 3.0 gm/cm.sup.2 was assumed. A tritium breeding ratio of 1.0 is obtained in the fall alone, and the total tritium breeding ratio is 1.7. The excess tritium produced in this reactor could in fact supply fuel for other laser fusion applications where tritium breeding is difficult or impractical. Two examples of such applications are radiolytic hydrogen production where it is desirable to deposit the neutron energy in stream rather than lithium blankets and propulsion applications where weight and volume considerations are extremely important. As previously stated, the vacuum condition required for laser beam propagation is on the order of 0.1 torr. FIG. 8 shows that the vapor pressure of lithium is orders of magnitude less than 0.1 torr at reasonably high temperatures. Corrosion considerations require that, for use with stainless steel, lithium temperatures must be limited to less than 500.degree. C. The vapor pressure at this temperature is less than 5.times.10.sup.-3 torr. Each microexplosion will vaporize a certain amount of lithium thus increasing the chamber pressure above the required 0.1-torr vacuum condition. The amount of lithium vaporized and the resulting chamber pressure will depend on the initial system conditions and in what way the fusion energy could conceivably be used to heat liquid lithium, vaporize lithium, and heat lithium vapor. If the chamber pressure equals the liquid vapor pressure prior to the microexplosion, even deposited neutron energy could vaporize lithium. The debris and x-ray energy, which is deposited over a very short range at the fluid surface, is expected to effectively blow off lithium vapor. After the initial transient events, a certain amount of lithium vapor will exist in the chamber. The resulting quasi-equilibrium pressure will most certainly be higher than the required vacuum condition of 0.1 torr and must therefore be reduced prior to the next microexplosion. There is so much liquid lithium in the chamber at the time of the microexplosion that the mixed-mean temperature rise of the fall per pulse is quite small. The vapor will therefore be in a supersaturated or superheated condition and proceed to recondense on the liquid lithium in the chamber. In effect, the liquid lithium waterfall acts as a condensing vacuum pump for the chamber. The effectiveness of the lithium liquid in condensing the vaporized lithium will depend on the condition of the fall (bulk temperature, surface temperature, whether or not it is disassembled) and on the condition of the vapor (temperature, pressure) shortly after the microexplosion. The liquid fall conditions are important for determining the vapor pressure of the liquid and the sticking coefficient, defined as the probability that a gas molecule incident on the liquid surface will stick. If the sticking coefficient is great than .about.0.5 (which is almost certainly the case for a liquid metal) the vaporized lithium will be driven by a pressure gradient to the condensing liquid surface at the local sonic velocity. Assuming adiabatic, frictionless flow of an ideal monatomic gas, the pressure decays according to ##EQU1## where A=condensing surface area, m.sup.2 P.sub.0 is the quasi-equilibrium pressure immediately after the microexplosion but before recondensation begins. FIG. 9 shows the pressure decay as vaporized lithium is recondensed by the fall. These curves represent worst-case calculations in which all the fusion energy (600 MJ per microexplosion) is used to vaporize lithium at 700.degree. K. While an increase in vapor temperature would result in a higher value of P.sub.0, the sonic velocity of the gas would also be higher and the gas would condense more quickly. The sticking coefficient of 0.5 is pessimistically low for a liquid metal vapor. Also, if the fall is disassembled by the neutron energy deposition, the condensing surface area will be much larger than the original surface area of the fall assumed here. Even with these assumptions, the vacuum conditions return to the required 0.1 torr in .about.0.1 sec for the 5-m chamber and in .about.0.12 sec for the 10-m chamber. Flow considerations will limit operation to a pulse repetition rate of a few times per second. Thus while the exact time-dependent chamber conditions have not been determined, this analysis indicates that vacuum conditions can be maintained and are in fact aided by the presence of the liquid lithium fall. We now turn attention to what is required in terms of pumping power and flow rates to maintain the thick fluidized wall of lithium. FIG. 10 shows the model and constraints used to calculate the pumping power required to recirculate the lithium fall. The fall protects the cylindrical portion of the chamber wall, which has a height-to-diameter ratio of 1. The flow inlet forms an annulus of thickness .DELTA.R.sub.0 with the inner edge of the ring at one-half the chamber wall radius R.sub.w. The fall is injected vertically downward with an inlet velocity V.sub.0. The pumping power is then estimated on the basis of the kinetic and static head requirements. EQU P.P.=(1/2V.sub.0.sup.2 +gH.sub.0).rho.V.sub.0 A.sub.0 where: EQU H.sub.0 =fall height=2R.sub.w ##EQU2## Preliminary calculations indicate that the fall will be disassembled by the microexplosion. If this is the case, the inlet velocity must be sufficient to allow the fall to reestablish itself prior to the next microexplosion. A clearing ratio of unity should be adequate. The first constraint is therefore: EQU H.sub.0 =2R.sub.w .ltoreq.V.sub.0 .tau.+1/2g.tau..sup.2, EQU 1/.tau.=pulse repetition rate. As previously noted, 50 cm of lithium is thick enough to provide adequate protection to the first structural wall up to 3.0 MW/m.sup.2. The second constraint is therefore that source neutrons must be attenuated by at least 50 cm of lithium at any point through the fall. Flow continuity requires that the thickness of the fall decrease as the fluid is accelerated by gravity. The minimum path length for neutrons actually occurs slightly below the horizontal midplane, but it is a very shallow minimum. The constraint on minimum effective thickness has, therefore, been taken at the midplane to simplify the calculations. The second constraint is: EQU .DELTA.R.sub.m .gtoreq.50 cm. The third constraint is that the inlet thickness must be less than one-half the chamber wall radius. This is a constraint of our selected geometry. EQU .DELTA.R.sub.0 .ltoreq.1/2R.sub.w For small chambers the clearing ratio constraint is less important than the midplane thickness and total inlet thickness constraints. For example, a 3-m radius chamber requires an inlet velocity of only .about.1 m/sec. At this velocity, however, the midplane thickness is only 13% of the inlet thickness which would thus have to be over 3.8 m thick to provide 50 cm of protection at the midplane. Because this is larger than 1/2 the chamber radius, a higher inlet velocity must be used in calculating the pumping power. The pumping power required to recirculate the liquid lithium waterfall is given in FIG. 11 as a function of vacuum chamber radius. (The gross power at pulse repetition rates of 1 Hz and 2 Hz is .about.270 MW.sub.e and 540 MW.sub..e for a pellet gain, Q=600, respectively). As indicated, at 1 Hz the pumping power is less than 5 percent of the gross power up to a chamber radius of .about.7 m. At the higher repetition rate the inlet velocity required to cover the length of the chamber in the 1/2 second between microexplosions is quite large. In this case, the fraction of the gross power used becomes substantial and, in fact, prohibitive for large chambers. The use of liquid lead with a low concentration of Li results in an increase in the pumping power due to the .about.20-fold increase in density. Lower thickness requirements with the Pb-Li alloy somewhat reduce this disadvantage. Depending on the specific case, pumping powers are thus a factor of 5-10 greater. The possible advantages of using multiple falls (waterfalls) entering the chamber at different vertical positions have also been considered. The primary advantage would be a reduction in the velocity head required to obtain a clearing ratio of 1. Each fall would be required to reestablish only to the inlet of the next lower fall. This is especially important at higher repetition rates. A second advantage is that the static head requirement is reduced for the fraction of the flow delivered to the lower falls. Also, if the injected fall should tend to break up into separate streams instead of forming a continuous curtain, additional lower falls could replenish the primary fall. A similar calculation model was used for a system with two falls, one inletting at the top of the wall surface as before and one at the midplane of the wall surface. The constraints on the fall are the same as for the single-fall case, with the actual positions of minimum effective thickness used to determine the required inlet thicknesses. As before, the inlet thickness (now equal to the sum of .DELTA.R.sub.10 and .DELTA.R.sub.20) must be less than 1/2 R.sub.w. The pumping power required to recirculate the double fall is shown in FIG. 12. Note particularly the substantial reduction in pumping power for the larger chambers at 2 Hz (50 MW.sub.e compared to 130 MW.sub.e for the 8 m chamber). The advantages of reduced velocity and static heads have been offset by larger flow area requirements for the smaller chambers at a repetition rate of 1 Hz. It is pointed out that the double-fall geometry is not by any means optimal. It is presented only to illustrate the possible advantages of multiple falls. Also note that a theoretical pumping power has been calculated that does not include the efficiency of the pump or drive motor. These factors will depend on the specific design, but for large axial flow pumps the combined efficiency could be .about.80%. The above evaluation of the fluidized wall, and particularly the liquid lithium waterfall approach of this invention has shown that the protection afforded by the thick fluidized curtain of lithium will allow first-wall and blanket structures to retain their integrity for the life of the system. Since the lithium waterfall preferably stands away from the first-wall a certain distance, the shock in the waterfall material which is initiated by the target debris striking said material will not be directly transmitted to the first-wall. The use of a waterfall reduces the neutron damage to the exposed structures to the point where the target or fusion chamber will have a life of .about.30 years for wall loadings .gtorsim.3 Megawatt/m.sup.2, where by contrast, an unprotected structure of stainless steel will last only about 3 years at a loading of 1 Megawatt/m.sup.2. Thus, with existing technology, the present invention provides an effective means for protection of the first or inner wall of inertial confinement fusion chambers from the microexplosion debris, x-rays, neutrons, etc. produced by the implosion of fusion fuel targets within such chamber. With the technology to protect the chamber, as provided by the present invention, efforts for the development of an overall inertial fusion power plant has been substantially advanced by the use of a blanket which involves a fluidized wall, particularly in the form of a lithium waterfall of either liquid lithium, along or with selected additives, or solid pellets of ceramic-lithium. The present invention thus provides a fluidized wall of at least lithium (liquid or solid pellets) within a target or fusion chamber of an inertial confinement system which: (1) protects the exposed first wall from the microexplosion products, (2) moderates the neutrons kinetic energy and converts it to useful thermal energy, and (3) provide a source of tritium, which is bred by reaction of neutrons with the lithium. While use of the present invention in a fusion power plant will be in the near future, the invention has utility in its use in the production of high-energy neutrons, x-rays, etc., for currently known applications, such as in the field of neutron crystallography, means of achieving crystal dislocations, initiation of some action such as a switch or random number generator upon receipt of a neutron pulse by a detector, calibration of diagnostics for other apparatus, fluor studies and as a source of strong shock waves for high pressure testing. While particular embodiments of the invention have been illustrated or described, modifications will become apparent to those skilled in the art, and it is intended to cover in the appended claims all such modifications that come within the spirit and scope of the invention. |
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043200281 | abstract | In a system and method for disposal of nuclear waste, the waste is mixed with glass components, melted, drawn into fibers and formed into a cable, the diameter of the fibers and the arrangement of the fibers in the cable being such that the cable can be wound onto a support. The cable is then fed through an underground conduit to buffer storage and eventually to a storage space deep beneath the surface of the earth, where it is wound onto a support for extended storage. The cable is equipped with a monitoring system based on optical glass fibers and a leader with which it can be retrieved from the extended-storage space in the event that the integrity of the storage space is threatened or breached. The cable can also be withdrawn for harvesting of specific nuclides in the fiber or for reconstituting with additional nuclear waste material. |
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