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abstract | A technique for deploying expandables is provided. The technique comprises actuating an expansion tool such that the expansion tool imparts an outwardly directed radial force on an expandable tubular. More specifically, the expansion tool imparts radial expansion forces against an interior surface of the tubular thereby allowing the tubular to be deployed in a wellbore environment. |
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description | The present application is claims priority of Chinese patent application Serial No. 200510086455.0, filed Sep. 22, 2005, the content of which is hereby incorporated by reference in its entirety. 1. Technical Field of the Invention The present invention generally relates to a ray inspection system for making safety inspection of articles, such as a liquid articles, by using rays, more particularly, to a ray beam guiding apparatus. 2. Description of the Related Art In conventional ray inspection systems, X-rays are generated by using an X-ray tube. When electrons strike onto a tungsten target at an accelerative speed under a high voltage between the anode and cathode of the X-ray tube, X-rays are generated. Generally, an X-ray beam is emitted from a focal spot and in the form of conic shape. The profile (e.g. size and/or shape) of the X-ray beam should be limited to different forms according to its different applications. In the prior art, the profile or contour of the X-ray beam is limited by using a collimator for changing the X-ray beam into a line shape, a collimator for changing the X-ray beam into a rectangular shape, or combinations thereof. The conventional ray beam guiding apparatus for limiting the X-ray beam to a substantial fan shape can neither change accurately width of the ray beam and nor adjust conveniently centering of the ray beam. Therefore, the X-rays tend to be deflected and scattered, so that the inspection quality is deteriorated and the thickness of the ray shielding layer inside the guiding apparatus is needed to be increased. An aspect of the present invention is to solve all or at least parts of the above problems occurring in the prior art. According to a first aspect of the present invention, there is provided a ray beam guiding apparatus provided integrally with first and second collimators. The ray beam guiding apparatus according to the embodiments of the present invention can adjust not only the profile (e.g. sizes, and shapes) of the ray beam but also centering of the ray beam relative to the detector array, so that the inspection quality can be improved and the thickness of the ray shielding layer inside the ray beam guiding apparatus can be reduced. The ray beam guiding apparatus according to the present invention is applicable to a ray inspection system for making security inspection of liquid matters. The ray beam guiding apparatus according to a first aspect of the present invention comprises: a ray beam guiding box which has substantially fan-shaped top and bottom surfaces, defines an inner space, and has open wide and narrow ends; an engaging member joined to the narrow end of the ray beam guiding box; a first collimator mounted to the ray beam guiding box adjacent to the narrow end for adjusting a profile of the ray beam in a first direction and a second direction perpendicular to the first direction; a second collimator having a calibration slit or grill and mounted to the ray beam guiding box adjacent to the wide end; and an adjusting member connecting the engaging member and a ray generator for emitting rays for adjusting a distance between the ray generator and the ray beam guiding box. The ray beam guiding box comprises a lower box body, including a substantially fan-shaped bottom plate, first and second side plates extending upwardly from two sides of the bottom plate and perpendicular to the bottom plate, and an extension plate portion extending from the wide end of the ray beam guiding box and lying in the same plane as that of the bottom plate, in which a detector support with a detector array is disposed on the extension plate portion; and an upper cover covering a top of the lower box body. According to an embodiment of the invention, a ray shielding layer is disposed inside the ray beam guiding box, and a through aperture is formed at a substantial center of the ray beam guiding box so as to penetrate through the bottom plate and the upper cover. Further, the ray beam guiding apparatus further comprises: a first boss disposed on the bottom plate adjacent to the narrow end, for engaging and mounting a first calibration member having a calibration slit; a second boss disposed on the bottom plate adjacent to the wide end, for engaging a second calibration member having a calibration slit, in which the through aperture is located between the first and second bosses and the first and second calibration members are cooperated to calibrate the ray beam emitted from the ray generator and passing through the ray beam guiding box; and a third boss disposed on the extension plate portion the detector support and adjust a distance between the detector support and a target spot of the ray generator, wherein the detector support has a support body and a detector arm moveable with respect to the support body. According to an embodiment of the invention, the first collimator comprises: first and second sliding stoppers which are engaged in a first slide groove formed in the ray beam guiding box, and slidable along the first slide groove in the first direction so as to adjust a first distance therebetween; and third and forth sliding stoppers which are engaged in a second slide groove formed in the ray beam guiding box, and slidable along the second slide groove in the second direction so as to adjust a second distance therebetween. In addition, the first collimator further comprises; a first graduator connected to the first and second sliding stoppers to adjust sliding of the first and second sliding stoppers along the first slide groove; and a second graduator connected the third and forth sliding stoppers to adjust sliding of the third and forth sliding stoppers along the second slide groove. Further, the ray beam guiding apparatus further comprises: a first adjusting screw disposed on the adjusting plate to adjust a position of the ray generator with respect to the ray beam guiding box in the first direction; and a second adjusting screw disposed on the adjusting plate to adjust a position of the ray generator with respect to the ray beam guiding box in the second direction. According to second aspect of the invention, there is provided a ray inspection system comprising the ray beam guiding box according to the first aspect of the invention. According to a second aspect of the present invention, there is provided a ray inspection system comprising the ray beam guiding apparatus according to a first aspect of the present invention. With the ray beam guiding apparatus having the above mentioned structures, the following advantages can be achieved with the present invention. 1. The field angle of the ray beam can be controlled easily and effectively. 2. It is possible to adjust the width and height (i.e. sizes in the horizontal and vertical directions) of the ray beam. 3. It is easy to center the target spot of the ray generator (ray source). 4. The thickness of the ray shielding layer of the ray beam guiding apparatus can be reduced, with the inspection quality improved as well. 5. The ray beam guiding apparatus of the present application is applicable to a ray inspection system, particularly to a ray inspection system for performing security inspection of liquid articles. Embodiments of the present invention will be described in detail with reference to drawings, the same elements are denoted by like reference numerals throughout the descriptions. The embodiments described herein are explanatory and illustrative and shall not be construed to limit the present invention. As shown in FIGS. 1 to 3, the ray beam guiding apparatus according to an embodiment of the present application comprises a ray beam guiding box 1, a first collimator 8 (right collimator in FIGS. 1-3), a second collimator 10 (left collimator in FIGS. 1-3), an engaging member/plate 4, and an adjusting member/plate 5. The ray beam guiding box 1 defines an inner space S, and has substantially fan-shaped top and bottom surfaces as well as an open wide end (left end in FIGS. 1-4) and an open narrow end (right end in FIGS. 1-4). The engaging plate 4 is joined to the narrow end of the ray beam guiding box 1, and the adjusting plate 5 connects the engaging plate 4 with a ray generator (ray source) 7 and is adapted to adjust a distance between the ray generator 7 and the ray beam guiding box 1. The first collimator 8, adaptable to adjust sizes and/or shapes (i.e. contour) of the ray beam in first and second directions (e.g. in horizontal and vertical directions in FIG. 1), is mounted to a narrow end portion (right side end in FIG. 1) of the ray beam guiding box 1, that is, the first collimator 8 is mounted to the ray beam guiding box 1 so as to be adjacent to the narrow end. The second collimator 10 having a calibration slit or grill 10a is mounted to a wide end portion (left side end in FIG. 1) of the ray beam guiding box 1, that is, the second collimator 10 is mounted to the ray beam guiding box 1 so as to be adjacent to the wide end. In FIG. 1, the calibration slit or grill 10a is shown as a horizontal slit. More particularly, the ray beam guiding box 1 comprises a lower box body 3 and an upper cover 2 for covering a top of the lower box body 3, thus defining the inner space S. The lower box body 3 includes a substantially fan-shaped bottom plate 3a, a first side plate 3c and a second side plate 3d which are extended upwardly from two sides of the bottom plate 3a and perpendicular to the bottom plate 3a, and an extension plate portion 3b which is extended from the wide end of the ray beam guiding box 1 and lies in the same plane as that of the bottom plate 3a. A detector support 11 provided with detector array (not shown) is disposed on the extension plate portion 3b. As described above, the ray beam guiding portion (i.e. the inner space S surrounded by the bottom plate 3a, the first and second side plates 3c, 3d, and the upper cover 2) Of the ray beam guiding box 1 has a fan shape. Preferably, a ray shielding layer 12 is disposed inside the ray beam guiding box 1, in other words, the ray shielding layer 12 is attached to the bottom plate 3a, the first and second side plates 3c, 3d, and the upper cover 2 respectively. A through aperture 14 is formed at a substantial center of the ray beam guiding box 1 so as to penetrate through the bottom plate 3a and the upper cover 2, and the through aperture 14 is preferably circular. The articles to be inspected, such as a bottle containing liquid can pass through the ray beam guiding box 1 via the through aperture 14, thereby the articles are inspected by the ray beam guided by the ray beam guiding box 1. Further, as shown in FIG. 4, a first boss 19 is disposed on the bottom plate 3a at one side (right side in FIG. 4) of the through aperture 14 adjacent to the narrow end, and adapted to engage and mount a first calibration member 22 having a calibration slit 22a. A second boss 20 is disposed on the bottom plate 3a at the other side (left side in FIG. 4) of the through aperture 14 adjacent to the wide end and adapted to engage and mount a second calibration member 23 having a calibration slit 23a. As shown in FIG. 4, the first calibration member 22 is inserted and engaged between two first bosses 19, and the second calibration member 23 is inserted and engaged between two second bosses 20, so that the first calibration member 22 and the second calibration member 23 can cooperate with each other to calibrate the rays emitted from the ray generator 7 and passing through the ray beam guiding box 1. In the embodiment shown in FIG. 4, the first calibration member 22 has three vertical calibration slits 22a and the second calibration member 23 has one vertical calibration slit 23a, but the present invention is not limited to this, and the calibration slit 22a and the calibration slit 23a can be of any suitable number and can be horizontal. At a side of the second boss 20 adjacent to the wide end of the ray beam guiding box 1, a third boss 21 is disposed on the extension plate portion 3b. The detector support 11 can be moved along the third boss 21 so as to adjust a distance L between the detector support 11 and a target spot P of the ray generator 7. The target spot P is a point where the rays are emitted. The detector support 11 has a support body 11a and a detector arm 11b which is moveable with respect to the support body 11a in the second direction. The detector arm 11b is provided with a detector array (not shown) for receiving the rays emitted from the ray generator 7. Preferably, the first collimator 8, i.e. the right collimator in FIG. 1, comprises first and second sliding stoppers/blocks 9a, 9b which are engaged in a first slide groove (vertical slide groove in FIG. 1) 17a formed in the ray beam guiding box 1, and slidable in the first direction (upward and downward directions in FIG. 1) so as to adjust a first distance (a distance in the first direction) therebetween. Further, the first collimator 8 comprises third and forth sliding stoppers/blocks 6a, 6b which are engaged in a second slide groove 17b (transversal slide groove in FIG. 1) formed in the ray beam guiding box 1, and slidable in the second direction (upward and downward directions in FIG. 2) perpendicular to the first direction so as to adjust a second distance (a distance in the second direction) therebetween. By adjusting the first distance between the first and second sliding stoppers 9a, 9b and the second distance between the third and forth sliding stoppers 6a, 6b, the profile, i.e. sizes/shapes, of the ray beam passing through the first collimator 8 in the first and second directions can be adjusted. Preferably, the first distance between the first and second sliding stoppers 9a, 9b is adjusted by a first graduator 13, and the first graduator 13 is preferably in the form of a graduation rod and connected to the first and second sliding stoppers 9a, 9b. More particularly, the first graduator 13 is rotated, thereby the first and second sliding stoppers 9a, 9b slide along the first slide groove 17a so as to adjust the first distance therebetween. Similarly, the second distance between the third and forth sliding stoppers 6a, 6b is adjusted by a second graduator 16, and the second graduator 16 is preferably in the form of a graduation rod and connected to the third and forth sliding stoppers 6a, 6b. More particularly, the second graduator 16 is rotated, thereby the third and forth sliding stoppers 6a, 6b slide along the second slide groove 17b so as to adjust the second distance therebetween. Preferably, the ray beam guiding apparatus according to the present invention further comprises a first adjusting screw 18 disposed on the adjusting plate 5 so as to adjust a position of the ray generator 7 with respect to the ray beam guiding box 1 in the first direction (in upward and downward directions in FIG. 1), and a second adjusting screw 15 disposed on the adjusting plate 5 so as to adjust a position of the ray generator 7 with respect to the ray beam guiding box 1 in the second direction (in upward and downward directions in FIG. 2). However, the present invention is not limited to this, and the position of the ray generator 7 with respect to the ray beam guiding box 1 in the first and second directions can be adjusted by using any suitable adjusting means in the art. Before using the ray beam guiding apparatus, the alignment between the target spot P of the ray generator 7 and the center of the detector array should be performed according to the relationship between the calibration slit 22a of the first calibration member 22 and the calibration slit 23a of the second calibration member 23. As shown in FIG. 4, the first calibration member 22 having three vertical calibration slits 22a is inserted between two first bosses 19, and the second calibration member 23 having one vertical calibration slits 23a is inserted between two second bosses 20. According to the ray beam, the distance L between the detector array on the detector support 11 and the target spot P of the ray generator 7 is determined by moving the detector support 11 along the third boss 21. In addition, centering the detector array with respect to the ray beam is performed by moving the detector support arm 11b along the relative the detector support body 11a. After determining the position of the detector array, the ray beam guiding apparatus can be used, for example, to perform the ray inspection of a bottle containing liquid matter. Operations of inspecting the bottle are similar to those in the prior art, so that detailed descriptions thereof are omitted herein. The above described ray beam guiding apparatus is applicable to a ray inspection system for ray-inspecting articles such as liquid articles. The other components and operations of the ray inspection system can be similar to those in the prior art, so that detailed descriptions of the ray inspection system are omitted herein. Although several preferred embodiments have been shown and described, it would be appreciated by a person skilled in the art that changes can be made to the present invention without departing from its substantial spirit or essential principle. All the changes occurring within the scope of this invention or within the equivalent scope are included in this invention. |
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claims | 1. A radiation detecting attachment comprising one or more radiation detectors configured to detect radiation from an object of detection, attached removably to a working machine, whereinthe radiation detecting attachment is supported by the working machine movable when the radiation detecting attachment is attached to the working machine, when a plurality of the radiation detectors are included, comprising a frame body supporting the plurality of the radiation detectors, and a distance between at least one radiation detector and another radiation detector is changeable by moving of a moving element included the frame body. 2. The radiation detecting attachment according to claim 1, comprising a plurality of claw members capable of coming closer to each other and separating from each other. 3. The radiation detecting attachment according to claim 1, wherein the radiation detecting attachment is supported by an arm body of the working machine swingably. 4. The radiation detecting attachment according to claim 1, wherein at least one of the radiation detectors is supported via an elastic member. 5. The radiation detecting attachment according to claim 1, comprising one or more discharge nozzles disposed to be capable of replacing pre-replacement air between the radiation detector and a detection area of the object of detection, whereinpost-replacement air having a reduced amount of a radioactive substance contained in the pre-replacement air is discharged from the one or more discharge nozzles. 6. The working machine to which the radiation detecting attachment according to claim 1 is attached. 7. The working machine according to claim 6, comprising a driver's cab equipped with a display device capable of displaying an amount of radiation based on an output of the radiation detector. 8. The working machine according to claim 7, wherein the display device can display mapping associating the amount of radiation with the detection area of the object of detection. 9. A radiation detecting attachment comprising one or more radiation detectors configured to detect radiation from an object of radiation, attached removably to a working machine, whereinthe radiation detecting attachment is supported by the working machine movably when the radiation detecting attachment is attached to the working machine, comprising a plurality of claw members capable of coming closer to each other and separating from each other. 10. The radiation detecting attachment according to claim 9, whereinwhen a plurality of the radiation detectors are included, a distance between at least one radiation detector and another radiation detector is changeable. 11. A sorting method for sorting, with a radiation detecting attachment including one or more radiation detectors configured to detect radiation from an object of detection, attached removably to a working machine, the object of detection on a basis of an amount of the radiation, the method comprising:a step of supporting the radiation detecting attachment by the working machine and bringing the radiation detecting attachment closer to each of detection areas of the object of detection;a step of obtaining an amount of radiation of the detection area on a basis of an output of the radiation detector;a step of detaching the radiation detecting attachment from the working machine and attaching a working attachment, capable of dividing the object of detection into each of the detection areas, to the working machine; anda step of dividing the object of detection with the working attachment in accordance with the obtained amount of radiation. 12. The sorting method according to claim 11, comprising: a step of replacing pre-replacement air between the radiation detector and the detection area with post-replacement air having a reduced amount of a radioactive substance contained in the pre-replacement air before obtaining the amount of radiation of the detection area. 13. The sorting method according to claim 11, comprising a step of displaying mapping associating the obtained amount of radiation with the detection area. 14. A sorting method for sorting, with a radiation detecting attachment including one or more radiation detectors configured to detect radiation from an object of detection, attached removably to a working machine, the object of detection on a basis of an amount of the radiation, the method comprising:a step of preparing a conveying unit capable of conveying a plurality of the objects of detection;a step of detaching the radiation detecting attachment from the working machine and disposing the radiation detecting attachment near the conveying unit so that radiation of the plurality of the objects of detection moving on the conveying unit can be detected sequentially; anda step of obtaining an amount of radiation for each of the plurality of the objects of detection moving on the conveying unit. 15. The sorting method according to claim 14, comprising a step of replacing, when obtaining the amount of radiation for each of the objects of detection, pre-replacement air between the radiation detector and the object of detection with post-replacement air having a reduced amount of a radioactive substance contained in the pre-replacement air. 16. The sorting method according to claim 14, comprising:a step of detaching the radiation detecting attachment from the working machine and attaching a working attachment, capable of supporting the object of detection, to the working machine; anda step of disposing the plurality of the objects of detection on the conveying unit with the working attachment in order to detect the radiation. 17. The sorting method according to claim 14, comprising, when the working machine simultaneously includes the radiation detecting attachment and a working attachment capable of supporting the object of detection, a step of disposing the plurality of the objects of detection on the conveying unit with the equipped working attachment in order to detect the radiation. 18. The sorting method according to claim 14, comprising a step of displaying the obtained amount of radiation for each of the objects of detection. 19. The sorting method according to claim 14, comprising a step of sorting the object of detection with the working attachment in accordance with the obtained amount of radiation. |
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claims | 1. An electron beam irradiation apparatus having an electron accelerator for accelerating electrons emitted from an electron beam source to irradiate a target, and a power supply for supplying power of direct current having a high voltage to said electron accelerator, said power supply comprising: an inverter device for transforming a commercial AC power output into an AC power output of a variable voltage; a DC power supply including a step-up transformer for stepping up a voltage of said AC power output of said inverter device, and a rectifying device for rectifying said stepped up voltage to a high DC voltage, said DC power supply applying said high DC voltage to said electron accelerator; a voltage detector for detecting said high DC voltage; a current detector for detecting an output current of said inverter device; and a feedback control circuit for controlling said power output of said inverter device by detecting said DC high voltage and said output current. 2. An electron beam irradiation apparatus according to claim 1 , further comprising a LC filter circuit disposed between said inverter device and said DC power supply. claim 1 3. An electron beam irradiation apparatus according to claim 1 , further comprising a step-down transformer disposed between said inverter device and a commercial AC power supply, said step-down transformer having a delta-star connection and a delta-delta connection and connected to a converter disposed in said inverter device. claim 1 4. An electron beam irradiation apparatus according to claim 1 , wherein said inverter device controls a power output with a pulse-width modulation on the basis of a result of comparing the voltage or current fed back at each cycle of a carrier frequency signal with a set value. claim 1 5. An electron beam irradiation apparatus according to claim 1 , wherein said feedback control circuit controls said inverter device so as to keep said high DC voltage at a certain set value when an input voltage of a commercial AC power supply varies. claim 1 6. An electron beam irradiation apparatus according to claim 1 , wherein said feedback control circuit controls said inverter device so as to stop said power output thereof when said current detector detects an abnormal current produced by an electric discharge or a short-circuit in parts to which a high voltage is applied. claim 1 |
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description | The present application is related to and claims the benefit of the earliest available effective filing date(s) from the following listed application(s) (the “Related Applications”) (e.g., claims earliest available priority dates for other than provisional patent applications or claims benefits under 35 USC §119(e) for provisional patent applications, for any and all parent, grandparent, great-grandparent, etc. applications of the Related Application(s)). For purposes of the USPTO extra-statutory requirements, the present application constitutes a continuation-in-part of United States Patent Application entitled METHOD, SYSTEM, AND APPARATUS FOR THE THERMAL STORAGE OF NUCLEAR REACTOR GENERATED ENERGY, naming RODERICK A. HYDE, MURIEL Y. ISHIKAWA, CLARENCE T. TEGREENE, JOSHUA C. WALTER, LOWELL L. WOOD, JR., AND VICTORIA Y. H. WOOD as inventors, filed Feb. 18, 2010, application Ser. No. 12/660,025, which is currently co-pending, or is an application of which a currently co-pending application is entitled to the benefit of the filing date. For purposes of the USPTO extra-statutory requirements, the present application constitutes a continuation-in-part of United States Patent Application entitled METHOD, SYSTEM, AND APPARATUS FOR THE THERMAL STORAGE OF NUCLEAR REACTOR GENERATED ENERGY, naming RODERICK A. HYDE, MURIEL Y. ISHIKAWA, CLARENCE T. TEGREENE, JOSHUA C. WALTER, LOWELL L. WOOD, JR., AND VICTORIA Y. H. WOOD as inventors, filed Feb. 19, 2010, application Ser. No. 12/660,157, which is currently co-pending, or is an application of which a currently co-pending application is entitled to the benefit of the filing date. The United States Patent Office (USPTO) has published a notice to the effect that the USPTO's computer programs require that patent applicants reference both a serial number and indicate whether an application is a continuation or continuation-in-part. Stephen G. Kunin, Benefit of Prior-Filed Application, USPTO Official Gazette Mar. 18, 2003, available at http://www.uspto.gov/web/offices/com/sol/og/2003/week11/patbene.htm. The present Applicant Entity (hereinafter “Applicant”) has provided above a specific reference to the application(s) from which priority is being claimed as recited by statute. Applicant understands that the statute is unambiguous in its specific reference language and does not require either a serial number or any characterization, such as “continuation” or “continuation-in-part,” for claiming priority to U.S. patent applications. Notwithstanding the foregoing, Applicant understands that the USPTO's computer programs have certain data entry requirements, and hence Applicant is designating the present application as a continuation-in-part of its parent applications as set forth above, but expressly points out that such designations are not to be construed in any way as any type of commentary and/or admission as to whether or not the present application contains any new matter in addition to the matter of its parent application(s). The present disclosure generally relates to the thermal storage and subsequent utilization of nuclear reactor generated energy. In one aspect, a method includes but is not limited to diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, diverting at least one additional selected portion of energy from a portion of at least one additional nuclear reactor system of the plurality of nuclear reactor systems to the at least one auxiliary thermal reservoir, and supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. In addition to the foregoing, other method aspects are described in the claims, drawings, and text forming a part of the present disclosure. In one or more various aspects, related systems include but are not limited to circuitry and/or programming for effecting the herein-referenced method aspects; the circuitry and/or programming can be virtually any combination of hardware, software, and/or firmware configured to effect the herein-referenced method aspects depending upon the design choices of the system designer. In one aspect, a system includes but is not limited to means for diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, means for diverting at least one additional selected portion of energy from a portion of at least one additional nuclear reactor system of the plurality of nuclear reactor systems to the at least one auxiliary thermal reservoir, and means for supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. In addition to the foregoing, other system aspects are described in the claims, drawings, and text forming a part of the present disclosure. In one aspect, an apparatus includes but is not limited to a first energy transfer system configured to divert a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, at least one additional energy transfer system configured to divert at least one additional selected portion of energy from a portion of at least one additional nuclear reactor system of the plurality of nuclear reactor systems to the at least one auxiliary thermal reservoir, and a heat supply system configured to supply at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. In addition to the foregoing, other system aspects are described in the claims, drawings, and text forming a part of the present disclosure. In addition to the foregoing, various other method and/or system and/or program product aspects are set forth and described in the teachings such as text (e.g., claims and/or detailed description) and/or drawings of the present disclosure. The foregoing is a summary and thus may contain simplifications, generalizations, inclusions, and/or omissions of detail; consequently, those skilled in the art will appreciate that the summary is illustrative only and is NOT intended to be in any way limiting. Other aspects, features, and advantages of the devices and/or processes and/or other subject matter described herein will become apparent in the teachings set forth herein. In the following detailed description, reference is made to the accompanying drawings, which form a part hereof. In the drawings, similar symbols typically identify similar components, unless context dictates otherwise. The illustrative embodiments described in the detailed description, drawings, and claims are not meant to be limiting. Other embodiments may be utilized, and other changes may be made, without departing from the spirit or scope of the subject matter presented here. Referring now to FIG. 1, a system 100 for storing and subsequently utilizing energy generated by a plurality of nuclear reactor systems 102 is described in accordance with the present disclosure. A first energy transfer system 104 may divert energy (e.g., thermal energy or electrical energy) from a portion (e.g., first nuclear reactor 108 or first energy conversion system 110) of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to one or more heat storage materials 111 of one or more auxiliary thermal reservoirs 112, and a second energy transfer system 104 may divert energy from a portion (e.g., second nuclear reactor 108 or second energy conversion system 108) of a second nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to the one or more heat storage materials 111 of the one or more auxiliary thermal reservoirs 112. Further, an additional energy transfer system, up to and including an Nth energy transfer system 104, may divert energy from a portion (e.g., Nth nuclear reactor 108 or Nth energy conversion system 110) of an Nth nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to the one or more heat storage materials 111 of the one or more auxiliary thermal reservoirs 112. Then, one or more heat supply systems 114 (e.g., first heat supply system 114, second heat supply system, or Nth heat supply system 114) may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to one or more energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. For instance, the energy conversion system 110 may include, but is not limited to, a first energy conversion system 110 associated with the first nuclear reactor system 106, a second energy conversion system 110 associated with the second nuclear reactor system 106, or an Nth energy conversion system 110 associated with the Nth nuclear reactor system 106. It is further contemplated that the labeling of the various nuclear reactor systems 106 as the first nuclear reactor system 106, the second nuclear reactor system 106, the third nuclear reactor system 106, and the Nth nuclear reactor system 106 is for illustrative purposes only. As such, the first nuclear reactor system 106, the second nuclear reactor system 106, the third nuclear reactor system 106 and the Nth nuclear reactor system 106 are substantially interchangeable for the purposes described within the present disclosure. Similarly, it is contemplated that the labeling of the various energy conversion systems 110 as the first energy conversion system 110, the second energy conversion system 110, and the Nth energy conversion system 110 is for illustrative purposes only and, therefore, the first energy conversion system 110, the second energy conversion system 110, and the Nth energy conversion system 110 are substantially interchangeable for the purposes described in the present disclosure. Additionally, it is contemplated that the labeling of the various heat supply systems 114 as the first heat supply system 114, the second heat supply system 114, and the Nth heat supply system 114 is for illustrative purposes only and, therefore, the first heat supply system 114, the second heat supply system 114, and the Nth heat supply system 114 are substantially interchangeable for the purposes described in the present disclosure. It is further contemplated that the labeling of the various energy transfer systems 104 as the first energy transfer system 104, the second energy transfer system 104, and the Nth energy transfer system 104 is for illustrative purposes and therefore the first energy transfer system 104, the second energy transfer system 104, and the Nth energy transfer system 104 are substantially interchangeable for the purposes described in the present disclosure. Referring now to FIG. 2, one or more of the nuclear reactors 108 (i.e., the first nuclear reactor, the second nuclear reactor, or the Nth nuclear reactor) of one or more of the nuclear reactor systems 106 (i.e., first nuclear reactor system, second nuclear reactor system, or Nth nuclear reactor system) of the plurality of nuclear reactor systems 102 may include, but are not limited to, one or more thermal spectrum nuclear reactors 202, one or more fast spectrum nuclear reactors 204, one or more multi-spectrum nuclear reactors 206, one or more breeder nuclear reactors 208, or one or more traveling wave nuclear reactors 210. For example, the energy produced by a thermal spectrum nuclear reactor 202 of a nuclear reactor system 106 may be diverted from the thermal spectrum nuclear reactor 202 to one or more auxiliary thermal reservoirs 112 using an energy transfer system 104. Then, one or more heat supply systems 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to one or more energy conversion systems 110 (e.g., the first energy conversion system, the second energy conversion system, or the Nth energy conversion system) of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. By way of further example, the energy produced by a traveling wave nuclear reactor 210 of a nuclear reactor system 106 may be diverted from the traveling wave nuclear reactor 210 to one or more auxiliary thermal reservoirs 112 using an energy transfer system 104. Then, one or more heat supply systems 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to one or more energy conversion systems 110 of the nuclear reactor systems 106. Further, it will be recognized by those skilled in the art that the first nuclear reactor 108, the second nuclear reactor 108, and the Nth nuclear reactor 108 need not consist of the same type of nuclear reactor. For instance, the first nuclear reactor 108 may include a traveling wave nuclear reactor 210, the second nuclear reactor 108 may include a breeder nuclear reactor 208, and the Nth nuclear reactor 108 may include a thermal spectrum nuclear reactor 202. In another aspect, one or more of the energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactors 102 may include, but are not limited to, one or more primary energy conversion systems 212, one or more auxiliary energy conversion systems 214, or one or more emergency energy conversion systems 216. For example, a heat supply system 114 may supply a portion of the thermal energy stored in the heat storage material 111 of the auxiliary thermal reservoir 112 to one or more primary energy conversion systems 212 of the one or more nuclear reactor systems 106 (e.g., the first nuclear reactor system, the second nuclear reactor system or the Nth nuclear reactor system) of the plurality of nuclear reactor systems 102 For instance, the primary energy conversion system 212 may include a turbine 218 coupled to an electric generator used to supply electrical power to the primary load 220 (e.g., electrical power grid) of one or more nuclear reactor systems 106. By way of an additional example, a heat supply system 114 may supply a portion of the thermal energy stored in the heat storage material 111 of the auxiliary thermal reservoir 112 to one or more auxiliary energy conversion systems 214 of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. For instance, the auxiliary energy conversion system 214 may include an energy conversion system that supplements or replaces the output of the primary energy conversion system 212. For example, the auxiliary energy conversion system 214 may include a turbine 218 coupled to an electric generator used to provide supplemental or backup electric power to the primary load 220 of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. By way of a further example, a heat supply system 114 may supply a portion of the thermal energy stored in the heat storage material 111 of the auxiliary thermal reservoir 112 to one or more emergency energy conversion systems 216 of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. For instance, the emergency energy conversion system may include a turbine 218 coupled to an electric generator used to supply electric power to an operation system 222 (e.g., monitoring system, safety system, control system, coolant system or security system) of one or more nuclear reactor systems 106 (e.g., first nuclear reactor, second nuclear reactor, or Nth nuclear reactor) of the plurality of nuclear reactor systems 102. It will be appreciated by those skilled in the art that the emergency energy conversion system 216 may be configured to operate at temperatures lower than the operational temperature of the primary energy conversion system 212, allowing the emergency energy conversion system 216 to supply electrical energy to various operation systems 222 of one or more nuclear reactors 106 of the plurality of nuclear reactors 102 during emergency situations when grid power is unavailable. Further, it will be recognized by those skilled in the art that the first energy conversion system 110, the second energy conversion system 110, and the Nth energy conversion system need not consist of the same type of energy conversion system. For instance, the first energy conversion system 110 may include a primary energy conversion system 212, the second energy conversion system 110 may include an auxiliary energy conversion system 214, and the Nth energy conversion system 110 may include an emergency energy conversion system 216. In another aspect, one or more of the energy conversion systems 110 may include, but are not limited to, one or more turbines 218 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. For example, a heat supply system 114 may supply a portion of the thermal energy stored in the heat storage material 111 of the auxiliary thermal reservoir 112 to one or more turbines 218 of one or more nuclear reactors 106 of the plurality of nuclear reactors 102. By way of further example, a heat supply system 114 may supply a portion of the thermal energy stored in the heat storage material 111 of the auxiliary thermal reservoir 112 to a working fluid 224 of one or more turbines 218 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. For instance, the heat supply system 114 may supply a portion of the thermal energy stored in the heat storage material 111 of the auxiliary thermal reservoir 112 to a pressurized steam working fluid 224 of one or more turbines 218 of the one or more nuclear reactor systems 106. It will be appreciated by those skilled in the art that the thermal energy supplied from the auxiliary thermal reservoir 112, via the one or more heat supply systems 114, to the working fluid 224 of one or more turbines 218 of the one or more nuclear reactor systems 106 may be used to augment the thermal energy supplied to the working fluid 224 of the one or more turbines 218 from the one or more nuclear reactors 108 of the one or more nuclear reactor systems 106. In another aspect, one or more of the energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 may include, but are not limited to, one or more topping cycles 226, one or more bottoming cycles 228, or one or more low grade heat dumps 230. For example, a heat supply system 114 may supply a portion of the thermal energy stored in the heat storage material 111 of the auxiliary thermal reservoir 112 to one or more topping cycles 226 of one or more of the nuclear reactor systems 106. By way of another example, a heat supply system 114 may supply a portion of the thermal energy stored in the heat storage material 111 of the auxiliary thermal reservoir 112 to one or more bottoming cycles 228 of one or more of the nuclear reactor systems 106. By way of further example, a heat supply system 114 may supply a portion of the thermal energy stored in the heat storage material 111 of the auxiliary thermal reservoir 112 to one or more low grade heat dumps of one or more of the nuclear reactor systems 106. For instance, the low grade heat dump may include a portion of the surrounding environment (e.g., surrounding soil or atmosphere). It will be recognized by those skilled in the art that the low grade environmental heat dump serves as the ultimate heat sink, allowing for the effective removal of reactor core decay heat in the event the primary heat removal system(s) fail. In this context, the auxiliary thermal reservoir may serve as a thermal capacitor, residing upstream of the more thermally resistive low grade heat dump, such as the surrounding soil or surrounding atmosphere. As the reactor decay heat falls of exponentially, the auxiliary thermal reservoir, acting as a thermal capacitor, may act to absorb the high initial heat load, while the heat is dissipated at a lower rate to the low grade environmental heat dump. Further, it will be recognized by those skilled in the art that the first energy conversion system 110, the second energy conversion system 110, and the Nth energy conversion system 110 need not consist of the same type of energy conversion system. For instance, the first energy conversion system 110 may include a topping cycle 226 of the first nuclear reactor system 106, the second energy conversion system 110 may include a bottoming cycle 228 of the second nuclear reactor system 106, and the Nth energy conversion system 110 may include a low grade heat dump 230 of the Nth nuclear reactor system 106. In another aspect, a heat supply system 114 may supply a portion of the thermal energy stored in the heat storage material 111 of the auxiliary reservoir 112 to one or more boiling loops 232 of the one or more nuclear reactor systems 106, wherein the one or more boiling loops 232 of the one or more nuclear reactor systems 106 are in thermal communication with one or more energy conversion systems 110 of the one or more nuclear reactor systems 106. For example, a heat supply system 114 may supply a portion of the thermal energy stored in the heat storage material 111 of an auxiliary reservoir 112 to a boiling loop 232 in thermal communication with a turbine 218 of one or more nuclear reactor systems 106. By way of further example, the boiling loop 232 may be in thermal communication with one or more topping cycles 226, one or more bottoming cycle 228 or one or more low grade heat dumps 230 of the one or more nuclear reactor systems 106. It will be appreciated by those skilled in the art that the thermal energy supplied to the boiling loop 232 of the one or more nuclear reactor systems 106 from the one or more auxiliary thermal reservoirs 112 may be used to augment the thermal energy supplied to the one or more boiling loops 232 from the one or more nuclear reactors 108 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Referring now to FIG. 3, one or more of the nuclear reactors 108 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 may include a nuclear reactor having a liquid coolant 302. For example, the liquid coolant 302 of one or more of the nuclear reactors 108 may include, but is not limited to, a liquid metal salt coolant 304 (e.g., lithium fluoride, beryllium fluoride or other fluoride salts), a liquid metal coolant 306 (e.g., sodium, lead, or lead bismuth), a liquid organic coolant 308 (e.g., diphenyl with diphenyl oxide), or a liquid water coolant 310. For instance, an energy transfer system 104 may divert energy from a portion of a liquid sodium cooled nuclear reactor of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. In another instance, the energy transfer system 104 may divert energy from a portion of a liquid water cooled nuclear reactor 220 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. In an additional instance, the energy transfer system 104 may divert energy from a portion of a lithium fluoride cooled nuclear reactor of a nuclear reactor system 106 of the plurality of the nuclear reactor systems to an auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the auxiliary thermal reservoir 112 to one or more energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another aspect, one or more of the nuclear reactors 108 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 may include one or more nuclear reactors having a pressurized gas coolant 312. For example, the pressurized gas coolant 222 may include, but is not limited to, pressurized helium gas or pressurized carbon dioxide gas. For instance, the energy transfer system 104 may divert energy from a portion of a pressurized helium cooled nuclear reactor 312 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the auxiliary thermal reservoir 112 to one or more energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another aspect, one or more of the nuclear reactors 108 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 may include one or more nuclear reactors having a mixed phase coolant 314. For example, the mixed phase coolant 314 may include, but is not limited to, a gas-liquid mixed phase material (e.g., steam water-liquid water). For instance, the energy transfer system 104 may divert energy from a portion of a steam water-liquid water cooled nuclear reactor 314 of a nuclear reactor system 106 of the plurality of nuclear reactors 102 to an auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the auxiliary thermal reservoir 112 to one or more energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Referring now to FIG. 4A, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a liquid heat storage material 402 of one or more auxiliary thermal reservoirs 112. For example, the liquid heat storage material 402 may include, but is not limited to, an organic liquid 404 (e.g., diphenyl with diphenyl oxide), a liquid metal salt 406 (e.g., lithium fluoride, beryllium fluoride or other fluoride salts), a liquid metal 408 (e.g., sodium, lead, or lead bismuth), or liquid water 410. For instance, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a mass of liquid sodium of an auxiliary thermal reservoir 112. In another instance, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems to a mass of liquid water 410 of an auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the liquid heat storage material 402 of the one or more auxiliary thermal reservoirs 112 to one or more energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another embodiment, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a pressurized gas heat storage material 412 of one or more auxiliary thermal reservoirs 112. For example, the pressurized gas material 412 suitable for heat storage may include, but is not limited to, pressurized helium gas or pressurized carbon dioxide gas. For instance, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a mass of pressurized helium of an auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the pressurized gas material 412 of the one or more auxiliary thermal reservoirs 112 to one or more energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another embodiment, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a solid heat storage material 414 of one or more auxiliary thermal reservoirs 112. In one aspect, the solid heat storage material 414 may include a continuous solid material forming a solid object 416. For example, the solid object 416 suitable for heat storage may include, but is not limited to, a three dimensional monolithic object (e.g., a brick), a three dimensional porous object (e.g., a brick containing pores suitable for fluid flow), a three dimensional channeled object (e.g. a brick containing channels suitable for fluid flow), or a three dimensional engineered object (e.g., an object containing an engineered honeycomb pattern for increased heat transfer). For instance, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to one or more solid monolithic objects, such as a brick, a rod, or a sheet of material. In another instance, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a solid engineered object, such as an object constructed of a high heat capacity honeycomb structured material. Further, the solid object 416 may include, but is not limited to, a ceramic solid object, such as a carbide ceramic (e.g., titanium carbide or silicon carbide) or a boride ceramic, a metal solid (e.g., iron or steel) object, or an environmentally present solid (e.g., rock or stone) object. For example, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a ceramic solid object. By way of further example, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to an environmentally preexisting rock or stone structure located in close proximity to one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another aspect, the solid heat storage material 414 may include a particulate solid material 418. For example, the particulate solid material 418 may include, but is not limited to, a granular material (e.g. sand) or a powder material. For instance, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a mass of sand located in close proximity to one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Further, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a mass of sand via heat pipes, wherein one portion of the heat pipes is in thermal communication with a portion of one or more nuclear reactors 108 of one or more nuclear reactor systems 106 and a second portion of the heat pipes is embedded in a volume of sand located in close proximity to one or more nuclear reactor systems 106. It will be recognized by those skilled in the art that the volume of the sand, and like solid materials, need not be constrained by the volume of a reservoir containment system 122, in that uncontained sand, stone, and like heat trapping materials surrounding one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 may serve as a suitable heat storage material 111. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the solid heat storage material 414 of the one or more auxiliary thermal reservoirs 112 to one or more energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another embodiment, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a mixed phase heat storage material 420 of one or more auxiliary thermal reservoir 112. For example, the mixed phase material 420 suitable for heat storage may include, but is not limited to a gas-liquid mixed phase material (e.g., steam water-liquid water) or a liquid-solid mixed phase material (e.g. liquid sodium-solid sodium). For instance, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a mass of steam water-liquid water. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the mixed phase heat storage material 420 of the one or more auxiliary thermal reservoirs 112 to one or more energy conversion systems 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another embodiment, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a mass of a heat storage material having a phase transition within the operating temperature 422 of the auxiliary thermal reservoir 112. For example, an auxiliary thermal reservoir 112 having a heat storage material 116 with a phase transition at approximately 100° C. may continuously operate at temperatures above and below the phase transition at 100° C. Those skilled in the art will recognize that this allows the heat supply system 114 to supply thermal energy from the auxiliary thermal reservoir 112 to one or more energy conversion systems 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 at reservoir temperatures above, below and at the phase transition temperature of the heat storage material 111. For instance, given that sodium has an approximate melting temperature of 97.7° C., a sodium based auxiliary thermal reservoir 112 may operate in the liquid phase at temperatures above 97.7° C. and in the solid phase at temperatures below 97.7° C. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the heat storage material 111 having a phase transition within the operating temperature 422 of the auxiliary thermal reservoir 112 to one or more energy conversion systems 110 of one or more the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Referring now to FIG. 4B, one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a mass of a heat storage material 111 contained in a reservoir containment system 424. For example, the reservoir containment system 424 may include, but is not limited to, an external vessel 426 or an external pool 432. By way of further example, the external vessel 426 may include, but is not limited to an external liquid vessel 428 or an external high pressure gas vessel 430. For instance, the one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a mass of liquid metal 408 (e.g. liquid sodium) contained in an external liquid vessel 428. In another instance, the one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a mass of pressurized gas 412 (e.g. pressurized helium) contained in an external high pressure vessel 430. By way of further example, the external pool 432 may include, but is not limited to, a liquid pool 434. For instance, the one or more energy transfer systems 104 may divert energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to a mass of liquid metal 408 (e.g. liquid sodium) contained in an external liquid pool 434. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the heat storage material 111 contained in the reservoir containment system 424 to one or more energy conversion systems 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Referring now to FIG. 4C, the one or more auxiliary thermal reservoirs 112 may store the energy diverted from the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 in the form of a temperature change 436 in the heat storage material 111 of the one or more auxiliary thermal reservoirs 112. For example, the energy diverted from the one or more nuclear reactor systems 106 to the heat storage material 111 of an auxiliary thermal reservoir 112 may cause the temperature of the heat storage material 111 to increase. For instance, the energy diverted from the one or more nuclear reactor systems 106 to the heat storage material 111 of an auxiliary thermal reservoir 112 may cause the temperature of the heat storage material 111, such as a liquid metal 408 (e.g., liquid sodium), to increase from an initial temperature of approximately 100° C. to a temperature of approximately 500° C. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the heat storage material 111 as a temperature increase 436 to one or more energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another aspect, the one or more auxiliary thermal reservoirs 112 may store the energy diverted from the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 in the form of a phase change 438 in the heat storage material 111 of the one or more auxiliary thermal reservoirs 112. For example, the phase change 438 in the heat storage material 111 may include a solid-liquid phase change 440 or a liquid-gas phase change 442. For instance, the energy diverted from the one or more nuclear reactor systems 106 to a solid heat storage material 414 of an auxiliary thermal reservoir 112 may be stored in the heat storage material 111 by melting the heat storage material 111. For example, the energy diverted from the one or more nuclear reactor systems 106 to a mass of solid sodium may liquefy the mass of sodium via a melting transition at approximately 97.7° C., thus storing a portion of the diverted energy in the liquid phase of the mass of sodium. It will be appreciated by those skilled in the art that the energy required to transform the heat storage material 111 from one phase (e.g. solid) to a new phase (e.g., liquid) is the heat of transformation (i.e., the “latent heat”). Then, a heat supply system 114 may supply a portion of the heat of transformation stored as thermal energy in the heat storage material 111 to one or more energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactors 102. Referring now to FIG. 4D, the operational status of the auxiliary thermal reservoir 112 may be monitored using one or more reservoir monitoring systems 444. For example, the reservoir monitoring system 444 may include a temperature monitoring system 446, a pressure monitoring system 448, a system configured to determine the amount of energy stored in the thermal reservoir 450 or a system configured to determine the amount of available energy capacity of the thermal reservoir 452. For instance, a system configured to determine the amount of energy stored in the thermal reservoir 450 may include thermal and pressure monitoring devices configured to probe the temperature and pressure of the heat storage material 111 of the auxiliary thermal reservoir 112. Further, the thermal and pressure monitoring devices may be interfaced with a computer processing system configured to apply an established algorithm (e.g., established equation-of-state for the storage material in question) to the data outputs of the thermal and pressure monitoring devices, thus relating the temperature and pressure of the heat storage material 111 to the internal energy of the heat storage material 111 (e.g., liquid metal or pressurized gas). In another aspect, the temperature of the auxiliary thermal reservoir 112 may be controlled using a reservoir temperature control system 454. For example, the reservoir temperature control system 454 may be used to increase or decrease the temperature of the auxiliary thermal reservoir 112. For instance, in situations where the internal temperature of the auxiliary thermal reservoir reaches levels outside the predefined operational limits, the reservoir temperature control system 454 may signal the heat supply system 114 to transfer a portion of the thermal energy stored in the auxiliary thermal reservoir 112 to the one or more energy conversion systems 110 of the nuclear reactor systems 106, such as a turbine 218 or a low grade heat dump 230. Referring now to FIG. 5A, one or more of the energy transfer systems 104 may include, but are not limited to, an energy transfer system configured to transfer thermal energy 502 from a portion of one or more nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102 to one or more auxiliary thermal reservoirs 112. For example, an energy transfer system configured to transfer thermal energy 502 from a portion (e.g., primary coolant system) of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to the auxiliary thermal reservoir 112 may divert thermal energy from a portion of a nuclear reactor system 106 to an auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the auxiliary thermal reservoir 112 to one or more energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Further, one or more of the energy transfer systems configured to transfer thermal energy 502 from a portion of one or more nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102 to one or more auxiliary thermal reservoirs 112 may include, but are not limited to, one or more heat transfer systems 504. For example, a heat transfer system 504 may transfer thermal energy from a portion of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. For instance, the heat transfer system 504 may transfer thermal energy from a portion of a nuclear reactor system 106 to an auxiliary thermal reservoir 112 via thermal convection 506 (e.g., natural convection or forced convection via coolant pump(s)). In another instance, the heat transfer system 504 may transfer thermal energy from a portion of a nuclear reactor system 106 to an auxiliary thermal reservoir 112 via thermal conduction 508 (e.g., using a heat exchanger). Those having skill in the art will recognize that the one or more heat transfer systems 504 may be configured to transfer thermal energy from a portion of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to one or more auxiliary thermal reservoirs 112 using both thermal conduction 506 and thermal convection 508. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to one or more energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Further, the one or more heat transfer systems 504 may include, but are not limited to, one or more direct fluid exchange heat transfer systems 510. For example, a direct fluid exchange heat transfer system 510 may transfer thermal energy from a portion of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. The direct fluid exchange heat transfer system 510 may include a system configured to intermix the coolant of a nuclear reactor 108 of a nuclear reactor system 106 with the fluidic heat storage material 111 contained in the reservoir containment system 424. For instance, a fluid carrying loop may couple a primary coolant system of a nuclear reactor system 106 and the reservoir fluid containment system 424, allowing for the intermixing of the two fluids. The rate of reactor coolant-reservoir fluid intermixing may be controlled by the direct fluid exchange transfer system 510. For instance, a valve system and/or fluid pumps (e.g., mechanical pumps or magnetohydrodynamic pumps) may be employed to volumetrically limit the exchange of material between the reactor coolant system of a nuclear reactor system 106 and the reservoir fluid containment system 424. Moreover, the reservoir fluid and the reactor coolant may consist of identical or substantially similar materials. For example, both the reservoir fluid and the reactor coolant may consist of an identical liquid metal, such as liquid sodium. Additionally, the reservoir fluid and the reactor coolant may consist of different materials. For example, the reservoir fluid may consist of a liquid organic, such as diphenyl with diphenyl oxide, while the reactor coolant may consist of liquid sodium. Further, the one or more heat transfer systems 504 may include, but are not limited to, one or more reactor-reservoir heat exchangers 514. For example, a reactor-reservoir heat exchanger 514 may transfer thermal energy from a portion of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. For instance, the reactor-to-reservoir heat exchanger 514 may include a heat exchanger 515 having a first portion in thermal communication with the primary coolant system of the nuclear reactor system 106 and a second portion in thermal communication with the auxiliary thermal reservoir 112. Further, the heat transfer system 504 may include more than one reactor-reservoir heat exchanger 514. For example, a first portion of a first heat exchanger may be in thermal communication with the primary coolant system of the nuclear reactor system 106, while a second portion of the first heat exchanger may be in thermal communication with a heat exchange loop. Further, a first portion of a second heat exchanger may be in thermal communication with the auxiliary thermal reservoir 112, while a second portion of the second heat exchanger may be in thermal communication with the heat exchange loop. Collectively, the first heat exchanger-heat exchange loop-second heat exchanger system acts to transfer thermal energy from the primary coolant system of the nuclear reactor system 106 to the auxiliary thermal reservoir 112. In another aspect, one or more of the energy transfer systems 104 may include, but are not limited to, an energy transfer system configured to transfer electrical energy 503 from a portion of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102 to one or more auxiliary thermal reservoirs 112. For example, an energy transfer system configured to transfer electrical energy 503 from a portion of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to the auxiliary thermal reservoir 112 may transfer electrical energy from a portion (e.g., energy conversion system 110) of the nuclear reactor system 106 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the auxiliary thermal reservoir 112 to one or more energy conversion system 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Further, one or more of the energy transfer systems configured to transfer electrical energy 503 from a portion of one or more nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102 to one or more auxiliary thermal reservoirs 112 may include, but are not limited to, an electrical energy-to-thermal energy conversion system 516. For example, an electrical energy-to-thermal energy conversion system 516, such as a resistive heating device 517 (e.g., a heating coil 518), may convert a portion of the electrical energy produced by an energy conversion system 110 of a nuclear reactor system 106 to thermal energy. It will be recognized by those skilled in the art that the system for transferring electrical energy 503 from a portion of a nuclear reactor system 106 to an auxiliary thermal reservoir 112 may be utilized to convert excess electrical energy produced by an energy conversion system 110 of the nuclear reactor system 106 to thermal energy. Subsequently, a portion of that thermal energy may be transferred to and stored in the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the auxiliary thermal reservoir 112 to one or more energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Referring now to FIG. 5B, one or more heat transfer systems 504 may transfer thermal energy from a portion of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to one or more auxiliary thermal reservoirs 112, wherein the portion of a nuclear reactor system 106 is in thermal communication with a heat source 522 of the nuclear reactor system 106. For example, a heat transfer system 504 may transfer thermal energy from a portion of a nuclear reactor system 106 in thermal communication with the nuclear reactor core 524 of a nuclear reactor 108 of the nuclear reactor system 106 to an auxiliary thermal reservoir 112. Further, the portion of the nuclear reactor system 106 in thermal communication with the nuclear reactor core 524 may include, but is not limited to, a portion of the primary coolant system 526 (e.g., portion of the primary coolant loop 528 or portion of the primary coolant pool 530). For example, a heat transfer system 504 may transfer thermal energy from a primary coolant system 526 of a nuclear reactor system 106 to an auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the auxiliary thermal reservoir 112 to one or more energy conversion systems 110 of the one ore more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Referring now to FIG. 5C, one or more heat transfer systems 504 may transfer thermal energy from a primary coolant system 526 of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112, wherein the primary coolant system 526 is in thermal communication (e.g., thermally communicating via a primary coolant system-secondary coolant system heat exchanger 536) with a secondary coolant system not in thermal communication 532 with the auxiliary thermal reservoir 112. For example, the auxiliary thermal reservoir 112 may be thermally coupled via a heat transfer system 504 to a primary coolant loop 528 of the primary coolant system 526. By way of further example, the auxiliary thermal reservoir 112 may be thermally coupled via a heat transfer system 504 to a primary coolant pool 530 of the primary coolant system 526. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the auxiliary thermal reservoirs 112 to one or more energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems. Referring now to FIG. 5D, one or more heat transfer systems 504 may transfer thermal energy from a primary coolant system 526 of one or more nuclear reactor systems 106 to one or more auxiliary thermal reservoirs 112, wherein the primary coolant system 526 and a secondary coolant system 532 of the one or more nuclear reactor systems 106 are both in thermal communication with the auxiliary thermal reservoir 112. For example, the auxiliary thermal reservoir 112 may be thermally coupled to both a primary coolant loop 528 of the primary coolant system 526 of a nuclear reactor system 106 and a secondary coolant loop 534 of a secondary coolant system 532 of the nuclear reactor system 106, such that the thermal path coupling the primary coolant loop 526, the auxiliary thermal reservoir 112, and the secondary coolant loop 532 is parallel to the thermal path coupling the primary coolant loop 526, the primary-secondary coolant system heat exchanger 536, and the secondary coolant loop 532. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the auxiliary thermal reservoir 112 to one or more energy conversion systems 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Referring now to FIG. 6, the heat supply system 114 may include, but is not limited to, a heat exchange loop 602. For example, a first portion of a heat exchange loop 602 may be in thermal communication with a portion of the auxiliary thermal reservoir 112 and a second portion of the heat exchange loop 602 may be in thermal communication with an energy conversion system 110 of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Then, in response to a shutdown event of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102, the heat exchange loop 602 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to at least one energy conversion system 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another aspect, the heat supply system 114 may include, but is not limited to, one or more heat pipes 604. For example, a first portion of a heat pipe 604 may be in thermal communication with a portion of the auxiliary thermal reservoir 112 and a second portion of the heat pipe 604 may be in thermal communication with an energy conversion system 110 of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Then, in response to a shutdown event of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102, the heat pipe 604 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to at least one energy conversion system 110 of the one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another aspect, the heat supply system 114 may include, but is not limited to, one or more heat exchangers 606. For example, a first portion of a first heat exchanger 608 may be in thermal communication with a portion of the auxiliary thermal reservoir 112 and a second portion of the first heat exchanger 606 may be in thermal communication with an energy conversion system 110 of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Then, the heat pipe 604 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to at least one energy conversion system 110 of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. It will be recognized by those skilled in the art that a combination of heat exchange loops 602, heat exchangers 606, and heat pipes 604 may be used in conjunction to supply heat from the auxiliary thermal reservoir 112 to an energy conversion system 110 of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. For example, a first heat exchanger 606, containing a number of heat pipes 604, may be used to thermally couple the auxiliary thermal reservoir 112 and a first portion of a heat exchange loop 602. Moreover, a second heat exchanger 606, also containing numerous heat pipes 604, may be used to thermally couple a portion of an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to the heat exchange loop 602. Then, thermal energy may be supplied from the auxiliary thermal 112 reservoir to the energy conversion system 110 via the heat exchange loop-heat exchanger circuit. In another aspect, the heat supply system 114 may include, but is not limited to, one or more thermoelectric devices 608. For example, a first portion of a thermoelectric device 608 (e.g., p-type/n-type semiconductor thermoelectric junction) may be placed in thermal communication with the auxiliary thermal reservoir 112, while a second portion of the thermoelectric device 608 may be placed in thermal communication with a cold reservoir (e.g., an environmental reservoir or any portion of the nuclear reactor system at a temperature lower than the auxiliary thermal reservoir) of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Then, the electrical power produced by the thermoelectric conversion of the thermal energy stored in the auxiliary thermal reservoir 112 may be used to supplement or replace the electrical output of an energy conversion system 110 of one or more nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Referring now to FIG. 7, an additional energy source 702 may supplement the auxiliary thermal reservoir 112 with an additional portion of energy. For example, excess energy from the load 220 (e.g., the external grid 703) of one or more of the nuclear reactor systems 106 may be used to provide supplemental energy to the auxiliary thermal reservoir 112. For instance, when grid requirements are such that an energy conversion system 110 is producing excess electrical power, the excess power may be converted to thermal energy via an electrical-to-thermal energy conversion process (e.g., heating coil) and transferred to the auxiliary thermal reservoir 112 using a supplementary energy transfer system 704, thus supplementing the energy transferred to the auxiliary thermal reservoir 112 via the energy transfer systems 104 during normal operation. By way of another example, the additional energy source 702 may include, but is not limited to, a non-nuclear reactor energy source 708, such as coal powered generator, a solar array, or wind powered turbine. For instance, electrical energy produced from a coal powered generator may be converted to thermal energy via an electrical-to-thermal energy conversion process and transferred to the auxiliary thermal reservoir 112 using a supplementary energy transfer system 704, thus supplementing the energy transferred to the auxiliary thermal reservoir 112 via the energy transfer systems 104 during normal operation. In another instance, excess electrical energy from a solar array or wind powered turbine may be converted to thermal energy via an electrical-to-thermal energy conversion process and transferred to the auxiliary thermal reservoir 112 using a supplementary energy transfer system 704, thus supplementing the energy transferred to the auxiliary thermal reservoir 112 via the energy transfer systems 104 during normal operation. In an additional instance, thermal energy produced by a coal generator may be transferred directly to the auxiliary thermal reservoir 112 via a supplementary energy transfer system 704, thus supplementing the energy transferred to the auxiliary thermal reservoir 112 via the primary energy transfer systems 104 during normal operation. It will be recognized by those skilled in the art that the supplemental energy supplied to the auxiliary thermal reservoir 112 by an additional energy source may be used to superheat the reservoir material of the auxiliary thermal reservoir to temperatures beyond normal operational capability. Referring now to FIG. 8A, one or more of the energy transfer systems 104 may include, but are not limited to, an energy transfer system responsive to a condition 802. The conditions with which the energy transfer system is responsive may include, but are not limited to, nuclear reactor operational conditions (e.g., temperature, rate of change of temperature, pressure or rate of change of pressure, nuclear reactor capacity), power demand on the one or more nuclear reactor systems (e.g., electrical power requirements of the grid), nuclear reactor system operation system conditions (e.g., control system, monitoring system, or safety system (e.g., heat sink status or coolant pump status)). For example, in response to a coolant pump malfunction of one of the nuclear reactor systems 106, an energy transfer system 104 may divert energy from a portion of the nuclear reactor system 106 to the auxiliary thermal reservoir 112. By way of further example, at or near a specified operating temperature of a portion of a nuclear reactor system 106 (e.g., nuclear reactor core or nuclear reactor coolant fluid), an energy transfer system 104 may initiate transfer of thermal energy from the nuclear reactor 108 of the nuclear reactor system 106 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to one or more of the energy conversion systems 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Further, an energy transfer system associated with a first nuclear reactor 106 of the plurality of nuclear reactor systems 102 may include an energy transfer system responsive to a condition of a first nuclear reactor system 804. For example, in response to a coolant pump malfunction of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102, an energy transfer system configured to respond to a condition of the first nuclear reactor system 804 may divert energy from a portion of the first nuclear reactor system 106 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to one or more of the energy conversion systems 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Additionally, an energy transfer system associated with a first nuclear reactor 106 of the plurality of nuclear reactor systems 102 may include an energy transfer system responsive to a condition of an additional nuclear reactor system 806 of the plurality of nuclear reactor systems 102. For example, in response to a drop in the energy output of a second nuclear reactor system 106 of the plurality of nuclear reactor systems, the energy transfer system configured to respond to a condition of an additional nuclear reactor system 806 may divert energy from a portion of the first nuclear reactor system 106 of the plurality of nuclear reactor system 102 to the auxiliary thermal reservoir 112. By way of further example, in response to a drop in the energy output of the second and third nuclear reactor systems 106 (e.g., a drop in both the individual outputs of the second nuclear reactor system and third nuclear reactor system or a drop in the collective output of the second and third nuclear reactors systems) of the plurality of nuclear reactor systems 102, the energy transfer system configured to respond to a condition of an additional nuclear reactor system 806 may divert energy from a portion of the first nuclear reactor system 106 of the plurality of nuclear reactor system 102 to the auxiliary thermal reservoir 112. Further, in response to a drop in the energy output of the Nth nuclear reactor system 106 of the plurality of nuclear reactor systems 102, the energy transfer system configured to respond to a condition of an additional nuclear reactor system 806 may divert energy from a portion of the first nuclear reactor system 106 of the plurality of nuclear reactor system 102 to the auxiliary thermal reservoir 112. More generally, in response to a condition of the Nth nuclear reactor system 106 of the plurality of nuclear reactor systems 102, the corresponding energy transfer systems configured to respond to a condition of an additional nuclear reactor system 806 may divert energy from a portion of the first nuclear reactor system 106, the second nuclear system 106, or up the (N−1) nuclear reactor system 106 of the plurality of nuclear reactor system 102 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to one or more of the energy conversion systems 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another aspect, one or more of the energy transfer systems 104 may include, but are not limited to, an energy transfer system responsive to the determination of excess capacity 808 of one or more of the nuclear reactor systems of the plurality of the nuclear reactor systems 102. For example, in the event one or more of the nuclear reactor systems 106 is producing more energy than is required by the load (e.g., external electrical power grid) of the energy conversion system 110 of the nuclear reactor system 106, the energy transfer system may initiate transfer of thermal or electrical energy from a portion of one or more of the nuclear reactor systems 106 (e.g., a first nuclear reactor system 106, a second nuclear reactor system 106 or a Nth nuclear system 106) to the auxiliary thermal reservoir 112. For instance, in the event a first nuclear reactor system 106 is producing more energy than is required by the load (e.g., external electrical power grid) of the energy conversion system 110 of the first nuclear reactor system 106, the energy transfer system 104 may initiate transfer of thermal or electrical energy from a portion of the first nuclear reactor system 106, the second nuclear reactor system 106 or the Nth nuclear reactor system 106 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to at least one energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102. In an additional aspect, one or more of the energy transfer systems 104 may include, but are not limited to, an energy transfer system responsive to an operation system 810 of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102. For example, the energy transfer system responsive to an operation system 810 may include, but is not limited to, an energy transfer system responsive to a signal from an operation system 812. For example, in response to a signal, such as a remote wireless signal (e.g., radio frequency signal) or remote wireline signal (e.g., copper wire signal or fiber optic cable signal), from an operation system (e.g., shutdown system, warning system, or security system) of one or more of the nuclear reactor systems 106, an energy transfer system responsive to a signal from an operation system 812 may initiate transfer of energy from a portion of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102 to the auxiliary thermal reservoir 112. Further, the energy transfer system responsive to an operation system 810 may include, but is not limited to, an energy transfer system responsive to a monitoring system 808 (e.g., temperature monitoring system or pressure monitoring system), an energy transfer system responsive to a control system 810, or an energy transfer system responsive to safety system 812. For instance, in response to a signal from a monitoring system 814 of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102, one or more of the energy transfer systems 104 may initiate transfer of energy from a portion of one or more of the nuclear reactor systems 106 to the auxiliary thermal reservoir 112. In another instance, in response to a signal from a control system 816 of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102, one or more of the energy transfer systems 104 may initiate transfer of energy from a portion of one or more of the nuclear reactor systems 106 to the auxiliary thermal reservoir 112. Further, in response to a signal from a safety system 818 of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102, one or more of the energy transfer systems 104 may initiate transfer of energy from a portion of one or more of the nuclear reactor systems 106 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to one or more of the one energy conversion systems 110 of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102. Further, an energy transfer system associated with a first nuclear reactor 106 of the plurality of nuclear reactor systems 102 may include an energy transfer system responsive to a signal from an operation system of the first nuclear reactor system. For example, in response to a signal from an operation system of the first nuclear reactor system 106 of the plurality of nuclear reactor systems 102, an energy transfer system configured to respond to a signal of an operation system of the first nuclear reactor system may divert energy from a portion of the first nuclear reactor system 106 to the auxiliary thermal reservoir 112. For instance, in response to a signal from the monitoring system of the first nuclear reactor system the energy transfer system may divert energy from a portion of the first nuclear reactor system 106 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to one or more of the energy conversion systems 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Moreover, an energy transfer system associated with a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 may include an energy transfer system responsive to a signal from an operation system of an additional nuclear reactor system. For example, in response to a signal from an operation system of an additional nuclear reactor system 106 (e.g., second nuclear reactor system 106, third nuclear reactor system 106, or Nth nuclear reactor system 106) of the plurality of nuclear reactor systems 102, the energy transfer system configured to respond to a signal from an operation system of an additional nuclear reactor system may divert energy from a portion of the first nuclear reactor system 106 of the plurality of nuclear reactor system 102 to the auxiliary thermal reservoir 112. For instance, in response to a signal from a monitoring system of an additional nuclear reactor system, the energy transfer system may divert energy from a portion of the first nuclear reactor system 106 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to one or more of the energy conversion systems 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another aspect, one or more of the energy transfer systems 104 may include, but are not limited to, an energy transfer system responsive to a signal from an operator 820 of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102. For example, in response to a signal from an operator (e.g., human user or human controlled system, such as a programmed computer system), one or more energy transfer systems responsive to a signal from an operator 820 may initiate transfer of energy from a portion of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102 to the auxiliary thermal reservoir 112. For instance, an energy transfer system responsive to a signal from an operator 820, in response to a remote signal, such as a wireline or wireless signal from a computer terminal controlled by an operator, may initiate transfer of thermal energy from a nuclear reactor 108 of one or more of the nuclear reactor systems 106 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to one or more of the energy conversion systems 110 of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102. In an additional aspect, one or more of the energy transfer systems 104 may include, but are not limited to, an energy transfer system responsive to a pre-selected diversion start time 822. For example, the pre-selected diversion start time may include a time of elapse (e.g., time of elapse measured relative to a specific event, such as a shutdown event or satisfaction of grid demand requirements) or an absolute time. For instance, an energy transfer system responsive to a pre-selected diversion start time 822, at a pre-selected absolute time (e.g., 2:00 a.m. eastern standard time) may initiate transfer of energy from a nuclear reactor system 106 of the plurality of the nuclear reactor systems 102 to the auxiliary thermal reservoir 112. It will be recognized by those skilled in the art that historical grid power demand data may be utilized to determine the appropriate time in which to begin diversion of nuclear reactor generated energy to the auxiliary thermal reservoir 112. In another instance, the energy transfer system responsive to a pre-selected diversion start time 822, upon elapse of a pre-selected amount of time from a specific event, such as a nuclear reactor shutdown or achievement of power production in excess of external demand, may initiate transfer of energy from a portion of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to at least one energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102. In another aspect, one or more of the energy transfer systems 104 may include, but are not limited to, an energy transfer system responsive to a reservoir operation system 824 of one or more auxiliary thermal reservoirs 112. For example, an energy transfer system responsive to a reservoir operation system 824 may include, but is not limited to, an energy transfer system responsive to a signal from a reservoir operation system 826. For example, in response to a signal, such as a remote wireless signal (e.g., radio frequency signal) or remote wireline signal (e.g., copper wire signal or fiber optic cable signal), from a reservoir operation system of the auxiliary thermal reservoir 112, the energy transfer system responsive to a signal from a reservoir operation system 826 may initiate transfer of energy from a portion of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. Further, the energy transfer system responsive to a reservoir operation system 824 may include, but is not limited to, an energy transfer system responsive to a reservoir monitoring system 828 (e.g., temperature monitoring system, pressure monitoring system, system for monitoring amount of stored energy, or system for monitoring the amount of available storage capacity), an energy transfer system responsive to a reservoir control system 830, or an energy transfer system responsive to a reservoir safety system 832. For instance, in response to a signal from a reservoir monitoring system, the energy transfer system responsive to a reservoir monitoring system 828 may initiate transfer of energy from a portion of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to the auxiliary thermal reservoir 112. In another instance, in response to a signal from a reservoir control system, the energy transfer system responsive to a reservoir control system 830 may initiate transfer of energy from a portion of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to the auxiliary thermal reservoir 112. Further, in response to a signal from a reservoir safety system, the energy transfer system responsive to a reservoir safety system 8832 may initiate transfer of energy from a portion of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to at least one energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Referring now to FIG. 8B, one or more of the energy transfer systems 104 may include, but are not limited to, an energy transfer system responsive to a shutdown event 834 of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102. For example, an energy transfer system responsive to a shutdown event 834 of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102 may include, but is not limited to, an energy transfer system responsive to a scheduled shutdown event 834 of one or more of the nuclear reactor systems 106 or an energy transfer system responsive to an emergency shutdown event 838 of one or more of the nuclear reactor systems 106. For instance, in response to a schedule shutdown event (e.g., routine maintenance), one or more of the energy transfer systems responsive to a scheduled shutdown event 836 of one or more of the nuclear reactors 106 may initiate transfer of energy from a portion of one or more of the nuclear reactor system 106 to the auxiliary thermal reservoir 112. In another instance, in response to an emergency shutdown event (e.g., SCRAM), one or more of the energy transfer systems responsive to an emergency shutdown event 838 of one or more of the nuclear reactors 106 may initiate transfer of energy from a portion of one or more of the nuclear reactor system 106 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the auxiliary thermal reservoir 112 to at least one energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. It will be recognized by those skilled in the art that, in response to a shutdown event of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102, energy may be diverted from a portion of the nuclear reactor system 106 to the auxiliary thermal reservoir 112 prior to, during, and following the shutdown of the nuclear reactor 108 of the nuclear reactor system 106, as part of the steps required to facilitate the nuclear reactor system 106 shutdown. Further, an energy transfer system associated with a first nuclear reactor 106 of the plurality of nuclear reactor systems 102 may include an energy transfer system responsive to a shutdown event of the first nuclear reactor system. For example, in response to a shutdown event of the first nuclear reactor system 106 of the plurality of nuclear reactor systems 102, an energy transfer system configured to respond to a shutdown event of the first nuclear reactor system may divert energy from a portion of the first nuclear reactor system 106 to the auxiliary thermal reservoir 112. For instance, in response to an emergency shutdown event of the first nuclear reactor system the energy transfer system may divert energy from a portion of the first nuclear reactor system 106 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to one or more of the energy conversion systems 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Moreover, an energy transfer system associated with a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 may include an energy transfer system responsive to a shutdown event of an additional nuclear reactor system. For example, in response to a shutdown event of an additional nuclear reactor system 106 (e.g., second nuclear reactor system 106, third nuclear reactor system 106, or Nth nuclear reactor system 106) of the plurality of nuclear reactor systems 102, the energy transfer system configured to respond to a shutdown event of an additional nuclear reactor system may divert energy from a portion of the first nuclear reactor system 106 of the plurality of nuclear reactor system 102 to the auxiliary thermal reservoir 112. For instance, in response to a scheduled shutdown event of an additional nuclear reactor system, the energy transfer system may divert energy from a portion of the first nuclear reactor system 106 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to one or more of the energy conversion systems 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another aspect, one or more of the energy transfer systems 104 may include, but are not limited to, an energy transfer system suitable for establishing thermal communication between a nuclear reactor system and the auxiliary thermal reservoir 840. For example, in response to a condition, the energy transfer system suitable for establishing thermal communication between the nuclear reactor system and the auxiliary thermal reservoir 840 may establish a thermal pathway between a portion of a nuclear reactor 108 (e.g., primary coolant system) of the nuclear reactor system 106 and the auxiliary thermal reservoir 112. For instance, in the case of a direct fluid exchange heat transfer system 510, a control valve may be used to initiate the intermixing of the reactor coolant and reservoir fluid. In another instance, in the case of a heat transfer system employing a reactor-reservoir heat exchanger 514, a control valve may be used to initiate reactor coolant flow through the heat exchanger. In another aspect, one or more of the energy transfer systems 104 may include, but are not limited to, an energy transfer system responsive to the determination of the amount of energy stored in the auxiliary thermal reservoir 842. For example, in response to the determination of energy currently stored in the auxiliary thermal reservoir 112, the energy transfer system responsive to the determination of the amount of energy stored in the auxiliary thermal reservoir 842 may initiate transfer of energy from a portion of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102 to the auxiliary thermal reservoir 112. Further, the energy transfer system responsive to the determination of the amount of energy stored in the auxiliary thermal reservoir 842 may include an energy transfer system responsive to the determination of the percentage of energy stored, relative to the overall storage capacity, in the auxiliary thermal reservoir 844. For example, in response to a determination of a set percentage level of stored energy (e.g., 25% of energy storage capacity is being utilized), the energy transfer system responsive to the determination of the percentage of stored energy 842 may initiate transfer of energy from a portion a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to at least one energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In an additional aspect, one or more of the energy transfer systems 104 may include, but are not limited to, an energy transfer system responsive to the determination of the amount of available storage capacity in the auxiliary thermal reservoir 846. For example, in response to the determination of available energy storage capacity, the energy transfer system responsive to the determination of the amount of available storage capacity in the auxiliary thermal reservoir 846 may initiate transfer of energy from a portion of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to the auxiliary thermal reservoir 112. Further, the energy transfer system responsive to the determination of the amount of available storage capacity in the auxiliary thermal reservoir 846 may include an energy transfer system responsive to the determination of the percentage of available energy storage capacity in the auxiliary thermal reservoir 848. For example, in response to a determination of a set level of available energy storage (e.g., 75% storage capacity remains), the energy transfer system responsive to the determination of the percentage of available energy storage capacity 848 may initiate transfer of energy from a portion of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to the auxiliary thermal reservoir 112. Then, a heat supply system 114 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to at least one energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Referring now to FIG. 8C, one or more of the energy transfer systems 104 may include, but are not limited to, an energy transfer system suitable for diverting excess energy from a nuclear reactor system of the plurality of nuclear reactor systems to an auxiliary thermal reservoir 850. For example, an energy transfer system suitable for diverting excess energy from a nuclear reactor system to an auxiliary thermal reservoir 850 may transfer energy exceeding operational demand of an energy conversion system 852. For instance, in the event a turbine-generator system of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 is producing electrical power in excess of grid demand, the energy transfer system 104 may transfer energy (e.g., thermal or electrical) from a portion of a nuclear reactor system 106 to an auxiliary thermal reservoir 112. Further, one or more of the energy transfer systems 104 may include an energy transfer system configured to divert a specified percentage of the energy output of a nuclear reactor system to an auxiliary thermal reservoir 854. For example, a control system or operator may choose to transfer a pre-selected percentage of a nuclear reactor system 106 output and transfer at least a portion of that energy to the auxiliary thermal reservoir 112. It will be recognized by those skilled in the art that the level of energy output pre-selected to be transferred to the auxiliary thermal reservoir may be a function of time and may be derived from historic external power demand curves. For example, in times of day or times of year historically displaying relatively low grid demand, the control system or operator may choose to divert a larger percentage of the output of one or more of the nuclear reactor systems 106 to the auxiliary thermal reservoir than the percentage transferred during periods of higher demand. Referring now to FIG. 9A, one or more of the heat supply systems 114 may include, but are not limited to a heat supply system responsive to a condition 902. The conditions with which one or more of the heat supply systems are responsive may include, but are not limited to, nuclear reactor operational conditions (e.g., temperature, rate of change of temperature, pressure or rate of change of pressure, nuclear reactor capacity), power demand on the one or more nuclear reactor systems (e.g., electrical power requirements of the grid), nuclear reactor system operation system conditions (e.g., control system, monitoring system, or safety system (e.g., heat sink status or coolant pump status)), or reservoir operational conditions (e.g., temperature, rate of change of temperature, pressure or rate of change of pressure). For example, in response to a condition of one or more of the nuclear reactor systems 106, a heat supply system configured to respond to a condition of one or more of the nuclear reactor systems 904 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. For instance, in response to heightened power demand on a the nuclear reactor systems 106, a heat supply system responsive to heightened power demand on one or more of the nuclear reactor systems 906 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another aspect, one or more of the heat supply systems responsive to a condition 902 may include, but are not limited to a heat supply system responsive to a shutdown event 908. For example, in response to an emergency shutdown event (e.g., SCRAM), a heat supply system responsive to an emergency shutdown event 910 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. By way of another example, in response to a scheduled shutdown event (e.g., routine maintenance), a heat supply system responsive to a schedule shutdown event 912 may supply a portion of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. It will be recognized by those skilled in the art that, in response to a shutdown event of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102, the thermal energy stored in the auxiliary thermal reservoir 112 may be transferred from the auxiliary thermal reservoir 112 to an energy conversion system 110 of one or more nuclear reactor systems 106 prior to, during, and following the shutdown of a nuclear reactor system 106 as part of the steps required to facilitate the nuclear reactor system 106 shutdown. In another aspect, one or more of the heat supply systems responsive to a shutdown event 908, may include, but are not limited to, a heat supply system responsive to a shutdown event established by an operation system 914. For example, in response to a shutdown event established by an operation system (e.g., shutdown system) of one or more of the nuclear reactor systems 106, a heat supply system responsive to a shutdown event established by an operation system 914 may initiate transfer of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. By way of further example, a heat supply system responsive to a shutdown event established by a reactor control system 916 may initiate transfer of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems of the plurality of nuclear reactor systems. Further, the reactor control system may include a reactor control system responsive to a signal from one or more reactor safety systems 918. For example, a heat supply system responsive to a shutdown event established by a reactor control system responsive to a signal from a safety system 918 may initiate transfer of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of the nuclear reactor systems 102. Even further, the safety system may include a safety system responsive to one or more sensed conditions of one or more of the nuclear reactor systems 106 (e.g., external conditions or internal conditions) 920. For instance, a safety system of one o more of the nuclear reactor systems 106, upon sensing a loss of heat sink, may send a signal to a reactor control system of one of the nuclear reactor systems 106. In turn, the reactor control system may establish a nuclear reactor system 106 shutdown and send a corresponding signal to a heat supply system responsive to a shutdown event established by a reactor control system. Then, the heat supply system responsive to a shutdown event established by a reactor control system may initiate transfer of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Referring now to FIG. 9B, one or more of the heat supply systems responsive to a condition 902 may include, but are not limited to, a heat supply system responsive to an operation system 922 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. For example, in response to a signal, such as a remote wireless signal (e.g., radio frequency signal) or remote wireline signal (e.g., copper wire signal or fiber optic cable signal), from an operation system (e.g., control system, safety system, monitoring system, shutdown system, warning system, or security system) of one or more of the nuclear reactor systems, the heat supply system responsive to a signal from an operation system 924 may initiate transfer of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. For instance, upon receiving a signal from a monitoring system a nuclear reactor system 106 indicating the shutdown of the nuclear reactor system 106, a heat supply system responsive to a signal from an operation system 924 of one or more of the nuclear reactor systems 106 may initiate transfer of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another aspect, one or more of the heat supply systems responsive to a condition 902 may include, but are not limited to, a heat supply system responsive to a reservoir operation system 926 of one or more of the auxiliary thermal reservoirs 112. For example, in response to a signal, such as a remote wireless signal (e.g., radio frequency signal) or remote wireline signal (e.g., copper wire signal or fiber optic cable signal), from a reservoir operation system (e.g., control system, safety system, monitoring system) of one or more of the auxiliary thermal reservoirs 112, a heat supply system responsive to a signal from a reservoir operation system 928 may initiate transfer of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. For instance, upon receiving a signal from a monitoring system of an auxiliary thermal reservoir 112 indicating the shutdown of a nuclear reactor system 106 (e.g., energy no longer being diverted to thermal reservoir), the heat supply system responsive to a signal from a reservoir operation system 928 may initiate transfer of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one of more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In an additional aspect, one or more of the heat supply systems responsive to a condition 902 may include, but are not limited to, a heat supply system responsive to an operator 930 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. For example, in response to a signal from an operator (e.g., human user or human controlled system, such as a programmed computer system), a heat supply system responsive to a signal from an operator 932 may initiate transfer of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. For instance, a heat supply system responsive to a signal from an operator 932, in response to a remote signal, such as wireline or wireless signal from a computer terminal controlled by an operator, may initiate transfer of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Referring now to FIG. 9C, one or more of the heat supply systems responsive to a condition 902 may include, but are not limited to, a heat supply system responsive to a pre-selected supply start time 934. For example, the pre-selected supply start time may include the amount of elapsed time relative to a specific event (e.g., shutdown event) or an absolute time. For instance, a heat supply system responsive to a pre-selected supply start time 934, at a pre-selected absolute time (e.g., 5:00 p.m. eastern standard time), may initiate transfer of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. It will be recognized by those skilled in the art that historical grid power demand data may be utilized to determine the appropriate time in which to begin transfer of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to at least one energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another instance, a heat supply system responsive to a pre-selected supply start time 934, upon elapse of a pre-selected amount of time from a specific event, such as a nuclear reactor 108 shutdown, may initiate transfer of the thermal energy stored in the one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In another aspect, one or more of the heat supply systems responsive to a condition 902 may include, but are not limited to, a heat supply system responsive to the determination of the amount of energy stored in one or more of the auxiliary thermal reservoirs 936. For example, in response to the determination of energy currently stored in an auxiliary thermal reservoir 112, a heat supply system responsive to the determination of the amount of energy stored in the auxiliary thermal reservoir 936 may initiate transfer of the thermal energy stored in the auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Further, the heat supply system responsive to the determination of the amount of energy stored in the auxiliary thermal reservoir 936 may include a heat supply system responsive to the determination of the percentage of energy stored, relative to the overall storage capacity, in the auxiliary thermal reservoir 938. For example, in response to the determination of a set percentage level of stored energy (e.g., 80% of energy storage capacity is being utilized), a heat supply system responsive to the determination of the percentage of stored energy 938 may initiate transfer of the thermal energy stored in one or more auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. In an additional aspect, one or more of the heat supply systems responsive to a condition 902 may include, but are not limited to, a heat supply system responsive to the determination of the amount of available storage capacity in one or more of auxiliary thermal reservoirs 940. For example, in response to the determination of available energy storage capacity, a heat supply system responsive to the determination of the amount of available storage capacity in an auxiliary thermal reservoir 940 may initiate transfer of the thermal energy stored the auxiliary thermal reservoirs 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Further, the heat supply system responsive to the determination of the amount of available storage capacity in the auxiliary thermal reservoir 940 may include a heat supply system responsive to the determination of the percentage of available energy storage capacity in an auxiliary thermal reservoir 942. For example, in response to the determination of a set percentage level of available energy storage (e.g., 20% storage capacity remains), a heat supply system responsive to the determination of the percentage of available energy storage capacity 942 of an auxiliary thermal reservoir 112 may initiate transfer of the thermal energy stored in the auxiliary thermal reservoir 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Referring now to FIG. 9D, one or more of the heat supply systems 114 may include, but are not limited to, a heat supply system suitable for supplying a specified portion of the energy stored in one or more of auxiliary thermal reservoirs to an energy conversion system of one or more of the nuclear reactor systems of the plurality of nuclear reactor systems 944. For example, a heat supply system suitable for supplying a specified portion of the energy stored in an auxiliary thermal reservoir 944 may be utilized to transfer a specified amount of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. For instance, the amount of energy transferred from an auxiliary thermal reservoir 112 to an energy conversion system 110 may be based on current load demand (e.g., grid demand), where a control system or operator may choose the amount of energy to be transferred to the energy conversion system based on the level of demand that the energy conversion system is currently undergoing. Further, the heat supply system suitable for supplying a specified portion of the energy stored in an auxiliary thermal reservoir to the energy conversion system 944 may include a heat supply system suitable for supplying a specified percentage of the energy stored in the auxiliary thermal reservoir to the energy conversion system 946. For example, a heat supply system suitable for supplying a specified percentage of the energy stored in the auxiliary thermal reservoir to the energy conversion system 946 may be utilized by a control system or operator to transfer a chosen percentage (e.g., 50% of the stored energy) of the energy stored in the auxiliary thermal reservoir 112 to an energy conversion system 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems 102. Following are a series of flowcharts depicting implementations. For ease of understanding, the flowcharts are organized such that the initial flowcharts present implementations via an example implementation and thereafter the following flowcharts present alternate implementations and/or expansions of the initial flowchart(s) as either sub-component operations or additional component operations building on one or more earlier-presented flowcharts. Those having skill in the art will appreciate that the style of presentation utilized herein (e.g., beginning with a presentation of a flowchart(s) presenting an example implementation and thereafter providing additions to and/or further details in subsequent flowcharts) generally allows for a rapid and easy understanding of the various process implementations. In addition, those skilled in the art will further appreciate that the style of presentation used herein also lends itself well to modular and/or object-oriented program design paradigms. FIG. 10 illustrates an operational flow 1000 representing example operations related to the storage and utilization of energy generated by a plurality of nuclear reactor systems In FIG. 10 and in following figures that include various examples of operational flows, discussion and explanation may be provided with respect to the above-described examples of FIGS. 1 through 9D, and/or with respect to other examples and contexts. However, it should be understood that the operational flows may be executed in a number of other environments and contexts, and/or in modified versions of FIGS. 1 through 9D. Also, although the various operational flows are presented in the sequence(s) illustrated, it should be understood that the various operations may be performed in other orders than those which are illustrated, or may be performed concurrently. After a start operation, the operational flow 1000 moves to a first diverting operation 1010. The first diverting operation 1010 depicts diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, a first energy transfer system 104 may transfer energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to one or more auxiliary thermal reservoirs 112. Then, the additional diverting operation 1020 depicts diverting at least one additional selected portion of energy from a portion of at least one additional nuclear reactor system of the plurality of nuclear reactor systems to the at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, a second energy transfer system 104 may transfer energy from a portion of a second nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to the one or more auxiliary thermal reservoirs 112. More generally, an Nth energy transfer system 104 may transfer energy from a portion of an Nth nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to the one or more auxiliary thermal reservoirs 112. Then, the supplying operation 1030 depicts supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, one or more heat supply systems 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to one or more of energy conversion systems 110 of one or more of the nuclear reactor systems 106 of the plurality of nuclear reactor systems. FIG. 11 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 11 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 1102, and/or an operation 1104. Operation 1102 illustrates diverting at least a first portion of excess energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, a first energy transfer system 104 may transfer excess energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. Further, the operation 1104 illustrates diverting at least a first portion of energy exceeding operational demand of at least one energy conversion system from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, a first energy transfer system 104 may transfer energy exceeding operational demand (e.g., energy in excess of grid requirements) of an energy conversion system associated with a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 from a portion of the first nuclear reactor system 106 to an auxiliary thermal reservoir 112. FIG. 12 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 12 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 1202. Operation 1202 illustrates diverting a specified percentage of the energy output of a portion of a first nuclear reactor system of a plurality of nuclear reactor systems from a portion of the first nuclear reactor system of the plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, a first energy transfer system 104 may transfer a specified percentage of the energy output of a portion (e.g., nuclear reactor core or portion of nuclear reactor system in thermal communication with the nuclear reactor core, such as the primary coolant system) of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 from a portion of the first nuclear reactor system 106 to an auxiliary thermal reservoir 112. FIGS. 13A and 13B illustrate alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 13 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 1302, an operation 1304, an operation 1306, an operation 1308, an operation 1310, an operation 1312, and/or an operation 1314. Operation 1302 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one energy transfer system. For example, as shown in FIGS. 1 through 9D, a first energy transfer system 104 may transfer energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. Further, the operation 1304 illustrates diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one energy transfer system. For example, as shown in FIG. 5A, one or more of the energy transfer systems 104 may be suitable for transferring thermal energy 502. For instance, as shown in FIGS. through 9D, a first energy transfer system 104 may transfer thermal energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. Further, the operation 1306 illustrates diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one heat transfer system. For example, as shown in FIG. 5A, one or more of energy transfer systems 104 may include a heat transfer system 504. For instance, as shown in FIG. 1 through 9D, a first heat transfer system 504 may transfer thermal energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems to an auxiliary thermal reservoir 112. Further, the operation 1308 illustrates diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one heat transfer system, the portion of the first nuclear reactor in thermal communication with at least one heat source of the first nuclear reactor system. For example, as shown in FIG. 5B, heat may be transferred from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112, wherein the portion of the first nuclear reactor system 106 is in thermal communication with a heat source 522 of the first nuclear reactor system 106. For instance, as shown in FIGS. 1 through 9D, a first heat transfer system 504 may transfer thermal energy from a portion of a first nuclear reactor system 106 (e.g., coolant system of the nuclear reactor system) in thermal communication with a heat source 522 of the first nuclear reactor system 106 to an auxiliary thermal reservoir 112. Further, the operation 1310 illustrates diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one heat transfer system, the portion of the first nuclear reactor system in thermal communication with at least one nuclear reactor core of the first nuclear reactor system. For example, as shown in FIG. 5B, the heat source 522 of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 may include a nuclear reactor core 524. For instance, as shown in FIGS. 1 through 9D, a heat transfer system 504 of a first nuclear reactor system 106 may transfer thermal energy from a portion of the first nuclear reactor system 106 in thermal communication with the nuclear reactor core 524 of the first nuclear reactor system 106 to an auxiliary thermal reservoir 112. Further, the operation 1312 illustrates diverting a first selected portion of thermal energy from a portion of at least one primary coolant system of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one heat transfer system. For example, as shown in FIG. 5B, the portion of the first nuclear reactor system 106 in thermal communication with the nuclear reactor core 524 of the first nuclear reactor system 106 may include a portion of the primary coolant system 526 of the first nuclear reactor system 106. For instance, as shown in FIGS. 1 through 9D, a first heat transfer system 504 may transfer thermal energy from a portion of a primary coolant system 526 of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. Further, the operation 1314 illustrates diverting a first selected portion of thermal energy from a portion of at least one primary coolant loop of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one heat transfer system. For example, as shown in FIG. 5B, the portion of the primary coolant system of the first nuclear reactor system 106 may include a portion of a primary coolant loop 528 of the first nuclear reactor system 106. For instance, as shown in FIGS. 1 through 9D, a first heat transfer system 504 may transfer thermal energy from a portion of a primary coolant loop 528 of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to the auxiliary thermal reservoir 112. FIGS. 14A and 14B illustrate alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 14 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 1402. Operation 1402 illustrates diverting a first selected portion of thermal energy from at least one coolant pool of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one heat transfer system. For example, as shown in FIG. 5B, the portion of the primary coolant system of the first nuclear reactor system 106 may include a portion of a primary coolant pool 530, such as a liquid metal pool (e.g. liquid sodium) or a liquid metal salt pool (e.g., lithium fluoride pool), of the first nuclear reactor system 106. For instance, as shown in FIGS. 1 through 9D, a first heat transfer system 504 may transfer thermal energy from a portion of a primary coolant pool 530 of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to the auxiliary thermal reservoir 112. FIGS. 15A and 15B illustrate alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 15 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 1502. Operation 1502 illustrates diverting a first selected portion of thermal energy from a portion of at least one primary coolant system of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, the at least one primary coolant system of the first nuclear reactor system in thermal communication with the at least one auxiliary thermal reservoir and at least one secondary coolant system of the first nuclear reactor system, the at least one auxiliary thermal reservoir and the at least one secondary coolant system not in thermal communication. For example, as shown in FIG. 5C, the primary coolant system 526 of the first nuclear reactor system 106 may include a primary coolant system 526 in thermal communication with both an auxiliary thermal reservoir 112 and a secondary coolant system 532 of the first nuclear reactor system 106, wherein the auxiliary thermal reservoir 112 and the secondary coolant system 532 are not in thermal communication with each other. For instance, a first heat transfer system 504 may transfer thermal energy from a portion of a primary coolant system 526 of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112, wherein the primary coolant system 526 is in thermal communication with both the auxiliary thermal reservoir 112 and a secondary coolant system 532 of the first nuclear reactor system 106, while the auxiliary thermal reservoir 112 and the at least one secondary coolant 532 system are not in thermal communication. FIGS. 16A and 16B illustrate alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 16 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 1602. Further, operation 1602 illustrates diverting a first selected portion of thermal energy from a portion of at least one primary coolant system of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, the at least one auxiliary thermal reservoir in thermal communication with the at least one primary coolant system of the first nuclear reactor system and at least one secondary coolant system of the first nuclear reactor system. For example, as shown in FIG. 5D, the primary coolant system 526 of the first nuclear reactor system 106 may include a primary coolant system in thermal communication with both an auxiliary thermal reservoir 112 and a secondary coolant system 532 of the first nuclear reactor system 106, wherein the auxiliary thermal reservoir 112 is in thermal communication with the primary coolant system 526 of the nuclear reactor system 106 and the secondary coolant system 532 of the nuclear reactor system 106. For instance, a first heat transfer system 504 may transfer thermal energy from a portion of a primary coolant system 526 of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112, wherein the auxiliary thermal reservoir 112 is in thermal communication with both the primary coolant system 526 of the nuclear reactor system 106 and a secondary coolant system 532 of the nuclear reactor system 106. FIGS. 17A and 17B illustrate alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 17 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 1702, an operation 1704, and/or an operation 1706. Further, the operation 1702 illustrates diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one direct fluid exchange heat transfer system. For example, as shown in FIG. 5A, a first energy transfer system 104 of a first nuclear reactor system 106 may include a direct fluid exchange heat transfer system 510. For instance, as shown in FIGS. 1 through 9D, a first direct fluid exchange system 510 may transfer thermal energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. Further, the operation 1704 illustrates intermixing at least one reservoir fluid of at least one auxiliary thermal reservoir with at least one coolant of a first nuclear reactor system of a plurality of nuclear reactor systems using at least one direct fluid exchange heat transfer system. For example, as shown in FIG. 5A, a first direct fluid exchange system 510 of a first nuclear reactor system 106 may include a system configured to intermix 511 the reservoir fluid of an auxiliary thermal reservoir 112 and the coolant of a nuclear reactor 108 of the first nuclear reactor system 106. For instance, as shown in FIGS. 1 through 9D, a system for intermixing 511 the reservoir fluid of an auxiliary thermal reservoir 112 and the reactor coolant of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 may transfer thermal energy from the first nuclear reactor system 106 to the auxiliary thermal reservoir 112 by directly mixing the two fluids. Further, the operation 1706 illustrates intermixing at least one reservoir fluid of at least one auxiliary thermal reservoir with at least one coolant of a first nuclear reactor system of a plurality of nuclear reactor systems using at least one direct fluid exchange heat transfer system, the at least one reservoir fluid substantially similar to the at least one coolant. For example, as shown in FIG. 5A, the auxiliary thermal reservoir fluid and the coolant of the first nuclear reactor system 106 may be substantially similar 512. For instance, the reservoir fluid and the nuclear reactor coolant of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 may both comprise the same liquid metal, such as liquid sodium, liquid lead, or liquid lead bismuth. In another instance, the reservoir fluid and the nuclear reactor coolant may both comprise the same liquid organic, such as diphenyl with diphenyl oxide. FIGS. 18A and 18B illustrate alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 18 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 1802. Further, the operation 1802 illustrates intermixing at least one reservoir fluid of at least one auxiliary thermal reservoir with at least one coolant of at a first nuclear reactor system of a plurality of nuclear reactor systems using at least one direct fluid exchange heat transfer system, the at least one reservoir fluid different from the at least one coolant. For example, as shown in FIG. 5A, the auxiliary thermal reservoir fluid and the coolant of the first nuclear reactor system 106 may be different 513. For instance, the reservoir fluid may comprise a liquid organic fluid (e.g., diphenyl with diphenyl oxide), while the nuclear reactor coolant of a first nuclear reactor system 106 of a plurality of nuclear reactor systems may comprise a liquid metal coolant (e.g., liquid sodium, lead, or lead bismuth). Similarly, the reservoir fluid may comprise a first liquid metal coolant, such as liquid sodium, while the nuclear reactor coolant may comprise a second liquid metal coolant, such as liquid lead. FIGS. 19A and 19B illustrate alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 19 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 1902, and/or an operation 1904. Operation 1902 illustrates diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one heat exchanger. For example, as shown in FIG. 5A, a first energy transfer system configured to transfer thermal energy 502 may transfer thermal energy from a portion of the nuclear reactor system 101 to the auxiliary thermal reservoir 112 using one or more reactor-to-reservoir heat exchangers 514. Further, operation 1904 illustrates diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one heat exchanger, a first portion of the at least one heat exchanger in thermal communication with a portion of at least one primary coolant system of the first nuclear reactor system and a second portion of the at least one heat exchanger in thermal communication with a portion of the at least one auxiliary thermal reservoir. For example, the reactor-to-reservoir heat exchanger 514 may include a heat exchanger 515 having a first portion in communication with a primary coolant system of the first nuclear reactor system 106 and a second portion in thermal communication with an auxiliary thermal reservoir 112. For instance, the energy transfer system configured to transfer thermal energy 502 may transfer energy from the first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112 using a heat exchanger 515 having a first portion in communication with the primary coolant system of the first nuclear reactor system 106 and a second portion in thermal communication with the auxiliary thermal reservoir 112. FIG. 20 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 20 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 2002, an operation 2004, and/or an operation 2006. Operation 2002 illustrates diverting a first selected portion of electrical energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one energy transfer system. For example, as shown in FIG. 5A, a first energy transfer system 104 may include an energy transfer system configured to transfer electrical energy 503 from a portion of a first nuclear reactor system 106 (e.g., an energy conversion system 110 of the first nuclear reactor system 106) to an auxiliary thermal reservoir 112. For instance, an energy transfer system configured to transfer electrical energy 503 from a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112 may be used to transfer electrical energy from a portion of the first nuclear reactor system 106 to the auxiliary thermal reservoir 112. Those skilled in the art will recognize that in the transfer process the electrical energy originating from a portion of the first nuclear reactor system 106 must be converted to thermal energy in order to be stored in the auxiliary thermal reservoir 112. Further, the operation 2004 illustrates diverting a first selected portion of electrical energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one electrical-to-thermal conversion system. For example, as shown in FIG. 5A, the energy transfer system suitable for transferring electrical energy 503 from a first nuclear reactor system 106 to an auxiliary thermal reservoir 112 may include an electrical energy-to-thermal energy conversion device 516. For instance, an electrical energy-to-thermal energy conversion device 516 may be used to convert electrical energy produced by a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102. The thermal energy may then be transferred to the auxiliary thermal reservoir 112. Further, the operation 2006 illustrates diverting a first selected portion of electrical energy from at least one energy conversion system of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one electrical-to-thermal conversion system. For example, as shown in FIG. 5A, the energy transfer system suitable for transferring electrical energy 503 from a first nuclear reactor system 106 to an auxiliary thermal reservoir 112 may include an electrical energy-to-thermal energy conversion device configured to transfer electrical energy from an energy conversion device 110 of the first nuclear reactor system 106 to the auxiliary thermal reservoir 112. For instance, an electrical energy-to-thermal energy conversion device configured to transfer electrical energy from an energy conversion device 110 to the auxiliary thermal reservoir 112 may be used to convert electrical energy from the electrical output of an energy conversion device 110 (e.g., turbine-generator system) of the first nuclear reactor system 106 to thermal energy. The thermal energy may then be transferred to the auxiliary thermal reservoir 112. FIG. 21 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 21 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 2102, and/or an operation 2104. Operation 2102 illustrates diverting a first selected portion of electrical energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one resistive heating device. For example, as shown in FIG. 5A, the electrical energy-to-thermal energy conversion device may include one or more than one resistive heating devices 517. For instance, a resistive heating device 517 may be utilized to convert electrical energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to thermal energy. The thermal energy may then be transferred to an auxiliary thermal reservoir 112. Further, the operation 2104 illustrates diverting a first selected portion of electrical energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir using at least one heating coil. For example, as shown in FIG. 5A, the resistive heating device 517 may include one or more heating coils. For instance, a heating coil 518 may be used to convert electrical energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to thermal energy. The thermal energy may then be transferred to an auxiliary thermal reservoir 112. FIG. 22 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 22 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 2202, and/or an operation 2204. Operation 2202 illustrates, responsive to at least one condition, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a condition (e.g., power demands on a nuclear reactor system, state of readiness of auxiliary thermal reservoir, thermal properties of nuclear reactor or thermal properties of reservoir), an energy transfer system responsive to a condition 802 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. Further, the operation 2204 illustrates responsive to at least one condition of a first nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of the first nuclear reactor system of the plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a condition of a first nuclear reactor system, an energy transfer system responsive to a condition of the first nuclear reactor system 804 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 23 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 23 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 2302. Operation 2302 illustrates, responsive to at least one condition of at least one additional nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of the first nuclear reactor system of the plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a condition of an additional nuclear reactor system, an energy transfer system responsive to a condition of an additional nuclear reactor system 806, such as a 2nd nuclear reactor system, a 3rd nuclear reactor system, or up to and including an Nth nuclear reactor system, may initiate transfer of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 24 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 24 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 2402. Operation 2402 illustrates, responsive to determination of excess capacity of at least one nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of the plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to the determination of excess capacity of at least one nuclear reactor system 106 of a plurality of nuclear reactor systems 102 (e.g., determination that current nuclear reactor power production exceeds current grid demand), an energy transfer system responsive to the determination of excess nuclear reactor capacity 808 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 25 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 25 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 2502, and/or an operation 2504. Operation 2502 illustrates, responsive to at least one operation system of at least one nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of the plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to an operation system (e.g., warning system, security system, or shutdown system) of a nuclear reactor system 106 (e.g., first nuclear reactor system 106, second nuclear reactor system 106, or up to and including the Nth nuclear reactor system 106) of a plurality of nuclear reactor systems 102, an energy transfer system responsive to an operation system 810 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. Further, the operation 2504 illustrates, responsive to at least one monitoring system of at least one nuclear reactor system of a plurality of nuclear reactor systems, diverting first a selected portion of thermal energy from a portion of a first nuclear reactor system of the plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a monitoring system of a nuclear reactor system 106 (e.g., first nuclear reactor system 106, second nuclear reactor system 106, or up to and including the Nth nuclear reactor system 106) of a plurality of nuclear reactor systems 102, an energy transfer system responsive to a monitoring system 814 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 26 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 26 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 2602. Operation 2602 illustrates, responsive to at least one control system of at least one nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of the plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a control system of a nuclear reactor system 106 (e.g., first nuclear reactor system 106, second nuclear reactor system 106, or up to and including the Nth nuclear reactor system 106) of a plurality of nuclear reactor systems 102, an energy transfer system responsive to a control system 816 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 27 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 27 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 2702. Operation 2702 illustrates, responsive to at least one safety system of at least one nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of the plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a safety system of a nuclear reactor system 106 (e.g., first nuclear reactor system 106, second nuclear reactor system 106, or up to and including the Nth nuclear reactor system 106) of a plurality of nuclear reactor systems 102, an energy transfer system responsive to a safety system 818 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 28 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 28 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 2802, and/or an operation 2804. Operation 2802 illustrates, responsive to at least one signal from at least one operation system of at least one nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of the plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a signal (e.g., a digital wireline signal, an analog wireline signal, a digital wireless signal, or an analog wireless signal) from an operation system of a nuclear reactor system 106 (e.g., first nuclear reactor system 106, second nuclear reactor system 106, or up to and including the Nth nuclear reactor system 106) of a plurality of nuclear reactor systems 102, an energy transfer system responsive to a signal from an operation system 812 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. Further, the operation 2804 illustrates, responsive to at least one signal from at least one operation system of a first nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of the first nuclear reactor system of the plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a signal from an operation system of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102, an energy transfer system responsive to a signal from an operation system 812 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 29 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 29 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 2902. Operation 2902 illustrates, responsive to at least one signal from at least one operation system of at least one additional nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of the plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a signal from an operation system of an additional nuclear reactor system 106 (e.g., second nuclear reactor system 106, third nuclear reactor system, or up to an including the Nth nuclear reactor system 106) of a plurality of nuclear reactor systems 102, an energy transfer system responsive to a signal from an operation system 812 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 30 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 30 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 3002, and/or an operation 3004. Operation 3002 illustrates, responsive to at least one reservoir operation system of at least one auxiliary thermal reservoir, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to the at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a reservoir operation system (e.g., monitoring system, warning system, or control system) of an auxiliary thermal reservoir, an energy transfer system responsive to a reservoir operation system 824 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. Further, the operation 3004 illustrates, responsive to at least one signal from at least one reservoir operation system of at least one auxiliary thermal reservoir, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to the at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a signal (e.g., a digital wireline signal, an analog wireline signal, a digital wireless signal, or an analog wireless signal) from a reservoir operation system of an auxiliary thermal reservoir, an energy transfer system responsive to a signal from a reservoir operation system 826 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 31 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 31 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 3102. Operation 3102 illustrates, responsive to at least one reservoir monitoring system of at least one auxiliary thermal reservoir, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to the at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a reservoir monitoring system (e.g., thermal monitoring system) of an auxiliary thermal reservoir, an energy transfer system responsive to a reservoir monitoring system 828 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 32 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 32 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 3202. Operation 3202 illustrates, responsive to at least one signal from at least one operator of at least one nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to at least one signal from an operator of a nuclear reactor system 106 of a plurality of nuclear reactor systems 102, an energy transfer system responsive to a signal (e.g., wireless or wireline signal) from an operator 820 (e.g., human user or human controlled programmable computer system) may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 33 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 33 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 3302. Further, the operation 3302 illustrates, upon a preselected diversion start time, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, upon a preselected diversion start time (e.g., absolute time or time of elapse relative to the occurrence of a predetermined event), an energy transfer system responsive to a preselected diversion start time 822 may initiate the transfer of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 34 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 34 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 3402, and/or an operation 3404. Operation 3402 illustrates, responsive to a shutdown event of at least one nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a shutdown event of a nuclear reactor system 106 (e.g., first nuclear reactor system 106, second nuclear reactor system 106, or up to and including the Nth nuclear reactor system 106) of a plurality of nuclear reactor systems 102, an energy transfer system responsive to a shutdown event 834 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. Further, the operation 3404 illustrates, responsive to a scheduled shutdown event of at least one nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of the first nuclear reactor system of the plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a scheduled shutdown event (e.g., shutdown for routine maintenance) of a nuclear reactor system 106 of a plurality of nuclear reactor systems 102, an energy transfer system responsive to a scheduled shutdown event 836 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 35 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 35 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 3502. Operation 3502 illustrates, responsive to an emergency shutdown event of at least one nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of the first nuclear reactor system of the plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to an emergency shutdown event (e.g., SCRAM) of a nuclear reactor system 106 of a plurality of nuclear reactor systems 102, an energy transfer system responsive to an emergency shutdown event 838 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 36 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 36 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 3602. Operation 3602 illustrates, responsive to a shutdown event of a first nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of the first nuclear reactor system of the plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a shutdown event of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102, an energy transfer system responsive to a shutdown event 834 of the first nuclear reactor system 106 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 37 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 37 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 3702. Operation 3702 illustrates, responsive to a shutdown event of at least one additional nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of the plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a shutdown event of an additional nuclear reactor system 106 (e.g., the second nuclear reactor system, the third nuclear reactor system 106, or up to an including the Nth nuclear reactor system 106) of a plurality of nuclear reactor systems 102, an energy transfer system responsive to a shutdown event 834 of the additional nuclear reactor system 106 may initiate transfer of energy from a portion of the first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 38 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 38 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 3802. Further, the operation 3802 illustrates, responsive to a shutdown event of at least one nuclear reactor system of a plurality of nuclear reactor systems, establishing thermal communication between a portion of a first nuclear reactor system of the plurality of nuclear reactor systems and at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a shutdown event of a nuclear reactor system 106 of a plurality of nuclear reactor systems 102, an energy transfer system configured to establish thermal communication between a first nuclear reactor system and an auxiliary thermal reservoir 840 may establish thermal communication between a portion of the first nuclear reactor system (e.g., primary coolant system) of the plurality of nuclear reactor systems and the auxiliary thermal reservoir 112. FIG. 39 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 39 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 3902. Operation 3902 illustrates, preceding shutdown of at least one nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, prior to shutdown of a nuclear reactor 108 of a nuclear reactor system 106, an energy transfer system responsive to a shutdown event 834 of the nuclear reactor system 106 may initiate the transfer of energy from a portion of a nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to the auxiliary thermal reservoir 112. FIG. 40 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 40 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 4002, and/or an operation 4004. Operation 4002 illustrates, during shutdown of at least one nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, during shutdown of a nuclear reactor 108 of a nuclear reactor system 106, an energy transfer system responsive to a shutdown event 834 of the nuclear reactor system 106 may initiate the transfer of energy from a portion of a nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to the auxiliary thermal reservoir 112. Further, the operation 4004 illustrates, following shutdown of at least one nuclear reactor system of a plurality of nuclear reactor systems, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, following shutdown of a nuclear reactor 108 of a nuclear reactor system 106, an energy transfer system responsive to a shutdown event 834 of the nuclear reactor system 106 may initiate the transfer of energy from a portion of a nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to the auxiliary thermal reservoir 112. FIG. 41 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 41 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 4102, and/or an operation 4104. Operation 4102 illustrates, responsive to determination of the amount of energy stored in at least one auxiliary thermal reservoir, diverting a first selected portion of thermal energy from a portion of the first nuclear reactor system of the plurality of nuclear reactor systems to the at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system responsive to determination of the amount of energy stored in an auxiliary thermal reservoir 842 may initiate the transfer of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. Further, the operation 4104 illustrates, responsive to determination of the percentage of energy stored in at least one auxiliary thermal reservoir, diverting a first selected portion of thermal energy from a portion of the first nuclear reactor system of the plurality of nuclear reactor systems to the at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system responsive to determination of the percentage of energy stored in an auxiliary thermal reservoir 844 may initiate the transfer of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 42 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 42 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 4202, and/or an operation 4204. Operation 4202 illustrates, responsive to determination of the amount of available energy storage capacity of at least one auxiliary thermal reservoir, diverting a first selected portion of thermal energy from a portion of the first nuclear reactor system of the plurality of nuclear reactor systems to the at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system responsive to determination of the amount of available energy storage capacity of at least one auxiliary thermal reservoir 846 may initiate the transfer of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. Further, the operation 4204 illustrates, responsive to determination of the percentage of available energy storage capacity of at least one auxiliary thermal reservoir, diverting a first selected portion of thermal energy from a portion of the first nuclear reactor system of the plurality of nuclear reactor systems to the at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system responsive to determination of the percentage of available energy storage capacity of at least one auxiliary thermal reservoir 848 may initiate the transfer of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 43 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 43 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 4302, and/or an operation 4304. The operation 4302 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to a mass of at least one heat storage material of at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of heat storage material 111 of an auxiliary thermal reservoir 112. Further, the operation 4304 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to a mass of at least one solid heat storage material of at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of a solid heat storage material 414, such a solid object (e.g., solid ceramic object, solid metal object, or solid stone object) or a particulate solid (e.g., sand), of an auxiliary thermal reservoir 112. FIG. 44 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 44 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 4402, and/or an operation 4404. Operation 4402 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to a mass of at least one liquid heat storage material of at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of liquid material 402 (e.g., liquid metal, liquid metal salt, liquid organic, or liquid water) of an auxiliary thermal reservoir 112. Further, the operation 4404 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to a mass of at least one organic liquid heat storage material of at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of liquid organic material 404 (e.g., diphenyl with diphenyl oxide) of an auxiliary thermal reservoir 112. FIG. 45 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 45 illustrates example embodiments where the diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 4502. Further, the operation 4502 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to a mass of at least one liquid metal salt heat storage material of at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of liquid metal salt 406 (e.g., lithium fluoride) of an auxiliary thermal reservoir 112. FIG. 46 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 46 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 4602. Operation 4602 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to a mass of at least one liquid metal heat storage material of at least one auxiliary thermal reservoir.]. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of liquid metal 408 (e.g., sodium) of the auxiliary thermal reservoir 112. FIG. 47 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 47 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 4702. Operation 4702 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to a mass of liquid water of at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of liquid water 410 of an auxiliary thermal reservoir 112. FIG. 48 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 48 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 4802. Further, the operation 4802 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to a mass of at least one pressurized gaseous mass of material of at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of pressurized gaseous material 412 (e.g., pressurized helium or pressurized carbon dioxide) of the auxiliary thermal reservoir 112. FIG. 49 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 49 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 4902. Operation 4902 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to a mass of at least one mixed phase material of at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of a mixed phase material 420 (e.g., steam water-liquid water) of the auxiliary thermal reservoir 112. FIG. 50 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 50 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 5002. Further, the operation 5002 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to a mass of at least one material of at least one auxiliary thermal reservoir, the mass of at least one material having a phase transition within the operating temperature of the at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of a material having a phase transition within the operating temperature 422 of the auxiliary thermal reservoir 112. FIG. 51 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 51 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 5102, an operation 5104, and/or an operation 5106. Operation 5102 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to a mass of at least one heat storage material contained in a reservoir containment system. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of heat storage material 111 of an auxiliary thermal reservoir 112 contained in a reservoir containment system 424 (e.g., vessel). Further, the operation 5104 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to a mass of at least one heat storage material contained in at least one external vessel. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of heat storage material 111 of an auxiliary thermal reservoir 112 contained in an external vessel 426. Further, the operation 5106 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to a mass of at least one heat storage material contained in at least one external high pressure gas vessel. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of heat storage material 111 of an auxiliary thermal reservoir 112 contained in a high pressure gas vessel 430. For instance, the energy transfer system 104 may transfer a selected portion of energy from a portion a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of high pressurized gaseous helium contained in an external high pressure helium vessel. FIG. 52 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 52 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 5202. Operation 5202 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to a mass of at least one heat storage material contained in at least one external liquid vessel. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of heat storage material 111 of the auxiliary thermal reservoir 112 contained in an external liquid vessel 428. For instance, the energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of liquid water contained in an external water vessel. FIG. 53 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 53 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 5302. Further, the operation 5302 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to a mass of at least one heat storage material contained in at least one external liquid pool. For example, as shown in FIGS. 1 through 9D, the energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of liquid heat storage material 402 of the auxiliary thermal reservoir 112 contained in an external liquid pool 434. For instance, the energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to a mass of liquid sodium contained in an external liquid sodium pool. FIG. 54 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 54 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 5402. Operation 5402 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, the at least one auxiliary thermal reservoir storing the selected portion of energy in the form of a temperature change in at least one heat storage material of the auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112, where the auxiliary thermal reservoir stores the energy in the form of an increase in temperature of the heat storage material 436. For instance, the energy transferred to the auxiliary thermal reservoir 112 may cause a liquid heat storage material 402 to increase in temperature from 100° C. to 200° C. FIG. 55 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 55 illustrates example embodiments where the operation 1010 may include at least one additional operation. Additional operations may include an operation 5502, and/or an operation 5504. Operation 5502 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, the at least one auxiliary thermal reservoir storing the selected portion of energy in the form of a phase change in at least one heat storage material of the auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112, where the auxiliary thermal reservoir stores the energy in the form of a phase change in the heat storage material 438. For instance, the energy transferred to the auxiliary thermal reservoir 112 may cause a solid reservoir material to undergo a phase change into a liquid reservoir material, where the energy is stored in the reservoir material as a latent heat of transformation. Further, the operation 5504 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, the at least one auxiliary thermal reservoir storing the selected portion of energy in the form of a solid-liquid phase change in at least one heat storage material of the auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112, where the auxiliary thermal reservoir 112 stores the energy in the form of a solid-liquid phase change 440 (e.g., solid sodium-liquid sodium phase change). FIG. 56 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 56 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 5602. Operation 5602 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, the at least one auxiliary thermal reservoir storing the selected portion of energy in the form of a liquid-gas phase change in at least one heat storage material of the auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, the energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112, where the auxiliary thermal reservoir 112 stores the energy in the form of a liquid-gas phase change 442 (e.g., liquid water-steam water phase change). FIG. 57 illustrates an operational flow 5700 representing example operations related to the storage and utilization of energy generated by a plurality of nuclear reactor systems. FIG. 57 illustrates an example embodiment where the example operational flow 1000 of FIG. 10 may include at least one additional operation. Additional operations may include an operation 5710, and/or an operation 5712. After a start operation, a first diverting operation 1010, an additional diverting operation 1020, and a supplying operation 1030, the operational flow 5700 moves to a temperature maintaining operation 5710. Operation 5710 illustrates maintaining the temperature of at least one heat storage material of at least one auxiliary thermal reservoir above a selected temperature. For example, as shown in FIG. 4D, the temperature of a heat storage material 111 of an auxiliary thermal reservoir 112 may be maintained with a reservoir temperature control system 454 (e.g., thermostat). The operation 5712 illustrates maintaining the temperature of at least one heat storage material of at least one auxiliary thermal reservoir above the melting temperature of the at least one heat storage material. For example, as shown in FIG. 4D, the temperature of a heat storage material 111 of an auxiliary thermal reservoir 112 may be maintained with a reservoir temperature control system 454 above a specified temperature, such as the melting temperature of the heat storage material 111. FIG. 58 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 58 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 5802, and/or an operation 5804. The operation 5802 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, the first nuclear reactor system of the plurality of nuclear reactor systems having at least one liquid coolant. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first liquid cooled 302 nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. Further, the operation 5804 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, the first nuclear reactor system of the plurality of nuclear reactor systems having at least one liquid metal salt coolant. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first liquid metal salt (e.g., lithium fluoride or other fluoride salts) cooled 304 nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 59 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 59 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 5902. Operation 5902 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, the first nuclear reactor system of the plurality of nuclear reactor systems having at least one liquid metal coolant. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first liquid metal (e.g., liquid sodium or liquid lead) cooled 306 nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 60 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 60 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 6002. Operation 6002 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, the first nuclear reactor system of the plurality of nuclear reactor systems having at least one liquid organic coolant. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first liquid organic (e.g., diphenyl with diphenyl oxide) cooled 308 nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 61 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 61 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 6102. Operation 6102 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, the first nuclear reactor system of the plurality of nuclear reactor systems having at least one liquid water coolant. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first liquid water cooled 310 nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 62 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 62 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 6202, and/or an operation 6204. Operation 6202 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, the first nuclear reactor system of the plurality of nuclear reactor systems having at least one pressurized gas coolant. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first pressurized gas (e.g., pressurized helium or carbon dioxide) cooled 312 nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. The operation 6204 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, the first nuclear reactor system of the plurality of nuclear reactor systems having at least one mixed phase coolant. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first mixed phase (e.g., liquid water-steam water) cooled 314 nuclear reactor system 106 of a plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 63 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 63 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 6302, and/or an operation 6304. Operation 6302 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, at least one of the nuclear reactor systems of the plurality of nuclear reactor systems having a thermal spectrum nuclear reactor. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 202, where at least one of the nuclear reactor systems 106 (e.g., the first nuclear reactor system 106, the second nuclear reactor system 106, or up to and including the Nth nuclear reactor system 106) of the plurality of nuclear reactor systems 102 includes a thermal spectrum nuclear reactor 202. The operation 6304 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, at least one of the nuclear reactor systems of the plurality of nuclear reactor systems having a fast spectrum nuclear reactor. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 202, where at least one of the nuclear reactor systems 106 (e.g., the first nuclear reactor system 106, the second nuclear reactor system 106, or up to and including the Nth nuclear reactor system 106) of the plurality of nuclear reactor systems 102 includes a fast spectrum nuclear reactor 204. FIG. 64 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 64 illustrates example embodiments where the diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 6402, and/or an operation 6404. Operation 6402 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, at least one of the nuclear reactor systems of the plurality of nuclear reactor systems having a multi-spectrum nuclear reactor. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 202, where at least one of the nuclear reactor systems 106 (e.g., the first nuclear reactor system 106, the second nuclear reactor system 106, or up to and including the Nth nuclear reactor system 106) of the plurality of nuclear reactor systems 102 includes a multi-spectrum nuclear reactor 206. The operation 6404 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, at least one of the nuclear reactor systems of the plurality of nuclear reactor systems having a breeder nuclear reactor. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 202, where at least one of the nuclear reactor systems 106 (e.g., the first nuclear reactor system 106, the second nuclear reactor system 106, or up to and including the Nth nuclear reactor system 106) of the plurality of nuclear reactor systems 102 includes a breeder nuclear reactor 208. FIG. 65 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 65 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 6502. Operation 6502 illustrates diverting a first selected portion of energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to at least one auxiliary thermal reservoir, at least one of the nuclear reactor systems of the plurality of nuclear reactor systems having a traveling wave nuclear reactor. For example, as shown in FIGS. 1 through 9D, an energy transfer system 104 may transfer a selected portion of energy from a portion of a first nuclear reactor system 106 of a plurality of nuclear reactor systems 202, where at least one of the nuclear reactor systems 106 (e.g., the first nuclear reactor system 106, the second nuclear reactor system 106, or up to and including the Nth nuclear reactor system 106) of the plurality of nuclear reactor systems 102 includes a traveling wave nuclear reactor 210. FIG. 66 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 66 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 6602, and/or an operation 6604. Operation 6602 illustrates supplying at least a portion of thermal energy from a first auxiliary thermal reservoir and a portion of thermal energy from at least a second thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, thermal energy stored in a first auxiliary reservoir 112 and thermal energy stored in an additional thermal reservoir (e.g., second thermal reservoir, third thermal reservoir, or up to and including an Nth thermal reservoir) may be supplied to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. For instance, a first heat supply system 114 may supply thermal energy stored in the first auxiliary thermal reservoir 112 to an energy conversion system 110 and a second heat supply system 114 may supply thermal energy stored in the second auxiliary thermal reservoir 112 to the energy conversion system 110. Further, the operation 6604 illustrates supplying at least a portion of thermal energy from a first auxiliary thermal reservoir and a portion of thermal energy from at least a second thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems, the first auxiliary thermal reservoir and the at least a second thermal reservoir not in thermal communication. For example, as shown in FIGS. 1 through 9D, thermal energy stored in a first auxiliary reservoir 112 and thermal energy stored in an additional thermal reservoir (e.g., second thermal reservoir, third thermal reservoir, or up to and including an Nth thermal reservoir) may be supplied to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102, wherein the first thermal reservoir 112 and the second thermal reservoir 112 are not in thermal communication. FIG. 67 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 67 illustrates example embodiments where the supply operation 1030 may include at least one additional operation. Additional operations may include an operation 6702. Operation 6702 illustrates supplying at least a portion of thermal energy from a first auxiliary thermal reservoir and a portion of thermal energy from at least a second thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems, the first auxiliary thermal reservoir and the at least a second thermal reservoir in thermal communication. For example, as shown in FIGS. 1 through 9D, thermal energy stored in a first auxiliary reservoir 112 and thermal energy stored in an additional thermal reservoir (e.g., second thermal reservoir, third thermal reservoir, or up to and including an Nth thermal reservoir) may be supplied to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102, wherein the first thermal reservoir 112 and the second thermal reservoir 112 are in thermal communication. It will be recognized by those skilled in the art that even though the first thermal reservoir 110 and the second thermal reservoir 110 are thermally coupled the two reservoirs can for practical purposes be treated as two distinct thermal reservoirs under non-steady state conditions. FIG. 68 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 68 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 6802, and/or an operation 6804. The operation 6802 illustrates supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems using at least one heat supply system. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 (e.g., topping cycle 226 or turbine 218) of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. Further, the operation 6804 illustrates supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems using at least one heat exchange loop. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 via one or more heat exchange loops 602. FIG. 69 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 69 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 6902. Operation 6902 illustrates supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems using at least one heat exchange pipe. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 via one or more heat pipes 604. FIG. 70 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 70 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 7002. Operation 7002 illustrates supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems using at least one heat exchanger. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 via one or more heat exchangers 606. For instance, a first portion of a heat exchanger 606 may be in thermal communication with an auxiliary thermal reservoir 112, while the second portion of the heat exchanger 606 may be in thermal communication with an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 71 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 71 illustrates example embodiments where the supply operation 1030 may include at least one additional operation. Additional operations may include an operation 7102. Operation 7102 illustrates supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems using at least one thermoelectric device. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 via one or more thermoelectric devices 608. For instance, a first portion of a thermoelectric device 608 may be in thermal communication with an auxiliary thermal reservoir 112 and a second portion of the thermoelectric device 608 may be in thermal communication with a heat sink (e.g., environmental heat sink) of a nuclear reactor system 106 of the plurality of the nuclear reactor systems 102. FIG. 72 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 72 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 7202, and/or an operation 7204. Operation 7202 illustrates supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one primary energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to a primary energy conversion system 212 (e.g., energy conversion system coupled to the primary boiling loop) of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. The operation 7204 illustrates supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one auxiliary energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an auxiliary energy conversion system 214 (e.g., energy conversion system coupled to a non-primary boiling) of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 73 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 73 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 7302, and/or an operation 7304. Operation 7302 illustrates supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one emergency energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an emergency energy conversion system 216 (e.g., energy conversion system supplying electric power to various operation systems of the nuclear reactor system) of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. The operation 7304 illustrates supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one boiling loop of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to a boiling loop 232 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 74 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 74 illustrates example embodiments where the supply operation 1030 may include at least one additional operation. Additional operations may include an operation 7402, and/or an operation 7404. The operation 7402 illustrates supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one turbine of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to a turbine 218 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. Further, the operation 7404 illustrates [supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one working fluid of at least one turbine of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to the working fluid of a turbine 224 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 75 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 75 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 7502, and/or an operation 7504. The operation 7502 illustrates supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one low grade heat dump. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to a low grade heat dump 230 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. The operation 7504 illustrates supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one topping cycle. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to a topping cycle 226 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 76 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 76 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 7602, and/or an operation 7604. The operation 7602 illustrates supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one bottoming cycle. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to a bottoming cycle 228 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. The operation 7604 illustrates supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of the first nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 77 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 77 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 7702. The operation 7702 illustrates supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of the at least one additional nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system 114 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of an additional nuclear reactor system 106 (e.g., a second nuclear reactor system 106, a third nuclear reactor system 106 or up to and including an Nth nuclear reactor system 106) of the plurality of nuclear reactor systems 102. FIG. 78 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 78 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 7802, an operation 7804, and/or an operation 7806. The operation 7802 illustrates, responsive to at least one condition, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to a condition 902 (e.g., grid demand, thermal properties of one or more of the auxiliary thermal reservoirs) may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. Further, the operation 7804 illustrates, responsive to at least one condition of at least one nuclear reactor system of the plurality of nuclear reactor systems, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to a condition of one or more of the nuclear reactor systems 904 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. Further, the operation 7806 illustrates, responsive to heightened power demand on at least one nuclear reactor system of the plurality of nuclear reactor systems, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to heightened power demand of one or more of the nuclear reactor systems 906 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 79 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 79 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 7902, and/or an operation 7904. Operation 7902 illustrates, responsive to at least one operation system of at least one nuclear reactor system of the plurality of nuclear reactor systems, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to an operation system (e.g., monitoring system, control system, safety system, or security system) of a nuclear reactor system 922 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. Further, the operation 7904 illustrates, responsive to at least one signal from at least one operation system of at least one nuclear reactor system of the plurality of nuclear reactor systems, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to a signal (e.g., wireless or wireline) from an operation system of a nuclear reactor system 924 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 80 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 80 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 8002, and/or an operation 8004. Operation 8002 illustrates, responsive to at least one reservoir operation system of the at least one auxiliary thermal reservoir, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to a reservoir operation system 926 (e.g., reservoir monitoring system, reservoir control system, or reservoir safety system) may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. Further, the operation 8004 illustrates responsive to at least one signal from at least one reservoir operation system of at least one auxiliary thermal reservoir, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to a signal (e.g., wireless or wireline) from a reservoir operation system 928 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 81 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 81 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 8102, and/or an operation 8104. Operation 8102 illustrates, responsive to at least one operator of at least one nuclear reactor system of the plurality of nuclear reactor systems, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to an operator (e.g., human or human programmed computer control system) of a nuclear reactor system of the plurality of nuclear reactor systems 930 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. Further, the operation 8104 illustrates responsive to at least one signal from at least one operator of at least one nuclear reactor system of the plurality of nuclear reactor systems, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to a signal from an operator of a nuclear reactor system of the plurality of nuclear reactor systems 932 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 82 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 82 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 8202, and/or an operation 8204. Operation 8202 illustrates, responsive to a shutdown event of at least one nuclear reactor system of the plurality of nuclear reactor systems, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to a shutdown event of a nuclear reactor system of the plurality of nuclear reactor systems 908 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. Further, the operation 8204 illustrates, responsive to a scheduled shutdown event of at least one nuclear reactor system of the plurality of nuclear reactor systems, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to a scheduled shutdown event (e.g., shutdown for routine maintenance) of a nuclear reactor system of the plurality of nuclear reactor systems 912 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 83 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 83 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 8302. Operation 8302 illustrates, responsive to an emergency shutdown event of at least one nuclear reactor system of the plurality of nuclear reactor systems, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to an emergency shutdown event (e.g., SCRAM) of a nuclear reactor system of the plurality of nuclear reactor systems 910 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 84 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 84 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 8402. Operation 8402 illustrates, preceding shutdown of at least one nuclear reactor system of the plurality of nuclear reactor systems, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, preceding the shutdown of a nuclear reactor system 106, a heat supply system responsive to a shutdown event of a nuclear reactor system of the plurality of nuclear reactor systems 908 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 85 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 85 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 8502. Operation 8502 illustrates, following shutdown of at least one nuclear reactor system of the plurality of nuclear reactor systems, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, following the shutdown of a nuclear reactor system 106, a heat supply system responsive to a shutdown event of a nuclear reactor system of the plurality of nuclear reactor systems 908 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIGS. 86A and 86B illustrate alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 86 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 8602, an operation 8604, an operation 8606, and/or an operation 8608. Operation 8602 illustrates, responsive to a shutdown event established by at least one operation system of at least one nuclear reactor system of the plurality of nuclear reactor systems, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to a shutdown event of a nuclear reactor system of the plurality of nuclear reactor systems established by an operation system of a nuclear reactor system 914 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. Further, the operation 8604 illustrates, responsive to a shutdown event established by at least one reactor control system of at least one nuclear reactor system of the plurality of nuclear reactor systems, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to a shutdown event of a nuclear reactor system of the plurality of nuclear reactor systems established by a reactor control system of a nuclear reactor system 916 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. Further, the operation 8606 illustrates, responsive to a shutdown event established by at least one reactor control system responsive to at least one signal from at least one safety system of at least one nuclear reactor system of the plurality of nuclear reactor systems, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to a shutdown event of a nuclear reactor system of the plurality of nuclear reactor systems established by a reactor control system that is responsive to a safety system of a nuclear reactor system 918 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. Further, the operation 8608 illustrates, responsive to a shutdown event established by at least one reactor control system responsive to at least one signal from at least one safety system of at least one nuclear reactor system of the plurality of nuclear reactor systems, the safety system responsive to at least one sensed condition, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system 920 responsive to a shutdown event of a nuclear reactor system established by a reactor control system that is responsive to a safety system, where the safety system is responsive to a sensed condition (e.g., external condition or internal condition) of a nuclear reactor system, may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 87 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 87 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 8702. Operation 8702 illustrates, upon a pre-selected supply start time, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to the elapse of a preselected supply start time 934 (e.g., time of elapse measured relative to the initiation of a nuclear reactor system or system shutdown event or absolute time) may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 88 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 88 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 8802, and/or an operation 8804. Further, the operation 8802 illustrates, responsive to determination of the amount of energy stored in at least one auxiliary thermal reservoir, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to determination of the amount of energy stored in an auxiliary thermal reservoir 936 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. Further, the operation 8804 illustrates, responsive to determination of the percentage of energy stored in at least one auxiliary thermal reservoir, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to determination of the percentage of energy stored in an auxiliary thermal reservoir 938 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 89 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 89 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 8902, and/or an operation 8904. Operation 8902 illustrates, responsive to determination of the amount of available energy storage capacity of at least one auxiliary thermal reservoir, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to determination of the amount of available energy storage capacity of an auxiliary thermal reservoir 940 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. Further, the operation 8904 illustrates, responsive to determination of the percentage of available energy storage capacity of at least one auxiliary thermal reservoir, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system responsive to determination of the percentage of available energy storage capacity of an auxiliary thermal reservoir 942 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 90 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 90 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 9002, and/or an operation 9004. The operation 9002 illustrates supplying a specified portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system suitable for supplying a specified portion of the energy stored in an auxiliary thermal reservoir to an energy conversion system 944 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 may initiate the transfer of a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. Further, the operation 9004 illustrates supplying a specified percentage of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, a heat supply system suitable for supplying a specified percentage of the energy stored in an auxiliary thermal reservoir to an energy conversion system 946 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102 may initiate the transfer of a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 91 illustrates an operational flow 9100 representing example operations related to storage and utilization of energy generated by a plurality of nuclear reactor systems. FIG. 91 illustrates an example embodiment where the example operational flow 1000 of FIG. 10 may include at least one additional operation. Additional operations may include an operation 9110, and/or an operation 9112. After a start operation, a first diverting operation 1010, an additional diverting operation 1020, and a supplying operation 1030, the operational flow 9100 moves to a supplementing operation 9110. Operation 9110 illustrates supplementing the at least one auxiliary thermal reservoir with an additional portion of thermal energy from at least one additional energy source. For example, as shown in FIG. 7, the thermal energy stored in an auxiliary thermal reservoir 112 may be supplemented with an additional portion of energy transferred from an additional energy source 702, such as a non-nuclear energy source (e.g., coal powered generator, diesel powered generator, or solar cell array) via a supplementary energy transfer system 706. The operation 9112 illustrates supplementing the at least one auxiliary thermal reservoir with an additional portion of energy from at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIG. 7, the thermal energy stored in an auxiliary thermal reservoir 112 may be supplemented with an additional portion of energy transferred from an energy conversion device 110 of a nuclear reactor system of the plurality of nuclear reactor systems 102 the via a supplementary energy transfer system 706. FIG. 92 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 92 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 9202. Operation 9202 illustrates, responsive to at least one reservoir control system of at least one auxiliary thermal reservoir, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to the at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a reservoir control system (e.g., thermal control system) of an auxiliary thermal reservoir, an energy transfer system responsive to a reservoir control system 830 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 93 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 93 illustrates example embodiments where the first diverting operation 1010 may include at least one additional operation. Additional operations may include an operation 9302. Operation 9302 illustrates, responsive to at least one reservoir safety system of at least one auxiliary thermal reservoir, diverting a first selected portion of thermal energy from a portion of a first nuclear reactor system of a plurality of nuclear reactor systems to the at least one auxiliary thermal reservoir. For example, as shown in FIGS. 1 through 9D, in response to a reservoir safety system of an auxiliary thermal reservoir, an energy transfer system responsive to a reservoir safety system 832 may initiate transfer of energy from a portion of a first nuclear reactor system 106 of the plurality of nuclear reactor systems 102 to an auxiliary thermal reservoir 112. FIG. 94 illustrates alternative embodiments of the example operational flow 1000 of FIG. 10. FIG. 94 illustrates example embodiments where the supplying operation 1030 may include at least one additional operation. Additional operations may include an operation 9402. Further, the operation 9402 illustrates, during shutdown of at least one nuclear reactor system of the plurality of nuclear reactor systems, supplying at least a portion of thermal energy from the at least one auxiliary thermal reservoir to at least one energy conversion system of at least one nuclear reactor system of the plurality of nuclear reactor systems. For example, as shown in FIGS. 1 through 9D, during the shutdown of a nuclear reactor system 106, a heat supply system responsive to a shutdown event of a nuclear reactor system of the plurality of nuclear reactor systems 908 may supply a portion of the thermal energy stored in an auxiliary thermal reservoir 112 to an energy conversion system 110 of a nuclear reactor system 106 of the plurality of nuclear reactor systems 102. FIG. 95 illustrates an operational flow 9500 representing example operations related to storage and utilization of energy generated by a plurality of nuclear reactor systems. FIG. 95 illustrates an example embodiment where the example operational flow 1000 of FIG. 10 may include at least one additional operation. Additional operations may include an operation 9510, an operation 9512, and/or an operation 9514. After a start operation, a first diverting operation 1010, an additional diverting operation 1020, and a supplying operation 1030, the operational flow 9500 moves to a monitoring operation 9510. Operation 9510 illustrates monitoring at least one condition of the at least one auxiliary thermal reservoir. For example, as shown in FIG. 4D, a condition, such as the operational status (e.g., state of readiness, temperature pressure, or storage capacity), of an auxiliary thermal reservoir 112 may be monitored. Further, the operation 9512 illustrates monitoring at least one condition of the at least one auxiliary thermal reservoir using at least one reservoir monitoring system. For example, as shown in FIG. 4D, a reservoir monitoring system 444 maybe used to monitor a condition of an auxiliary thermal reservoir 112. Further, the operation 9514 illustrates monitoring the temperature of the at least one auxiliary thermal reservoir. For example, as shown in FIG. 4D, a reservoir temperature monitoring system 446 maybe used to monitor the temperature of an auxiliary thermal reservoir 112. FIG. 96 illustrates alternative embodiments of the example operational flow 9500 of FIG. 95. FIG. 96 illustrates example embodiments where the monitoring operation 9510 may include at least one additional operation. Additional operations may include an operation 9602, and/or an operation 9604. The operation 9602 illustrates monitoring the pressure of the at least one auxiliary thermal reservoir. For example, as shown in FIG. 4D, a reservoir pressure monitoring system 448 maybe used to monitor the pressure of an auxiliary thermal reservoir 112. The operation 9604 illustrates determining the amount of energy stored in the at least one auxiliary thermal reservoir. For example, as shown in FIG. 4D, a system configured to determine the amount of stored energy 450 in an auxiliary thermal reservoir 112 may be utilized to monitor the energy storage level in the auxiliary thermal reservoir 112. FIG. 97 illustrates alternative embodiments of the example operational flow 9500 of FIG. 95. FIG. 97 illustrates example embodiments where the monitoring operation 9510 may include at least one additional operation. Additional operations may include an operation 9702. The operation 9702 illustrates determining the amount of available energy storage capacity in the at least one auxiliary thermal reservoir. For example, as shown in FIG. 4D, a system configured to determine the amount of available energy storage capacity 452 in an auxiliary thermal reservoir 112 may be utilized to monitor the available energy storage capacity of the auxiliary thermal reservoir 112. Those having skill in the art will recognize that the state of the art has progressed to the point where there is little distinction left between hardware, software, and/or firmware implementations of aspects of systems; the use of hardware, software, and/or firmware is generally (but not always, in that in certain contexts the choice between hardware and software can become significant) a design choice representing cost vs. efficiency tradeoffs. Those having skill in the art will appreciate that there are various vehicles by which processes and/or systems and/or other technologies described herein can be effected (e.g., hardware, software, and/or firmware), and that the preferred vehicle will vary with the context in which the processes and/or systems and/or other technologies are deployed. For example, if an implementer determines that speed and accuracy are paramount, the implementer may opt for a mainly hardware and/or firmware vehicle; alternatively, if flexibility is paramount, the implementer may opt for a mainly software implementation; or, yet again alternatively, the implementer may opt for some combination of hardware, software, and/or firmware. Hence, there are several possible vehicles by which the processes and/or devices and/or other technologies described herein may be effected, none of which is inherently superior to the other in that any vehicle to be utilized is a choice dependent upon the context in which the vehicle will be deployed and the specific concerns (e.g., speed, flexibility, or predictability) of the implementer, any of which may vary. Those skilled in the art will recognize that optical aspects of implementations will typically employ optically-oriented hardware, software, and or firmware. In some implementations described herein, logic and similar implementations may include software or other control structures. Electronic circuitry, for example, may have one or more paths of electrical current constructed and arranged to implement various functions as described herein. In some implementations, one or more media may be configured to bear a device-detectable implementation when such media hold or transmit device-detectable instructions operable to perform as described herein. In some variants, for example, implementations may include an update or modification of existing software or firmware, or of gate arrays or programmable hardware, such as by performing a reception of or a transmission of one or more instructions in relation to one or more operations described herein. Alternatively or additionally, in some variants, an implementation may include special-purpose hardware, software, firmware components, and/or general-purpose components executing or otherwise invoking special-purpose components. Specifications or other implementations may be transmitted by one or more instances of tangible transmission media as described herein, optionally by packet transmission or otherwise by passing through distributed media at various times. Alternatively or additionally, implementations may include executing a special-purpose instruction sequence or invoking circuitry for enabling, triggering, coordinating, requesting, or otherwise causing one or more occurrences of virtually any functional operations described herein. In some variants, operational or other logical descriptions herein may be expressed as source code and compiled or otherwise invoked as an executable instruction sequence. In some contexts, for example, implementations may be provided, in whole or in part, by source code, such as C++, or other code sequences. In other implementations, source or other code implementation, using commercially available and/or techniques in the art, may be compiled/implemented/translated/converted into a high-level descriptor language (e.g., initially implementing described technologies in C or C++ programming language and thereafter converting the programming language implementation into a logic-synthesizable language implementation, a hardware description language implementation, a hardware design simulation implementation, and/or other such similar mode(s) of expression). For example, some or all of a logical expression (e.g., computer programming language implementation) may be manifested as a Verilog-type hardware description (e.g., via Hardware Description Language (HDL) and/or Very High Speed Integrated Circuit Hardware Descriptor Language (VHDL)) or other circuitry model which may then be used to create a physical implementation having hardware (e.g., an Application Specific Integrated Circuit). Those skilled in the art will recognize how to obtain, configure, and optimize suitable transmission or computational elements, material supplies, actuators, or other structures in light of these teachings. The foregoing detailed description has set forth various embodiments of the devices and/or processes via the use of block diagrams, flowcharts, and/or examples. Insofar as such block diagrams, flowcharts, and/or examples contain one or more functions and/or operations, it will be understood by those within the art that each function and/or operation within such block diagrams, flowcharts, or examples can be implemented, individually and/or collectively, by a wide range of hardware, software, firmware, or virtually any combination thereof. In one embodiment, several portions of the subject matter described herein may be implemented via Application Specific Integrated Circuits (ASICs), Field Programmable Gate Arrays (FPGAs), digital signal processors (DSPs), or other integrated formats. However, those skilled in the art will recognize that some aspects of the embodiments disclosed herein, in whole or in part, can be equivalently implemented in integrated circuits, as one or more computer programs running on one or more computers (e.g., as one or more programs running on one or more computer systems), as one or more programs running on one or more processors (e.g., as one or more programs running on one or more microprocessors), as firmware, or as virtually any combination thereof, and that designing the circuitry and/or writing the code for the software and or firmware would be well within the skill of one of skill in the art in light of this disclosure. In addition, those skilled in the art will appreciate that the mechanisms of the subject matter described herein are capable of being distributed as a program product in a variety of forms, and that an illustrative embodiment of the subject matter described herein applies regardless of the particular type of signal bearing medium used to actually carry out the distribution. Examples of a signal bearing medium include, but are not limited to, the following: a recordable type medium such as a floppy disk, a hard disk drive, a Compact Disc (CD), a Digital Video Disk (DVD), a digital tape, a computer memory, etc.; and a transmission type medium such as a digital and/or an analog communication medium (e.g., a fiber optic cable, a waveguide, a wired communications link, a wireless communication link (e.g., transmitter, receiver, transmission logic, reception logic, etc.), etc.). In a general sense, those skilled in the art will recognize that the various embodiments described herein can be implemented, individually and/or collectively, by various types of electro-mechanical systems having a wide range of electrical components such as hardware, software, firmware, and/or virtually any combination thereof; and a wide range of components that may impart mechanical force or motion such as rigid bodies, spring or torsional bodies, hydraulics, electro-magnetically actuated devices, and/or virtually any combination thereof. Consequently, as used herein “electro-mechanical system” includes, but is not limited to, electrical circuitry operably coupled with a transducer (e.g., an actuator, a motor, a piezoelectric crystal, a Micro Electro Mechanical System (MEMS), etc.), electrical circuitry having at least one discrete electrical circuit, electrical circuitry having at least one integrated circuit, electrical circuitry having at least one application specific integrated circuit, electrical circuitry forming a general purpose computing device configured by a computer program (e.g., a general purpose computer configured by a computer program which at least partially carries out processes and/or devices described herein, or a microprocessor configured by a computer program which at least partially carries out processes and/or devices described herein), electrical circuitry forming a memory device (e.g., forms of memory (e.g., random access, flash, read only, etc.)), electrical circuitry forming a communications device (e.g., a modem, communications switch, optical-electrical equipment, etc.), and/or any non-electrical analog thereto, such as optical or other analogs. Those skilled in the art will also appreciate that examples of electro-mechanical systems include but are not limited to a variety of consumer electronics systems, medical devices, as well as other systems such as motorized transport systems, factory automation systems, security systems, and/or communication/computing systems. Those skilled in the art will recognize that electro-mechanical as used herein is not necessarily limited to a system that has both electrical and mechanical actuation except as context may dictate otherwise. In a general sense, those skilled in the art will recognize that the various aspects described herein which can be implemented, individually and/or collectively, by a wide range of hardware, software, firmware, and/or any combination thereof can be viewed as being composed of various types of “electrical circuitry.” Consequently, as used herein “electrical circuitry” includes, but is not limited to, electrical circuitry having at least one discrete electrical circuit, electrical circuitry having at least one integrated circuit, electrical circuitry having at least one application specific integrated circuit, electrical circuitry forming a general purpose computing device configured by a computer program (e.g., a general purpose computer configured by a computer program which at least partially carries out processes and/or devices described herein, or a microprocessor configured by a computer program which at least partially carries out processes and/or devices described herein), electrical circuitry forming a memory device (e.g., forms of memory (e.g., random access, flash, read only, etc.)), and/or electrical circuitry forming a communications device (e.g., a modem, communications switch, optical-electrical equipment, etc.). Those having skill in the art will recognize that the subject matter described herein may be implemented in an analog or digital fashion or some combination thereof. Those skilled in the art will recognize that at least a portion of the devices and/or processes described herein can be integrated into a data processing system. Those having skill in the art will recognize that a data processing system generally includes one or more of a system unit housing, a video display device, memory such as volatile or non-volatile memory, processors such as microprocessors or digital signal processors, computational entities such as operating systems, drivers, graphical user interfaces, and applications programs, one or more interaction devices (e.g., a touch pad, a touch screen, an antenna, etc.), and/or control systems including feedback loops and control motors (e.g., feedback for sensing position and/or velocity; control motors for moving and/or adjusting components and/or quantities). A data processing system may be implemented utilizing suitable commercially available components, such as those typically found in data computing/communication and/or network computing/communication systems. One skilled in the art will recognize that the herein described components (e.g., operations), devices, objects, and the discussion accompanying them are used as examples for the sake of conceptual clarity and that various configuration modifications are contemplated. Consequently, as used herein, the specific exemplars set forth and the accompanying discussion are intended to be representative of their more general classes. In general, use of any specific exemplar is intended to be representative of its class, and the non-inclusion of specific components (e.g., operations), devices, and objects should not be taken limiting. Although a user is shown/described herein as a single illustrated figure, those skilled in the art will appreciate that the user may be representative of a human user, a robotic user (e.g., computational entity), and/or substantially any combination thereof (e.g., a user may be assisted by one or more robotic agents) unless context dictates otherwise. Those skilled in the art will appreciate that, in general, the same may be said of “sender” and/or other entity-oriented terms as such terms are used herein unless context dictates otherwise. With respect to the use of substantially any plural and/or singular terms herein, those having skill in the art can translate from the plural to the singular and/or from the singular to the plural as is appropriate to the context and/or application. The various singular/plural permutations are not expressly set forth herein for sake of clarity. The herein described subject matter sometimes illustrates different components contained within, or connected with, different other components. It is to be understood that such depicted architectures are merely exemplary, and that in fact many other architectures may be implemented which achieve the same functionality. In a conceptual sense, any arrangement of components to achieve the same functionality is effectively “associated” such that the desired functionality is achieved. Hence, any two components herein combined to achieve a particular functionality can be seen as “associated with” each other such that the desired functionality is achieved, irrespective of architectures or intermedial components. Likewise, any two components so associated can also be viewed as being “operably connected”, or “operably coupled,” to each other to achieve the desired functionality, and any two components capable of being so associated can also be viewed as being “operably couplable,” to each other to achieve the desired functionality. Specific examples of operably couplable include but are not limited to physically mateable and/or physically interacting components, and/or wirelessly interactable, and/or wirelessly interacting components, and/or logically interacting, and/or logically interactable components. In some instances, one or more components may be referred to herein as “configured to,” “configurable to,” “operable/operative to,” “adapted/adaptable,” “able to,” “conformable/conformed to,” etc. Those skilled in the art will recognize that such terms (e.g., “configured to”) can generally encompass active-state components and/or inactive-state components and/or standby-state components, unless context requires otherwise. While particular aspects of the present subject matter described herein have been shown and described, it will be apparent to those skilled in the art that, based upon the teachings herein, changes and modifications may be made without departing from the subject matter described herein and its broader aspects and, therefore, the appended claims are to encompass within their scope all such changes and modifications as are within the true spirit and scope of the subject matter described herein. It will be understood by those within the art that, in general, terms used herein, and especially in the appended claims (e.g., bodies of the appended claims) are generally intended as “open” terms (e.g., the term “including” should be interpreted as “including but not limited to,” the term “having” should be interpreted as “having at least,” the term “includes” should be interpreted as “includes but is not limited to,” etc.). It will be further understood by those within the art that if a specific number of an introduced claim recitation is intended, such an intent will be explicitly recited in the claim, and in the absence of such recitation no such intent is present. For example, as an aid to understanding, the following appended claims may contain usage of the introductory phrases “at least one” and “one or more” to introduce claim recitations. However, the use of such phrases should not be construed to imply that the introduction of a claim recitation by the indefinite articles “a” or “an” limits any particular claim containing such introduced claim recitation to claims containing only one such recitation, even when the same claim includes the introductory phrases “one or more” or “at least one” and indefinite articles such as “a” or “an” (e.g., “a” and/or “an” should typically be interpreted to mean “at least one” or “one or more”); the same holds true for the use of definite articles used to introduce claim recitations. In addition, even if a specific number of an introduced claim recitation is explicitly recited, those skilled in the art will recognize that such recitation should typically be interpreted to mean at least the recited number (e.g., the bare recitation of “two recitations,” without other modifiers, typically means at least two recitations, or two or more recitations). Furthermore, in those instances where a convention analogous to “at least one of A, B, and C, etc.” is used, in general such a construction is intended in the sense one having skill in the art would understand the convention (e.g., “a system having at least one of A, B, and C” would include but not be limited to systems that have A alone, B alone, C alone, A and B together, A and C together, B and C together, and/or A, B, and C together, etc.). In those instances where a convention analogous to “at least one of A, B, or C, etc.” is used, in general such a construction is intended in the sense one having skill in the art would understand the convention (e.g., “a system having at least one of A, B, or C” would include but not be limited to systems that have A alone, B alone, C alone, A and B together, A and C together, B and C together, and/or A, B, and C together, etc.). It will be further understood by those within the art that typically a disjunctive word and/or phrase presenting two or more alternative terms, whether in the description, claims, or drawings, should be understood to contemplate the possibilities of including one of the terms, either of the terms, or both terms unless context dictates otherwise. For example, the phrase “A or B” will be typically understood to include the possibilities of “A” or “B” or “A and B. With respect to the appended claims, those skilled in the art will appreciate that recited operations therein may generally be performed in any order. Also, although various operational flows are presented in a sequence(s), it should be understood that the various operations may be performed in other orders than those which are illustrated, or may be performed concurrently. Examples of such alternate orderings may include overlapping, interleaved, interrupted, reordered, incremental, preparatory, supplemental, simultaneous, reverse, or other variant orderings, unless context dictates otherwise. Furthermore, terms like “responsive to,” “related to,” or other past-tense adjectives are generally not intended to exclude such variants, unless context dictates otherwise. |
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abstract | According to one embodiment, reactor oscillation power ranges monitor includes: a receiving unit which receives LPRM signals; an exclusion processing unit which searches the LPRM signals allocated to the cell for an LPRM signal corresponding to an exceptional condition; an averaging unit which averages the allocated LPRM signals; a time averaging unit which calculates a time average of the average flux value; a normalized value calculation unit which divides the average flux value by the time averaged flux value; an initialization unit which outputs an initialization signal identifying the cell allocated to an LPRM signal which is changed to correspond or not correspond to the exceptional condition; and a determination unit which derives at least one of amplitude and cycle of a power oscillation from the normalized value. |
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claims | 1. An apparatus, comprising:a first spooler section;a segmented rolling floor comprising a first end attached to said first spooler section, said first spooler section configured to wind up a section of said segmented rolling floor during use; anda second spooler section attached to a second end of said segmented rolling floor. 2. The apparatus of claim 1, further comprising:a nozzle system penetrating through an aperture in said segmented rolling floor. 3. The apparatus of claim 2, further comprising:a beam path from an accelerator, along a beam transport line, though a rotatable section of said beam transport line, and through said nozzle system. 4. The apparatus of claim 3, further comprising:a gantry attached to said rotatable section of said beam transport line; anda mechanical connection forcing movement of said segmented rolling floor with rotation of said gantry. 5. The apparatus of claim 3, said segmented rolling floor further comprising:an upper section directly above a patient position, said patient position in said beam path. 6. The apparatus of claim 5, said segmented rolling floor further comprising:a vertical section at least five feet in length. 7. The apparatus of claim 5, further comprising:an X-ray imaging system penetrating through said segmented rolling floor. 8. The apparatus of claim 1, said first spooler section positioned above a plane of a floor coplanar with a walkable plane surface section of said segmented rolling floor, said second spooler section positioned below the plane of the floor. 9. An apparatus, comprising:a first spooler section;a segmented rolling floor comprising a first end attached to said first spooler section and a curved rolling wall section, said first spooler section configured to wind up a section of said segmented rolling floor during use. 10. A method, comprising the steps of:providing a segmented rolling floor; andspooling said segmented rolling floor onto a first rolling floor spool, said first rolling floor spool attached to a first end of said segmented rolling floor; andunwinding said segmented rolling floor from a second rolling floor spool, attached to a second end of said segmented rolling floor, during said step of spooling. 11. The method of claim 10, further comprising the step of:positioning a nozzle end of a nozzle system through an aperture in said segmented rolling floor. 12. The method of claim 11, further comprising the step of:co-moving said nozzle system and said segmented rolling floor. 13. The method of claim 12, further comprising the step of:sequentially transporting positively charged particles from an accelerator, along a rotatable beam transport line, and through said nozzle system. 14. The method of claim 13, further comprising the step of:a gantry rotating said rotatable beam transport line at a rate of said step of spooling said segment rolling floor. |
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description | Now, Embodiment 1 of the invention will be described with reference to the accompanying drawings. FIG. 1 is a process flowchart of a lithography pattern data generation method of Embodiment 1, and FIGS. 2(a) through 2(c) show the layout of design data in respective procedures in the lithography pattern data generation method of this embodiment. First, in a design data preparing process ST01 of FIG. 1, plural design data 11 through 16 corresponding to design patterns to be formed on a substrate are prepared as is shown in FIG. 2(a). These design data 11 through 16 are arranged in a data arranging area 10 correspondingly to a design pattern formation area. Next, in a first area dividing process ST02 of FIG. 1, the data arranging area 10 is divided into a first stripe area group 61 including three stripe areas 61a through 61c. Each of the stripe areas 61a through 61c has a width corresponding to, for example, that of a partial exposure area of the exposure beam of an electron beam lithography system, and hence, the width is controlled in accordance with the maximum deflection width of a primary deflection device of the electron beam lithography system. Herein, the first stripe area group 61 has a stripe width of 5 mm. Then, in a first data group extracting process ST03 of FIG. 1, design data each falling within any of the stripe areas 61a through 61c, namely, design data not extending over any boundary of the stripe areas 61a through 61c, are extracted from the plural design data 11 through 16 as a first design data group. Accordingly, in FIG. 2(b), since each of the design data 11 through 15 falls within any of the stripe areas 61a through 61c, the first design data group consists of the design data 11 through 15. Subsequently, first lithography pattern data 1 of each stripe area of the first stripe area group 61 are generated from the design data 11 through 15 belonging to the first design data group. Next, in a second data group extracting process ST04 of FIG. 1, design data each extending over plural stripe areas of the first stripe area group 61 are extracted as a second design data group. In FIG. 2(b), since the design data 16 extends over the adjacent stripe areas 61a and 61b, the second design data group consists of the design data 16. For the sake of explanation, the design data 11 through 16 are herein subjected to the two extracting processes, that is, the first design data group extracting process ST03 and the second design data group extracting process ST04, but needless to say, when one of the first and second design data groups is extracted from the design data 11 through 16, the other data group is naturally determined. Next, in a second area dividing process ST05 of FIG. 1, the data arranging area 10 is divided, as is shown in FIG. 2(c), to obtain a second stripe area 62, which can cover the design data 16 and is differently divided from the first stripe are a group 61. A second lithography pattern data 2 is generated from the design data 16 falling within the second stripe area 62. Herein, the width of the second stripe area 62 can be the same as or different from the stripe width of the first stripe area group 61. Thus, among the design data 11 through 16, the design data 16 extending over the boundary in the first stripe area group 61 falls within the second stripe area 62 different from the first stripe area group 61 in this embodiment. Therefore, the second lithography pattern data 2 generated from the design data 16 is free from a connection error. As a result, an exposed pattern obtained based on the second lithography pattern data 2 can be improved in the accuracy. Also, the exposure beam of charged particles is used as an electron beam in this embodiment, but the generation of lithography pattern data does not depend upon the kind of charged particles. Specifically, the invention is effective in generating lithography pattern data when the drawing (exposure) area is smaller than a design pattern formation area and hence the lithography pattern data corresponding to design patterns should be divided into plural drawing areas. Furthermore, it is assumed in this embodiment that the lithography pattern data are generated for drawing design patterns on a semiconductor substrate of a semiconductor integrated circuit device, but the invention is not limited to this application but is effectively applicable to a mask pattern of an exposure mask used in fabrication of a semiconductor integrated circuit device, a design pattern on a display substrate of a liquid crystal display device and a design pattern on a magnetic head of a thin film magnetic head device. Now, Modification 1 of Embodiment 1 of the invention will be described with reference to the accompanying drawings. FIGS. 3(a) through 3(c) show the layout of design data in respective procedures in a lithography pattern data generation method of this modification. In FIGS. 3(a) through 3(c), like reference numerals are used to refer to like elements used in FIGS. 2(a) through 2(c), so as to omit the description. As is shown in FIG. 3(a), a design data 17 most of which falls within the stripe area 61c of the first stripe area group 61 has a tip portion extending over the stripe area 61b and the stripe area 61c. As a characteristic of this modification, in the first design data group extracting process ST03 of FIG. 1, each stripe area of the first stripe area group 61 is enlarged by a predetermined width merely in the extraction, so as to extract the first design data group by using the stripe areas 61a through 61c each with the enlarged width. In FIG. 3(a), the enlarged predetermined width is indicated as an extraction width 61d, which is set to, for example, approximately 0.5 xcexcm. In this manner, the design data 17 can be determined to be covered by the stripe area 61c in the data extraction of the first design data group extracting process ST03 as is shown in FIG. 3(b), and hence is extracted as the first design data group. Although the maximum deflection area of the primary deflection device is assumed as 5 mm, the primary deflection device of a general electron beam lithography system has a deflection width having a margin of several microns. Therefore, even when an exposed pattern protrudes to the stripe area 61b as in the design data 17 of FIG. 3(b), the pattern can be drawn. In this manner, in addition to the characteristic of Embodiment 1, the number of design data belonging to the first design data group is increased and the number of design data belonging to the second design data group is decreased in this modification. Therefore, in the case where the process for dividing the area into stripe areas is repeatedly conducted, the repeated processes can be easily converged. Although the extraction width 61d is set to 0.5 xcexcm in this modification, an appropriate value can be selected depending upon the used electron beam lithography system. Modification 2 of Embodiment 1 will now be described with reference to the accompanying drawings. FIGS. 4(a) through 4(c) show the layout of design data in respective procedures in a lithography pattern data generation method of this modification. In FIGS. 4(a) through 4(c), like reference numerals are used to refer to like elements used in FIGS. 2(a) through 2(c), so as to omit the description. As is shown in FIG. 4(a), among plural design data arranged on the data arranging area 10, design data 16 and 18 extend over the boundaries of the adjacent stripe areas. Also, a portion of the design data 16 crossing the boundary of the stripe areas is assumed to be in the size of 0.4 xcexcm, and a portion of the design data 18 crossing the boundary is assumed to be in the size of 1.2 xcexcm. As a characteristic of this modification, in the first design data group extracting process ST03 of FIG. 1, among design data each extending over any boundary of the adjacent stripe areas, one having a portion crossing the boundary in the size of 1.0 xcexcm or more is extracted as the first design data group, and one having a portion crossing the boundary in the size smaller than 1.0 xcexcm is extracted as the second design data group. Although a connection error caused in adjacently connecting divided data corresponds to approximately 50 nm, the design data 18 has the crossing portion in the size of 1 xcexcm or more, and hence is minimally affected by the connection error but can attain sufficient accuracy in the connected exposed pattern. Therefore, the design data 18 can be included in the first design data group. Accordingly, in addition to the characteristic of Embodiment 1, the number of design data belonging to the second design data group can be decreased. Therefore, in the case where the process for dividing the area into stripe areas is repeatedly conducted, the repeated processes can be easily converged. Although merely a design data having a portion crossing the boundary of the stripe areas in the size smaller than 1 xcexcm is extracted as the second design data group in this modification, the range of the size is preferable to be optimized depending upon the accuracy of the lithography system and the process conditions. Also, when the size of the portion crossing the boundary of the stripe areas accords with the predetermined value, for example, 1 xcexcm, the corresponding design data is included in the first design data group in this embodiment, but it can be included in the second design data group. Embodiment 2 of the invention will now be described with reference to the accompanying drawings. FIG. 5 is a process flowchart of a lithography pattern data generation method of Embodiment 2, and FIGS. 6(a) through 6(e) show the layout of design data in respective procedures in the lithography pattern data generation method of Embodiment 2. First, in a design data preparing process ST11 of FIG. 5, plural design data 21 through 26 corresponding to design patterns to be formed on a substrate are prepared on a data arranging area 10 as is shown in FIG. 6(a). Next, in a first area dividing process ST12 of FIG. 5, the data arranging area 10 is divided into a first stripe area group 61 including stripe areas 61a through 61c each with a width of approximately 5 mm. Then, in a first data group extracting process ST13 of FIG. 5, design data each falling within any of the stripe areas are extracted from the plural design data as a first design data group. Accordingly, in FIG. 6(b), since each of the design data 21 through 23 falls within any of the stripe areas 61a and 61c, the first design data group consists of the design data 21 through 23. Subsequently, first lithography pattern data 1 of each stripe area of the first stripe area group 61 are generated from the design data 21 through 23 belonging to the first design data group. Next, in a second data group extracting process ST14 of FIG. 5, design data each extending over the plural stripe areas of the first stripe area group 61 are extracted as a second design data group. As is shown in FIG. 6(a), the design data 24 extends over the adjacent stripe areas 61a and 61b, and the design data 25 and 26 extend over the adjacent stripe areas 61b and 61c. Therefore, the second design data group consists of the design data 24 through 26. Then, in a second area dividing process ST15 of FIG. 5, as is shown in FIG. 6(c), the data arranging area 10 is divided into a second stripe area group 62 including stripe areas 62a and 62b, which cover the design data 24 and 25, respectively and are differently divided from the first stripe area group 61. Next, in a third data group extracting process ST16 of FIG. 5, design data each falling within any of the stripe areas of the second stripe area group 62 are extracted from the second design data group as a third design data group. In FIG. 6(d), the design data 24 and 25 fall within the stripe areas 62a and 62b, respectively, and hence the third design data group consists of the design data 24 and 25. Accordingly, second lithography pattern data 2 of each stripe area of the second stripe area group 62 are generated from the design data 24 and 25 belonging to the third design data group. Then, in a fourth data group extracting process ST17 of FIG. 5, design data each extending over adjacent stripe areas of the second stripe area group 62 are extracted as a fourth design data group. As is shown in FIG. 6(c), the design data 26 extends over the adjacent stripe areas 62a and 62b, and the fourth design data group consists of the design data 26. Next, in a third area dividing process ST18 of FIG. 5, the data arranging area 10 is divided to obtain a third stripe area 63 as is shown in FIG. 6(e), which covers the design data 26 belonging to the fourth design data group and is differently divided from the second stripe area group 62. Subsequently, in a fifth data group extracting process ST19 of FIG. 5, design data each falling within the third stripe area is extracted from the fourth design data group as a fifth design data group. In FIG. 6(e), the design data 26 falls within the third stripe area 63, and hence the fifth design data group consists of the design data 26. Accordingly, a third lithography pattern data 3 of the third stripe area 63 is generated from the design data 26 belonging to the fifth design data group. Then, in a sixth data group extracting process ST20 of FIG. 5, design data each extending over the boundary of the third stripe area (group) 63 are extracted as a sixth design data group, and fourth lithography pattern data of each stripe area of the third stripe area (group) 63 are generated from design data belonging to the extracted sixth design data group. However, in the exemplified layout shown in FIG. 6(e), there is no design data extending over the boundary of the third stripe area 63, and hence, the sixth design data group is not generated in this case. In this embodiment, merely the basic concept of the invention is described, and the elements of the fourth and fifth design data groups are the design data 26 alone. The number of design data used in an actual semiconductor device is huge, and therefore, it seems that there remain a large number of design data extending over the boundaries of the adjacent stripe areas after the three dividing processes for obtaining the third stripe area (group) 63. Accordingly, actual generation of lithography pattern data is carried out as the data process using a computer, and the dividing process is repeatedly conducted with the dividing positions of the stripe area groups successively changing so as to decrease the number of design data extending over adjacent stripe areas, preferably until there is no design data extending over adjacent stripe areas. However, when the number of times of repeating the dividing process is too large, the through-put time can be largely increased depending upon the scale of the design data, and therefore, the number of times of repeating the dividing process should be naturally controlled. Also in this embodiment, in the first data group extracting process ST13, the third data group extracting process ST16 or the fifth data group extracting process ST19, each stripe area can be provided with an extraction width of 0.5 xcexcm as in Modification 1 of Embodiment 1. Furthermore, in the first data group extracting process ST13, the third data group extracting process ST16 or the fifth data group extracting process ST19, a design data having a portion crossing a boundary of stripe areas in the size exceeding a predetermined value-can be extracted as the first design data group, the third design data group or the fifth design data group as in Modification 2 of Embodiment 1. In this manner, in this embodiment, the data arranging area 10 is divided by using one stripe area group corresponding to a partial exposure area controlled by the deflectable width of the exposure beam. Then, merely design data extending over the adjacent stripe areas are extracted, and the data arranging area 10 is divided again by using another stripe area group so that at least one of the design data extending over the stripe areas can be made not to extend over new stripe areas. Such a dividing process is repeated until none of the prepared plural design data extends over the stripe areas, so that the resultant lithography pattern data can be free from a connection error. Accordingly, exposed patterns drawn on the basis of the lithography pattern data can be remarkably improved in the accuracy. Modification 1 of Embodiment 2 will now be described with reference to the accompanying drawings. FIGS. 7(a) through 7(d) and 8(a) through 8(c) show the layout of design data in respective procedures in a lithography pattern data generation method of this modification. In FIGS. 7(a) through 7(d) and 8(a) through 8(c), like reference numerals are used to refer to like elements used in FIGS. 6(a) through 6(e), so as to omit the description. In the data generation of this modification, prepared design data includes a design data 27 having a size (length) along the widthwise direction of a stripe area larger than the width of the stripe area and including a wide portion 27a as is shown in FIG. 7(a). First, as is shown in FIG. 7(b), the first design data group consists of the design data 21 through 23, and the second design data group consists of the design data 24 through 28. Then, as is shown in FIG. 7(c), since the data arranging area 10 is divided into the second stripe area group 62 so that each of the design data 24 and 25 belonging to the second design data group falls within one stripe area, the fourth design data group consists of the design data 27 and 28 as is shown in FIG. 7(d). Next, as is shown in FIGS. 8(a) and 8(b), when the data arranging area 10 is divided so that the design data 28 belonging to the fourth design data group can fall within one stripe area of the third stripe area group 63, the sixth design data group consists of the design data 27 because it extends over the boundary of the stripe areas. Then, as is shown in FIG. 8(c), a design data having a size, along the widthwise direction of the stripe area, larger than the width of the stripe area is extracted from the design data belonging to the sixth design data group. The design data having a size, along the widthwise direction of the stripe area, larger than the width of the stripe area can never fall within one stripe area, and hence, the dividing process of the data arranging area 10 cannot be converged no matter how many times it is repeated. Accordingly, in this modification, the design data 27 is extracted, which has the wide portion 27a with a length, perpendicular to the exposure direction corresponding to the extending direction of the stripe area, smaller than the width of the stripe area and a width scarcely causing a connection error, for example, of 1 xcexcm or more. Subsequently, the data arranging area 10 is divided into a fourth stripe area group 64 including stripe areas 64a and 64b so that the boundary therebetween can be positioned on the wide portion 27a of the design data 27. When the data arranging area 10 is thus divided so that the wide portion 27a of the design data 27 with a width exceeding the predetermined value can be positioned on the boundary in the fourth stripe area group 64, a connection error is minimally caused in the resultant exposed pattern even when it is obtained from the design data divided between the stripe areas. Modification 2 of Embodiment 2 will now be described with reference to the accompanying drawings. FIGS. 9(a) through 9(d) and 10(a) through 10(c) show the layout of design data in respective procedures in a lithography pattern data generation method of this modification. In FIGS. 9(a) through 9(d) and 10(a) through 10(c), like reference numerals are used to refer to like elements used in FIGS. 7(a) through 7(d) and 8(a) through 8(c), so as to omit the description. In Modification 1 of Embodiment 2, as is shown in FIGS. 8(b) and 8(c), after the design data 27 having the size, along the widthwise direction of the stripe area, larger than the width of the stripe area is extracted from the design data belonging to the sixth design data group, since the extracted design data 27 has the wide portion 27a with a width exceeding the predetermined value, the data arranging area 10 is divided into the fourth stripe area group 64 so that the wide portion 27a can be positioned on the boundary. In this modification, as is shown in FIG. 10(c), a design data 29 having the size, along the widthwise direction of the stripe area, larger than the width of the stripe area does not have a wide portion as that of the design data 27 of Modification 1. Therefore, an auxiliary pattern data 30 for preventing deformation of an exposed pattern is positively added onto a portion of the design data 29 crossing the boundary in the fourth stripe area group 64. In this manner, although the design data 29 does not have a wide portion as that in Modification 1, a connection error is scarcely caused in the resultant exposed pattern. Furthermore, as a third modification, the design data 29 extending over the plural stripe areas can be subjected to multiple exposure used in writing a photomask. When the multiple exposure is conducted on the design data 29 extending over the boundary between the stripe areas, a connection error can be avoided. In this manner, a highly accurate exposed pattern can be obtained by adding the auxiliary pattern data 30 to the design data 29 extending over the boundary of the stripe areas or by conducting the multiple exposure. In addition, the addition of the auxiliary pattern data or the multiple exposure is carried out on the specified design data alone, and hence, the degradation of the through-put can be prevented. Embodiment 3 of the invention will now be described with reference to the accompanying drawings. FIGS. 11(a) through 11(e) show the layout of design data in respective procedures in a lithography pattern data generation method of Embodiment 3. This embodiment is characterized by classifying prepared design data in accordance with a pattern width. First, as is shown in FIG. 11(a), plural design data 31 through 36 are arranged on a design data arranging area 10. Among these design data, the design data 32 and 35 are in the shape of a composite figure formed by connecting a figure with a comparatively large width and another figure with a comparatively small width. Next, as is shown in FIG. 11(b), the design data are classified into a first design data group with a pattern width exceeding, for example, 1 xcexcm and a second design data group with a pattern width of 1 xcexcm or less. Furthermore, the design data 32 and 35 are herein divided into figure units. Accordingly, the first design data group consists of the design data 31, 32A, 34, 35A and 36, and the second design data group consists of the design data 32B, 33 and 35B. Then, as is shown in FIG. 11(c), the data arranging area 10 where the first design data group is arranged is divided into a first stripe area group 61 including stripe areas 61a through 61c each with a width of approximately 5 mm. First exposed patterns of each stripe area of the first stripe area group 61 are generated from the design data 31, 32A, 34, 35A and 36 belonging to the first design data group. At this point, the design data 36 extends over the boundary between the stripe areas 61b and 61c, but a connection error is scarcely caused because it has a pattern width larger than 1 xcexcm. Similarly, the data arranging area 10 where the second design data group is arranged is divided into the first stripe area group 61 including the stripe areas 61a and 61b. Next, as is shown in FIG. 11(d), the design data 32B and 33 each falling within any of the stripe areas of the first stripe area group 61 are extracted as a third design data group, and the design data 35B extending over the boundary is extracted as a fourth design data group. Subsequently, second lithography pattern data of each stripe area of the first stripe area group 61 are generated from the design data 32B and 33 belonging to the third design data group. Then, as is shown in FIG. 11(e), the data arranging area 10 is divided to obtain a second stripe area 62 which covers the design data 35B and is differently divided from the first stripe area group 61. Subsequently, a third lithography pattern data of the second stripe area 62 is generated from the design data 35B belonging to the fourth design data group. At this point, the width of the second stripe area 62 can be the same as or different from the stripe width of the first stripe area group 61. In this manner, the design data 31 through 36 are classified in accordance with a pattern width in this embodiment before dividing the data arranging area 10 into the first stripe area group 61. Accordingly, the number of design data having a pattern width smaller than the predetermined value and extending over the boundary of the stripe areas can be largely decreased. As a result, there is less fear of a connection error caused in exposed patterns, and when the process for dividing the area into the stripe areas is repeatedly conducted, the repeated processes can be more rapidly converged, resulting in improving the through-put. Although the design data 32 and 35 are divided into figure units in this embodiment, the process for dividing a design data is not always necessary. Embodiment 4 of the invention will now be described with reference to the accompanying drawings. FIG. 12 shows the functional structure of an electron beam lithography system of Embodiment 4. As is shown in FIG. 12, respective units of the electron beam lithography system 90 of this embodiment are operated under control of a control CPU 91. A lithography pattern data generation unit 92 functions in accordance with the lithography pattern data generation method of this invention, namely, a software program for realizing the lithography pattern data generation method described in any of Embodiments 1 through 3. The lithography pattern data generation unit 92 includes an area dividing part for dividing a data arranging area corresponding to a pattern formation area on a substrate into plural partial exposure areas each in the shape of a stripe corresponding to the deflection width of an exposure beam; a data group extracting part for extracting, from design data stored in a data storage unit 93, design data each falling within any of the plural partial exposure areas as a first design data group and extracting design data each extending over two or more of the partial exposure areas as a second design data group; and a data generating part for generating lithography pattern data of the respective partial exposure areas from the design data belonging to the first design data group and the second design data group. A reference numeral 100 denotes an electron optical lens barrel, in which an electron gun 104 is disposed in the upper portion and a movable stage for supporting a substrate to be exposed is disposed in a position for receiving an electron beam. The structure of the electron optical lens barrel 100 will be described in detail below. A lithography control unit 94 serving as charged particle controlling means, substrate position controlling means and beam shape controlling means controls blanking of the electron gun 104 by adjusting its output state on the basis of the lithography pattern data generated by the lithography pattern data generation unit 92. In addition, the lithography control unit 94 instructs a stage position control unit 95 to adjust the relative position of the movable stage for supporting the substrate against the electron gun 104, and instructs a deflection control unit 96 to control the shape of the electron beam by adjusting the deflection state of the electron beam. A mechanism control unit 97 adjusts lithography environments, for example, adjusts the pressure in the electron optical lens barrel 100. FIG. 13 schematically shows the structure of the electron optical lens barrel 100 of this embodiment. As is shown in FIG. 13, above a substrate 102 supported by a movable stage 101 serving as substrate supporting means is disposed an electron gun 104, serving as charged particle producing means, for emitting an electron beam 103 toward the substrate 102. Between the movable stage 101 and the electron gun 104, a first aperture 105, serving as beam shaping means, having a first opening 105a in a square shape; a selective deflection device 106, serving as beam shaping means, for appropriately deflecting the electron beam 103 having passed through the first opening 105a; a second aperture 107, serving as beam shaping means, having a second opening 107a in a square shape; and a reducing lens 108 for reducing an exposure beam with a square section, that is, the electron beam having passed through the second opening 107a are disposed in this order in the direction from the electron gun 104 to the movable stage 101. On the inside of the reducing lens 108, a primary deflection device 109A for deflecting the exposure beam is disposed, and on the inside of the primary deflection device 109A, a secondary deflection device 109B and a tertiary deflection device 109C are disposed in the upper portion and the lower portion, respectively. The operation of the electron lithography system having the aforementioned structure will now be simply described. First, as is shown in FIG. 13, the substrate 102 coated with a photosensitive material sensitive to the electron beam is supported by the movable stage 101. Next, the electron gun 104 supplied with an acceleration voltage of approximately 50 kV emits the electron beam (exposure beam). The emitted electron beam 103 is shaped to have a square section by the first opening 105a of the first aperture 105. The electron beam 103 shaped into a square section is deflected by the selective deflection device 106 before reaching the second opening 107a, so that the electron beam 103 passing through the second opening 107a can be shaped to have a rectangular section. The thus shaped electron beam 103 is allowed to irradiate a predetermined area on the substrate 102 by the deflection devices 109A, 109B and 109C, so as to successively draw the exposed patterns in accordance with the design data. Since the deflection devices are thus provided in plural stages, higher deflection accuracy is attained. Now, a lithography pattern fabrication method using the electron beam lithography system with the aforementioned structure will be described. FIGS. 14(a) through 14(d) show the layout of exposed patterns in respective procedures in the lithography pattern fabrication method of this embodiment. In a memory space of the data storage unit 93 of FIG. 12, a data arranging area 10 corresponding to an exposure area on a substrate is formed as is shown in FIG. 14(a), and plural design data 41A through 46A corresponding to design patterns to be formed on the substrate are prepared on the data arranging area 10. The design data 41A through 46A are herein in the same positions and in the same shapes as the design data described in Embodiment 2. First, lithography pattern data are generated based on the design data 41A through 46A. At this point, the lithography pattern data are assumed in this embodiment to be generated by the generation method of Embodiment 2. Accordingly, the first design data group consists of the design data 41A, 42A and 43A, and the first lithography pattern data are generated from the first design data group. The third design data group consists of the design data 44A and 45A, and the second lithography pattern data are generated from the third design data group. The fourth design data group consists of the design data 46A, and the third lithography pattern data is generated from the fourth design data group. The thus generated first through third lithography pattern data are stored in the data storage unit 93. Next, the first lithography pattern data stored in the data storage unit 93 are drawn as is shown in FIG. 14(b). Specifically, lithography pattern data 41B through 41C corresponding to the design data 41A through 43A each falling within any of the stripe areas 61a through 61c of the first stripe area group 61, each in the shape of a stripe with a width of approximately 5 mm, are drawn on a pattern formation area 70 on the substrate successively in respective areas 71a, 71b and 71c of a first partial exposure area group 71. Then, the second lithography pattern data stored in the data storage unit 93 are drawn as is shown in FIG. 14(c). Specifically, lithography pattern data 44B and 45B corresponding to the design data 44A and 45A each falling within any of the stripe areas 62a and 62b of the second stripe area group 62 are drawn on the pattern formation area 70 on the substrate successively in respective areas 72a and 72b of a second partial exposure area group 72. Next, the third lithography pattern data stored in the data storage unit 93 is drawn as is shown in FIG. 14(d). Specifically, a lithography pattern data 46B corresponding to the design data 46A falling within the third stripe area 63 is drawn on the pattern formation area 70 on the substrate in a third partial exposure area 73. Herein, the data arranging area 10 is assumed to be in the same size as the pattern formation area 70 for simplification, and hence, each stripe area for dividing the data arranging area 10 is assumed to have the same width as each partial exposure area for dividing the pattern formation area 70. An exposed pattern can be, however, generally reduced or enlarged with keeping the relative relationship in position and size between the design data and the exposed pattern. Although the first lithography pattern data, the second lithography pattern data and the third lithography pattern data are drawn in this order in this embodiment, but the order is not herein specified as far as all the lithography pattern data can be ultimately drawn on the substrate. Similarly, the lithography pattern data are drawn in each of the partial exposure area groups 71 and 72 successively in the rightward direction in this embodiment, but the direction is not herein specified. However, since the movable stage 101 of FIG. 13 should be moved in order to proceed the lithography process from one stripe area to another stripe area, the lithography pattern data can be more efficiently drawn in a manner that adjacent stripe areas are successively exposed. In the conventional lithography system and method, merely one dividing process is carried out for drawing all the design data. Therefore, one exposed pattern extends over plural partial exposure areas, and hence, the exposed pattern is divided. As a result, a connection error is easily caused in the lithography. However, the number of exposed patterns extending over adjacent partial exposure areas can be decreased in this embodiment, resulting in reducing connection errors. Accordingly, highly accurate exposed patterns can be obtained. Since the electron beam lithography system 90 of this embodiment is operated on the basis of lithography pattern data generated by the lithography pattern data generation unit 92 of FIG. 12, the following auxiliary functions described in Embodiments 1 through 3 and their modifications can be reflected in lithography pattern data to be generated: (1) To increase the number of data belonging to a design data group extracted at an earlier stage by providing each stripe area used for extracting the data group with a predetermined extraction width; (2) To increase the number of data belonging to a design data group extracted at an early stage by extracting, among design data extending over stripe areas, a data having a portion crossing the boundary of stripe areas in a predetermined size or larger as a design data group obtained at an earlier stage; (3) To repeat the process for dividing the data arranging area into stripe areas until none of design data extends over the stripe areas; (4) With respect to a design data which is unavoidably divided because it has a size, along the widthwise direction of a stripe area, larger than the width of the stripe area and which has a wide portion crossing the boundary of stripe areas in a predetermined size or larger, to set the boundary between the stripe areas on the wide portion; (5) With respect to a design data which is unavoidably divided because it has a size, along the widthwise direction of a stripe area, larger than the width of the stripe area, to add, onto a portion crossing the boundary, an auxiliary pattern data having a predetermined size or larger in the crossing portion; and (6) To conduct the multiple exposure on a design data which is unavoidably divided because it has a size, along the widthwise direction of a stripe area, larger than the width of the stripe area. The electron beam lithography system 90 of this embodiment uses an electron beam as the exposure beam, which can be replaced with an ion beam. Also, the width of each stripe area or each partial exposure area is set to approximately 5 mm in this embodiment, but the width can be appropriately set depending upon the electron gun 104, the conditions for controlling the electron gun and the design data. Furthermore, the stripe width of each stripe area group (or partial exposure area group) adopted for repeated division can be appropriately selected with respect to each strip area within a deflectable range of the exposure beam. Now, as a specific example of the effect (6) described above, Modification 1 of Embodiment 4 will be described with reference to the accompanying drawings. FIGS. 15(a) through 15(d) show the layout of exposed patterns in respective procedures in a lithography pattern fabrication method of this modification. In FIGS. 15(a) through 15(d), like reference numerals are used to refer to like elements shown in FIGS. 14(a) through 14(d), so as to omit the description. As is shown in FIG. 15(a), a design data 47A has a size, in a direction crossing the extending direction of a stripe area of the first stripe area group 61, larger than the width of the stripe area. Therefore, the design data 47A cannot fall within one stripe area but is unavoidably divided, and hence belongs to the fourth design data group. Accordingly, as is shown in FIG. 15(d), in drawing the third lithography pattern data obtained from the fourth design data group, a lithography pattern data 47B extending over partial exposure areas 73a and 73b of a third partial exposure area group 73 is subjected to the multiple exposure. As a result, a connection error is scarcely caused in the resultant pattern, and thus, the accuracy of the unavoidably divided exposed pattern can be improved. Furthermore, the multiple exposure is conducted merely on the lithography pattern data unavoidably divided, and hence, degradation in the through-put can be minimized. Embodiment 5 of the invention will now be described with reference to the accompanying drawings. FIG. 16 is a process flowchart of a lithography pattern data generation method of Embodiment 5, and FIGS. 17(a) through 17(c) show the layout of exposed patterns obtained in respective procedures in a lithography pattern fabrication method using the lithography pattern data generation method of this embodiment. First, in a design data preparing process ST21 of FIG. 16, plural design data 41A through 46A corresponding to design patterns to be formed on a substrate are prepared as shown in FIG. 17(a). The design data 41A through 46A are arranged on a data arranging area 10 correspondingly to a design pattern formation area. Next, in an area dividing process ST22 of FIG. 16, the data arranging area 10 is divided into a first stripe area group 61 including three stripe areas 61a through 61c. Then, in a first data group extracting process ST23 of FIG. 16, design data each falling within any of the stripe areas 61a through 61c, namely, not extending over any of the boundaries of the stripe areas 61a through 61c, are extracted from the design data 41A through 46A as a first design data group. Subsequently, based on the design data 41A through 43A belonging to the first design data group, first lithography pattern data 41B through 43B of each of the stripe areas 61a through 61c are generated. Next, in a second data group extracting process ST24 of FIG. 16, design data each extending over the plural stripe areas of the first stripe area group 61 are extracted as a second design data group. Subsequently, based on the design data 44A through 46A belonging to the second design data group, second lithography pattern data 44B through 46B of each of the stripe areas 61a through 61c are generated. Then, the first lithography pattern data 41B through 43B thus generated are transferred onto a pattern formation area 70 as is shown in FIG. 17(b). Subsequently, the second lithography pattern data 44B through 46B thus generated are subjected to the multiple exposure so as to transfer them onto the pattern formation area 70 as is shown in FIG. 17(c). In this manner, merely the second lithography pattern data 44B through 46B extending over the stripe areas are subjected to the multiple exposure in this embodiment. Therefore, connection errors can be easily reduced. Also in this embodiment, the order of conducting the lithography process on the first lithography pattern data and the second lithography pattern data is not particularly specified. Wherein, in the first design data group extracting process ST23, as described in MODIFICATION 1 of EMBODIMENT 1 of the present invention, each stripe area of the fist stripe area group 61 may be enlarged by a predetermine width merely in the extraction. In each of the aforementioned embodiments, the countermeasure against connection errors between stripe areas mainly derived from primary deflection is described, but the invention can exhibit the same effect also on a connection error derived from secondary or tertiary deflection. |
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042785014 | description | Referring now to the drawing, there is shown in FIG. 1 a perspective view as seen from the top onto the head 1 of a fuel assembly formed of a multiplicity of fuel rods 2. The head member 1 of this fuel assembly is connected to a non-illustrated base member by guide tubes 3 and, together with spacers 21, forms the rigid frame or skeleton of such a fuel assembly. The fuel rods 2 per se are held at a mutual spacing within this frame by the spacers 21 but can expand freely in axial direction. To control this nuclear reactor, control rods formed at least in part of neutron-absorbing materials are immersed in the guide tubes 3. In the interst of clarity, these control rods are not shown in detail, however, especially since illustration thereof is unnecessary for an understanding of the invention. The head 1 of the fuel assembly if formed in the conventional embodiment of FIGS. 1 and 2, by a laterally cut-out frame and has, at an end face thereof, four bores 5 for receiving adjusting or dowel pins which are fastened to a non-illustrated hold-down plate. A spring element 4 protrudes from the frame on both sides of each bore 5 for resiliently bracing or supporting the frame at the non-illustrated hold-down plate. The subject of the invention of the instant application is a spring element for holding down fuel assemblies which is an improvement over the conventional spring element illustrated in FIG. 1 which is mounted in the fuel assembly head. The conventional spring element of FIG. 1 is shown in enlarged and greater detail in FIG. 2. As is clearly shown in FIGS. 1 and 2, the conventional spring element 4 is formed of a pin which is slidably mounted in bores 11 and 12 provided in the fuel assembly head. A compression spring 41 is braced against the bottom of the fuel assembly head and forces the pin 4 upwardly in the illustrated manner. The upper end of this spring 41 rests, for this purpose, against stop 42 which is laterally inserted into a transverse slot formed in the pin 4. In FIG. 3, there is shown an embodiment of a new spring element 6 according to the invention, partly in longitudinal cross section, and in FIG. 4, the spring element 6 is shown, likewise in longitudinal section, but further in inserted position thereof in a fuel assembly head in a manner similar to the view of the conventional device of FIG. 2. In this regard, it is noted that FIGS. 2 and 4 represent longitudinal sections through the fuel assembly head of FIG. 1 taken along the line II--II in direction of the arrows. It can be seen, especially from FIG. 3, that the spring element 6 according to the invention is a part separate from the fuel assembly, and is inserted as a whole into the fuel assembly head 1 from the top, as shown in FIG. 4. The fuel element 6 can therefore also be withdrawn again as a whole; an operation which presents no difficulties even if performed by remote control. Also readily evident in FIG. 4 is the gain in "resilient length" of the spring element 6 for the same overall length that it has with respect to the coventional spring element 4 of FIG. 2. The spring element 6 is assembled telescopically of three parts, namely, an upper sleeve 6a, a screw bolt 6b and a lower sleeve 6c. The upper sleeve 6a is provided with a collar or shoulder 62, and the lower sleeve with a collar or shoulder 63 which serve as stops for the spring 61. The threaded bolt 6b is screwed into the lower sleeve 6c and thereby adjustable in length; the lower projecting end 64 of the threaded bolt 6b is provided for engagement in a bore 14 formed in the fuel assembly head 1. Movement of the projecting bolt end 64 in the fuel assembly head 1 is not possible, and interference thereof with fuel rods may possibly be located therebelow which is therefore precluded. The upper sleeve 6a is formed with an internal shoulder 66 by which it is braced against the head 67 of the threaded bolt 6b and can be forced downwardly on the latter telescopically against the force of the spring 61. To prevent rotation of the lower sleeve 6c on the threaded bolt 6b, which would result in a change in length of the entire spring element 6 and also to a change in the initial stress or pre-tensioning force thereof, a conventional screw securing device such as a pin 65 or a spot weld or any other rotation-preventing device known in the prior art is provided. This new spring element 6 according to the invention has not only the advantages of simplicity of construction, manufacture without problems, individual adjustibility with respect to length and initial spring stress, as well as ease of installation in the fuel-assembly head, but also the great advantage that it can be used instead of heretofore-known spring elements according to the state of the art, such as the elements 4 of FIGS. 1 and 2. For this purpose, it is necessary merely to enlarge the bore 11 (note FIG. 2) so that a bore 13 (FIG. 4) having a somewhat larger bore diameter is formed. This is an operation which can be performed without any problem on all fuel assembly heads employed heretofore, so that the new spring elements 6 according to the invention also can be exchanged for the heretofore employed ones, such as the spring elements 4 of FIG. 2, especially if there is a change in the operating condition. It is also possible, of course, to use these spring elements 6 according to the invention in fuel assembly heads of different construction than that shown in FIG. 4, since such heads need only to be provided with the bores 14 and 13 for receiving this spring element 6. In the illustrated embodiment of the instant application, eight such spring elements 6 have been provided for each fuel assembly head 1, but this number is in no way characteristic or limiting, since the required spring force for each fuel assembly is able to be adjusted or set by suitably selecting the strength of the individual springs 61 as well as the number of spring elements 6 themselves. In conclusion, yet another advantage of this new construction should be noted, namely that even in the event of a break in the spring 61, no loose parts can result, which would otherwise require retrieval at all costs at great expense for specialized apparatus for this purpose and in time consumption. |
043307116 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS Referring more specifically to the drawings the combination container consists of a removable inner container 1 having a cover 3, which optionally consist of several component parts and an outer container 2 having a cover 4. The bottom 10 and the jacket 11 of the outer container are so dimensioned in their thicknesses that they either completely or at least preponderantly shield off the gamma and neutron radiation of the container contents from the outside. The inner container 1, moreover, is fixed axially in the outer container 2 in such a manner that the inner container cover 3 and the outer container 4 do not contact. The radial position of the inner container 1 in the outer container 2 is fixed by a narrowed cross section 12 of the inner space of the outer container 2 proceeding downwardly to the bottom 10 wherein this narrowing preferably is produced by a correspondingly wedge shaped constructed profile 13 to the outer wall 14 of the inner container 1 and the inner wall 15 of the outer container 2. The outer wall 14 of the inner container 1 is tightly sealed against the inner wall 15 of the outer container 2 by means of sealing element 5. By dimensioning of the outer container 2 to the shielding wall thickness this container has a very high mechanical strength and so is substantially protected against damage in the case of accidents during transportation. Besides because of the omission of a heat transferring gas or liquid between inner container 1 and outer container 2 it is without pressure in normal operation so that no sealing problems occur. The heat transfer takes place over the small gap between the two containers, preferably, however, via the contacting profile 13 to the outer wall 14 of the inner container 1 and the inner wall 15 of the outer container 2. The gap between outer container 2 and the inner container 1 is so sealed through the sealing element 5 that in loading the container there cannot take place contamination of the outside of the inner container. The inner container is fixed axially in the jacket 11 of the outer container, preferably via holding down device 6, which is received in corresponding recesses 16 of the outer container jacket 11 in such manner that the cover 4 and therewith the seal 17 of the outer container 2 is not loaded. The combination container of the invention has a relatively thin-walled inner container which is produced simply and cheaply and on a large scale, for example it is made of commercial tubular material. There are placed high requirements on a stored container in regard to the sealability. However, as is known it is difficult to carry out the customary examination such as X-ray and supersonic testing in thick walled container. With relatively thin walled containers according to the invention this test procedure causes no problem. The outer container 2 in combination with the transportation cover 4 fulfills all of the requirements for a Type B container in regard to handling mechanical integrity, heat transfer, tightness and shielding in normal transportation and in the case of accident. This outer container 2 can be made of all work materials and combination of work materials known in the practice and literature, such as wrought iron, cast iron, lead, depleted uranium, copper or synthetic resin. Since the outer container 2 is only employed for transportation and can be utilized for a great number of inner containers 1 there is required in regard to the storage only a very limited number of the outer containers 2. Therefore there can be placed especially high requirements on the selection of material, construction, manufacture and testing without mentionably increasing the total cost of the storage. These safety reserves in the transportation are very valuable to this most risky part in the entire storage strategy. Especial requirements are placed on the sealing in a storage container. This seal should have constant good sealing properties during the entire storage time since during the collecting of many storage containers even leak rates which are admissible for individual transportation containers would lead to the release of noteworthy activity. Prerequisite for applying such a seal is good accessibility of the seal. The present container combination permits this accessibility in an outstanding manner through the fact that the cover 3 of the inner container 1 acts as a shielding cover. Therefore so long as the inner container 1 is still located in the outer container 2 the place of sealing is freely accessible and the permanent seal required for the storage can be installed without requiring for this purpose remote control devices and a hot cell or a water tank for radiological protection. The inner container 1 and the shielding cover 3 additionally are made tight with a seal 8. This seal 8 above all is effective through the weight of the stationary container itself and prevents a contamination of the space between the outer container cover 4 and the inner container cover 3. Therefore it is possible to insert the permanent seal required for storage first at the place of storage. This has the advantage that this important operation for the security of the storage always can be undertaken at a stationary device by the same crew, not every nuclear power plant must be equipped with devices and the routine loading in the nuclear power plant is not hindered. This sealing of the gap between the inner container 1 and the outer container 2 through the seal 5 has the advantage that the inner container 1 need not be decontaminated before its insertion in the storage shield. Therefore there is not accumulated any secondary waste in the provided operation which would require additional apparatus to attend to and therewith high operating costs. For the permanent sealing for the storage the inner container 1 and shielding cover 3 can each be provided with a welded end 9 on which they can be welded or soldered for gas tight storage. The gap between inner container 1 and cover 3, however, can also be so formed that it can be filled with a low melting metal. The emptying and washing of the inner container is carried out in known manner. During the transportation the inner container 1 must be so fixed in the outer container 2 that even under conditions of an accident the transportation cover 4 and its sealing system 17 is not burdened by the inner container 1 and its content. This is solved according to the invention with a holding down device 6 at whose periphery distributed brackets fit in corresponding recesses 16 of the jacket 11 of the outer container 2 and through twisting according to the bayonnet principle is secured on the container. The security against twisting is reached through screwable tension elements 7 which simultaneously shuts the shielding cover. The tension element 7 can contain a pack of springs to compensate for the longitudinal tension of the box. The outer container 2 advantageously can have cooling couplings 18 which can be joined with spirally arranged cooling channels 19 on the inner surface of the outer container 2. Thus it is possible to cool the container contents before emptying in a reprocessing plant without needing to open the inner container 1. The cooling connection 18 can also be connected to a cooling circuit during the transportation so that the fuel element temperature is lowered in the transportation. As shown in FIGS. 2-5 it is particularly advantageous to form the profile 13 hollowly on the inner wall 15 of the outer container 2 and to integrate it in the cooling circuit via the cooling channel 19. The cover 3 of the inner container 1 can be erected of several parts and for example can consist of a thin walled true cover portion and a thick walled shielding portion. Through this the shielding portion can also be used repeatedly since it is not needed in the storage in corresponding warehousing. The heat is transferred from the inner container 1 of the outer container 2 by free convection and radiation. There is not needed an additional heat transfer medium which might fail in the case of an accident. After the loading of the inner container 1 the inner container heats up first more quickly than the outer container 2 so that the gap between inner and outer container which is conditional by the manufacture is smaller and therewith the heat transfer is better. Still better heat transfer is produced if the inner container 1 has wedge shaped profiles 13 (see FIGS. 2, 3) over the entire length which fit into corresponding wedge shaped profiles 13' in the outer container 2 so that there is always metallic contact between inner and outer container and therewith metallic heat conductance occurs. The tolerance between inner and outer container is then obtained through different positions of inner and outer container and must be compensated for via seal 5. The entire disclosure of German priority application Ser. No. P2915376.2 is hereby incorporated by reference. |
summary | ||
abstract | The system and method provide security and cargo handling personnel a versatile tool to rapidly check cargo for hidden radiological materials, explosives, drugs, and chemical weapons material. Gamma ray emission is stimulated by a pulsed neutron source. The gamma ray signature is used to classify the material. Passive gamma ray analysis can be used to detect and identify radiological material. The method of determining the contents of a target includes irradiating a target; detecting at least one spectrum emitted from the target; performing a primary analysis to extract a first set of indicators; and performing a secondary analysis to decide the contents of the target. The primary analysis utilizes either a least squares analysis or principal component analysis. The secondary analysis utilizes a generalized likelihood ratio test or support vector machines. |
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052971759 | abstract | The present invention is made to provide an assembling machine for nuclear fuel assembly, having a simple mechanical construction, in which the time required for inserting the fuel rod can be reduced.. In the prior-art machine, the pull-in rod and its drive mechanism and the like are required to perform the operation for inserting the fuel rod into the supporting grid, thus, the mechanical construction must be complicated. Further, a relatively long time must be required to perform the operation for holding the fuel rod by the pull-in rod, which makes the whole working hours longer. In the machine according to the present invention, a drive mechanism is provided to drive a stopper, by which the regulation of the stopper for regulating the fall-down movement of the fuel rod can be released with ease. Thus, the fuel rod is fallen down by its own weight, so that the fuel rod can be inserted into the grid cell of the supporting grid which is supported by the supporting post. Therefore, the machine construction can be simplified, and it is not necessary to perform the prior-art operation for holding the fuel rod by the pull-in rod, so that the time required for performing the inserting operation of the fuel rod can be reduced. |
040653528 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS Hydrogen-absorbing metal material used in a state sealed in a nuclear fuel element according to this invention includes zirconium and alloys thereof such as zircalloy, and titanium and alloys thereof, or preferably zirconium, zircalloy and the aforesaid Ni-Ti-Zr getter alloy. As used herein, "absorption of hydrogen" is defined broadly to mean reaction with hydrogen to produce hydrogenated material. A hydrogen-permeable metal member used to enclose the above-mentioned hydrogen-absorbing metal is formed of palladium, palladium alloys such as silver-palladium alloy, rhenium or preferably palladium or alloys thereof. Said hydrogen-permeable metal member allows the passage of hydrogen alone but not water vapor, oxygen and other gases. FIG. 2 presents a preferred concrete embodiment of a nuclear fuel element according to this invention. A nuclear fuel element 1 is constructed by packing a zircalloy cladding tube 2 with a large number of pellets 3 of uranium dioxide as fuel material and further loading said cladding tube 2 with a hydrogen getter 4 consisting of zirconium particles 5 enclosed in an evacuated palladium tube 6. FIG. 2 only shown the upper portion of the nuclear fuel element 1, wherein the cladding tube 2 is tightly sealed with an end plug 7, and a spring 9 is inserted into a plenum 8 securely to set the uranium dioxide pellets 3 in place. FIG. 3 is a hydrogen getter 10 according to another embodiment of this invention, wherein zircalloy particles 12 are received in a blind metal tube 11 and the opening 13 of said tube is sealed in vacuum with a palladium plug 14. The process of enclosing hydrogen-absorbing metal material in a hydrogen-permeable metal member may be carried out either by sealing a certain amount of the hydrogen-absorbing metal material in the hydrogen-permeable metal member in vacuum by means of, for example, electron beams or by vacuum evaporating or electroplating a hydrogen-permeable metal film on individual hydrogen-absorbing metal particles. The hydrogene absorbing metal material is preferred to be formed of particles about 1 to 2 mm in diameter for quick absorption of hydrogen, that is, to provide a large contact area between the metal mateial and hydrogen. Hydrogen-absorbing metal material such as zirconium, titanium or alloys thereof has the surface coated with a protective film when exposed to air or water vapor and ceases to absorb hydrogen at lower temperature than 400.degree. C. According to this invention, therefore, particles of hydrogen-absorbing metal material such as zirconium are enclosed in a hydrogen-permeable metal member of, for example, palladium. Thereafter the metal assembly is baked at high temperature before fitted into a nuclear fuel element. It is preferred that said baking be carried out in vacuum at higher temperature than 600.degree. C for about 1 to 3 hours or more preferably at about 700.degree. C for about 1 hour. This baking causes a protective film possibly formed up to this point on the hydrogen-absorbing metal particles due to reaction with water vapor to disappear quickly. Since hydrogen-absorbing metal particles are enclosed in a hydrogen-permeable metal member, any other gas than hydrogen does not enter said hydrogen-permeable metal member, eliminating the possibility of the above-mentioned protective film being again formed on the hydrogen-absorbing metal material. Thus the hydrogen getter of this invention displays a more prominent capacity of absorbing hydrogen than the prior art hydrogen getter formed of a single element. The following experiments were carried out to compare the effect of the hydrogen getter of this invention and that of the conventional hydrogen getter. Measurement was made of an amount (mg) of hydrogen absorbed per gram of zircalloy perticles enclosed in a hydrogen-permeable palladium member and these not enclosed therein an atmosphere consisting of hydrogen having a partial pressure of 760 mm Hg and water vapor having a partial pressure of 14 mmHg at 300.degree. C. Throughout the experiments, the zircalloy had a composition of 98.357% of zirconium (Zr), 1.500% of tin (Sn), 0.112% of iron (Fe), 0.001% of nickel (Ni) and 0.030% of chronium (Cr). As apparent from FIG. 4 showing the results of experiments, zircalloy particles not enclosed in a hydrogen-permeable metal member substantially failed to absorb hydrogen due to a protective film being formed on said particles. In contrast, zircalloy particles of this invention which are covered with said hydrogen-permeable metal element efficiently absorbed hydrogen until the zirconium is converted into zirconium dihydride. Further, measurement was made of an amount (mg) of hydrogen absorbed per gram of a zircalloy hydrogen getter of this invention enclosed in a hydrogen-permeable palladium member and the Ni-Ti-Zr getter alloy not enclosed in any hydrogen-permeable metal member in an atmosphere at 300.degree. C consisting of hydrogen having a partial pressure of 15 mmHg and water vapor having a partial pressure of 15 mmHg. The Ni-Ti-Zr getter alloy used had a composition of 84.42% of zirconium (Zr), 8.70% of Titanium (Ti) and 6.88% of nickel (Ni). FIG. 5 giving the results of experiments proves that the hydrogen getter of this invention exclusively absorbed hydrogen even in the above-mentioned oxidizing atmosphere, whereas the Ni-Ti-Zr getter alloy not enclosed in any hydrogen-permeable metal member absorbed only an extremely small amount of hydrogen. This undesirable event is supposed to result from a protective film deposited on the surface of the Ni-Ti-Zr getter alloy not enclosed in a hydrogen-permeable metal member. A nuclear fuel element according to this invention provided with the above-mentioned hydrogen getter enclosed in a hydrogen-permeable metal member has been found little liable to be hydrogenated and become brittle with the possibility of being eventually destroyed during the run of a nuclear reactor. |
description | This application claims the benefit of Provisional Application No. 61/138,140, filed on Dec. 17, 2008 and entitled, “Core Shroud Corner Joints.” This invention was made with government support under Contract No. DE-FC07-05ID14636 awarded by the Department of Energy. The government has certain rights in this invention. 1. Field The disclosed concept relates generally to nuclear reactors and, more particularly, to core shrouds for nuclear reactors. The disclosed concept also relates to an associated method of assembling core shrouds. 2. Background Information The primary side of nuclear reactor power generating systems which are cooled with water under pressure, comprises a closed circuit that is isolated from and in heat-exchange relationship with a secondary side for the production of useful energy. FIG. 1 shows a simplified nuclear reactor primary system, including a generally cylindrical reactor pressure vessel 10 having a closure head 12 (also shown in FIG. 2) enclosing a nuclear core 14. A liquid reactor coolant, such as water, is pumped into the vessel 10 by pumps 16 through the core 14 where heat energy is absorbed and is discharged to a heat exchanger 18, typically referred to as a steam generator, in which heat is transferred to a utilization circuit (not shown), such as a steam-driven turbine generator. The reactor coolant is then returned to the pump 16, completing the primary loop. Typically, a plurality of the above-described loops are connected to a single reactor vessel 10 by reactor coolant piping 20. FIGS. 2 and 3 show simplified side elevation and top plan views, respectively, of the pressure vessel 10, and both show portions of the pressure vessel 10 in section view. The core 14 is comprised of a plurality of parallel, vertical co-extending fuel assemblies 22, only two of which are shown in FIG. 2 for ease of illustration. For purposes of this description, the other vessel internal structures can be divided into the lower internals 24 and the upper internals 26 (both shown in FIG. 2). In conventional designs, the lower internals 24 function to support and align the core and guide instrumentation, as well as direct flow within the vessel 10. The upper internals 26 restrain or provide a secondary restraint for the fuel assemblies 22, and support and guide instrumentation and core components, such as control rods 28. In operation, coolant enters the vessel 10 through one or more inlet nozzles 30, flows downward through an annulus between the vessel 10 and the core barrel 32, is turned about 180° in a lower plenum 34, passes upwardly through a lower core support plate 37 and a lower core plate 36 upon which the fuel assemblies 22 are seated, and through and about the fuel assemblies 22. In some designs the lower core support plate 37 and lower core plate 36 are replaced by a single structure. The coolant flow through the core and surrounding area 38 is typically large, on the order of 400,000 gallons per minute at a velocity of approximately 20 feet per second. The resulting pressure drop and frictional forces tends to cause the fuel assemblies to rise, which movement is restrained by the upper internals 26, including a circular upper core plate 40. Coolant exiting the core 14 flows along the underside of the upper core plate 40 and upwardly through a plurality of perforations 42. The coolant then flows upwardly and radially to one or more outlet nozzles 44. As shown in simplified form in FIG. 3, a core shroud 17 is positioned inside the circular core barrel 32, and includes a plurality of vertically extending plates 19 that convert the inner profile of the core barrel 32 to a stepped circumferential profile that generally matches the peripheral outline of the fuel assemblies 22 (shown in simplified form in FIG. 3) within the core 14. The simplified cross-section view of FIG. 3 also shows a thermal shield 15, which is interposed between the pressure vessel 10 and core barrel 32. Some plants have neutron pads in lieu of the thermal shield. Typically, the plates 19 that form the stepped circumferential profile are substantially flat and abut at right angles at intersecting, corner, locations. As a result of machining and/or forming, some reactor vessel internals, however, include atypical corner joints. By way of example, these atypical corner joints can be characterized as being round for outside corner locations, being “key-like” (e.g., without limitation, having a groove) for interior locations and/or having relatively large pockets of open areas. Each atypical corner joint provides an area for flow to bypass the adjacent fuel assembly due to the low hydraulic resistance in these corners. In fact, flow calculations have shown a relatively high axial velocity in atypical core shroud corners. Among other disadvantages, this may result in unacceptable fuel rod vibration, which leads to fuel assembly grid-to-rod fretting, and may also cause elevated cross-flow velocities in this region. There is, therefore, room for improvement in core shrouds and corner joints therefor. These needs and others are met by embodiments of the disclosed concept, which provide an improved design and assembly method for nuclear reactor core shrouds wherein, among other benefits, the corners of the core shroud assembly are preferably of a unitary design comprising one single continuous piece of material that is devoid of any seems or associated gaps or voids. As one aspect of the disclosed concept, a core shroud is provided. The core shroud comprises: a number of planar members; a number of unitary corners; and a number of subassemblies each comprising a combination of the planar members and the unitary corners. Each of the unitary corners may be substantially identical. Each of the unitary corners may comprise a unitary extrusion including a first planar portion and a second planar portion disposed perpendicularly with respect to the first planar portion. At least one of the subassemblies may comprises a plurality of the unitary corners, wherein the first planar portion of one of the unitary corners is joined to the second planar portion of another one of the unitary corners, in order that the unitary corners are disposed side-by-side in an alternating opposing relationship. As another aspect of the disclosed concept, a nuclear reactor is provided which comprises: a pressure vessel; an annular core barrel seated within and supported by the pressure vessel; and a core shroud supported within the core barrel, the core shroud comprising: a number of planar members, a number of unitary corners, and a number of subassemblies each comprising a combination of the planar members and the unitary corners. The core shroud may further comprise a number of flow deflectors, and each of the unitary corners of the core shroud may include a curved interior junction and a curved exterior junction. Each of the flow deflectors may include a curved portion and a number of substantially flat portions disposed opposite the curved portion, wherein the curved portion is structured to mate with a corresponding one of the curved interior junction and the curved exterior junction. The nuclear reactor may further comprise a number of grids disposed within the core shroud, wherein the substantially flat portions of the flow deflectors are structured to engage and support a portion of a corresponding one of the grids. As a further aspect of the disclosed concept, a method of assembling a core shroud is provided. The method comprises: providing a number of planar members; extruding a plurality of unitary corners; joining a combination of the planar members and the unitary corners to form a number of subassemblies; and joining a plurality of the subassemblies to form at least a portion of the core shroud. The method may further comprise joining a plurality of the subassemblies to form a quarter perimeter segment of the core shroud, and may still further comprise joining four of the quarter perimeter segments to form the core shroud. For purposes of illustration, embodiments of the disclosed concept will be described as applied to core shrouds although it will become apparent that they could also be applied to replace or otherwise eliminate corner joints between joined components of other internals assemblies (e.g., without limitation, battle-former assemblies) to address and overcome problems associated therewith (e.g., without limitation, baffle-jetting). Directional phrases used herein, such as, for example, interior, exterior, inside, outside, top, bottom and derivatives thereof, relate to the orientation of the elements shown in the drawings and are not limiting upon the claims unless expressly recited therein. As employed herein, the term “unitary” shall mean one single continuous piece of material that is devoid of any seems, joints or connections and which may be formed using any known or suitable method or process such as, for example and without limitation, an extrusion process. As employed herein, the term “number” shall mean one or an integer greater than one (i.e., a plurality). FIG. 4 shows a core shroud 100. Several fuel assemblies 102 are shown in locations (generally indicated as A, B and C in FIG. 4) on the perimeter 106 of the core shroud 100 where they are adjacent to either an inner core shroud corner 108 or an outer core shroud corner 110. For example and without limitation, in the non-limiting example of FIG. 4, there are 44 locations (only 10 locations are shown for ease of illustration) where fuel assemblies 102 will be adjacent to either inside corners 108, outside corners 110, or both inside and outside corners 108,110. FIG. 5 shows an enlarged view of an inside corner 108 and an outside corner 110 of the core shroud 100. As shown in FIGS. 6-12, the disclosed concept entails the fabrication of each core shroud corner as a unitary corner 200 such as, for example and without limitation, a unitary extrusion having the desired corner shape. The open areas associated with conventional inside and outside corners 108, 110 (FIGS. 4 and 5) are, therefore, eliminated. Such open areas can result from forming (e.g., without limitation, bending) and/or from welding the seems (not shown in FIGS. 4 and 5) where two planar portions abut and are joined to make the corner joint. Recent changes in manufacturing extrusion technology have made this possible. For example and without limitation, up to about 4.75 inch×4.75 inch×1 inch or larger full-length (e.g., without limitation, see length, L, of FIG. 6 which may be, for example and without limitation, about 180 inches (not shown to scale)) extruded shapes are possible. Thus, as shown in FIG. 6, each unitary corner 200 (two are shown) includes a first planar portion 202 and a second planar portion 204 disposed perpendicular to the first planar portion 202, and without any seem or other connection therebetween. This is particularly advantageous where, as in the example of FIG. 6, the width (e.g., without limitation, about 4.75 inch) when welded to a similar extrusion 200, is sufficiently wide to accommodate a perimeter location having a single fuel assembly 102 (see, for example, fuel assemblies 102 in locations “A” in FIG. 4). The unitary extruded shapes 200 (e.g., unitary corners) can then be laser welded or otherwise suitably joined, for example, to form a subassembly 210, as shown in FIG. 6. To complete the perimeter of the core shroud near the pair of fuel assemblies 102 designated as “A” in FIG. 4, continued welding of three pairs of the unitary corners 200 is completed to form a subassembly 212, as shown in FIG. 7. For the other two typical perimeter locations, which have two or three fuel assemblies in a row (see, for example, fuel assemblies 102 at the locations designated as “B” and “C” in FIG. 4), planar members 214 (FIG. 8), 216 (FIG. 9) of differing widths 218 (FIG. 8), 220 (FIG. 9), are welded to a unitary corner 200 to form subassemblies 222, 224, which are shown in FIGS. 8 and 9, respectively. Specifically, FIGS. 8 and 9 show planar members 214, 216 of widths 218, 220, respectively, that can be laser welded or otherwise suitably joined to a unitary corner 200 to accommodate core fuel assembly locations “B” and “C,” respectively, of FIG. 4. In the example of FIG. 8, the first planar member 214 includes opposing edges 215, 217 and a first width 218 measured by the distance therebetween, whereas the second planar member 216 of FIG. 9 has opposing edges 219, 221 and a second width 220 therebetween, which is greater than the first width 218 of first planar member 214. The exemplary method of forming the perimeter 106 of the core shroud 100 (FIG. 4), involves welding or otherwise suitably joining two of the subassembly 222, shown in FIG. 8, to the subassembly 212, shown in FIG. 7, resulting in the subassembly 226, shown in FIG. 10. Lastly, to complete a quarter perimeter segment, or subassembly 228, of the core shroud 100 (FIG. 4; also referred to generally as 300 in FIG. 12), the unitary corner subassembly 210 of FIG. 6 and the subassembly 224, shown in FIG. 9, are welded or otherwise suitably joined to subassembly 226 of FIG. 10. In the example shown and described herein, each of the quarter perimeter segments 228 includes eleven of the unitary corners 200, two of the first planar members 214, and one of the second planar members 216, as shown in FIG. 11. Finally, four of the quarter perimeter segments 228 are welded or otherwise suitably joined together to form the complete core shroud 300, shown in FIG. 12 (see also core shroud 100 of FIG. 4). It will be appreciated that any known or suitable alternative configuration, number and/or assembly sequence of components (e.g., without limitation, unitary corners 200; first planar members 214; second planar members 216) and subassemblies (e.g., without limitation 210, 212, 222, 224, 226, 228) could be employed, without departing from the scope of the disclosed concept. It will also be appreciated that while components (e.g., without limitation, unitary corners 200; first planar members 214; second planar members 216) are preferably welded using laser technology, as generally indicated in simplified form in FIG. 6, that any known or suitable alternative method, process or mechanism could be employed to suitably join the core shroud components (e.g., without limitation, unitary corners 200; first planar members 214; second planar members 216) and/or subassemblies (e.g., without limitation 210, 212, 222, 224, 226, 228). Among other advantages, the disclosed concept eliminates the inside and outside corner void areas associated with the known core shroud designs. The void areas are the result of extensive machining, bending and/or forming operations. The disclosed concept provides a unitary corner and therefore eliminates all seems at the corner, and also eliminates welding corner joints that are difficult to inspect due to lack of accessibility. The only welds are at locations away from the corners, which are substantially flat and/or relatively easy to access and facilitate inspection. Additionally, in stages, each unitary corner extrusion 200 can be laser welded or otherwise suitably joined to like extrusions 200 (see, for example, FIGS. 6 and 7) and/or associated planar members 214, 216 (see, for example, FIGS. 8 and 9). Each of the unitary corners 200 may also be substantially identical, with the first planar portion 202 of one unitary corner 200 being joined to the second planar portion 204 of another unitary corner 200, in order that the unitary corners 200 are disposed side-by-side in an alternating opposing relationship (best shown in FIG. 7). Moreover, it will be appreciated that the assembly method can employ rigid-like tooling and/or fixtures (not shown) to hold the extruded unitary corners 200, planar members 214, 216 and/or subassemblies 210, 212, 222, 224, 226, 228. Therefore, weld distortion, which can be caused for example by heat, is minimized. Furthermore, compared to conventional welding processes that use consumable electrodes, laser welding results in minimal heat input, which should further result in better dimensional control. This will improve core cavity dimensions, particularly at the final stage of welding the four quarter perimeter segments 228 (FIGS. 11 and 12), whereas the current process results in significant machining being required after welding, in order to meet core cavity dimensional requirements. The extruded unitary corner design of the disclosed concept eliminates the welding of corner joints altogether and, therefore, eliminates a significant amount of machining work and time and cost associated therewith. Additionally, due to the possibility of less machining being necessary after welding, a further savings may be available by way of the ability to potentially use thinner extruded shapes. The possibility exists for the aforementioned subassemblies 210, 212, 222, 224, 226, 228 to be fabricated by qualified suppliers, in order that partial or full core shroud assemblies could be shipped to a designated manufacturing facility for final assembly or completion of core shroud 100 (FIG. 4; see also core shroud assembly 300 of FIG. 12). It will also be appreciated that the potential exists to incorporate the disclosed unitary corner concept with respect to other reactors internals assemblies (e.g., without limitation, baffle-former assembly corner joints (not shown)). For example and without limitation, an extruded unitary corner could be retro-fitted to an existing baffle-former design to eliminate corner joints between baffle plates (not shown) of a baffle-former assembly (not shown). Thus, the potential for undesirable “baffle jetting” is eliminated. “Baffle jetting is a result of water jetting from inside the baffle-former core cavity toward the direction of the core as a result of gaps or openings in the corner joint. FIGS. 13A-15 and 16A-18 respectively show interior flow deflectors 302 and exterior flow deflectors 304, 304′, in accordance with non-limiting example alternative embodiments of the disclosed concept. In the example of FIGS. 13A-15, the flow deflector 302, sometimes referred to as a “hockey puck” type of insert, is structured to be installed in the interior junction 112 of the core shroud corner 108, as shown in FIGS. 14 and 15. Specifically, each of the flow deflectors 302 includes a curved portion 306 and a number of substantially flat portions 308 (two are shown in FIGS. 13A-15). The curved portion 306 is structured to mate with the curved interior junction 112 of the core shroud interior corner 108, as shown in FIGS. 14 and 15. Thus, it will be appreciated that the dimensional characteristics of the flow deflector 302 (e.g., without limitation, height of the flow deflector 302) can be established to be consistent with, for example, the height of fuel assembly grids 46 (e.g., without limitation, fuel assembly grids 46, partially shown in phantom line drawing in FIG. 15). Thus, not only do the flow deflectors 302 deflect the flow of coolant so as to reduce axial velocity and resist undesired flow bypass, for example, by filling an open area in the corner (e.g., 108), but they can also serve to further support the grids 46. To secure the flow deflector 302 to the core shroud 100, a fillet weld (indicated generally be reference numeral 310 in FIG. 15) may be used, as shown for example, in FIG. 15. Implementation of the aforementioned exterior flow deflector 304 and 304′ for exterior corners 110 of the core shroud 100 is shown in the non-limiting examples of FIGS. 16A-17 and 18, respectively. As with the interior flow deflectors 302, previously discussed, the height of the exterior flow deflectors 304, 304′ can be made to be consistent with the height of the fuel assembly grids 46 (partially shown in phantom line drawing in FIG. 15). Each exterior flow deflector 304, 304′ preferably includes a curved portion 312 and a number of substantially flat portions 314 (two are shown) disposed generally opposite the curved portion 312. The curved portion 312 is structured to mate with, and be suitably joined (e.g., without limitation, welded) to, the curved exterior junction 114 of the exterior corner 110 of the core shroud, as shown in FIGS. 17 and 18. It will, however, be appreciated that any known or suitable alternative number, shape and/or configuration of flow deflectors (not shown) other than, or in addition to, those that are shown and described herein, could be employed without departing from the scope of the disclosed concept. For example and without limitation, FIG. 17 shows two different exterior flow deflectors 304, 304′ disposed on the exterior corner 110 of the core shroud 100, each of which has a different length. Additionally, as shown in exaggerated form in FIG. 18, the flow deflector 304 may include a chamfer 316. Such chamfer 316, which may for example, be formed as part of a fillet weld, would help, for example, to avoid undesired interaction (e.g., without limitation, a “snag”) with a grid during fuel assembly loading and unloading. It will also be appreciated that such flow deflectors (e.g., without limitation, 302, 304, 304′) could be incorporated independently (e.g., as a separate solution from the disclosed unitary corner concept), for example, with an existing core shroud design, to improve the flow-related problems associated with the corner joints thereof. It will, therefore, further be appreciated that such flow deflectors (e.g., without limitation, 302, 304, 304′) are not required in a least some embodiments in accordance with the disclosed concept. While specific embodiments of the disclosed concept have been described in detail, it will be appreciated by those skilled in the art that various modifications and alternatives to those details could be developed in light of the overall teachings of the disclosure. Accordingly, the particular arrangements disclosed are meant to be illustrative only and not limiting as to the scope of the disclosed concept which is to be given the full breadth of the claims appended and any and all equivalents thereof. |
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description | Priority is claimed to U.S. provisional patent application No. 61/663,316, filed Jun. 22, 2012, the disclosure of which is incorporated herein by reference in its entirety. The field of the present invention relates to neutron absorbing apparatus and methods used to facilitate close packing, of spent nuclear fuel assemblies. Nuclear power plants currently store their spent fuel assemblies on site for a period after being removed from the reactor core. Such storage is typically accomplished by placing the spent fuel assemblies in closely packed fuel racks located at the bottom of on-site storage pools. The storage pools provide both radiation shielding and much needed cooling for the spent fuel assemblies. Fuel racks often contain a large number of closely arranged adjacent storage cells wherein each cell is capable of accepting a spent fuel assembly. In order to avoid criticality, which can be caused by the close proximity of adjacent fuel assemblies, a neutron absorbing material is positioned within the cells so that a linear path does not exist between any two adjacent cells (and thus the fuel assemblies) without passing through the neutron absorbing material. Early fuel racks utilized a layer of neutron absorbing material attached to the cell walls of the fuel rack. However, these neutron absorbing materials have begun to deteriorate as they have been submerged in water for over a decade. In order to either extend the period over which the fuel assemblies may be stored in these fuel racks, it is necessary to either replace the neutron absorber in the cell walls or to add an additional neutron absorber to the cell or the fuel assembly. In an attempt to remedy the aforementioned problems with the deteriorating older fuel racks, the industry developed removable neutron absorbing assemblies, such as those disclosed in U.S. Pat. No. 5,841,825; U.S. Pat. No. 6,741,669; and U.S. Pat. No. 6,442,227. Neutron absorbing assemblies such as these have become the primary means by which adjacent fuel assemblies are shielded from one another when supported in a submerged fuel rack. Thus, newer fuel racks are generally devoid of the traditional layer of neutron absorbing material built into the structure of the fuel rack itself that can degrade over time. Instead, fuel assembly loading and unloading procedures utilizing neutron absorbing assemblies have generally become standard in the industry. In older racks, the neutron absorbing assemblies are added over the older, often degrading, layer of neutron absorbing material. While the neutron absorbing assemblies disclosed in the prior art have proved to be preferable to the old fuel racks having the neutron absorbing material integrated into the cell walls, these neutron absorbing assemblies are less than optimal for a number of reasons, including without limitation complexity of construction, the presence of multiple welds, complicated securing mechanisms, and multi-layered walls that take up excessive space within the fuel rack cells. Additionally, with existing designs of neutron absorbing assemblies, the inserts themselves must be removed prior to or concurrently with the fuel assemblies in order to get the fuel assemblies out of the fuel rack. This not only complicates the handling procedure but also leaves certain cells in a potentially unprotected state. The present invention is directed toward a neutron absorbing apparatus for insertion into spent fuel cell transport and/or storage systems. In a first separate aspect of the present invention, a neutron absorbing apparatus includes a corner spine and first and second walls, each affixed to the corner spine to form a chevron shape. Each wall includes an absorption sheet and a guide sheet. The absorption sheet is formed from a metal matrix composite having neutron absorbing particulate reinforcement and is affixed to the corner spine. The guide sheet is affixed to and covers an upper portion of the absorption sheet, and it also extends over a top of the absorption sheet. The absorption sheet extends along the corner spine along a greater length than the guide sheet. In a second separate aspect of the present invention, a neutron absorbing apparatus includes a corner spine and first and second walls, each affixed to the corner spine to form a chevron shape. Each wall includes an absorption sheet and a guide sheet. The absorption sheet is formed from a metal matrix composite having neutron absorbing particulate reinforcement and is affixed to the corner spine. The guide sheet is affixed to and covers an upper portion of the absorption sheet, and it also extends over a top of the absorption sheet. At least one of the walls also includes a locking protuberance coupled to the respective guide sheet and protruding through an opening formed in the respective absorption sheet. In a third separate aspect of the present invention, a system for supporting spent nuclear fuel in a submerged environment includes a fuel rack, a fuel assembly, and a neutron absorbing apparatus. The fuel rack includes an array of cells, with each cell being separated from adjacent cells by a cell wall. The fuel assembly is positioned within one of the cells, and the neutron absorbing, apparatus is also disposed within that cell. The neutron absorbing apparatus includes a corner spine and first and second walls, each affixed to the corner spine to form a chevron shape. Each wall includes an absorption sheet and a guide sheet. The absorption sheet is formed from a metal matrix composite having neutron absorbing particulate reinforcement and is affixed to the corner spine. The guide sheet is affixed to and covers an upper portion of the absorption sheet, and it also extends over a top of the absorption sheet. At least one of the cell wall in which the fuel assembly is disposed, adjacent the first wall or the second wall of the neutron absorbing apparatus, and the first wall or the second wall include a locking protuberance positioned to retain the neutron absorbing apparatus in the first cell during removal of the fuel assembly from the first cell. In a fourth separate aspect of the present invention, a method of retrofitting a spent nuclear fuel cell storage system includes inserting a neutron absorbing apparatus into one cell of an array of cells, wherein each cell is separated from each adjacent cell by a cell wall. The neutron absorbing apparatus includes a corner spine and first and second walls, each affixed to the corner spine to form a chevron shape. Each wall includes an absorption sheet and a guide sheet. The absorption sheet is formed from a metal matrix composite having neutron absorbing particulate reinforcement and is affixed to the corner spine. The guide sheet is affixed to and covers an upper portion of the absorption sheet, and it also extends over a top of the absorption sheet. At least one of the walls also includes a first locking protuberance coupled to the respective guide sheet and protruding through an opening formed in the respective absorption sheet. The method further includes creating a second locking protuberance in a first cell wall adjacent the neutron absorbing apparatus, wherein the first locking protuberance and the second locking protuberance are positioned to interlock to retain the neutron absorbing apparatus in the one cell. In a fifth separate aspect of the present invention, any of the foregoing aspects may be employed in combination. Accordingly, an improved neutron absorption apparatus for spent nuclear fuel pools and casks is disclosed. Advantages of the improvements will be apparent from the drawings and the description of the preferred embodiment. The description of illustrative embodiments according, to principles of the present invention is intended to be read in connection with the accompanying drawings, which are to be considered part of the entire written description, in the description of embodiments of the invention 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,” “left,” “right,” “top” and “bottom” as well as derivatives 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 unless explicitly indicated as such. 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. Moreover, the features and benefits of the invention are illustrated by reference to the preferred embodiments. Accordingly, the invention expressly should not be limited to such preferred embodiments illustrating some possible non-limiting, combinations of features that may exist alone or in other combinations of features, the scope of the invention being defined by the claims appended hereto. Turning in detail to the drawings, FIG. 1 shows a fuel rack 101 having an array of cells 103 into which spent nuclear fuel assemblies may be inserted. The fuel rack 101 may be part of a submerged storage system for spent nuclear fuel, or it may be part of a transportation system for spent nuclear fuel, such as dry or wet spent fuel casks. As shown, the cell walls include a feature for interlocking with a locking protuberance included as part of a neutron absorbing assembly inserted, into one or more of the cells. This feature may be a complementary locking protuberance, or a complementary receptacle to receive the locking protuberance of the neutron absorbing assembly. The feature may be created by bending, punching, welding, riveting, or otherwise permanently deforming the cell walls of the rack or the fuel cask, or by securing attachments to the cell walls, for holding the absorption assembly in place once it is inserted into the fuel cell. In some embodiments, if the fuel rack 101 has too small of a cell opening to accommodate thickness of the fuel insert, the insert may be directly inserted into the guide tubes of the fuel assembly. FIGS. 2-5 show a neutron absorbing assembly 111 which may be used in conjunction with both PWR or BWR storage requirements. The neutron absorbing assembly 111 is configured to be slidably inserted at strategic locations within the cell array of a submerged fuel rack. However, the absorbing assembly can be used in any environment (and in conjunction with any other equipment) where neutron absorption is desirable. Furthermore, based on the disclosed process for bending a metal matrix composite having neutron absorbing particulate reinforcement for the resulting angled plate structure), an absorbing assembly may be configured for use in any environment and/or used to create a wide variety of structures, including without limitation fuel baskets, fuel racks, sleeves, fuels tubes, housing structures, etc. The neutron absorbing assembly 111 includes a corner spine 113, to which are fastened two walls 115 to form a chevron-shaped structure (when viewed from the top or bottom). For a cell with a square cross-sectional configuration, the corner spine 113 creates a relative angle between the two walls 115 of about 90 degrees. Other relative angles may also be used, primarily depending upon the cross-sectional configuration of the cell into which the neutron absorbing assembly 111 is to be inserted (e.g, triangular, pentagonal, hexagonal, etc.). Each wall has an absorption sheet 117, constructed from a neutron absorbing material, and a guide sheet 119. Since the walls may be mirror images of each other, the following addresses the configuration of only one of the walls, with the understanding that the second wall may be similarly configured. However, in one embodiment, one of the walls includes a locking feature, and one does not. In other embodiments, both walls include a locking feature. In certain embodiments, additional corner spines and walls may be added to provide neutron absorption on more than two sides of a cell. The absorption sheet 117 is affixed to and extends much the length of the corner spine 113, and it may extend the entire length or only part of the length, depending upon the requirements for neutron absorption within the cell, e.g., the linear space within the cell occupied by the spent fuel rods. The absorption sheet 117 may be affixed to the corner spine 113 using any suitable fastener, such as rivets. The bottom edge 118 of the absorption sheet 117 has a skewed shape to facilitate ease of insertion of the neutron absorbing assembly 111 into a cell of a fuel rack. Specifically, the bottom edge 118 of the absorption sheet 117 tapers upward and away from the corner spine 113. The guide sheet 119 is affixed to only a top portion of the absorption sheet 117 by suitable fasteners, such as rivets, and the guide sheet 119 extends along less of a length of the corner spine 113 than the absorption sheet 117. The edge of the guide sheet 119 abuts up against the edge of the corner spine 113 along a common edge 121 to help reduce the overall thickness of the assembly. As shown in FIG. 2, the absorption sheet extends along most of the length of the corner spine 113, and the guide sheet 113 extends along a short top portion of the corner spine 113. The difference in lengths reflects the difference in functions between the absorption sheet 117 and the guide sheet 119. Where the absorption sheet 117 is included for neutron absorption, the guide sheet 119 is included, at least to aid in guiding a spent nuclear fuel assembly into the cell after the absorption assembly 111 is in place within the cell, to protect the top edge of the absorption assembly from damage, to provide a support surface for a locking protuberance, and to provide a structure by which the absorption assembly 111 may be supported during installation into the cell. The guide sheet 119 also includes an extension portion 123 which extends over and above the top edge 125 of the absorption sheet 117. This extension portion 123 provides a surface to aid in guiding a spent fuel assembly into a cell in which the absorption assembly is 111 placed. The extension portion 123 also protects the top edge 125 of the absorption sheet 117 from damage during the process of loading a spent fuel assembly into the cell. The top portion of each absorption sheet 117 includes a cut-out 126, and a tab 127 (which is a locking protuberance in the embodiment shown) extends from the guide sheet 119, through the cut-out 126, and beyond the outer surface of the absorption sheet 117. The tab 127 includes a lower part 129 affixed to the guide sheet, using any suitable fastener, such as rivets, and an upper part 131 which is bent away from the guide sheet 119 to extend through the cut-out 126. A locking protuberance may be formed in any other manner to provide the same locking functionality as described in connection with the tab herein. In addition, a locking protuberance may be included on both the absorption assembly 111 and the cell wall (See FIG. 6), or in other embodiments it may be included on only one of the absorption assembly 111 and the cell wall. As seen in FIG. 5, one suspension aperture 135 is included at the top of the corner spine 113, and one suspension aperture 137 is included in the extension portion 123 of each guide sheet 119. These suspension apertures 135, 137 are included to facilitate robotically placing the absorption assembly 111 in a cell within a submerged storage system. The shape and positioning of the suspension apertures is a matter of design choice. A single cell 151 for receiving a spent nuclear fuel assembly and an absorption assembly is shown in FIG. 6. Two walls of the cell 151 each include a feature 153 near the top of the cell wall 155, and the feature 153 is configured to engage the absorption assembly to retain the absorption assembly when the spent nuclear fuel assembly is removed from the cell. This feature 153 may be an indentation, a cut-out, or a protuberance, depending upon what type of corresponding locking, feature is included on the absorption assembly. The type of feature and its configuration are a matter of design choice. A detailed cross-sectional view of the locking features of the absorption assembly 111 and the cell 151 are shown in FIG. 7. As described above, the locking feature may be a tab, and such a tab 127 is shown with its top portion 131 in locking engagement with a second tab 161, this second tab 161 being formed in the cell wall 155. When manufacturing the absorption assembly for a fuel rack that has not yet been placed in service the order of making the locking protuberances, the type of locking protuberance used, and even whether one or both of the cell wall and the absorption assembly include a locking protuberance, are anticipated to be variables that may be addressed by design decisions for a particular configuration. However, when retrofitting a fuel rack or cask that is already in use, and a tab is used in the cell wall as a locking protuberance, preferably the absorption assembly is first manufactured and placed into the cell before the tab in the cell wall is created. This permits maximization of space use within a pool or cask by minimizing the space requirements of the absorption assembly, because the tab effectively reduces the overall nominal width of the cell. When retrofitting an existing and in-use fuel rack or cask, the tab 161 in the cell wall may be formed just above the position of the tab in the absorption assembly as a half-shear using a C-shaped tool which spans the extension portion 123 of the guide sheet 119. With such a tool, a double-acting hydraulic cylinder may be used to push a wedge-shaped piece of the tool into the cell wall, thereby creating the half-sheared tab 161 extending toward the inner space of the cell. The cell 151 has an overall length L, and the corner spine is configured to have approximately the same length, as shown in FIG. 8. As shown, the corner spine 113 clears the top 157 of the cell by a sufficient amount to make the suspension aperture 135 of the corner spine 113 accessible, even when the spent nuclear fuel assembly 159 is placed within the cell 151. The length of the corner spine 113 is such that the bottom edge 162 rests against the bottom 163 of the cell 151. The absorption sheet 117 need not extend all the way to the bottom 163 of the cell 151, as the length of the absorption sheet 117 may extend as far down into the cell as needed so that it shields adjacent fuel assemblies from one another. This is because adjacent spent nuclear fuel rods may not extend the entire length of the cell either, and the length of the absorption sheet 117 need only be as long as the spent nuclear fuel rods within the spent nuclear fuel assembly 159, although they may be longer if desired. Since there is a need to maximize space use within a fuel pond or cask, it is desirable that the absorption assembly 111 take up as little room as possible in the cell of the fuel rack. To this end, the absorption sheets 117 are preferably constructed of an aluminum boron carbide metal matrix composite material having a percentage of boron carbide greater than 25%. While the addition of boron carbide particles to the aluminum matrix alloy increases the ultimate tensile strength, increases yield strength, and dramatically improves the modulus of elasticity (stiffness) of the material, it also results in a decrease in the ductility and fracture toughness of the material compared to monolithic aluminum alloys. The boron carbide aluminum matrix composite material of which the absorption sheets are constructed includes a sufficient amount of boron carbide so that the absorption sheets can effectively absorb neutron radiation emitted from a spent fuel assembly, and thereby shield adjacent spent fuel assemblies in a fuel rack from one another. The absorption sheets may be constructed of an aluminum boron carbide metal matrix composite material that is about 20% to about 40% by volume boron carbide. Of course, other percentages may also be used. The exact percentage of neutron absorbing particulate reinforcement which is in the metal matrix composite material, in order to make an effective neutron absorber for an intended application, will depend on a number of factors, including the thickness (i.e., gauge) of the absorption sheets 117, the spacing between adjacent cells within the fuel rack, and the radiation levels of the spent fuel assemblies. Other metal matrix composites having, neutron absorbing particulate reinforcement may also be used. Examples of such materials include, without limitation, stainless steel boron carbide metal matrix composite. Of course, other metals, neutron absorbing particulate and combinations thereof may be used including without limitation titanium (metal) and carborundum (neutron absorbing, particulate). Suitable aluminum boron carbide metal matrix composites are sold under the trade names Metamic® and Boralyn®. The center spine, the guide sheets, and the locking protuberance may be formed from steel or other materials, or they may alternatively be formed from nonmetallic materials. When the locking protuberance is configured as a tab affixed to the guide sheet of the absorption assembly, the tab is preferably formed from a sheet of 301 stainless spring steel, tempered to about 3/4 hard. In a preferred embodiment, the tab is about 0.035 inches thick, about 0.7 inches wide, and about 1.7 inches long, with the upper portion of the tab being about 1.09 inches long and bent to extend beyond the outer side of the absorption layer by between 0.125 inches to 0.254 inches, depending upon how thick the absorption layer is and whether the absorption assembly is being placed over an existing absorption layer within the cell. In the latter instance, the tab should be configured so that the upper portion extends beyond the existing absorption layer. The extent to which the tab extends beyond the absorption layer is a matter of design choice, as it depends upon several factors such as the type of locking feature included on the cell wall, how much the tab needs to deflect upon insertion, and how much removal force the tab should be able to withstand. For example, with a tab extending 0.125 inches beyond the absorption layer, it may be desirable to have the tab be able to deflect by approximately 0.124 inches upon insertion. Such a configuration is anticipated to withstand at least a 200 lb removal force once the tab is interlocked with a second tab formed in the cell wall. It should be noted that the tab will remain in a substantially deflected state once the absorption assembly is inserted into cell wall. While the invention has been described with respect to specific examples including presently preferred modes of carrying out the invention, those skilled in the art will appreciate that there are numerous variations and permutations of the above described systems and techniques. It is to be understood that other embodiments may be utilized and structural and functional modifications may be made without departing from the scope of the present invention. Thus, the spirit and scope of the invention should be construed broadly as set forth in the appended claims. |
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047073306 | abstract | A metallic composite material and nuclear components such as fuel cladding, rod guide thimbles, grids and channels made therefrom. The metallic composite material comprises 90-60 volume percent of a metal matrix of zirconium or a zirconium alloy containing homogeneously incorporated, throughout the matrix, 10-40 volume percent of silicon carbide whiskers. |
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abstract | A photo-mask for use in extreme ultraviolet (EUV) lithography, in which the photo-mask has low coefficient of thermal expansion and high specific stiffness. |
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046831134 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS The present invention has been attained as a result of a minute study on the characteristics of the known fuel assembly disclosed in Japanese Patent Laid-Open No. 29878/1983, as will be understood from the following description. First of all, an explanation will be made with reference to the fuel assembly shown in U.S. Pat. No. 4,229,258 which provides a base for the fuel assembly shown in Japanese Patent Laid-Open No. 29878/1983. The fuel assembly of the U.S. Patent mentioned above exhibits a uniform power distribution along the axis thereof. In the fuel assembly used in a boiling water reactor, however, voids are generated within the channel box, whereas no void is generated outside the channel box. Therefore, in FIG. 4 attached to the U.S. Patent mentioned above, a non-uniform density distribution of water (moderator) appears in a vertical section which contains a corner of the channel box facing a control rod and another corner which is on the diagonal line passing the first-mentioned corner. More specifically, the water density is lower in the inside of the channel box than in the outside of the channel box. Therefore, the density of the thermal neutron flux .phi. exhibits such a distribution in the section containing above-mentioned two corners of the channel box that the density is low at the mid portion of the section and high outside the channel box, as will be seen from FIG. 1. The power of each of the fuel rods constituting the fuel assembly is given by the following formula: EQU P=.phi..multidot..sigma..sub.F .multidot.N (1) where, .phi. represents the thermal neutron flux density at the position of the fuel rod, .sigma..sub.F represents the fission cross-sectional area, and N represents the atomic number density of the fissile material in the fuel rod. In the known fuel assemblies such as that shown in U.S. Pat. No. 4,229,258, the atomic number density N (which is in proportion to enrichment e) of the fissile material in the fuel rods disposed in the peripheral region of the fuel assembly where the thermal neutron flux density .phi. is high is selected to be small as compared with that of the fuel rods in the central region of the fuel assembly as shown in FIG. 2, in order to flatten the power distribution of each fuel rod (referred to as local power distribution, hereinunder) thereby minimizing the local power peaking which is the ratio between the maximum power of the fuel rod and the mean power of the fuel assembly. For instance, in the fuel assembly shown in the above-mentioned U.S. Patent, the fuel rods adjacent the channel box have a mean enrichment of uranium 235 which is about 25 to 50% lower than that in the fuel rods in the central region of the fuel assembly. When a reactor core is charged with new fuel assemblies, the excess reactivity in the reactor core is so large as can never be controlled by the control rods solely. In order to suppress the excess reactivity in the beginning period of the burning, several fuel rods in the fuel assembly are made to contain gadolinea as a burnable poison as mentioned before. This burnable poison has an extremely large neutron absorption cross-sectional area, so that it is decreased more rapidly than the uranium 235 as the time elapses. Thus, the burnable poison is extinguished completely after a certain period of time so that the reactivity is not affected by the burnable poison in the later period of burning. FIG. 3 illustrates how the reactivity is suppressed by the use of gadolinea. More specifically, the full-line curve shows the infinite multiplication factor of the fuel assembly which contains gadolinea, while the broken-line curve shows the infinite multiplication factor of the fuel assembly which does not contain gadolinea. Thus, the reactivity suppressing effect produced by the gadolinea is shown as the difference between the values on both curves. When the fuel assembly contains the gadolinea, the infinite multiplication factor is linearly increased in accordance with the increment of the burn-up degree and, after exhibiting a peak at exposure around 10 GWd/t at which the gadolinea is burnt out, decreases linearly in accordance with the increase in of exposure, as shown by the full-line curve in FIG. 3. The period in which the infinite multiplication factor increased up to the peak will be referred to as "earlier burning period", while the period after the peak will be referred to as "later burning period". The fuel assembly shown in Japanese Patent Laid-Open No. 29878/1983 can provide a higher exposure without imparing the flat axial power distribution proposed by the above-mentioned U.S. Patent. Namely, the fuel assembly can be burnt for a longer period of time. FIG. 4 shows the cross-section of the fuel assembly as shown in Japanese Patent Laid-Open No. 29873/1983. This fuel assembly 23 is constituted by fuel rods 1 to 6, and G.sub.1 and G.sub.2. These fuel rods 1 to 6 and G.sub.1, G.sub.2 exhibit enrichment distributions and gadolinea density distributions as shown in FIG. 5. It will be seen that the fuel rods G.sub.1 and G.sub.2 have uniform gadolinea density distribution along the axis thereof. Both fuel rods G.sub.1 and G.sub.2 have an equal gadolinea density Gd.sub.0. The fuel rods 1 to 6 do not contain gadolinea. Each of the fuel rods 1 to 6 and G.sub.1, G.sub.2 has a clad tube charged with pellets of UO.sub.2 as the fissile material. The fuel rods 1 to 6 and G.sub.1, G.sub.2 have enrichments e.sub.1 to e.sub.6 as shown in FIG. 5. These enrichment are selected to meet the condition of e.sub.1 >e.sub.2 >e.sub.3 > e.sub.4 >e.sub.5 >e.sub.6. In this fuel assembly 23, fuel rods having mean enrichments greater than the mean enrichment of the fuel assembly are disposed in a large number in the peripheral region, whereas fuel rods having a mean enrichment lower than that of the fuel assembly are disposed in the central region of the fuel assembly. In this fuel assembly 23, the power of the fuel rods in the peripheral region is increased so that the power is increased locally in the peripheral region of the fuel assembly throughout the period of operation of the nuclear reactor. The infinite multiplication factor of the fuel assembly 23 is increased substantially in linear proportion to the increase of the local power in the peripheral region. Therefore, the increase in the infinite multiplication factor of the fuel assembly can be maximized by increasing the local power peaking. The maximum value of the local power peaking is restricted by the thermal condition of the fuel rod, so that the increase in the infinite multiplication factor is materially limited. For attaining a further increase in the infinite multiplication factor, it is necessary to increase the local powers of the fuel rods in the peripheral region of the fuel rods. This, however, must be done equally over all fuel rods in the peripheral region. Thus, the local power distribution and enrichment distribution which will maximize the increase in the reactivity are determined on condition that the mean enrichment and the maximum value of the local power peaking in the fuel assembly 23 are given. FIG. 7 shows an example of the optimum local power distribution in the peripheral region, particularly in the outermost peripheral region, of the fuel assembly when the local power peaking factor is 1.30. In this Figure, each square represents each fuel rod. The outer peripheral region of the fuel assembly 23 includes both the fuel rods having high enrichment intended for attaining high local power peaking and fuel rods having two axial regions of different enrichments intended for suppressing the axial power peaking. In this fuel assembly 23, the axial power peaking is suppressed such as to allow a corresponding increase in the local power in the peripheral region, thus attaining a greater reactivity gain. This fuel assembly 23, however, proved the following disadvantage. Namely, since the fuel rods 2 and 3 having axial regions of different enrichments are disposed in the peripheral region as shown in FIG. 4, the local power distribution is deviated from the optimum power distribution shown in FIG. 7 either in the upper region or the lower region of the fuel assembly 23. This makes it impossible to attain the local power distribution for maximizing the reactivity both in the upper and lower regions of the fuel assembly. FIGS. 8 and 9 show, respectively, the local power distributions at the upper and lower regions of the fuel assembly 23 in the beginning of burn-up. These local power distributions cannot provide a uniform power peaking of, for example, 1.30 over the peripheral region, particularly in the outermost region, of the fuel assembly, unlike the local power distribution shown in FIG. 7. In addition, a difference of local power distribution is produced in the outermost region between the upper and lower regions of the fuel assembly 23. As a result of an intense study on the characteristics of the fuel assembly 23, the present inventors have found that the above-explained problems of the known fuel assemblies can be obviated by providing a difference in the enrichment between the upper and lower regions of each fuel rod which contain a burnable poison over almost entire axial region thereof. More specifically, the power P of the fuel rod is proportional to the product of the enrichment e and the thermal neutron flux density .phi.. Namely, the condition of P=e.multidot..phi. is met. Therefore, if a fuel rod having a large thermal neutron flux density .phi. or a large power P is made to have such an enrichment distribution as being high in the upper region than in the lower region, the change in the power caused by a given change in the enrichment e becomes excessively large, resulting in a large difference in the power between the upper and lower regions of the fuel rod. In contrast to the above, in the case of a fuel rod which contains gadolinea over almost entire axial region thereof, the thermal neutron flux density .phi. is small due to the presence of gadolinea, so that only a small change in the power P is caused by a given change in the enrichment e. Thus, the fuel rod containing gadolinea over almost entire axial region thereof exhibits a comparatively small change in the axial power distribution. For this reason, when the difference in the enrichment between the upper and lower regions is provided in a fuel rod which contains gadolinea over almost entire axial region thereof, the local power distributions in the upper and lower regions of the fuel rod are substantially equalized. The present invention is based upon the discovery explained hereinbefore. The invention will be fully understood from the following description of the preferred embodiments. FIG. 10 shows a preferred embodiment of the fuel assembly in accordance with the invention. The fuel assembly 30 has a channel box 10, a lower tie plate 11, an upper tie plate 12, spacers 15 and fuel rods 16. The fuel rods 16 are held at their lower and upper ends by the lower tie plate 11 and the upper tie plate 12. A plurality of the spacers 15 are arranged in the axial direction such as to maintain predetermined gaps between adjacent fuel rods. The channel box 10 is secured to the upper tie plate 12 and surrounds the bundle of the fuel rods 16 held by the spacers 15. A channel fastener 13 is secured to the upper tie plate 12. FIG. 11 shows the detail of the fuel rod 16. The fuel rod 16 has a clad tube 20 charged with a multiplicity of fuel pellets 21 and closed at its upper and lower ends by means of upper and lower end plugs 17 and 18. The fuel pellets 21 are pressed by a spring 22 disposed in a gas plenum defined in the clad tube 20. FIG. 12 is a sectional view of the fuel assembly taken along the line XII--XII of FIG. 10. Fuel rods 16 are arranged in a lattice-like form within the channel box 10. Two water rods 14 are disposed in the central region of the channel box 10. There are several fuel rods 16 which contain gadolinea as a burnable poison. Water gaps are formed between adjacent fuel assemblies. These water gaps are adapted to receive control rods 19. The fuel rods 16 disposed in the fuel assembly 30 can be sorted into 6 (six) kinds: namely, fuel rods 31, 32, 33, 34, G.sub.4 and G.sub.5 which have enrichment distributions and gadolinea distributions as shown in FIG. 13. These fuel rods 31 to 34, G.sub.4 and G.sub.5 are disposed within the channel box 10 in a pattern shown in FIG. 12. The fuel rods 31 to 34 contain fuel pellets 21 of UO.sub.2 as the nuclear fuel. These fuel pellets contain uranium 235 as the fissile material. These fuel rods 31 to 34 do not contain gadolinea as the burnable poison. On the other hand, the fuel rods G.sub.3 and G.sub.4 have UO.sub.2 fuel pellets which contain gadolinea together with the uranium 235. The enrichments e.sub.1 to e.sub.4 in the fuel rods shown in FIG. 13 are determined to meet the conditions of e.sub.1 >e.sub.2 >e.sub.3 > e.sub.4. The fuel rods 31 to 34 and G.sub.4 have a uniform enrichment over the entire axial length thereof. The upper region of the fuel rod G.sub.4 above the level of 11/24 of the effective length of fuel as measured from the bottom of the effective length of fuel uniformly contains gadolinea, while the lower region below the abovementioned level does not contain gadolinea. The term "effective length of fuel" means the length or region of the fuel rod charged with the nuclear fuel material, i.e., the fuel pellets. The fuel rod G.sub.3 contains gadolinea uniformly over the entire axial length thereof. The fuel rods G.sub.3 and G.sub.4 have an equal density of gadolinea. In the fuel rod G.sub.3, the upper region above the level of 11/24 of the entire length as measured from the bottom of the fuel has a higher enrichment that the lower region below the above-mentioned level. Namely, the fuel rod G.sub.3 has upper and lower regions having different enrichments, but the enrichment is uniform in each of the upper and lower regions. The fuel assembly 30 having the fuel rods G.sub.3 and G.sub.4 naturally have two regions: namely, an upper region above the level of 11/24 of the effective fuel length as measured from the bottom of the effective fuel length and a lower region below the above-mentioned level. The mean enrichment in a plane perpendicular to the axis of the fuel assembly within the upper region is greater than that in a plane perpendicular to the axis of the fuel assembly within the lower region thereof. In addition, the amount of gadolinea contained by the upper region of the fuel assembly 30 is greater than that contained by the lower region of the same. It is to be understood also that the upper region of the fuel assembly 30 has a greater infinite multiplication factor than the lower region. Thus, the amount of gadolinea contained by the upper region is greater than that contained by the lower region. This axial gadolinea distributions serves to provide a smaller infinite multiplication factor than in the lower region of the fuel assembly 30. On the other hand, the upper region of the fuel assembly 30 has a greater mean enrichment than the lower region of the same. This enrichment distribution serves to provide a greater infinite multiplication factor in the upper region than in the lower region of the fuel assembly. In the fuel assembly of the invention, the mean enrichment in the upper region is selected to be large enough to compensate for any reduction in the infinite multiplication factor due to the presence of the gadolinea in the upper region, so that the fuel assembly as a whole exhibits a greater infinite multiplication factor in the upper region than in the lower region. Referring to FIG. 12 showing the fuel assembly 30 in a cross-section perpendicular to the axis thereof, two regions are assumed in this cross-section of the fuel assembly: namley, a peripheral region outside the one-dot-and-dash line L which is an annular region having two layers of fuel rods, and a central region inside the one-dot-and-dash line and having three and four layers of fuel rods. In the described embodiment of the fuel assembly, the mean enrichment in the peripheral portion is greater than that in the central region. As stated before, in the described embodiment of the fuel assembly, the axial enrichment distribution is created by providing an axial enrichment distribution in the fuel rods G.sub.3 which contain gadolinea over almost the entire axial region thereof and, therefore, the upper and lower regions of the fuel rod G.sub.3 has a substantially equal power distribution. Consequently, the difference in the local power between the peripheral region of the upper region and the peripheral region in the lower region is minimized. In fact, the local powers of these peripheral regions become substantially equal to each other. This effect is maximized because the fuel rods G.sub.3 are disposed in the portion of the peripheral region except the outermost portion. Consequently, the reactivity is increased and a higher fuel economy is attained. It is to be noted also that, while the known fuel assembly mentioned before employs 8 kinds of fuel rods, the described embodiment of the fuel assembly of the invention employs only six kinds of fuel rods, thus remarkably simplifying and facilitating the production of the fuel assembly. Furthermore, the described embodiment of the fuel ssembly provides the same advantage as that offered by the fuel assembly shown in FIG. 4 of Japanese Patent Laid-Open No. 26292/1983, i.e., a longer period of burning of the fuel assembly, because the mean enrichment is greater in the peripheral region than in the central region. The described embodiment of the fuel assembly also produces the same effect as that provided by the fuel assembly shown in FIG. 4 of U.S. Pat. No. 4,229,258, i.e., a flat or uniform axial power distribution of the fuel assembly, because the mean enrichment is higher in the upper region than in the lower region of the fuel assembly. This effect becomes appreciable after the burning of the gadolinea in the fuel assembly 30. This effect eliminates the use of control rods which are to be inserted only to small depth, and the power of the nuclear reactor can be controlled only by means of control rods which are to be inserted to a large depth. Consequently, the control operation for the control rods can be remarkably simplified. Preferably, the boundary between the upper and lower regions is positioned within the range between 1/3 and 7/12 of the fuel effective length as measured from the bottom of the fuel effective length. In the described embodiment of the fuel assembly, not only the enrichment but also the amount of gadolinea is greater in the upper region than in the lower region. This in turn produces an effect called "spectrum shift" which is stated in lines 7 to 27, page 10 of the specification of U.S. patent application No. 548,845 and shown in FIGS. 5 to 7 attached to this U.S. patent specification. This spectrum shift effect also contributes to an increase in the discharged exposure of fuel burn-up, i.e., to a prolongation of period of burning of the fuel. Another embodiment of the invention will be described hereinunder with reference to FIGS. 14 and 15. The fuel assembly 40 of this embodiment has six kinds of fuel rods, i.e., fuel rods 41 to 44, G.sub.5 and G.sub.6, which are arranged in a manner shown in FIG. 14 and having enrichments and gadolinea densities as shown in FIG. 15. As will be understood from a comparison between FIG. 13 and FIG. 15, the fuel rods 41 to 44, G.sub.5 and G.sub.6 used in this embodiment are similar to the fuel rods 31 to 34, G.sub.3 and G.sub.4 of the first embodiment shown in FIG. 13, except that they are provided at their one or both ends with layers of natural uranium e.sub.5. Usually, the power is not so large at each axial end of the fuel rod, so that only a small discharged exposure is attained at such axial ends even if these axial ends are charged with enriched uranium. Rather, the use of enriched uranium in such axial ends leads to a wasteful use of the uranium. From this point of view, the embodiment of the fuel assembly shown in FIGS. 14 and 15 employs layers of natural uranium in one or both ends of the fuel rods, thus minimizing the wasteful use of the uranium. The fuel rods G.sub.5 and G.sub.6 do not have the layer of natural uranium e.sub.5 in their upper ends. The enrichment e.sub.4 has a greater content of uranium 235 than the natural uranium e.sub.5. These fuel rods G.sub.5 and G.sub.6 contain gadolinea and, therefore, produces large volume of gases during the operation of the nuclear reactor. In this embodiment of the fuel assembly, a sufficiently large volume of gas plenum is provided on the upper end of each of the fuel rods G.sub.5 and G.sub.6 which are devoid of the layers of natural uranium e.sub.5. The effective fuel length of the fuel assembly 40 as a whole is equal to that of the fuel rods 41 to 44. The length of the region charged with the natural uranium is 1/24 of the effective fuel length. The fuel rods G.sub.5 and G.sub.6 containing gadolinea are sectioned axially into two regions: namely, an upper region above the level 11/24 of the fuel effective length as measured from the bottom of the fuel effective length and a lower region below the above-mentioned level. The fuel rod G.sub.5 contains gadolinea uniformly over almost the entire axial region thereof except the lower end constituted by the natural uranium e.sub.5. The enrichment in the most part of the lower region of the fuel rod G.sub.5 except the lower end portion having the natural uranium is lower than the enrichment in the most part of the upper region thereof. Each of the fuel rods 41 to 44 and G.sub.6 has a substantially uniform enrichment distribution over the most part of the axial region thereof except the portions charged with the natural uranium. The upper region of the fuel rod G.sub.6 uniformly contains gadolinea at a density equal to that in the fuel rod G.sub.5. The fuel assembly 40 of this embodiment is materially identical to the fuel assembly 30 of the first embodiment except that the fuel rods are charged at their one or both axial ends with natural uranium. The enrichments and the gadolinea densities of the fuel rods 41 to 44, G.sub.5 and G.sub.6 are shown in the following table. TABLE 1 ______________________________________ No. of Fuel Rods 41 42 43 44 G.sub.5 G.sub.6 ______________________________________ Upper Enrichment 4.1 3.8 3.2 2.5 3.8 2.5 region wt % Gadolinea 0 0 0 0 3.5 2.0 density wt % Lower Enrichment 4.1 3.8 3.2 2.5 2.5 2.5 region wt % Gadolinea 0 0 0 0 3.5 0 density wt % ______________________________________ FIG. 16 shows the local power distribution in the outermost portion of the upper region in the fuel assembly 40 in the beginning of burn-up, while FIG. 17 shows the local power distribution in the outermost portion of the lower region of the fuel assembly 40 in the beginning of burn-up. From these Figures, it will be seen that the local power in the outermost portion in the upper region is almost equal to that in the outermost portion in the lower region. Thus, the fuel assembly 40 of this embodiment offers the same advantage as that produced by the fuel assembly 30 of the first embodiment. FIG. 18 shows still another embodiment of the invention. The fuel assembly 50 of this embodiment has a construction similar to that of the fuel assembly 40 of the second embodiment, except that four water rods disposed in the central portion of the fuel assembly 40 is substituted by a single large water rod 51. Thus, the fuel assembly 50 of this embodiment produced substantially the same effect as that produced by the fuel assembly 40. FIG. 19 shows a further embodiment of the invention. The fuel assembly 60 of this embodiment employs the fuel rods 31, 32, 33 and 34 used in the first embodiment explained in connection with FIG. 13. In this embodiment, however, these fuel rods are arranged in a manner shown in FIG. 19. It will be seen also that the fuel assembly 60 of this embodiment employs a fuel rod G.sub.7 in place of the fuel rod G.sub.3 used in the first embodiment. The fuel rod G.sub.7 is materially identical to the fuel rod G.sub.3 except that its lower region has a mean enrichment e.sub.3 in contrast to the fuel rod G.sub.3 which has a mean enrichment e.sub.4 in its lower region. The fuel assembly 60 of this embodiment has a uniform distribution of gadolinea because it is devoid of the fuel rod G.sub.4 shown in FIG. 13. This embodiment, therefore, cannot produce the spectrum shift effect which is obtained with the fuel assembly 30 of the first embodiment. Therefore, the fuel assembly 60 of this embodiment produces all the advantages produced by the fuel assembly 30 other than the advantage derived from the spectrum shift effect. FIG. 20 shows a further embodiment of the invention. The fuel assembly 70 of this embodiment employs fuel rods 31 to 34 shown in FIG. 13 and fuel rods 75 and G.sub.8 shown in FIG. 21. These fuel rods are arranged in a manner shown in FIG. 20. The fuel rod 75 and G.sub.8 has a greater enrichment in the upper region thereof above the level 1/2 of the fuel effective length as measured from the bottom of the same than in the lower region below the above-mentioned level. The fuel rod 75 does not contain gadolinea. The fuel rod G.sub.8 has a greater gadolinea density Gd.sub.2 in its upper region above the level 1/2 of the fuel effective length as measured from the bottom of the same than that Gd.sub.3 in the lower region thereof below the above-mentioned level. Thus, the fuel assembly 70 of this embodiment employs lower enrichment fuel rods which do not contain gadolinea and which have upper and lower regions of different enrichments. Therefore, the difference in the local power between the upper and lower regions can be reduced as compared with that in the known fuel assembly 23, although the difference is larger than that in the fuel assembly 30 of the first embodiment. The fuel assembly 70 of this embodiment produces effects substantially the same as those produced by the fuel assembly 30 except the point mentioned above. FIG. 22 shows a further embodiment of the invention. The fuel assembly 80 of this embodiment employs the aforementioned fuel rods 31 to 34, 75 and G.sub.8 arranged in a manner shown in FIG. 22. This fuel assembly is similar to the fuel assembly 70 of the preceding embodiment except that some of the fuel rods 75 in the central region thereof are substituted by the fuel rods 34. The fuel assembly 80 of this embodiment exhibits a difference in the local power between the upper and lower regions which is reduced as compared with that in the fuel assembly 70 by an amount corresponding to the number of reduction of the fuel rods having difference of enrichment between their upper and lower regions. FIG. 23 shows a still further embodiment of the invention. The fuel assembly 90 of this embodiment employs fuel rods 41 to 44 and G.sub.6 shown in FIG. 15 and fuel rods G.sub.9 shown in FIG. 24. These fuel rods are arranged in a manner shown in FIG. 23. The fuel rod G.sub.9 has a length smaller than that of the fuel rods 41 to 44 by amount corresponding to the length of natural uranium layer e.sub.5 provided in the fuel rods 41 to 44. It is to be noted also that the fuel rod G.sub.9 has an upper-most region of a length within 3/24 of the fuel effective length (this equals to effective length of fuel rods 41 to 44). This uppermost region is charged with fuel pellets of low enrichment e.sub.4. Thus, the fuel rod G.sub.9 has three axial regions besides the lowermost region of natural uranium e.sub.5. This fuel assembly 90 has low enrichment at the upper ends of the fuel rods G.sub.9 so that the infinite multiplication factor in the cold state of the reactor can be suppressed effectively, so that a large reactor shut-down margin can be preserved. The axial distribution of enrichment in the fuel assembly 90 is created by providing a difference in the enrichment between the upper and lower regions of the fuel rods which do not contain gadolinea, so that the difference in the local power between different axial regions can be suppressed as in the case of the fuel assembly 40 explained before. FIG. 25 shows a still further embodiment of the invention. The fuel assembly 95 of this embodiment employs the fuel rods 41 to 44, G.sub.5 and G.sub.6 shown in FIG. 15 and the fuel rods 84 shown in FIG. 26. These fuel rods are arranged in a manner shown in FIG. 25. The fuel rod 84 has the total length of layer of natural uranium e.sub.5 greater than that in the fuel rods 41 to 44. In fact, the length of the layer of natural uranium in the fuel rod 84 reaches 1/6 of the fuel effective length of this fuel rod. The enrichment in the upper end of the fuel assembly and, hence, the reactor shut-down margin is increased also in this case. The difference in the local power distribution in the outer peripheral region between different cross-sections of the fuel assembly can be reduced provided that the fuel rods 84 are disposed in the portion of the cross-section of the fuel assembly other than the outer peripheral region and that the length of the layer of the natural uranium is less than 1/6 of the effective fuel length. Thus, the fuel assembly 95 of this embodiment produces substantially the same effect as those produced by the fuel assembly 40 explained before. The increase in the reactor shut-down margin through a reduction of the enrichment in the upper end portion of the fuel assembly and the constant local power distribution in the outer peripheral portion of the fuel assembly over almost the entire axial region of the fuel assembly are attainable also by a fuel assembly 100 of a still further embodiment of the invention shown in FIG. 27. The fuel assembly 100 employs fuel rods 41 to 44, G.sub.5 and G.sub.6 shown in FIG. 15 and fuel rods 85 shown in FIG. 28. These fuel rods are arranged in a manner shown in FIG. 27. The fuel rod 84 is provided at its upper and lower ends with layers of natural uranium e.sub.5 each having an axial length of 1/24 of the fuel effective length thereof. In addition, the fuel rod 84 is provided with a region of a small enrichment e.sub.4 (e.sub.4 <e.sub.3) which extends over a length of 1/8 of the fuel effective length downwardly from the lower end of the upper layer of the natural uranium e.sub.5. The axial enrichment distribution of the fuel rod 85 can be regarded as being materially constant, provided that the length of the region of reduced enrichment e.sub.4 is less than 1/8 of the fuel effective length. By arranging the fuel rods 85 in the central region of the cross-section of the fuel assembly, it is possible to increase the local power in the outer peripheral region of cross-sections of the fuel assembly as a mean and the difference in the local power distribution between different axial regions can be minimized as in the case of the fuel assembly 40. As has been described, according to the invention, it is possible to remarkably reduce the difference in the power distribution between different cross-sections taken at different positions along the axis of the fuel assembly, thereby attaining a higher discharged exposure of the fuel assembly and, hence, a higher fuel economy. |
047117605 | abstract | An improved nuclear reactor wherein the bolted connections between the baffle plates and baffle former and the baffle former and the core barrel are provided with a noval locking device that prevents loosening of the bolted connections. The locking device to prevent loosening of bolted connections where a threaded bolt is used to secure two components together uses a threaded lock nut over the bolt head, cooperative with a threaded wall in a recess in a component. The lock nut and wall of the recess have threads opposite the threads of the bolt. Deformable sections are formed in the base of the lock nut to engage an unsymmetrical cavity in the head of the bolt. Turning of the bolt in a direction that would tend to loosen the connection causes tightening of the lock nut. The wall of the lock nut may be threadedly engaged with the wall of the recess or the lock nut wall may have slits formed thereon which enable deformation of wall portions to engage the threaded wall of the recess. |
RE0298760 | abstract | A container for transporting radioactive materials utilizing a removable system of heat conducting fins is provided which permits a substantial reduction in the weight of the container during transport, increases the heat dissipation capability of the container and substantially reduces the scrubbing operation after loading and unloading the radioactive material from the container. The detachable fins are made of a light weight highly heat conductive metal such as aluminum or aluminum alloys. |
description | This is a national stage application under 35 U.S.C. § 371 of PCT/US2013/027663 filed Feb. 25, 2013, the entire contents of which are incorporated herein by reference and this application claims priority to U.S. Provisional Application No. 61/603,145 filed Feb. 24, 2012, the entire contents of which are hereby incorporated by reference. This invention was made with Government support under: W81XWH-10-1-1049, awarded by the U.S. Army, Medical Research and Materiel Command. The Government has certain rights in the invention. The field of the currently claimed embodiments of this invention relates to charged particle acceleration devices, and more particularly to triboelectric charged particle acceleration devices. Triboelectricity has been utilized in fundamental scientific research as a source of high electrostatic potential for over three centuries, from the early electrostatic apparatus of Haukesbee (F. Haukesbee, Physico-Mechanical experiments on various subjects (London: 1709)) through to the eponymous generators of van der Graaf, yet there remains a notable absence of a first principles approach to the subject (M. Stoneham, Modelling Simul. Mater. Sci. Eng. 17, 084009 (2009)). Electrostatic generators store the integrated charge that is developed when two materials are rubbed together in frictional contact. The materials are selected to be furthest apart in the triboelectric series—an empirically derived list showing both the propensity of the materials to charge and the polarity of charge (P. E. Shaw, Proc. R. Soc. Lond. A 94, 16 (1917)). At the point of contact between the two materials, the frictional electrification may be of such magnitude that it may ionize the gas surrounding it, creating triboluminescence. The triboluminescence observed during peeling pressure sensitive adhesive (PSA) tape has long attracted scientific attention (E. N. Harvey, Science 89, 460 (1939)) and has an electrostatic origin. When the tape is peeled, charge densities 1012 e cm−2 (where is e is the fundamental charge on the electron) are exposed on the surfaces of the freshly peeled region and subsequently discharge (C. G. Camara, J. V. Escobar, J. R. Hird and S. P. Putterman, Nature 455, 1089 (2008)). If the tape is peeled in vacuum ˜10 mTorr, it has been found that the triboluminescence produced extends to X-ray energies (V. V. Karasev, N. A. Krotova and B. W. Deryagin, Dokl. Akad. Nauk. SSR 88 777 (1953)). More recently (Camara, et al., id.), it was found that there are two timescales for tribocharging during the peeling of tape in vacuo: the first, common to electrostatic generators and classic electrostatic experiments (W. R. Harper, Contact and frictional electrification, (Oxford University Press, London, 1967)), is the long timescale process which results in an average charge density of 1010 e cm−2 being maintained on the surface of the tape and second, a nanosecond process with charge densities of 1012 e cm−2. In addition, it was found that the X-ray discharge from peeling tape was sufficiently self-collimated at the peel line to resolve the inter-phlangeal spacing of a human digit. The emission of nanosecond X-ray pulses allowed an estimate of the emission region to be calculated. Subsequent research on peeling PSA tape with a width of 1.5 mm has confirmed that the process takes place at dimensions less than 300 μm (C. G. Camara, J. V. Escobar, J. R. Hird and S. P. Putterman, Appl. Phys. B 99, 613 (2010)). Underpinning this recent work on triboelectricity is a resurgence of interest in how charge transfer occurs between different materials and particularly between polymers. Particularly intriguing is the report of like-polymers charging each other (M. M. Apodaca, P. J. Wesson, K. J. M. Bishop, M. A. Ratner and B. A. Grzybowski, Angew. Chem. Int. Ed. 49, 946 (2010)). More fundamentally, an open question is whether the transfer particle is an ion (L. McCathy and G. M. Whitesides, Angew. Chem. Int. Ed. 47, 2188 (2008)) or an electron (Harper, id.)—a matter that is still debated despite centuries of experimental research. Whether the charge carriers responsible for tribocharging are electrons or ions, what is clear is that very large charge densities are readily generated. For the most effective charging to occur, intimate contact between the materials and cleanliness of the contacting surfaces is important (R. Budakian, K. Weninger, R. A. Hiller and S. P. Putterman, Nature 391, 266 (1998)). Mechanoluminescent x-ray generators appear to have a fundamental limitation regarding the maximum energy of x-rays they can obtain (˜50 kV). Furthermore, the x-ray flux is limited via a poorly understood process whereby the polymer that acts as the ‘electron gun’ (in the x-ray tube sense) restricts the electron current that flows when the plates are separated. Therefore, there remains a need for improved triboelectric charged particle acceleration devices. A charged particle acceleration device according to some embodiments of the current invention includes a first triboelectric element, a second triboelectric element arranged proximate the first triboelectric element to be brought into contact with and separated from the first triboelectric element, an actuator assembly operatively connected to at least one of the first and second triboelectric elements to bring the first and second triboelectric elements into contact with each other and to separate the first and second triboelectric elements from each other, and a charged-particle source configured to provide charged particles in a gap between the first and second triboelectric elements. The first and second triboelectric elements include triboelectric materials that become charged with respect to each other by a triboelectric interaction such that an electric field is established between the first and second triboelectric elements when they are separated from each other by the actuator assembly. The charged-particle source is configured to provide the charged particles in the gap between the first and second triboelectric elements to be accelerated towards one of the first and second triboelectric elements by the electric field. A method of producing nuclear isotopes according to some embodiments of the current invention includes providing a triboelectric charged particle acceleration device that is configured to generate fusion reactions, providing a target material comprising nuclear elements to be transformed into heavier isotopes by exposure to neutrons, and operating the triboelectric charged particle acceleration device to generate fusion reactions so as to provide a source of neutrons resulting from said fusion reaction. The operating of the triboelectric charged particle acceleration device is performed such that the target material is exposed to neutrons from the source of neutrons resulting from the fusion reaction. Some embodiments of the current invention are discussed in detail below. In describing embodiments, specific terminology is employed for the sake of clarity. However, the invention is not intended to be limited to the specific terminology so selected. A person skilled in the relevant art will recognize that other equivalent components can be employed and other methods developed without departing from the broad concepts of the current invention. All references cited anywhere in this specification, including the Background and Detailed Description sections, are incorporated by reference as if each had been individually incorporated. As noted in the previous section, mechanoluminescent x-ray generators appear to have a fundamental limitation regarding the maximum energy of x-rays they can obtain (˜50 kV). Furthermore, the x-ray flux is limited via a poorly understood process whereby the polymer that acts as the ‘electron gun’ (in the x-ray tube sense) restricts the electron current that flows when the plates are separated. In other words, in these devices, triboelectric material 1 is responsible for providing both the high voltage field and the accelerated electrons. Replacing this electron source using a separate electron emitting element, according to an embodiment of the current invention, can circumvent these limitations. The vacuum pressure can now be lowered according to some embodiments of the current invention and the separation of materials increased. This increases the accelerating field and allows for the theoretical flux based on the triboelectric charge density to be realized. Greater energies can also be realized. Accordingly, some embodiments of the current invention are directed to a low voltage, compact source of high-energy electromagnetic radiation. Although x-ray generation is one application of the current invention, we stress that this invention is not mechanoluminescence in the proper sense of the word. Embodiments of this device use triboelectricity to produce a strong electric field which is then seeded with electrons and/or other charged particles using a separate element. In triboluminescence, the electrons come from the same materials which also set up the field. Applications may include, but are not limited to, areas such as those requiring fusion reactions, medicine, isotope manufacture, x-ray generation, and pulsed x-ray generation for x-ray movies. FIG. 1 is a schematic illustration of a charged particle acceleration device 100 according to an embodiment of the current invention. The charged particle acceleration device 100 includes a first triboelectric element 102, and a second triboelectric element 104 arranged proximate the first triboelectric element 102 to be brought into contact with and separated from the first triboelectric element 102. The charged particle acceleration device 100 also includes an actuator assembly 106 operatively connected to at least one of the first and second triboelectric elements (102, 104) to bring the first and second triboelectric elements (102, 104) into contact with each other and to separate the first and second triboelectric elements (102, 104) from each other. The charged particle acceleration device 100 also includes a charged-particle source 108 configured to provide charged particles (e.g., charged particle 110) in a gap between the first and second triboelectric elements (102, 104). The first and second triboelectric elements (102, 104) include triboelectric materials that become charged with respect to each other by a triboelectric interaction such that an electric field is established between the first and second triboelectric elements (102, 104) when they are separated from each other by the actuator assembly 106. The charged-particle source is configured to provide the charged particles in the gap between the first and second triboelectric elements (102, 104) to be accelerated towards one of the first and second triboelectric elements (102, 104) by the electric field. In some embodiments, the charged particle acceleration device 100 also includes a containment vessel 112 adapted to provide at least one of a vacuum or an atmosphere of a selected gas at a selected pressure. The charged particles, such as charged particle 110, can be either positively charged, or negatively charged particles. For example, the charged particles can be electrons or either positively charged or negatively charged ions. The term “ions” is intended to include the nuclei of hydrogen, deuterium and tritium (i.e., protons, deuterons and tritons). The general concepts of the current invention are not limited to only these examples. Heavier ions can also be provided in some embodiments of the current invention. In the case of negative charged particles 110, the first triboelectric element 102 contains a triboelectric material that becomes positively charged while the second triboelectric element 104 contains a triboelectric material that becomes negatively charged such that particles 110 accelerate to, and impinge upon, the first triboelectric element 102. In the case of positive charged particles 110, the first triboelectric element 102 contains a triboelectric material that becomes negatively charged while the second triboelectric element 104 contains a triboelectric material that becomes positively charged such that particles 110 accelerate to, and impinge upon, the first triboelectric element 102. The triboelectric materials can be selected from known triboelectric materials to suite the particular application. The actuator assembly 106 can be selected from a wide range of possible assemblies. For example, the actuator assembly 106 can use piezoelectricity, clockwork, electromechanical force, magnetostriction, or human energy or another means to effect motion. Either one of the first and second triboelectric elements (102, 104), or both, can be moved by the actuator assembly 106, depending on the particular application. In some embodiments, the charged-particle source 108 is an electron source. In some embodiments, the triboelectric element 104 can include a triboelectric material that charges with a negative charge by the triboelectric interaction to provide a cathode. The triboelectric material is a photoconductive material and the electron source includes a light source arranged to illuminate at least a portion of the triboelectric material such that the electrons provided by the electron source are provided by a photoelectric effect. The light source can be an ultraviolet light source in some embodiments. FIG. 2 is a schematic illustration of a charged particle acceleration device according to an embodiment of the current invention in which the electron source is a photoelectric element. FIG. 3 is a schematic illustration of a charged particle acceleration device according to an embodiment of the current invention in which the electron source includes a thermionic emitting element. In the embodiment of FIG. 2, the triboelectric photoelectron accelerator has two dissimilar materials, material 1 and material 2, which have the property of being able to exchange electrical charge when subject to a tribological interaction (rubbing, contacting, brushing etc.) in a vacuum. These materials are then separated by some means for example using an actuator (e.g. a device which uses piezoelectricity, clockwork, electromechanical force, magnetostriction, or human energy or another means to effect motion). Material 1 charges electrically in a −Ve sense relative to material 2 and a strong electric field is set up when they are separated. In some embodiments of the current invention, the Material 1 can be both triboelectrically active and photoconductive. The purpose of this is that when the material becomes charged negatively (through the process as described above) most polymers we have looked at appear to have an inherent property which inhibits the flow of charge to the anode. If, however, the polymer becomes conducting, this surface charge will agglomerate and start to field emit from sharp points on the surface towards the triboelectric anode. Some embodiments can include a high power LED or other light source which is directed at the cathode and illuminated when the desired separation between material 1 and 2 is reached. The surface of Material 1 can also contain a metallic tip such that charge will flow to this instead of encountering bottlenecks etc. In another embodiment, ultraviolet light from a source (e.g. ultraviolet light from a LED, super-luminescent diode, laser, high pressure lamp bulb) is focused onto a third material (material 3) positioned close to −Ve material 1. Material 3 is chosen to possess a work function that is lower than the energy of the ultraviolet light in order to induce that material to emit electrons via the photoelectric effect. Examples of such materials include alkali metal compounds, metals, polymers, ceramics, liquids, gases and combinations thereof. When the photoelectrons are emitted from material 3 they are immediately attracted by the strong electric field that exists between materials 1 and 2 and are accelerated towards material 2 which is +Ve. When they strike material 2, high-energy electromagnetic radiation is emitted. In another embodiment, ultraviolet light is directed at material 1 to release electrons trapped in surface states. In the embodiment of FIG. 3, a thermionic source of electrons, i.e a cathode from a material comprising either yttria, lanthanum hexaboride, tantalum, tungsten and barium, their oxides, compounds and ceramics thereof is used to seed the triboelectric field with electrons. However, the material of the cathode is not limited to only these examples. This source can be screened using an external bias when it is not required, enabling the electron (and thus the x-ray source) to be pulsed. Another embodiment of the current invention uses field emission as the source of electrons. In any of the embodiments of the device listed above, grounding of the cathode side to prevent trickle can be implemented. Having the cathode grounded also makes it easier to inject external electrons from a grounded source without losing any acceleration potential (otherwise, you lose half). So that the device works in a cycle, the cathode side should be somewhat electronically conducting. Rubber containing conducting dopants such as carbon or metal impurities can be utilized, or the cathode can be photoconductive. Devices according to some embodiments of the current invention can find application where high-energy electro-magnetic radiation is used and/or needed and may open up new market areas due it's low voltage requirements and compactness. In an embodiment, the triboelectric accelerator includes two dissimilar materials, tribo-material 1 and tribo-material 2, which have the property of being able to exchange electrical charge when subject to a tribological interaction (rubbing, contacting, brushing etc.) in a vacuum. These materials are then separated by some means for example using an actuator (e.g. a device which uses piezoelectricity, clockwork, electromechanical force, magnetostriction, or human energy or another means to effect motion). Tribo-material 1 charges electrically in a −Ve sense relative to tribo-material 2 and a strong electric field is set up when they are separated. An electron is then seeded into the field in one of a number of ways, e.g., via the photoelectric effect, by thermionic emission, field emission or discharge using a heat source. In the embodiment of FIG. 3, a thermoionic or a field emission source is used to introduced electrons into the electric field set up by the triboelectric effect. In this example, a metallic film is in close proximity to a thin film of the cathode polymer (triboelectric element 2), which pins the electric charge to the surface of the polymer preventing the self-emitted electrons (the trickle) from traversing the gap to the anode. The electron source is turned on or pulsed when the plates are at maximum separation. When the external source of electrons is not needed, a bias can be used to shield the electrons from the triboelectric field. This bias can be pulsed rapidly as needed and in synchrony to the separation of the triboelectric plates. In another embodiment, a source of heat (e.g. a radiant heat source (i.e. IR lamp, laser, filament), or other heater (i.e resistance, inductive, microwave) is used to drive electrons off the −Ve charged material. The following references are incorporated by reference. In the general area of contact electrification there have been several innovations emanating from Seth Putterman's lab. WO/2009/102784: Mechanoluminescent x-ray generator. Also, see Nature 455 1089-1092 (2008) and in Appl. Phys B. Lasers and Optics 99 613-617 (2010). Another embodiment of a triboelectric x-ray source is described in Hird, J. R., Camara, C. G. & Putterman, S. J. A triboelectric x-ray source, Appl. Phys. Lett. 98 133501 (2011). Additional results are described in UCLA Case No. 2011-425 (Application Ser. No. 61/451,694 filed Mar. 11, 2011); and UCLA Case No. 2011-707 (Application Ser. No. 61/482,031 filed May 3, 2011). In some embodiments, the source of charged particles is a source of ions such as, but not limited to, protons, deuterons and/or tritons (i.e., the nuclei of hydrogen, deuterium or tritium). The negative side can furthermore be made from a polymer or material which has deuterons or tritons replacing the hydrogen. So now when ions strike the surface they provide fusion. The neutrons from the fusion can then be used to make isotopes when they strike a third body. In one embodiment, the charged particle acceleration device includes an atmosphere of low pressure deuterium gas or tritium gas in the containment vessel. A tip can be mounted on the plus surface. The tip feels the potential of the surface and ionizes the gas and then the ion accelerates. Alternatively, a commercially available ion emitter can be used. Another embodiment of the current invention provides a method of producing nuclear isotopes. The method includes providing a triboelectric charged particle acceleration device that is configured to generate fusion reactions, providing a target material that includes nuclear elements to be transformed into heavier isotopes by exposure to neutrons, and operating the triboelectric charged particle acceleration device to generate fusion reactions so as to provide a source of neutrons resulting from said fusion reaction. The operating of the triboelectric charged particle acceleration device is performed such that the target material is exposed to neutrons from the source of neutrons resulting from the fusion reaction. Mechanoluminescent x-ray generators appear to have a fundamental limitation regarding the maximum energy of x-rays they can obtain (˜50 kV). Furthermore, the x-ray flux is limited via a poorly understood process whereby the polymer that acts as the ‘electron gun’ (in the x-ray tube sense) restricts the electron current that flows when the plates are separated. In other words, in these devices triboelectric material 1 is responsible for providing both the high voltage field and the accelerated electron. Replacing this electron source using a separate photoelectric, thermionic or field emission element, for example, can thus circumvent these limitations. The vacuum pressure can now be lowered to realize high value and so the separation of materials can be increased. This increases the accelerating field and allows for the theoretical flux based on the triboelectric charge density to be realized. In some embodiments, greater energies can also be realized. Additional uses can include fusion reactions, medicine, isotope manufacture, x-ray generation, pulsed x-ray generation for x-ray movies, for example. The embodiments illustrated and discussed in this specification are intended only to teach those skilled in the art how to make and use the invention. In describing embodiments of the invention, specific terminology is employed for the sake of clarity. However, the invention is not intended to be limited to the specific terminology so selected. The above-described embodiments of the invention may be modified or varied, without departing from the invention, as appreciated by those skilled in the art in light of the above teachings. It is therefore to be understood that, within the scope of the claims and their equivalents, the invention may be practiced otherwise than as specifically described. |
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056617682 | summary | FIELD OF THE INVENTION This invention relates generally to the handling of spent nuclear materials and more particularly to the handling of spent nuclear fuel rods for transportation to storage areas, inspection areas, or for further treatment. BACKGROUND OF THE INVENTION The generation of power from nuclear materials has been well known in the art for many decades. The nuclear material, after a period of use for power generation, is incapable of generating the energy necessary and must be removed from its nuclear reactor. The major component of used nuclear material is spent nuclear fuel (SNF) rods. The transportation of such spent nuclear fuel rods has been a troubling problem over the decades and one that has not been solved entirely satisfactorily. Spent nuclear fuel has the characteristic of emitting high amounts of radiation which is damaging to living tissue, particularly in humans. To handle the spent nuclear fuel safely, the fuel must be shielded with appropriate radiation shielding materials. Shielding materials, in general, are of a dense nature. To shield the radiation emitted from the spent nuclear fuel, shielded casks are used to maintain the exterior radiation levels at a sufficiently low level to prevent harm to personnel or the environment. For transferring spent nuclear fuel to a transportation or storage cask, the current commercial industrial practice is an underwater or wet fuel transfer process which includes the step of submerging a cask into the nuclear fuel storage pool. Then, through the use of cranes and grappling hooks and the like, the spent nuclear fuel rods are transferred while still underwater into the submerged cask. The water acts as a radiation shield to protect personnel performing the transfer. The cask is then lifted from the storage pool, the interior of the cask is drained and dried, the outside is decontaminated and the cask is sealed. An alternate method for the transfer of spent nuclear fuel rods is dry transfer. The current dry transfer system utilizes a transfer container to handle the spent nuclear fuel rods. A stand is placed under water in the spent nuclear fuel storage pool and the spent nuclear fuel rods are transferred to a position within the stand while still under water. The transfer container is landed on the stand and the bottom of the transfer container is opened by operating a translating gate. Then, a single spent fuel rod is raised through the bottom of and into the transfer container. This process is repeated to load the transfer container. Then, the transfer container is sealed by closing a translating gate and the transfer container is transported to a discharge stand. The transfer container is seated on the discharge stand and the translating gate is operated to open the bottom of the transfer container. A single spent fuel rod is lowered out of the transfer container and into a transportation cask located in the discharge stand. This process is repeated until all the spent nuclear fuel rods are removed from the transfer container and placed into the cask. Such method is disclosed in commonly assigned U.S. Pat. No. 5,319,686 to Pizzano et al, which patent is incorporated in its entirety herein by reference. The wet fuel transfer process utilizes casks which are too large to be handled at many fuel storage sites because of constraints on existing lifting and handling resources. In addition, the wet fuel transfer process requires the exterior of the submerged cask to be cleaned or decontaminated to remove radioactive particles which increases the process time and the possible exposure of operating personnel to radiation and radioactive contamination. The current dry fuel transfer process requires personnel to be located atop the transfer container to manipulate fuel handling tools and the like. Consequently, to protect the personnel atop the container, the transfer container must provide adequate shielding of the radiation being emitted by the spent nuclear fuel rods. This in turn increases the weight of the transfer container and its exterior dimensions which prevents its use at some fuel storage sites with limited lifting and handling capabilities. Also, grapple actuating tools penetrate the transfer container which hinders containment of any potential radioactive off-gases in the transfer container. In addition, the transfer container is only capable of raising or lowering one spent nuclear fuel (SNF) rod at a time. During the raising process, the SNF rods are not constrained from lateral motion after they are removed from the loading stand and prior to entering the transfer container. Also, during the lowering process, the SNF rods are not constrained from lateral motion after they are lowered from the transfer container and prior to entering the storage or transportation container. SUMMARY OF THE INVENTION The dry fuel transfer system of the present invention allows the transfer of spent nuclear fuel rods in the dry condition, while using a remote means to control and operate the system. The use of a remote control allows a reduction in personnel and allows personnel to remain a safe distance from the transfer container and thereby reduces the amount of radiation shielding material required and the radiation dosage to operating personnel. Consequently, the weight and overall dimensions of the transfer container are reduced to allow use of the present invention at all fuel storage sites without modification to their existing lifting and handling apparatus. Also, the dry fuel transfer system of the present invention permits raising or lowering of a number of nuclear fuel rods simultaneously. This greatly reduces the process time and thereby the radiation exposure of personnel. In addition, the proposed system does not require the penetration of any fuel handling tools or the like, which enables the transfer container to be sealed easily. The remote operation of the transfer container only requires one operator. Also, the improved system of the present invention ensures that the SNF assemblies are constrained from lateral motion during all of the fuel handling movements. |
description | This application is a National Phase filing under 35 U.S.C. §371 of PCT/JP2010/073708 filed on Dec. 28, 2010; and this application claims priority to Application No. 2010-002301 filed in Japan on Jan. 7, 2010 under 35 U.S.C. §119; the entire contents of all are hereby incorporated by reference. The present invention relates to a phase controller which is suitably used for a device serving to convert light having a high energy such as a soft X-ray from linearly polarized light to circularly polarized light, for example. Conventionally, there is provided a device for converting light from linearly polarized light to circularly polarized light. For example, a simple structure such as a transmission type polarizing plate or polarizing film is used for converting visible light or infrared light into circularly polarized light. Moreover, there is also provided an undulator for spirally meandering an electron beam to carry out a conversion into circularly polarized light by periodically applying a magnetic field in a horizontal or perpendicular direction with respect to an orbit of the electron beam (for example, see Patent Documents 1 and 2). Patent Document 1: Japanese Laid-Open Patent Publication No. 7-288200 Patent Document 2: Japanese Laid-Open Patent Publication No. 9-219564 An X-ray is included as a kind of light. The X-ray is an electromagnetic wave having a wavelength of approximately 1 [pm] to several tens [nm] which includes a hard X-ray and a soft X-ray. The hard X-ray is an X-ray having a high energy and a great transmission to a substance and is used for taking an X-ray photograph, for example. On the other hand, the soft X-ray is an X-ray having a lower energy than the hard X-ray, a high absorption into a substance and a small transmission. The soft X-ray converted into circularly polarized light is regarded to be easily absorbed into a substance because of a small transmission and to enable a detection of an electronic spin state in the substance, and therefore, is expected as effective means for an intravital test or a genetic analysis. In the case in which a soft X-ray is utilized for an intravital test, a genetic analysis or the like, it is required to be circularly polarized light. The circularly polarized light has a difference in an electronic spin state, for example, a difference between a counterclockwise direction and a clockwise direction, a difference between a parallelism and an antiparallelism, or the like. Therefore, the difference can be applied to an analysis of a nanomaterial. Since the soft X-ray basically appears as linearly polarized light (a superposition of two states of circularly polarized counterclockwise light and circularly polarized clockwise light), it is to be converted into circularly polarized light. However, the soft X-ray has a lower energy than the hard X ray and still has a high energy of 10 [eV] or more. In a region having a high energy of the soft X-ray which exceeds 10 [eV], a simple structure such as a polarizing plate cannot be used for converting the linearly polarized light into the circularly polarized light. For this reason, there is conventionally employed a method using an undulator for converting linearly polarized light of an electron beam into circularly polarized light. However, this method has a problem in that large-scale facilities referred to as a so-called synchrotron (synchronous circular accelerator) or linac (linear accelerator) are required. The synchrotron or linac serves to carry out a conversion into circularly polarized light in a principle for applying a cyclic magnetic field to periodically bend an electron beam when the electron beam passes through the undulator. The accelerated electron beam does not easily react to the magnetic field. For this reason, an electron orbit is to be meandered little by little by a very long magnetic array. In order to bend the orbit of the electron beam, moreover, a large magnetic field is required and a large-scale superconductive magnet or the like is to be used. In order to minimize an energy loss of the accelerated electron beam, furthermore, it is necessary to bring a vacuum state. Since the electron beam is to run by a long distance, however, large-scale facilities for bringing an ultrahigh vacuum state are required. For this reason, the synchrotron or the linac is to be large-scaled by mans of a small-scale device. The present invention has been made to solve the problem and has an object to phase control linearly polarized light of a soft X-ray, thereby enabling a conversion into circularly polarized light by means of a small-scale device. In order to solve the problem, in the present invention, a reflection surface constituted by a transition metal having a core level absorption edge in the vicinity of a wavelength of a soft X-ray is formed on an inside of a vacuum vessel, and furthermore, there is provided a magnet for generating a magnetic field in a perpendicular direction to a longitudinal direction of the vacuum vessel in a position of the reflection surface by which the soft X-ray is to be reflected. The soft X-ray incident on the vacuum vessel is reflected at least once over the reflection surface in the position where the magnetic field is applied so that the soft X-ray having a phase controlled is emitted from the vacuum vessel. According to the present invention constituted as described above, the soft X-ray has an energy in a wavelength which is close to the core level absorption edge of the transition metal forming the reflection surface. When the soft X-ray incident on the vacuum vessel is to be reflected by the reflection surface, therefore, magnetic scattering caused by the magnetic field applied in the position of the reflection surface is increased by a resonant effect of a magnetic circular dichroism. In other words, although a difference is made in a refractive index between circularly polarized counterclockwise light and circularly polarized clockwise light in the core level absorption edge causing the magnetic scattering, the difference in the refractive index leads to a phase difference between the circularly polarized counterclockwise light and the circularly polarized clockwise light. By varying the number of the reflection surfaces, a strength of the magnetic field or an angle of incidence, it is possible to control the phase difference. Moreover, the difference in the refractive index is increased by the resonant effect of the magnetic circular dichroism. Therefore, it is possible to obtain, at a time, the phase difference between the circularly polarized counterclockwise light and the circularly polarized clockwise light which constitute the linearly polarized light through a superposition. Consequently, it is possible to convert the linearly polarized light of the soft X-ray into the circularly polarized light by the reflection to be carried out at a few times. The linearly polarized light can be converted into the circularly polarized light at a small number of times of the reflection. Therefore, it is not necessary to lengthen the vacuum vessel and the magnetic array. Consequently, it is not necessary to employ large-scale facilities for bringing an ultrahigh vacuum state, and it is sufficient that the simple vacuum pump is used. Moreover, the magnetic scattering is increased by the resonant effect of the magnetic circular dichroism. Therefore, it is not necessary to use a large-scale superconductive magnet or the like, and it is sufficient that a small permanent magnet is provided. Accordingly, a size of the device for converting the linearly polarized light of the soft X-ray into the circularly polarized light can be reduced remarkably as compared with a synchrotron or the like. An embodiment of a phase controller according to the present invention will be described below with reference to the drawings. FIG. 1 is a view showing an example of a structure of a circularly polarized light converter carrying out a phase controller according to a first embodiment. FIG. 2 is a view showing an example of an arrangement of a reflection surface according to the first embodiment. FIG. 3 is a view showing an example of an arrangement of a permanent magnet according to the first embodiment. As shown in FIG. 1, a circularly polarized light converter 10 according to the first embodiment includes a hollow vacuum vessel 11 serving as a route for a soft X-ray which is emitted from a soft X-ray generator 100, a reflection surface 12 formed on an inside of the vacuum vessel 11, a permanent magnet 13 for generating a magnetic field, and a vacuum pump 14 for bringing a vacuum state in the vacuum vessel 11. As shown in FIG. 2, for example, the vacuum vessel 11 is an elliptically cylindrical vessel having an elliptical section and is constituted by glass or the like. A housing of the vacuum vessel 11 has a diameter of approximately 10 to 50 [mm], for example. Moreover, the housing has a length of approximately 10 to 50 [cm], for example. The reflection surface 12 includes a pair of reflection plates 12a and 12b formed in a longitudinal direction of the vacuum vessel 11, for example. The pair of reflection plates 12a and 12b are disposed to be perpendicularly opposed to each other in parallel with an average advancing direction of the soft X-ray (the longitudinal direction of the vacuum vessel 11). A void distance between the reflection plates 12a and 12b is approximately 1 to several [mm], for example. Moreover, full lengths of the reflection plates 12a and 12b are approximately 10 to 50 [cm], for example. The refection surface 12 is constituted by a transition metal having a core level absorption edge in the vicinity of a wavelength of a soft X-ray which is incident on the vacuum vessel 11. For example, the reflection surface 12 is constituted, as a transition metal having a 3p-3d core level absorption edge in the vicinity of the wavelength of the soft X-ray, by tungsten (W) if the wavelength of the soft X-ray is 2.8 [nm], cobalt (Co) if the wavelength of the soft X-ray is 19.8 [nm], nickel (Ni) if the wavelength of the soft X-ray is 17.9 [nm], manganese (Mn) if the wavelength of the soft X-ray is 24.3 [nm], titanium (Ti) if the wavelength of the soft X-ray is 25.8 [nm], a perovskite type 3d transition metal oxide (Y1-xCaxTiO3) if the wavelength of the soft X-ray is 26.9 [nm], and a ferrous superconductor (LaFeAsO) if the wavelength of the soft X-ray is 22.9 [nm]. The permanent magnet 13 serves to generate a magnetic field in a perpendicular direction to the longitudinal direction of the vacuum vessel 11 in a position where the soft X-ray is reflected by the reflection surface 12. A strength of a magnetism of the permanent magnet 13 is approximately 0.2 to 1 [T], for example. The permanent magnet 13 is constituted to include plural sets of magnet pairs 13a and 13b which are disposed to interpose the vacuum vessel 11 therebetween at an outside of the vacuum vessel 11. The pair of magnets 13a and 13b are disposed in such a manner that north and south poles are opposed to each other. Moreover, the plural sets of magnets 13a and 13b are disposed at an equal interval in the longitudinal direction of the vacuum vessel 11. Positions placed at the equal interval correspond to positions in which the soft X-ray is reflected by the reflection surface 12. It is sufficient that the permanent magnet 13 generates a magnetic field in a perpendicular direction to the longitudinal direction of the vacuum vessel 11 and whether the magnetic field is perpendicular to the reflection surface 12 does not matter. In other words, the permanent magnet 13 may be disposed in parallel with the reflection surface 12 as shown in FIG. 3(a) and the permanent magnet 13 may be disposed perpendicularly to the reflection surface 12 as shown in FIG. 3(b). In general, in the case in which an energy of an X-ray is close to a core level absorption edge of a magnetic atom, magnetic scattering is increased to be several times to 105 times as large as ordinary magnetic scattering by a resonant effect. According to the present embodiment, in order to utilize the resonant effect of a magnetic circular dichroism, the reflection surface 12 is constituted by a transition metal having a 3p-3d core level absorption edge in the vicinity of the wavelength of the soft X-ray and a magnetic field is thus applied to the reflection surface 12 by means of the permanent magnet 13. The soft X-ray to be linearly polarized light is incident in the vacuum vessel 11 set into the vacuum state by means of the vacuum pump 14 and is reflected at plural times over the reflection surface 12 in a position where the magnetic field is applied. According to the first embodiment thus constituted, when the soft X-ray incident on the vacuum vessel 11 is reflected by the reflection surface 12, the magnetic scattering is increased by the resonant effect of the magnetic circular dichroism. Therefore, a great difference is made in a refractive index between the circularly polarized counterclockwise light and the circularly polarized clockwise light which constitute the linearly polarized light of the soft X-ray, and a phase difference can be made between the circularly polarized counterclockwise light and the circularly polarized clockwise light at a time. Consequently, it is possible to convert the linearly polarized light of the soft X-ray into the circularly polarized light by carrying out the reflection at only several times and to then emit, from the vacuum vessel 11, the soft X-ray converted into the circularly polarized light. According to the present embodiment, moreover, it is possible to act on the soft X-ray itself which is generated in the soft X-ray generator 100, thereby converting the linearly polarized light into the circularly polarized light. To the contrary, it is also possible to reversibly return the circularly polarized light into the linearly polarized light. Although the conventional method using an electron beam can make a circularly polarized light component artificially, it cannot act on the soft X-ray itself at all. Thus, the linearly polarized light of the soft X-ray can be converted into the circularly polarized light at a small number of times of the reflection. Therefore, it is not necessary to lengthen the vacuum vessel 11 in the longitudinal direction. Consequently, it is not necessary to employ large-scale facilities for bringing an ultrahigh vacuum state, and it is sufficient that the simple vacuum pump 14 is used. Moreover, the magnetic scattering is increased by the resonant effect of the magnetic circular dichroism. Therefore, it is not necessary to use a large-scale superconductive magnet or the like, and it is sufficient that a few small permanent magnets 13 are used. Accordingly, a size of the device for converting the linearly polarized light of the soft X-ray into the circularly polarized light can be reduced remarkably as compared with a synchrotron or the like. Next, a second embodiment according to the present invention will be described with reference to the drawings. FIG. 4 is a view showing an example of a structure of a circularly polarized light converter carrying out a phase controller according to the second embodiment. In FIG. 4, components having the same reference numerals as those shown in FIG. 1 have the same functions and repetitive description will be omitted. As shown in FIG. 4, a circularly polarized light converter 20 according to the second embodiment includes a second reflection surface 22 in addition to the structure illustrated in FIG. 1. Moreover, a vacuum vessel 21 has a double length in the longitudinal direction as compared with the vacuum vessel 11 shown in FIG. 1. The second reflection surface 22 is disposed in a subsequent part to a reflection surface 12 at an inside of the vacuum vessel 21. A length of the second reflection surface 22 is equal to that of the reflection surface 12. In the same manner as the reflection surface 12, the second reflection surface 22 is also constituted by a pair of reflection plates 22a and 22b formed in a longitudinal direction of the vacuum vessel 21. The pair of reflection plates 22a and 22b are disposed to be perpendicularly opposed to each other in parallel with an average advancing direction of a soft X-ray (the longitudinal direction of the vacuum vessel 21). Moreover, the pair of reflection plates 22a and 22b are disposed in a perpendicular direction to a pair of reflection plates 12a and 12b. The second reflection surface 22 is formed by the same transition metal as the reflection surface 11. In other words, the second reflection surface 22 is also formed of tungsten (W) if the reflection surface 12 is formed of the tungsten (W), and the second reflection surface 22 is also formed of cobalt (Co) if the reflection surface 12 is formed of the cobalt (Co). In the second embodiment, a soft X-ray to be linearly polarized light is incident in the vacuum vessel 21 set into a vacuum state by means of a vacuum pump 14 and is reflected at plural times over the reflection surface 12 in a position where a magnetic field is applied by a permanent magnet 13, and then, the soft X-ray is further reflected at plural times over the second reflection surface 22. The number of times of the reflection over the reflection surface 12 is set to be equal to that of the reflection over the second reflection surface 22. In a polarizing state of the soft X-ray to be reflected by the reflection surface 12, a polarizing direction of the soft X-ray to be incident is represented as a sum of vectors of light (s polarized light) which is polarized in parallel with the reflection surface 12 and light (p polarized light) which is polarized perpendicularly to the reflection surface 12. However, a reflectance on the reflection surface 12 is varied between the s polarized light and the p polarized light. For this reason, an intensity of the s polarized light is different from that of the p polarized light. If phases of circularly polarized clockwise light and circularly polarized counterclockwise light are simply controlled, therefore, the soft X-ray is converted into elliptically polarized light which is not completely circularly polarized light. Therefore, the phase of the soft X-ray is controlled by the reflection at plural times over the reflection surface 12 to which a magnetic field is applied, and the reflection at equal times to that for the reflection surface 12 is then caused over the second reflection surface 22 to which the magnetic field is not applied. At this time, the s polarized light over the reflection surface 12 is set into the p polarized light over the second reflection surface 22 and the p polarized light over the reflection surface 12 is set into the s polarized light over the second reflection surface 22 so that the reflectances can be reversed and an intensity of the s polarized light and that of the p polarized light can be finally set to be equal to each other by the reflection at equal times to that for the reflection surface 12, since the second reflection surface 22 is disposed in a perpendicular direction to the reflection surface 12. Consequently, the soft X-ray converted into completely circularly polarized light can be emitted from the vacuum vessel 21. Although the description has been given to the example in which the transition metal having the 3p-3d core level absorption edge in the vicinity of the wavelength of the soft X-ray incident on the vacuum vessels 11 and 21 is used as the transition metal constituting the reflection surface 12 and the second reflection surface 22 in the first and second embodiments, the present invention is not restricted thereto. In other words, the 3p-3d based transition metal does not need to be utilized if there is used any transition metal having the core level absorption edge in the vicinity of the wavelength of the soft X-ray. For example, if the wavelength of the soft X-ray is 6.2 [nm], the reflection surface 12 and the second reflection surface 22 may be constituted by tungsten (W) having a 4s-4p core level absorption edge. Although the description has been given to the example in which the reflection surface 12 is constituted by the pair of reflection plates 12a and 12b and the second reflection surface 22 is constituted by the pair of reflection plates 22a and 22b in the first and second embodiments, moreover, the present invention is not restricted thereto. For example, a reflection sheet formed by a transition metal may be stuck onto inner surfaces of the vacuum vessels 11 and 21 or the transition metal may be deposited on the inner surfaces of the vacuum vessels 11 and 21. Although the description has been given to the example in which the light of the soft X-ray is converted from the linearly polarized light into the circularly polarized light in the embodiments, furthermore, the present invention is not restricted thereto. For example, by utilizing the same principle, it is also possible to convert the light of the soft X-ray from the circularly polarized light into the linearly polarized light. In addition, both of the first and second embodiments are only illustrative for materialization to carry out the present invention and the technical scope of the present invention should not be thereby construed to be restrictive. In other words, the present invention can be carried out in various forms without departing from the spirit or main features thereof. The phase controller according to the present invention is suitably used for a device which serves to convert light having a high energy such as a soft X-ray from linearly polarized light into circularly polarized light. Moreover, the phase controller according to the present invention can also be used for a device which serves to convert light having a high energy such as a soft X-ray from circularly polarized light into linearly polarized light. |
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041893476 | abstract | In a reactor vessel for pebble beds at high and varying temperatures having a side wall and a base formed of a multiplicity of stacked blocks of heat-resistant material and held together by an outer cylindrical or polygonal ring and supported on a foundation, the base and the side wall, respectively, being formed of a plurality of sectors having substantially vertical, radial parting lines therebetween, the sectors being supported with slight friction on the foundation and being braced against the outer ring, the sectors having boundary surfaces with a pebble bed, support surfaces on the foundation and abutment surfaces against the outer ring, the respective surfaces of the boundary surfaces, the support surfaces and the abutment surfaces having a convex mutual inclination whereby each of the sectors is held together in itself by external forces and is forced by its own weight and the weight of the pebble bed into a definite position. |
045086770 | summary | This invention relates to nuclear reactors for large-scale energy generation. In particular, a self-contained nuclear heat supply module is disclosed which is fabricated at a factory and which is assembled at a field location to form a nuclear reactor. The reactor module is shown in its factory construction, in its on-site assembly, and in the operation of the assembled power plant. PRIOR ART Typical nuclear reactor construction includes a heat supply containing a nuclear fuel, a primary coolant system, and a secondary coolant system. The primary coolant receives heat from the nuclear heat supply and delivers the received heat to the secondary coolant through a heat exchanger. The heat of the secondary coolant is delivered to a means for generating electricity, such as a steam turbine. For commercial energy generation nuclear generating facilities are typically large installations including one or more independent nuclear reactors, each of which can have an installed capacity as large as 1200 megawatts. Such reactor systems are typically housed within large containment buildings, which enclose the reactor and auxiliary and backup cooling and control systems as well as such facilities for refueling and servicing the reactor as will be needed over the lifetime of the plant. The containment building must, of course, be designed for entry by humans for servicing and refueling operations. Aside from the large-scale nuclear reactors intended to power commercial energy-generating plants, small-scale nuclear reactors have also been designed to operate in an entirely different environment--namely, aboard ship. A nuclear generating plant for shipboard use must necessarily be more compact than a land-base plant, but the power requirements are much less. Various designs for compact nuclear generating units suitable for shipboard use are disclosed in U.S. Pat. Nos. 3,170,846; 3,245,879; 3,255,088; 3,401,082; 3,941,187; 4,124,064; and 4,288,196. In the reactor units disclosed in these patents, the reactor core is contained within a close-coupled pressure vessel and containment vessel. For compactness a boiler or steam-generating unit is frequently included within the containment vessel. In typical shipboard construction the containment vessel and associated shielding are permanently assembled within the steam-generator compartment of the ship. The reactor core and associated cooling and heat-exchange systems are then lowered into the containment vessel. A nuclear generating unit for shipboard use must, accordingly, be constructed with the reactor core, steam-generating unit, heat exchangers and the like mounted so that they can all be removed through the top of the containment vessel for servicing and refueling. In contrast to the compact shipboard operating units, land-based nuclear power plants were initially constructed on a large scale because it was believed they would achieve an economy of scale. Basic geometrical considerations teach that as the reactor volume increases, the ratio of the reactor surface area to the reactor volume decreases, so that greater energy generation capability was expected per unit of shielding, cooling system, and containment vessel which had to be constructed. Simply stated, as the plant became larger, it was expected that the marginal amount of time, money, and effort devoted to the containment vessel, confinement building, site operating personnel and other necessary services would generally decrease. Unfortunately, these expectations have not generally been borne out in practice. Recent events have demonstrated that an unscheduled shutdown of one such unit can have very substantial economic consequences. Where a large reactor suffers radioactive contamination, plant capacity suffers substantially and cleanup costs conceivably outweigh the benefit otherwise derived from use of a nuclear fuel. Furthermore, government-imposed safety regulations require that nuclear power plants be constructed to the highest possible standard. With current construction methods this high standard must be "transported to" the remote construction sites in which nuclear plants are commonly located and imposed on construction operations in the field. Recruiting construction personnel and training them in the field to the high and exacting standards of modern nuclear licensing procedures has proved expensive, time-consuming, difficult--and sometimes impossible. In short, with current construction methods, quality assurance has been a problem. As it is ultimately disposed within a large nuclear energy-generation facility, the reactor relies upon active safety measures in the event of substantial malfunction. These measures include active heat rejection systems oftentimes coupled to auxiliary or backup systems through extensive piping networks. The integrity of such piping networks and their associated pumping stations during unpredictable seismic events has oftentimes been questioned. Additionally, such plant designs have essentially been unalterable once the plant is placed on-line. Plant capacity is fixed within a prescribed range for economical operation, making nuclear plants suitable for only large-scale base load power generation. A plant shutdown means complete interruption of base load enery supply. Finally, all such plants must be over-designed in an attempt to withstand possible malfunctions and must be repeatedly inspected for flawed components by X-ray techniques and the like to assure continued safety against such malfunctions. Some of the above-cited U.S. patents suggest that compact nuclear generating units may also play a role in land-based energy generation. U.S. Pat. No. 4,289,196, in particular, suggests that a number of auxiliary systems associated with typical land-based reactors can be eliminated if multiple small-scale modular nuclear steam generating units are connected to a single turbine for generating electricity. SUMMARY OF THE INVENTION The present invention efficiently merges into a large-scale land-based energy generating facility the concept of a compact nuclear generating unit. The invention provides a method by which all the critical elements of the compact nuclear generating unit are constructed and assembled at a central factory location, transported to a field location, outfitted with a biological shield, and incorporated into the generating plant. In particular, the invention provides a prefabricated nuclear heat supply module which includes all critical nuclear components of the compact generating unit and which can be completely nuclear-certified at the central factory and incorporated into the generating plant with only minimal non-critical assembly in the field. Briefly, the nuclear heat supply module as assembled at the factory includes the primary vessel which is surrounded by an outer vessel close-coupled to the primary vessel to define an interstitial region between the two for containing an inert gas. The outer vessel has dimensions which are sized to enable the factory-assembled module to be shipped on a railway car. An unloaded reactor core unit including a plurality of control rods is mounted within the primary vessel. For compactness without sacrificing power output, the reactor core unit is of the fast-breeder type. Also mounted within the primary vessel is a heat exchanger having an inlet and outlet for secondary coolant. Inlet means and outlet means are provided for communicating to the outer and primary vessels with the heat exchanger inlet and outlet. The inlet means and outlet means are adapted to be connected to the balance of a conventional secondary cooling system, which is constructed at the field location. A pump is mounted within the primary vessel for pumping a primary coolant through the reactor core unit and heat exchanger. The pump and the reactor core unit communicate directly with an inlet plenum so as to define a primary coolant flow path directly from the pump to the reactor core unit. Means are also provided within the primary vessel for defining a plenum-like primary coolant flow path from the reactor core unit to the heat exchanger and from the heat exchanger to the pump, the flow-path-defining means being contained entirely within the primary vessel and including no piping subject to leakage or rupture. A control rod drive unit is mounted within the outer vessel overlying the reactor core unit and is operatively connected through the primary vessel to the control rods within the reactor core. It is an object of the invention to provide a prefabricated nuclear heat supply module which contains within all critical nuclear components which are subject to compulsory nuclear-certification procedures, so that a fully certified module can be factory-produced and shipped to the generating plant field location. It is a feature of the invention that the nuclear heat supply module, as it leaves the factory, is sized especially to be transported on a conventional railway car. After the module is unloaded from the railway car at the field site, a segmented surrounding cylindrical shell is assembled about the module. Each segment of the shell defines an interstitial region for the pouring and curing of a concrete biological shield. Each segment also includes a plurality of passive, free-convection cooling loops, providing a shutdown heat-removal system. Shell assembly and pouring and curing of a cementatious biological shield is performed on a movable pad, for example, an airlift pad, and a concrete shield seal plug is poured at this time. In the referred plant design the assembled module with concrete shield is moved to a service building wherein the outer vessel head and primary vessel head are removed enabling the reactor core to be charged in a conventional open-head manner. The heads are then bolted and hermetically seal-welded in place, and the concrete biological seal plug is installed. The assembled and charged module is then moved into position for connection to a conventional steam generation circuit. OTHER OBJECTS, FEATURES AND ADVANTAGES An object of this invention is to provide a nuclear heat supply module which can be assembled at a factory location and thereafter shipped by railroad to a plant site in the field. Simply stated, maximum construction of the reactor occurs at the factory with minimum assembly occurring at the plant site. This disclosure hereafter sets forth: the reactor construction in the plant; the reactor module structure which is shipped from the factory; the process of shielded reactor assembly at the plant site; the resultant reactor at the plant which comprises a discrete steam generation unit; and an overall plant layout adapted to accommodate on a permanent site of infinite life the reactor disclosed herein. It is an object of the invention to disclose the construction of a nuclear heat supply module in which as many assembly steps as possible can be carried out in a factory under controlled condition by trained workers. An advantage of the disclosed apparatus is that where assembly occurs at a factory, quality assurance can be maximized. Critical nuclear components can be intimately inspected and verified before licensing. Proceeding in this manner, greater inspection can be carried out in a central location, thereby providing for more reliable certification with reduced administrative costs. Yet another advantage of the disclosed assembly of components is that the efficiencies of factory automation can be enjoyed. In effect, an almost completely hermetically sealed primary loop is shipped from the factory, with the primary-coolant pumps already installed in position. Only a minimum number of seal welds must be made on-site to complete the hermetic seal of the primary vessel after loading of the fuel assembies and charging of the interstitial regions with inert gas. Another object of the invention is to provide for easy assembly of a biological shield about the nuclear heat supply module at the field site. According to this aspect of the invention, a four-segment upstanding cylindrical annular shell is prefabricated at the factory and shipped to the site. The shell has discrete, vertically extending, serpentine free-convection loops mounted preferably eight to a shell segment. The inner, heat-receiving branches of each loop are in intimate contact with the outer vessel, and the heat-dissipating branches of each loop are exterior to the shell. According to one embodiment of the invention, the inner coolant branch of the shutdown heat removal circuits adjacent the outer vessel are supplemented with a water jacket vented to atmospheric pressure, which provides for greater passive heat removal capability in the event of shutdown. Venting the water jacket to the atmosphere assures that in the event of convective loop failure, the biological shield will not be exposed to temperatures above the boiling point of water so as to destroy the concrete. The jacket is conveniently refillable to provide for continual heat-removal capability in the event of total system failure. A further object of this invention is to provide a process of charging and refueling the reactor core unit. According to this aspect of the invention, the nuclear heat supply module with biological shield in place is transported to a service building. A concrete plug capping the biological shield is removed, followed by removal of the outer vessel head and primary vessel head. A conventional open-head charging or refueling operation is then carried out within the on-site service building after which the heads are re-attached and re-sealed to their respective vessel bodies. The biological shield cap is then re-positioned and the charged reactor module is transported to its operating location whereupon the secondary coolant inlet and outlet means are connected to the balance of the secondary cooling system. The disclosed process results in a steam generation unit of high integrity. All critical components of the reactor module are assembled without piping. Only the necesary flow of the intermediate heat exchanger occurs through a piping network, and that is exterior to the nuclear module. The disclosed module design takes advantage of the improved scram performance provided by gravity-actuated primary and secondary control rods to render the unit subcritical. No feedwater cooling or other active cooling system is relied upon in a scram condition. Instead the scrammed reactor is entirely passive providing for the dissipation of heat to the atmosphere through the free-convection coolant loops integrally included within the biological shield. An additional advantage of the disclosed construction is that the reactor module itself with its massive biological shield provides an ambient heat sink significantly contributing to the module's passive shutdown capability. An additional advantage of the invention is that during operation the shutdown heat-removal system accounts for less than three-tenths of a percent of the loss of the total reactor heat. The minimal heat loss through the shutdown system is sufficient to provide a continual flow in the sealed free-convection loops so as to maintain them in an operative condition, yet it does not reduce significantly the useful heat carried by the secondary coolant. The dimensions of the prefabricated nuclear heat supply module specifically set forth herein are directly related to the utility of the invention. The smaller reactor modules of the present invention taken in combination with one another render a significantly safer, more economical nuclear power plant which can be more thoroughly subjected to nuclear-certification procedures. The smaller units are amenable to the free-convection shutdown heat-removal loop disclosed herein for energy dissipation upon a scram condition. Moreover, a smaller unit can be constructed to take advantage of the thermal expansion of core material so as to aid in bringing the reactor to a subcritical disposition in the event of shutdown. Not only can a smaller unit be substantially assembled at the factory and charged at the field site, but additionally when subjected to casualty at the field site, it forms a manageable unit which can simply be moved away from the installation and left to cool down while the remainder of the plant continues to operate. The nuclear heat supply module achieving the above results will preferably be 14 feet in diameter and 70 feet in height. A nuclear heat supply module of these dimensions can be conveniently shipped to the field site on a conventional railway car. The unit assembled in the field with biological shield is 30 feet in diameter and 80 feet in height. The unit can be transported on a conventional air-lift pad and detachably connected to a steam generation system. An advantage of the disclosed reactor is that in the event of casualty, it can easily be moved off-site to an isolated area. The site is thus furnished with effectively infinite lifetime. The fuel elements can then be removed from the isolated reactor, or alternatively the reactor can be left intact for a radioactive decay period or more and then the fuel elements removed. |
claims | 1. A method for manufacturing a transmissive spectral purity filter configured to transmit extreme ultraviolet radiation, the method comprising:providing a semiconductor substrate having a top Si layer, a bottom Si layer and an intermediate etch stop layer in between the top Si layer and the bottom Si layer;etching a plurality of apertures in the substrate using an anisotropic etching process, the anisotropic etching process comprisingapplying a hard mask of an aperture pattern on the substrate,deep reactive ion etching the aperture pattern vertically through the top Si layer to the intermediate etch layer,removing the bottom Si layer, andremoving at least part of the etch stop layer to open up the plurality of apertures. 2. The method according to claim 1, wherein the etching creates textured sidewalls that define the apertures. 3. The method according to claim 2, further comprising depositing a metal or reflective layer on top of the substrate and depositing the metal or reflective layer on at least a part of each sidewall. 4. The method according to claim 1, further comprising removing all of the etch stop layer after having manufactured the apertures in the substrate. 5. A transmissive spectral purity filter configured to transmit extreme ultraviolet radiation, the transmissive spectral purity filter manufactured by the method according to claim 1, wherein the etching creates textured sidewalls that define the apertures. 6. The method according to claim 1, wherein the etch stop layer comprises SiO2. |
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description | FIG. 1 schematically illustrates the essential components of a layer-thickness measuring device 11, in which case the illustration of an evaluation unit, a screen for visualizing a measurement object recorded by a video camera, and also an input keyboard and printer has been dispensed with. This layer-thickness measuring device 11 is used for example for measuring bonding pads, contacts which are provided in part with a selective coating, conductor tracks and functional coatings on small areas. A layer-thickness measuring device 11 with the apparatus 12 according to the invention is preferably used to determine or check layer thicknesses whose measurement area or the functional areas are smaller than 100 xcexcmxc3x97100 xcexcm, in particular smaller than 50 xcexcmxc3x9750 xcexcm. X-rays are generated in an X-ray tube 13 and are directed via an anode 14 to a measurement object 16. The X-rays excite a fluorescent radiation in a layer of the measurement object 16. The intensity of this fluorescent radiation depending on the energy (spectrum) is a function of the layer thickness. This or the parameter of the layer system is utilized by the system of the emitted radiation being registered with the aid of a detector 17. The apparatus 12 according to the invention is provided between the X-ray tube 13 and the measurement object 16, which apparatus, in accordance with the exemplary embodiment, comprises two mutually opposite reflecting areas 18. These reflecting areas 18 serve for focussing rays and forwarding rays, with the result that the X-rays pass to the measurement area of the measurement object 16. The reflecting areas 18 are preferably arranged directly relative to the anode 14 or to an exit flange 21 near the anode 14. Furthermore, a collimator 23 is provided at the lower end 22 of the reflecting areas 18 which are assigned to one another, as a result of which it is possible to image a measurement region 24 as shown in FIG. 3 on a measurement object. The collimator 23 is advantageously a slit collimator whose slit width is adjustable. The reflecting areas 18 are designed as elongate, rectangular areas, as can be gathered from FIG. 1 and FIG. 2. The length of the reflecting areas 18 is essentially determined by the construction and also by the degree of total reflection. X-rays which do not run parallel between an axis of the measurement region 24 and the anode 14 are deflected at least once by total reflection. The width of the reflecting areas 18 is at least one and a half times as large as the maximum functional area to be checked. It is advantageous to use silicon wafers for the reflecting areas 18. This cost-effective base material can be adapted in a simple manner to the corresponding size of the apparatus 12 according to the invention. Further semiconductor materials such as, for example, germanium, gallium arsenide or the like are also suitable for the reflecting areas 18. The reflecting areas 18, which are preferably produced from a silicon wafer, are advantageously applied to holding elements 26, 27 as shown in FIG. 3. These are advantageously bonded on in a strain-free manner, so that the planarity of the reflecting area 18 can be maintained. As an alternative, the reflecting areas 18 can also be fixed in a stress-free manner on the holding elements 26, 27 by means of clamping or the like. As shown in FIG. 3, an adjusting unit 28 engages on one of the two holding elements 27, by means of which adjusting unit a holding element 27 can be adjusted relative to the stationary element 26. The holding element 26 advantageously accommodates the reflecting area 18 parallel to the central axis 29 of the apparatus 12. The slit width can be adjusted by the adjusting unit 28. It likewise becomes possible to adjust the angularity of the holding element 27 relative to the element 26. As an alternative, it is likewise possible to provide a mirror-inverted arrangement. Likewise, provision may alternatively be made for an adjusting unit 28 to be provided on each of the holding elements 26, 27, as a result of which the holding elements 26, 27 can be arranged either parallel to one another and/or at an angle to one another, thereby forming a uniform or tapering slit towards the measurement object 16. The adjusting unit 28 is designed in such a way that slit widths in a range of from 10 to 100 xcexcm, for example, can optionally be adjusted. For this purpose, it is possible to provide precision-mechanical adjusting mechanisms, piezo-electric actuators, and also electrically, hydraulically, pneumatically operated actuating drives. At an end pointing towards the measurement object 16, a flattened portion 31 is provided on the holding element 26. This flattened portion makes it possible for there to be a sufficient aperture width 32 available for the emitted fluorescent radiation in order to detect the emitted fluorescent radiation. The reflecting area 18 may, for example, have a noble metal vapour-deposited on it. This makes it possible to increase the critical angle for total reflection, which is 1.5 mrad for silicon, to 4.5 mrad by means of a platinum coating. This in turn has an advantageous effect on the transmission of the X-rays. As an alternative, in the case where coated reflecting areas are used, it is conceivable that the base material may comprise a quartz surface or a plastics material which satisfies the requirement of planarity and has a coating. The coating may advantageously be provided at least at the input of the reflecting areas 18, so that the number of captured and reflected rays is as large as possible. The coating may be continued completely over the course along the reflecting areas 18, or else be provided only partly. Likewise, the coating or the material of the coating may also change depending on the applications. By way of example, by reducing the critical angle for total reflection, it is possible to reduce the divergence at the output of the reflecting areas 18, which makes it possible to obtain focussing of the radiation and, as a result, an intensity increase on the measurement region 24 of the measurement object 16. To that end, it is conceivable, for example, for a coating not to be provided in a region near the lower end 22 of the reflecting area 18 or for a coating that prevents total reflection to be provided, as a result of which the radiation emerging below the reflecting area 18 is focussed precisely to the size of the measurement region 24 of the measurement object 16. The irradiation of edge regions outside the measurement region 24 can thereby be reduced considerably. The invention""s configuration of the apparatus 12 enables the measurement region to be adjusted depending on the measurement task. The collimator 23 can likewise be adapted to this measurement region, so that the focussing of the radiation enables an intensity increase on a predetermined measurement region. As an alternative, it may be provided that the reflecting areas 18 are designed to be at least slightly concave. Likewise, the concave design may taper towards the lower end 22, yielding a kind of meslithone-shaped configuration of the reflecting areas 18. In this case, however, account should be taken of the dimensions, which can also lie in the micrometer range. The aperture width of the reflecting areas 18 at the input of the apparatus 12 essentially corresponds to the outlet opening for the X-rays emitted via the anode. Likewise, it is also possible to provide a slightly larger or smaller aperture width relative to the diameter of the primary spot of the X-rays. Furthermore, the apparatus 12 may also have openings and receptacles which serve for arranging an optical system in order to visualize the measurement object 16 using a video camera. In accordance with the exemplary embodiment, the apparatus 12 is provided by two reflecting areas 18 which are arranged relative to one another and are arranged parallel or at an acute angle relative to one another. It may also be provided that, instead of these two reflecting areas 18, three or more reflecting areas are arranged in a suitable manner relative to one another in order to enable the transmission of X-rays to the measurement region 24 of a measurement object 16, so that an intensity increase is made possible by the focussing of the X-rays. However, in contrast to what is known from the prior art, it is not necessary to use a closed, tubular arrangement in order to focus the X-rays to the measurement region by total reflection. Further geometrical configurations of the reflecting areas 18 which enable the total reflection of the X-rays are likewise conceivable. |
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claims | 1. A device to implant impurities into a semiconductor wafer, comprising:a pressure compensation unit;a beam gun to shoot an ion beam at a semiconductor wafer;first and second ion gauges; anda switching device to selectively connect the first or second ion gauge to the pressure compensation unit. 2. A device to implant impurities into a semiconductor wafer according to claim 1, further comprising a current sensor to sense the number of implanted ions. 3. A device to implant impurities into a semiconductor wafer according to claim 1, whereinthe beam gun is positioned upstream from a process chamber,the first ion gauge is positioned outside of the process chamber, andthe second ion gauge is positioned within the process chamber. 4. A device to implant impurities into a semiconductor wafer according to claim 3, whereinthe device further comprises a cryo pump to minimize the pressure within the process chamber, andthe first ion gauge is positioned between the process chamber and the cryo pump. 5. A device to implant impurities into a semiconductor wafer according to claim 1, further comprising:a disk faraday to receive ions from the ion gun; anda current meter to count the number of electrons flowing to the disk faraday to neutralize the ions. 6. A device to implant impurities into a semiconductor wafer according to claim 1, wherein a portion of the ions are neutralized before reaching the semiconductor wafer. 7. A device to implant impurities into a semiconductor wafer according to claim 1, whereina plurality of wafers are arranged on a support disk which rotates about an axis of rotation substantially parallel to an ion beam path,at least some of the wafers have a resist layer, andresist outgassing occurs when the ion beam strikes the resist layer. 8. A device to implant impurities into a semiconductor wafer according to claim 7, whereina disk faraday receives ions from the ion gun,a current meter counts the number of electrons flowing to the disk faraday to neutralize the ions,the support disk has a radialy extending slot arranged among the wafers,the ion beam travels through the slot as the support disk rotates, andthe support disk is arranged between the disk faraday and the ion gun. 9. A device to implant impurities into a semiconductor wafer according to claim 1, whereinthe chamber has a wall, andthe second ion gauge extends through the wall of the chamber. 10. A device to implant impurities into a semiconductor wafer according to claim 1, whereina plurality of wafers are arranged on a support disk which rotates about an axis of rotation substantially parallel to an ion beam path,the support disk moves radially at a radial speed with respect to the ion beam, andthe radial speed is varied to control the amount of time the ion beam is focused at different radial positions on the support disk. 11. A device to implant impurities into a semiconductor wafer according to claim 10, wherein the radial speed decreases when it is determined that the number of impurities being implanted is relatively low. 12. A device to implant impurities into a semiconductor wafer according to claim 10, wherein:the first and second ion gauges sense pressure;a disk faraday to receive ions from the ion gun; andthe device further comprises:a current meter to count the number of electrons flowing to the disk faraday to neutralize the ions; andthe pressure compensation unit varies the radial speed of the support disk as a function of the current sensed by the current meter and as a function of the pressure sensed by the ion gauge connected to the pressure compensation unit. 13. A device to implant impurities into a semiconductor wafer according to claim 1, wherein the first ion gauge is used for high energy implants and the second ion gauge is used for low energy implants. 14. A device to implant impurities into a semiconductor wafer, comprising:a process chamber having a wall;a pressure compensation unit;a support disk to support a plurality of semiconductor wafers within the process chamber, the support disk having a radialy extending slot arranged among the wafers;a beam gun positioned upstream from the process chamber to shoot an ion beam at the semiconductor wafers;a cryo pump to minimize the pressure within the process chamber;first and second ion gauges, the first ion gauge being positioned between the process chamber and the cryo pump, the second ion gauge extending through the wall of the process chamber;a switching device to selectively connect the first or second ion gauge to the pressure compensation unit;a disk faraday to receive ions from the ion gun after the ions travel through the slot in the support disk; anda current meter to count the number of electrons flowing to the disk faraday to neutralize the ions. |
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description | This application claims the benefit of U.S. Provisional Application No. 60/659,738 filed Mar. 7, 2005, which application is incorporated herein by reference. The invention relates generally to the field of plasma physics, and, in particular, to methods and apparati for confining plasma to enable nuclear fusion and for converting energy from fusion products into electricity. Fusion is the process by which two light nuclei combine to form a heavier one. The fusion process releases a tremendous amount of energy in the form of fast moving particles. Because atomic nuclei are positively charged—due to the protons contained therein—there is a repulsive electrostatic, or Coulomb, force between them. For two nuclei to fuse, this repulsive barrier must be overcome, which occurs when two nuclei are brought close enough together where the short-range nuclear forces become strong enough to overcome the Coulomb force and fuse the nuclei. The energy necessary for the nuclei to overcome the Coulomb barrier is provided by their thermal energies, which must be very high. For example, the fusion rate can be appreciable if the temperature is at least of the order of 104 eV—corresponding roughly to 100 million degrees Kelvin. The rate of a fusion reaction is a function of the temperature, and it is characterized by a quantity called reactivity. The reactivity of a D-T reaction, for example, has a broad peak between 30 keV and 100 keV. Typical fusion reactions include:D+D→He3(0.8 MeV)+n(2.5 MeV),D+T→α(3.6 MeV)+n(14.1 MeV),D+He3→α(3.7 MeV)+p(14.7 MeV), andp+B11→3α(8.7 MeV),where D indicates deuterium, T indicates tritium, a indicates a helium nucleus, n indicates a neutron, p indicates a proton, He indicates helium, and B11 indicates Boron-11. The numbers in parentheses in each equation indicate the kinetic energy of the fusion products. The first two reactions listed above—the D-D and D-T reactions—are neutronic, which means that most of the energy of their fusion products is carried by fast neutrons. The disadvantages of neutronic reactions are that (1) the flux of fast neutrons creates many problems, including structural damage of the reactor walls and high levels of radioactivity for most construction materials; and (2) the energy of fast neutrons is collected by converting their thermal energy to electric energy, which is very inefficient (less than 30%). The advantages of neutronic reactions are that (1) their reactivity peaks are at a relatively low temperature; and (2) their losses due to radiation are relatively low because the atomic numbers of deuterium and tritium are 1. The reactants in the other two equations—D-He3 and p-B11—are called advanced fuels. Instead of producing fast neutrons, as in the neutronic reactions, their fusion products are charged particles. One advantage of the advanced fuels is that they create much fewer neutrons and therefore suffer less from the disadvantages associated with them. In the case of D-He3, some fast neutrons are produced by secondary reactions, but these neutrons account for only about 10 percent of the energy of the fusion products. The p-B11 reaction is free of fast neutrons, although it does produce some slow neutrons that result from secondary reactions but create much fewer problems. Another advantage of the advanced fuels is that their fusion products comprise charged particles whose kinetic energy may be directly convertible to electricity. With an appropriate direct energy conversion process, the energy of advanced fuel fusion products may be collected with a high efficiency, possibly in excess of 90 percent. The advanced fuels have disadvantages, too. For example, the atomic numbers of the advanced fuels are higher (2 for He3 and 5 for B11). Therefore, their radiation losses are greater than in the neutronic reactions. Also, it is much more difficult to cause the advanced fuels to fuse. Their peak reactivities occur at much higher temperatures and do not reach as high as the reactivity for D-T. Causing a fusion reaction with the advanced fuels thus requires that they be brought to a higher energy state where their reactivity is significant. Accordingly, the advanced fuels must be contained for a longer time period wherein they can be brought to appropriate fusion conditions. The containment time for a plasma is Δt=r2/D, where r is a minimum plasma dimension and D is a diffusion coefficient. The classical value of the diffusion coefficient is Dc=ai2/τie, where ai is the ion gyroradius and τie is the ion-electron collision time. Diffusion according to the classical diffusion coefficient is called classical transport. The Bohm diffusion coefficient, attributed to short-wavelength instabilities, is DB=( 1/16)ai2Ωi, where Ωi is the ion gyrofrequency. Diffusion according to this relationship is called anomalous transport. For fusion conditions, DB/Dc=( 1/16)Ωiτie≅108, anomalous transport results in a much shorter containment time than does classical transport. This relation determines how large a plasma must be in a fusion reactor, by the requirement that the containment time for a given amount of plasma must be longer than the time for the plasma to have a nuclear fusion reaction. Therefore, classical transport condition is more desirable in a fusion reactor, allowing for smaller initial plasmas. In early experiments with toroidal confinement of plasma, a containment time of Δt r2/DB was observed. Progress in the last 40 years has increased the containment time to Δt 1000 r2/DB. One existing fusion reactor concept is the Tokamak. For the past 30 years, fusion efforts have been focussed on the Tokamak reactor using a D-T fuel. These efforts have culminated in the International Thermonuclear Experimental Reactor (ITER). Recent experiments with Tokamaks suggest that classical transport, Δt≅r2/Dc, is possible, in which case the minimum plasma dimension can be reduced from meters to centimeters. These experiments involved the injection of energetic beams (50 to 100 keV), to heat the plasma to temperatures of 10 to 30 keV. See W. Heidbrink & G. J. Sadler, 34 Nuclear Fusion 535 (1994). The energetic beam ions in these experiments were observed to slow down and diffuse classically while the thermal plasma continued to diffuse anomalously fast. The reason for this is that the energetic beam ions have a large gyroradius and, as such, are insensitive to fluctuations with wavelengths shorter than the ion gyroradius (λ<ai). The short-wavelength fluctuations tend to average over a cycle and thus cancel. Electrons, however, have a much smaller gyroradius, so they respond to the fluctuations and transport anomalously. Because of anomalous transport, the minimum dimension of the plasma must be at least 2.8 meters. Due to this dimension, the ITER was created 30 meters high and 30 meters in diameter. This is the smallest D-T Tokamak-type reactor that is feasible. For advanced fuels, such as D-He3 and p-B11, the Tokamak-type reactor would have to be much larger because the time for a fuel ion to have a nuclear reaction is much longer. A Tokamak reactor using D-T fuel has the additional problem that most of the energy of the fusion products energy is carried by 14 MeV neutrons, which cause radiation damage and induce reactivity in almost all construction materials due to the neutron flux. In addition, the conversion of their energy into electricity must be by a thermal process, which is not more than 30% efficient. Another proposed reactor configuration is a colliding beam reactor. In a colliding beam reactor, a background plasma is bombarded by beams of ions. The beams comprise ions with an energy that is much larger than the thermal plasma. Producing useful fusion reactions in this type of reactor has been infeasible because the background plasma slows down the ion beams. Various proposals have been made to reduce this problem and maximize the number of nuclear reactions. For example, U.S. Pat. No. 4,065,351 to Jassby et al. discloses a method of producing counterstreaming colliding beams of deuterons and tritons in a toroidal confinement system. In U.S. Pat. No. 4,057,462 to Jassby et al., electromagnetic energy is injected to counteract the effects of bulk equilibrium plasma drag on one of the ion species. The toroidal confinement system is identified as a Tokamak. In U.S. Pat. No. 4,894,199 to Rostoker, beams of deuterium and tritium are injected and trapped with the same average velocity in a Tokamak, mirror, or field reversed configuration. There is a low density cool background plasma for the sole purpose of trapping the beams. The beams react because they have a high temperature, and slowing down is mainly caused by electrons that accompany the injected ions. The electrons are heated by the ions in which case the slowing down is minimal. In none of these devices, however, does an equilibrium electric field play any part. Further, there is no attempt to reduce, or even consider, anomalous transport. Other patents consider electrostatic confinement of ions and, in some cases, magnetic confinement of electrons. These include U.S. Pat. No. 3,258,402 to Farnsworth and U.S. Pat. No. 3,386,883 to Farnsworth, which disclose electrostatic confinement of ions and inertial confinement of electrons; U.S. Pat. No. 3,530,036 to Hirsch et al. and U.S. Pat. No. 3,530,497 to Hirsch et al. are similar to Farnsworth; U.S. Pat. No. 4,233,537 to Limpaecher, which discloses electrostatic confinement of ions and magnetic confinement of electrons with multi-pole cusp reflecting walls; and U.S. Pat. No. 4,826,646 to Bussard, which is similar to Limpaecher and involves point cusps. None of these patents consider electrostatic confinement of electrons and magnetic confinement of ions. Although there have been many research projects on electrostatic confinement of ions, none of them have succeeded in establishing the required electrostatic fields when the ions have the required density for a fusion reactor. Lastly, none of the patents cited above discuss a field reversed configuration magnetic topology. The field reversed configuration (FRC) was discovered accidentally around 1960 at the Naval Research Laboratory during theta pinch experiments. A typical FRC topology, wherein the internal magnetic field reverses direction, is illustrated in FIG. 3 and FIG. 5, and particle orbits in a FRC are shown in FIG. 6 and FIG. 9. Regarding the FRC, many research programs have been supported in the United States and Japan. There is a comprehensive review paper on the theory and experiments of FRC research from 1960-1988. See M. Tuszewski, 28 Nuclear Fusion 2033, (1988). A white paper on FRC development describes the research in 1996 and recommendations for future research. See L. C. Steinhauer et al., 30 Fusion Technology 116 (1996). To this date, in FRC experiments the FRC has been formed with the theta pinch method. A consequence of this formation method is that the ions and electrons each carry half the current, which results in a negligible electrostatic field in the plasma and no electrostatic confinement. The ions and electrons in these FRCs were contained magnetically. In almost all FRC experiments, anomalous transport has been assumed. See, e.g., Tuszewski, beginning of section 1.5.2, at page 2072. Thus, it is desirable to provide a fusion system having a containment system that tends to substantially reduce or eliminate anomalous transport of ions and electrons and an energy conversion system that converts the energy of fusion products to electricity with high efficiency. The present invention is directed to a system that facilitates controlled fusion in a magnetic field having a field-reversed topology and the direct conversion of fusion product energies to electric power. The system, referred to herein as a plasma-electric power generation (PEG) system, preferably includes a fusion reactor having a containment system that tends to substantially reduce or eliminate anomalous transport of ions and electrons. In addition, the PEG system includes an energy conversion system coupled to the reactor that directly converts fusion product energies to electricity with high efficiency. In one embodiment, anomalous transport for both ions and electrons tends to be substantially reduced or eliminated. The anomalous transport of ions tends to be avoided by magnetically confining the ions in a magnetic field of field reversed configuration (FRC). For electrons, the anomalous transport of energy is avoided by tuning an externally applied magnetic field to develop a strong electric field, which confines the electrons electrostatically in a deep potential well. As a result, fusion fuel plasmas that can be used with the present confinement apparatus and process are not limited to neutronic fuels, but also advantageously include advanced or aneutronic fuels. For aneutronic fuels, fusion reaction energy is almost entirely in the form of charged particles, i.e., energetic ions, that can be manipulated in a magnetic field and, depending on the fuel, cause little or no radioactivity. In a preferred embodiment, a fusion reactor's plasma containment system comprises a chamber, a magnetic field generator for applying a magnetic field in a direction substantially along a principle axis, and an annular plasma layer that comprises a circulating beam of ions. Ions of the annular plasma beam layer are substantially contained within the chamber magnetically in orbits and the electrons are substantially contained in an electrostatic energy well. In one preferred embodiment the magnetic field generator includes a current coil. Preferably, the magnetic field generator further comprises mirror coils near the ends of the chamber that increase the magnitude of the applied magnetic field at the ends of the chamber. The system also comprises one or more beam injectors for injecting neutralized ion beams into the magnetic field, wherein the beam enters an orbit due to the force caused by the magnetic field. In a preferred embodiment, the system forms a magnetic field having a topology of a field reversed configuration. In another preferred embodiment, an alternative chamber is provided that prevents the formation of azimuthal image currents in a central region of the chamber wall and enables magnetic flux to penetrate the chamber on a fast timescale. The chamber, which is primarily comprised of stainless steel to provide structural strength and good vacuum properties, includes axial insulating breaks in the chamber wall that run along almost the entire length of the chamber. Preferably, there are three breaks that are about 120 degrees apart from each other. The breaks include a slot or gap formed in the wall. An insert comprising an insulating material, preferably a ceramic or the like, is inserted into the slots or gaps. In the interior of the chamber, a metal shroud covers the insert. On the outside of the chamber, the insert is attached to a sealing panel, preferable formed from fiberglass or the like, that forms a vacuum barrier by means of an O-ring seal with the stainless steel surface of the chamber wall. In yet another preferred embodiment, an inductive plasma source is mountable within the chamber and includes a shock coil assembly, preferably a single turn shock coil, that is preferably fed by a high voltage (about 5-15 kV) power source (not shown). Neutral gas, such as Hydrogen (or other appropriate gaseous fusion fuel), is introduced into the source through direct gas feeds via a Laval nozzle. Once the gas emanates from the nozzle and distributes itself over the surface of the coil windings of the shock coil, the windings are energized. The ultra fast current and flux ramp-up in the low inductance shock coil leads to a very high electric field within the gas that causes breakdown, ionization and subsequent ejection of the formed plasma from the surface of the shock coil towards the center or mid-plane of the chamber. In a further preferred embodiment, a RF drive comprises a quadrupolar cyclotron located within the chamber and having four azimuthally symmetrical electrodes with gaps there between. The quadrupole cyclotron produces an electric potential wave that rotates in the same direction as the azimuthal velocity of ions, but at a greater velocity. Ions of appropriate speed can be trapped in this wave, and reflected periodically. This process increases the momentum and energy of the fuel ions and this increase is conveyed to the fuel ions that are not trapped by collisions. In another embodiment, a direct energy conversion system is used to convert the kinetic energy of the fusion products directly into electric power by slowing down the charged particles through an electro-magnetic field. Advantageously, the direct energy conversion system of the present invention has the efficiencies, particle-energy tolerances and electronic ability to convert the frequency and phase of the fusion output power of about 5 MHz to match the frequency of an external 60 Hertz power grid. In a preferred embodiment, the energy conversion system comprises inverse cyclotron converters (ICC) coupled to opposing ends of the fusion reactor. The ICC have a hollow cylinder-like geometry formed from multiple, preferably four or more equal, semi-cylindrical electrodes with small, straight gaps extending there between. In operation, an oscillating potential is applied to the electrodes in an alternating fashion. The electric field E within the ICC has a multi-pole structure and vanishes on the symmetry axes and increases linearly with radius; the peak value being at the gap. In addition, the ICC includes a magnetic field generator for applying a uniform uni-directional magnetic field in a direction substantially opposite to the applied magnetic field of the fusion reactor's containment system. At an end furthest from the fusion reactor power core the ICC includes an ion collector. In between the power core and the ICC is a symmetric magnetic cusp wherein the magnetic field of the containment system merges with the magnetic field of the ICC. An annular shaped electron collector is positioned about the magnetic cusp and electrically coupled to the ion collector. In yet another preferred embodiment, product nuclei and charge-neutralizing electrons emerge as annular beams from both ends of the reactor power core with a density at which the magnetic cusp separates electrons and ions due to their energy differences. The electrons follow magnetic field lines to the electron collector and the ions pass through the cusp where the ion trajectories are modified to follow a substantially helical path along the length of the ICC. Energy is removed from the ions as they spiral past the electrodes, which are connected to a resonant circuit. The loss of perpendicular energy tends to be greatest for the highest energy ions that initially circulate close to the electrodes, where the electric field is strongest. Other aspects and features of the present invention will become apparent from consideration of the following description taken in conjunction with the accompanying drawings. As illustrated in the figures, a plasma-electric power generation (PEG) system of the present invention preferably includes a colliding beam fusion reactor (CBFR) coupled to a direct energy conversion system. As alluded to above, an ideal fusion reactor solves the problem of anomalous transport for both ions and electrons. The solution to the problem of anomalous transport found herein makes use of a containment system with a magnetic field having a field reversed configuration (FRC). The anomalous transport of ions is avoided by magnetic confinement in the FRC in such a way that the majority of the ions have large, non-adiabatic orbits, making them insensitive to short-wavelength fluctuations that cause anomalous transport of adiabatic ions. In particular, the existence of a region in the FRC where the magnetic field vanishes makes it possible to have a plasma comprising a majority of non-adiabatic ions. For electrons, the anomalous transport of energy is avoided by tuning the externally applied magnetic field to develop a strong electric field, which confines them electrostatically in a deep potential well. Fusion fuel plasmas that can be used with the present confinement apparatus and process are not limited to neutronic fuels such as D-D (Deuterium-Deuterium) or D-T (Deuterium-Tritium), but also advantageously include advanced or aneutronic fuels such as D-He3 (Deuterium-helium-3) or p-B11 (hydrogen-Boron-11). (For a discussion of advanced fuels, see R. Feldbacher & M. Heindler, Nuclear Instruments and Methods in Physics Research, A271 (1988) JJ-64 (North Holland Amsterdam).) For such aneutronic fuels, the fusion reaction energy is almost entirely in the form of charged particles, i.e., energetic ions, that can be manipulated in a magnetic field and, depending on the fuel, cause little or no radioactivity. The D-He3 reaction produces an H ion and an He4 ion with 18.2 MeV energy while the p-B11 reaction produces three He4 ions and 8.7 MeV energy. Based on theoretical modeling for a fusion device utilizing aneutronic fuels, the output energy conversion efficiency may be as high as about 90%, as described by K. Yoshikawa, T. Noma and Y. Yamamoto in Fusion Technology, 19, 870 (1991), for example. Such efficiencies dramatically advance the prospects for aneutronic fusion, in a scalable (1-1000 MW), compact, low-cost configuration. In a direct energy conversion process of the present invention, the charged particles of fusion products can be slowed down and their kinetic energy converted directly to electricity. Advantageously, the direct energy conversion system of the present invention has the efficiencies, particle-energy tolerances and electronic ability to convert the frequency and phase of the fusion output power of about 5 MHz to match the frequency and phase of an external 60 Hertz power grid. Fusion Containment System FIG. 1 illustrates a preferred embodiment of a containment system 300 according to the present invention. The containment system 300 comprises a chamber wall 305 that defines therein a confining chamber 310. Preferably, the chamber 310 is cylindrical in shape, with principle axis 315 along the center of the chamber 310. For application of this containment system 300 to a fusion reactor, it is necessary to create a vacuum or near vacuum inside the chamber 310. Concentric with the principle axis 315 is a betatron flux coil 320, located within the chamber 310. The betatron flux coil 320 comprises an electrical current carrying medium adapted to direct current around a long coil, as shown, which preferably comprises parallel windings of multiple separate coils and, most preferably, parallel windings of about four separate coils, to form a long coil. Persons skilled in the art will appreciate that current through the betatron coil 320 will result in a magnetic field inside the betatron coil 320, substantially in the direction of the principle axis 315. Around the outside of the chamber wall 305 is an outer coil 325. The outer coil 325 produce a relatively constant magnetic field having flux substantially parallel with principle axis 315. This magnetic field is azimuthally symmetrical. The approximation that the magnetic field due to the outer coil 325 is constant and parallel to axis 315 is most valid away from the ends of the chamber 310. At each end of the chamber 310 is a mirror coil 330. The mirror coils 330 are adapted to produce an increased magnetic field inside the chamber 310 at each end, thus bending the magnetic field lines inward at each end. (See FIGS. 3 and 5.) As explained, this bending inward of the field lines helps to contain the plasma 335 in a containment region within the chamber 310 generally between the mirror coils 330 by pushing it away from the ends where it can escape the containment system 300. The mirror coils 330 can be adapted to produce an increased magnetic field at the ends by a variety of methods known in the art, including increasing the number of windings in the mirror coils 330, increasing the current through the mirror coils 330, or overlapping the mirror coils 330 with the outer coil 325. The outer coil 325 and mirror coils 330 are shown in FIG. 1 implemented outside the chamber wall 305; however, they may be inside the chamber 310. In cases where the chamber wall 305 is constructed of a conductive material such as metal, it may be advantageous to place the coils 325, 330 inside the chamber wall 305 because the time that it takes for the magnetic field to diffuse through the wall 305 may be relatively large and thus cause the system 300 to react sluggishly. Similarly, the chamber 310 may be of the shape of a hollow cylinder, the chamber wall 305 forming a long, annular ring. In such a case, the betatron flux coil 320 could be implemented outside of the chamber wall 305 in the center of that annular ring. Preferably, the inner wall forming the center of the annular ring may comprise a non-conducting material such as glass. As will become apparent, the chamber 310 must be of sufficient size and shape to allow the circulating plasma beam or layer 335 to rotate around the principle axis 315 at a given radius. The chamber wall 305 may be formed of a material having a high magnetic permeability, such as steel. In such a case, the chamber wall 305, due to induced countercurrents in the material, helps to keep the magnetic flux from escaping the chamber 310, “compressing” it. If the chamber wall were to be made of a material having low magnetic permeability, such as plexiglass, another device for containing the magnetic flux would be necessary. In such a case, a series of closed-loop, flat metal rings could be provided. These rings, known in the art as flux delimiters, would be provided within the outer coils 325 but outside the circulating plasma beam 335. Further, these flux delimiters could be passive or active, wherein the active flux delimiters would be driven with a predetermined current to greater facilitate the containment of magnetic flux within the chamber 310. Alternatively, the outer coils 325 themselves could serve as flux delimiters. As explained in further detail below, a circulating plasma beam 335, comprising charged particles, may be contained within the chamber 310 by the Lorentz force caused by the magnetic field due to the outer coil 325. As such, the ions in the plasma beam 335 are magnetically contained in large betatron orbits about the flux lines from the outer coil 325, which are parallel to the principle axis 315. One or more beam injection ports 340 are also provided for adding plasma ions to the circulating plasma beam 335 in the chamber 310. In a preferred embodiment, the injector ports 340 are adapted to inject an ion beam at about the same radial position from the principle axis 315 where the circulating plasma beam 335 is contained (i.e., around a null surface described below). Further, the injector ports 340 are adapted to inject ion beams 350 (See FIG. 17) tangent to and in the direction of the betatron orbit of the contained plasma beam 335. Also provided are one or more background plasma sources 345 for injecting a cloud of non-energetic plasma into the chamber 310. In a preferred embodiment, the background plasma sources 345 are adapted to direct plasma 335 toward the axial center of the chamber 310. It has been found that directing the plasma this way helps to better contain the plasma 335 and leads to a higher density of plasma 335 in the containment region within the chamber 310. Vacuum Chamber As described above, application of the containment system of a CBFR, it is necessary to create a vacuum or near vacuum inside the chamber. Since interactions (scattering, charge exchange) between neutrals and plasma fuel always present an energy loss channel, it is critical to limit the residual density in the reactor chamber. Furthermore, impurities resulting from poorly evacuated chambers can lead to contaminating side-reactions during operation and can drain an exorbitant amount of energy during startup as the system has to burn through these residuals. To achieve a good level vacuum usually involves the use of stainless steel chambers and ports as well as low outgassing materials. In the case of metals, the good vacuum properties are further paired with good structural characteristics. However, conductive materials such as stainless steel and the like, present various problems with regards to their electrical properties. Although these negative effects are all linked, they manifest themselves in different ways. Amongst the most negative characteristics are: Retarded diffusion of magnetic fields through chamber walls, accumulation of electrical charges on the surfaces, drastic alteration of response times of the system to transient signals as well as formation of image currents in the surfaces that impact the desired magnetic topology. Materials that do not have these undesirable characteristics and exhibit good vacuum properties are insulators such as ceramics, glass, quartz and to a lesser degree carbon-fibers. The primary problem with these materials is structural integrity as well as the potential for accidental damage. Fabrication problems such as poor machinability of ceramics are further limitations. In one embodiment, as depicted in FIGS. 2A, 2B, 2C and 2D, an alternative chamber 1310 is provided that minimizes these problems. The chamber 1310 of the CBFR is preferably primarily comprised of a metal, preferably stainless steel or the like, to provide structural strength and good vacuum properties. However, the cylindrical wall 1311 of the chamber 1310 includes axial insulating breaks 1360 in the wall 1311 that run along almost the entire length of the chamber 1310 in the central portion of the chamber 1310 or power core region of the CBFR. Preferably, as depicted in FIG. 2B, there are three breaks 1360 that are about 120 degrees apart from each other. The breaks 1360, as depicted in FIG. 2C, include a slot or gap 1362 in the wall 1311 of the chamber 1310 with a seal groove or seat 1369 formed about the periphery of the slot 1362. An O-ring seal 1367 is received in the groove 1369. The slots 1362, as depicted in FIG. 2D, extend almost the entire length of the chamber 1310 leaving sufficient stainless material forming an azimuthally continuous portion of the wall 1311 near the two ends to provide structural integrity and to allow for good quality vacuum seals at the ends. For improved structural integrity and the prevention of implosion, the chamber 1310, as depicted in FIG. 2A, preferably includes a plurality of sets of partial azimuthal ribs 1370 that are integrally formed with the chamber wall 1311 or coupled to the surface of the chamber wall 1311 by welding or the like. As depicted in FIG. 2C, the gap 1362 is filled with an insert 1364 formed of ceramic material. The insert 1364 extends slightly into the interior of the chamber 1310 and is covered on the inside by a metal shroud 1366 to prevent secondary plasma emission from collisions of primary plasma ions from the circulating plasma beam with the ceramic material. On the outside of the chamber 1310, the insert 1364 is attached to a sealing panel 1365 that forms a vacuum barrier by means of an O-ring seal 1367 with the stainless steel surface of the chamber wall 1311. To preserve desirable vacuum properties, the sealing panel 1365 is preferably formed from a substrate, preferably fiberglass or the like, which is more flexible and creates a tighter seal with the O-ring 1367 than a ceramic material would, especially when inward pressure slightly deforms the chamber 1310. The inserts or ceramic insulators 1364 inside the slots 1362 preferably prevent current from arching across the gaps 1362 and, thus, prevent the formation of azimuthal image currents in the chamber wall 1311. Image currents are a manifestation of Lenz's Law, which is nature's tendency to counteract any change in flux: for example, the change in flux that occurs in the flux coil 1320 during the formation of a FRC, as described below. Without slots 1362 in the cylindrical wall 1311 of the chamber 1310, the changing flux in the flux coil 1320 causes an equal and opposite inductively induced current to form in the stainless steel wall 1311 such as to cancel the magnetic flux change inside the chamber 1310. While the induced image currents would be weaker (due to inductive losses) than the current applied to the flux coil 1320, the image current tends to strongly reduce the applied or confinement magnetic field within the chamber 1310, which, when not addressed, tends to negatively impact the magnetic field topology and alter the confinement characteristics inside of the chamber 1310. The existence of the slots 1362 prevents azimuthal image currents from forming in the wall 1311 toward the mid-plane of the chamber 1310 away from the ends of the chamber 1310 in the azimuthally continuous portion of the wall 1311. The only image currents that can be carried by the chamber wall 1311 toward the mid-plane away from the ends of the chamber 1310 are very weak currents that flow parallel to the longitudinal axis of the slots 1362. Such currents have no impact on the axial magnetic confinement fields of the FRC as the magnetic image fields produced by the image currents longitudinally traversing the chamber wall 1311 only exhibit radial and azimuthal components. The azimuthal image currents formed in the azimuthally continuous conducting portion of the wall 1311 near the ends of the chamber 1310 tend not to negatively impact and/or alter the confinement characteristics inside of the chamber 1310 as the magnetic topology in this vicinity is not important to confinement of the plasma. In addition to preventing the formation of azimuthal image currents in the chamber wall 1311, the slots 1362 provide a way for magnetic flux from the field and mirror coils 1325 and 1330 to penetrate the chamber 1310 on a fast timescale. The slots 1362 enable sub-millisecond level fine-tuning and feedback control of the applied magnetic field as a result. Charged Particles in a FRC FIG. 3 shows a magnetic field of a FRC 70. The system has cylindrical symmetry with respect to its axis 78. In the FRC, there are two regions of magnetic field lines: open 80 and closed 82. The surface dividing the two regions is called the separatrix 84. The FRC forms a cylindrical null surface 86 in which the magnetic field vanishes. In the central part 88 of the FRC the magnetic field does not change appreciably in the axial direction. At the ends 90, the magnetic field does change appreciably in the axial direction. The magnetic field along the center axis 78 reverses direction in the FRC, which gives rise to the term “Reversed” in Field Reversed Configuration (FRC). In FIG. 4A, the magnetic field outside of the null surface 94 is in a first direction 96. The magnetic field inside the null surface 94 is in a second direction 98 opposite the first. If an ion moves in the direction 100, the Lorentz force 30 acting on it points towards the null surface 94. This is easily appreciated by applying the right-hand rule. For particles moving in the diamagnetic direction 102, the Lorentz force always points toward the null surface 94. This phenomenon gives rise to a particle orbit called betatron orbit, to be described below. FIG. 4B shows an ion moving in the counterdiamagnetic direction 104. The Lorentz force in this case points away from the null surface 94. This phenomenon gives rise to a type of orbit called a drift orbit, to be described below. The diamagnetic direction for ions is counterdiamagnetic for electrons, and vice versa. FIG. 5 shows a ring or annular layer of plasma 106 rotating in the ions' diamagnetic direction 102. The ring 106 is located around the null surface 86. The magnetic field 108 created by the annular plasma layer 106, in combination with an externally applied magnetic field 110, forms a magnetic field having the topology of a FRC (The topology is shown in FIG. 3). The ion beam that forms the plasma layer 106 has a temperature; therefore, the velocities of the ions form a Maxwell distribution in a frame rotating at the average angular velocity of the ion beam. Collisions between ions of different velocities lead to fusion reactions. For this reason, the plasma beam layer or power core 106 is called a colliding beam system. FIG. 6 shows the main type of ion orbits in a colliding beam system, called a betatron orbit 112. A betatron orbit 112 can be expressed as a sine wave centered on the null circle 114. As explained above, the magnetic field on the null circle 114 vanishes. The plane of the orbit 112 is perpendicular to the axis 78 of the FRC. Ions in this orbit 112 move in their diamagnetic direction 102 from a starting point 116. An ion in a betatron orbit has two motions: an oscillation in the radial direction (perpendicular to the null circle 114), and a translation along the null circle 114. FIG. 7A is a graph of the magnetic field 118 in a FRC. The horizontal axis of the graph represents the distance in centimeters from the FRC axis 78. The magnetic field is in kilogauss. As the graph depicts, the magnetic field 118 vanishes at the null circle radius 120. As shown in FIG. 7B, a particle moving near the null circle will see a gradient 126 of the magnetic field pointing away from the null surface 86. The magnetic field outside the null circle is in a first direction 122, while the magnetic field inside the null circle is in a second direction 124 opposite to the first. The direction of a gradient drift is given by the cross product {right arrow over (B)}×∇B, where ∇B is the gradient of the magnetic field; thus, it can be appreciated by applying the right-hand rule that the direction of the gradient drift is in the counterdiamagnetic direction, whether the ion is outside or inside the null circle 128. FIG. 8A is a graph of the electric field 130 in a FRC. The horizontal axis of the graph represents the distance in centimeters from the FRC axis 78. The electric field is in volts/cm. As the graph depicts, the electric field 130 vanishes close to the null circle radius 120. As shown if FIG. 8B, the electric field for ions is deconfining; it points in directions 132, 134 away from the null surface 86. The magnetic field, as before, is in opposite directions 122,124 inside and outside of the null surface 86. It can be appreciated by applying the right-hand rule that the direction of the {right arrow over (E)}×{right arrow over (B)} drift is in the diamagnetic direction 102, whether the ion is outside or inside the null surface 136. FIGS. 9A and 9B show another type of common orbit in a FRC, called a drift orbit 138. Drift orbits 138 can be outside of the null surface 114, as shown in FIG. 9A, or inside it, as shown in FIG. 9B. Drift orbits 138 rotate in the diamagnetic direction if the {right arrow over (E)}×{right arrow over (B)} drift dominates or in the counterdiamagnetic direction if the gradient drift dominates. The drift orbits 138 shown in FIGS. 9A and 9B rotate in the diamagnetic direction 102 from starting point 116. A drift orbit, as shown in FIG. 9C, can be thought of as a small circle rolling over a relatively bigger circle. The small circle 142 spins around its axis in the sense 144. It also rolls over the big circle 146 in the direction 102. The point 140 will trace in space a path similar to 138. FIGS. 10A and 10B show the direction of the Lorentz force at the ends of a FRC 151. In FIG. 10A, an ion is shown moving in the diamagnetic direction 102 with a velocity 148 in a magnetic field 150. It can be appreciated by applying the right-hand rule that the Lorentz force 152 tends to push the ion back into the region of closed field lines. In this case, therefore, the Lorentz force 152 is confining for the ions. In FIG. 10B, an ion is shown moving in the counterdiamagnetic direction with a velocity 148 in a magnetic field 150. It can be appreciated by applying the right-hand rule that the Lorentz force 152 tends to push the ion into the region of open field lines. In this case, therefore, the Lorentz force 152 is deconfining for the ions. Magnetic and Electrostatic Confinement in a FRC A plasma layer 106 (see FIG. 5) can be formed in a FRC by injecting energetic ion beams around the null surface 86 in the diamagnetic direction 102 of ions. (A detailed discussion of different methods of forming the FRC and plasma ring follows below.) In the circulating plasma layer 106, most of the ions have betatron orbits 112 (see FIG. 6), are energetic, and are non-adiabatic; thus, they are insensitive to short-wavelength fluctuations that cause anomalous transport. In a plasma layer 106 formed in a FRC and under equilibrium conditions, the conservation of momentum imposes a relation between the angular velocity of ions w, and the angular velocity of electrons ωe. The relation is ω e = ω i [ 1 - ω i Ω 0 ] , where Ω 0 = ZeB 0 m i c . ( 1 ) In Eq. 1, Z is the ion atomic number, mi is the ion mass, e is the electron charge, B0 is the magnitude of the applied magnetic field, and c is the speed of light. There are three free parameters in this relation: the applied magnetic field B0, the electron angular velocity ωe, and the ion angular velocity ωi. If two of them are known, the third can be determined from Eq. Because the plasma layer 106 is formed by injecting ion beams into the FRC, the angular velocity of ions ωi is determined by the injection kinetic energy of the beam Wi, which is given by W i = 1 2 m i V i 2 = 1 2 m i ( ω i r o ) 2 ( 2 ) Here, Vi=ωir0, where Vi is the injection velocity of ions, ωi is the cyclotron frequency of ions, and r0 is the radius of the null surface 86. The kinetic energy of electrons in the beam has been ignored because the electron mass me is much smaller than the ion mass mi. For a fixed injection velocity of the beam (fixed ωi), the applied magnetic field B0 can be tuned so that different values of ωe are obtainable. As will be shown, tuning the external magnetic field B0 also gives rise to different values of the electrostatic field inside the plasma layer. This feature of the invention is illustrated in FIGS. 11A and 11B. FIG. 11A shows three plots of the electric field (in volts/cm) obtained for the same injection velocity, ωi=1.35×107 s−1, but for three different values of the applied magnetic field B0: PlotApplied magnetic field (B0)electron angular velocity (ωe)154B0 = 2.77 kGωe = 0156B0 = 5.15 kGωe = 0.625 × 107 s−1158B0 = 15.5 kGωe = 1.11 × 107 s−1The values of ωe in the table above were determined according to Eq. 1. One can appreciate that ωe>0 means that Ω0>ωi in Eq. 1, so that electrons rotate in their counterdiamagnetic direction. FIG. 11B shows the electric potential (in volts) for the same set of values of B0 and ωe. The horizontal axis, in FIGS. 11A and 11B, represents the distance from the FRC axis 78, shown in the graph in centimeters. The electric field and electric potential depend strongly on ωe. The above results can be explained on simple physical grounds. When the ions rotate in the diamagnetic direction, the ions are confined magnetically by the Lorentz force. This was shown in FIG. 4A. For electrons, rotating in the same direction as the ions, the Lorentz force is in the opposite direction, so that electrons would not be confined. The electrons leave the plasma and, as a result, a surplus of positive charge is created. This sets up an electric field that prevents other electrons from leaving the plasma. The direction and the magnitude of this electric field, in equilibrium, is determined by the conservation of momentum. The electrostatic field plays an essential role on the transport of both electrons and ions. Accordingly, an important aspect of this invention is that a strong electrostatic field is created inside the plasma layer 106, the magnitude of this electrostatic field is controlled by the value of the applied magnetic field B0 which can be easily adjusted. As explained, the electrostatic field is confining for electrons if ωe>0. As shown in FIG. 11B, the depth of the well can be increased by tuning the applied magnetic field B0. Except for a very narrow region near the null circle, the electrons always have a small gyroradius. Therefore, electrons respond to short-wavelength fluctuations with an anomalously fast diffusion rate. This diffusion, in fact, helps maintain the potential well once the fusion reaction occurs. The fusion product ions, being of much higher energy, leave the plasma. To maintain charge quasi-neutrality, the fusion products must pull electrons out of the plasma with them, mainly taking the electrons from the surface of the plasma layer. The density of electrons at the surface of the plasma is very low, and the electrons that leave the plasma with the fusion products must be replaced; otherwise, the potential well would disappear. FIG. 12 shows a Maxwellian distribution 162 of electrons. Only very energetic electrons from the tail 160 of the Maxwell distribution can reach the surface of the plasma and leave with fusion ions. The tail 160 of the distribution 162 is thus continuously created by electron-electron collisions in the region of high density near the null surface. The energetic electrons still have a small gyroradius, so that anomalous diffusion permits them to reach the surface fast enough to accommodate the departing fusion product ions. The energetic electrons lose their energy ascending the potential well and leave with very little energy. Although the electrons can cross the magnetic field rapidly, due to anomalous transport, anomalous energy losses tend to be avoided because little energy is transported. Another consequence of the potential well is a strong cooling mechanism for electrons that is similar to evaporative cooling. For example, for water to evaporate, it must be supplied the latent heat of vaporization. This heat is supplied by the remaining liquid water and the surrounding medium, which then thermalize rapidly to a lower temperature faster than the heat transport processes can replace the energy. Similarly, for electrons, the potential well depth is equivalent to water's latent heat of vaporization. The electrons supply the energy required to ascend the potential well by the thermalization process that re-supplies the energy of the Maxwell tail so that the electrons can escape. The thermalization process thus results in a lower electron temperature, as it is much faster than any heating process. Because of the mass difference between electrons and protons, the energy transfer time from protons is about 1800 times less than the electron thermalization time. This cooling mechanism also reduces the radiation loss of electrons. This is particularly important for advanced fuels, where radiation losses are enhanced by fuel ions with an atomic number Z greater than 1; Z>1. The electrostatic field also affects ion transport. The majority of particle orbits in the plasma layer 106 are betatron orbits 112. Large-angle collisions, that is, collisions with scattering angles between 90° and 180°, can change a betatron orbit to a drift orbit. As described above, the direction of rotation of the drift orbit is determined by a competition between the {right arrow over (E)}×{right arrow over (B)} drift and the gradient drift. If the {right arrow over (E)}×{right arrow over (B)} drift dominates, the drift orbit rotates in the diamagnetic direction. If the gradient drift dominates, the drift orbit rotates in the counterdiamagnetic direction. This is shown in FIGS. 13A and 13B. FIG. 13A shows a transition from a betatron orbit to a drift orbit due to a 180° collision, which occurs at the point 172. The drift orbit continues to rotate in the diamagnetic direction because the {right arrow over (E)}×{right arrow over (B)} drift dominates. FIG. 13B shows another 180° collision, but in this case the electrostatic field is weak and the gradient drift dominates. The drift orbit thus rotates in the counterdiamagnetic direction. The direction of rotation of the drift orbit determines whether it is confined or not. A particle moving in a drift orbit will also have a velocity parallel to the FRC axis. The time it takes the particle to go from one end of the FRC to the other, as a result of its parallel motion, is called transit time; thus, the drift orbits reach an end of the FRC in a time of the order of the transit time. As shown in connection with FIG. 10A, the Lorentz force at the ends of the FRC is confining only for drift orbits rotating in the diamagnetic direction. After a transit time, therefore, ions in drift orbits rotating in the counterdiamagnetic direction are lost. This phenomenon accounts for a loss mechanism for ions, which is expected to have existed in all FRC experiments. In fact, in these experiments, the ions carried half of the current and the electrons carried the other half. In these conditions the electric field inside the plasma was negligible, and the gradient drift always dominated the {right arrow over (E)}×{right arrow over (B)} drift. Hence, all the drift orbits produced by large-angle collisions were lost after a transit time. These experiments reported ion diffusion rates that were faster than those predicted by classical diffusion estimates. If there is a strong electrostatic field, the {right arrow over (E)}×{right arrow over (B)} drift dominates the gradient drift, and the drift orbits rotate in the diamagnetic direction. This was shown above in connection with FIG. 13A. When these orbits reach the ends of the FRC, they are reflected back into the region of closed field lines by the Lorentz force; thus, they remain confined in the system. The electrostatic fields in the colliding beam system may be strong enough, so that the {right arrow over (E)}×{right arrow over (B)} drift dominates the gradient drift. Thus, the electrostatic field of the system would avoid ion transport by eliminating this ion loss mechanism, which is similar to a loss cone in a mirror device. Another aspect of ion diffusion can be appreciated by considering the effect of small-angle, electron-ion collisions on betatron orbits. FIG. 14A shows a betatron orbit 112; FIG. 14B shows the same orbit 112 when small-angle electron-ion collisions are considered 174; FIG. 14C shows the orbit of FIG. 14B followed for a time that is longer by a factor of ten 176; and FIG. 14D shows the orbit of FIG. 14B followed for a time longer by a factor of twenty 178. It can be seen that the topology of betatron orbits does not change due to small-angle, electron-ion collisions; however, the amplitude of their radial oscillations grows with time. In fact, the orbits shown in FIGS. 14A to 14D fatten out with time, which indicates classical diffusion. Formation of the FRC Conventional procedures used to form a FRC primarily employ the theta pinch-field reversal procedure. In this conventional method, a bias magnetic field is applied by external coils surrounding a neutral gas back-filled chamber. Once this has occurred, the gas is ionized and the bias magnetic field is frozen in the plasma. Next, the current in the external coils is rapidly reversed and the oppositely oriented magnetic field lines connect with the previously frozen lines to form the closed topology of the FRC (see FIG. 3). This formation process is largely empirical and there exists almost no means of controlling the formation of the FRC. The method has poor reproducibility and no tuning capability as a result. In contrast, the FRC formation methods of the present invention allow for ample control and provide a much more transparent and reproducible process. In fact, the FRC formed by the methods of the present invention can be tuned and its shape as well as other properties can be directly influenced by manipulation of the magnetic field applied by the outer field coils 325. Formation of the FRC by methods of the present inventions also results in the formation of the electric field and potential well in the manner described in detail above. Moreover, the present methods can be easily extended to accelerate the FRC to reactor level parameters and high-energy fuel currents, and advantageously enables the classical confinement of the ions. Furthermore, the technique can be employed in a compact device and is very robust as well as easy to implement—all highly desirable characteristics for reactor systems. In the present methods, FRC formation relates to the circulating plasma beam 335. It can be appreciated that the circulating plasma beam 335, because it is a current, creates a poloidal magnetic field, as would an electrical current in a circular wire. Inside the circulating plasma beam 335, the magnetic self-field that it induces opposes the externally applied magnetic field due to the outer coil 325. Outside the plasma beam 335, the magnetic self-field is in the same direction as the applied magnetic field. When the plasma ion current is sufficiently large, the self-field overcomes the applied field, and the magnetic field reverses inside the circulating plasma beam 335, thereby forming the FRC topology as shown in FIGS. 3 and 5. The requirements for field reversal can be estimated with a simple model. Consider an electric current IP carried by a ring of major radius r0 and minor radius a<<r0. The magnetic field at the center of the ring normal to the ring is Bp=2πIp/(cro). Assume that the ring current IP=Npe(Ω0/2π) is carried by Np ions that have an angular velocity Ω0. For a single ion circulating at radius r0=V0/Ω0, Ω0=eB0/mic is the cyclotron frequency for an external magnetic field B0. Assume V0 is the average velocity of the beam ions. Field reversal is defined as B p = N p e Ω 0 r 0 c > _ 2 B 0 , ( 3 ) which implies that Np>2 r0/αi, and I p > _ e V 0 πα i , ( 4 ) where αi=e2/mic2=1.57×10−16 cm and the ion beam energy is ½miV02. In the one-dimensional model, the magnetic field from the plasma current is Bp=(2π/c)ip, where ip is current per unit of length. The field reversal requirement is ip>eV0/πr0αi=0.225 kA/cm, where B0=69.3 G and ½miV02=100 eV. For a model with periodic rings and Bz is averaged over the axial coordinate Bz=(2π/c)(Ip/s) (s is the ring spacing), if s=r0, this model would have the same average magnetic field as the one dimensional model with ip=IP/s.Combined Beam/Betatron Formation Technique A preferred method of forming a FRC within the confinement system 300 described above is herein termed the combined beam/betatron technique. This approach combines low energy beams of plasma ions with betatron acceleration using the betatron flux coil 320. The first step in this method is to inject a substantially annular cloud layer of background plasma in the chamber 310 using the background plasma sources 345. Outer coil 325 produces a magnetic field inside the chamber 310, which magnetizes the background plasma. At short intervals, low energy ion beams are injected into the chamber 310 through the injector ports 340 substantially transverse to the externally applied magnetic field within the chamber 310. As explained above, the ion beams are trapped within the chamber 310 in large betatron orbits by this magnetic field. The ion beams may be generated by an ion accelerator, such as an accelerator comprising an ion diode and a Marx generator. (see R. B. Miller, An Introduction to the Physics of Intense Charged Particle Beams, (1982)). As one of skill in the art can appreciate, the applied magnetic field will exert a Lorentz force on the injected ion beam as soon as it enters the chamber 310; however, it is desired that the beam not deflect, and thus not enter a betatron orbit, until the ion beam reaches the circulating plasma beam 335. To solve this problem, the ion beams are neutralized with electrons and, as illustrated in FIG. 15, when the ion beam 350 is directed through an appropriate magnetic field, such as the unidirectional applied magnetic field within the chamber 310, the positively charged ions and negatively charged electrons separate. The ion beam 350 thus acquires an electric self-polarization due to the magnetic field. This magnetic field also may be produced by, e.g., a permanent magnet or by an electromagnet along the path of the ion beam. When subsequently introduced into the confinement chamber 310, the resultant electric field balances the magnetic force on the beam particles, allowing the ion beam to drift undeflected. FIG. 16 shows a head-on view of the ion beam 350 as it contacts the plasma 335. As depicted, electrons from the plasma 335 travel along magnetic field lines into or out of the beam 350, which thereby drains the beam's electric polarization. When the beam is no longer electrically polarized, the beam joins the circulating plasma beam 335 in a betatron orbit around the principle axis 315, as shown in FIG. 1 (see also FIG. 5). When the plasma beam 335 travels in its betatron orbit, the moving ions comprise a current, which in turn gives rise to a poloidal magnetic self-field. To produce the FRC topology within the chamber 310, it is necessary to increase the velocity of the plasma beam 335, thus increasing the magnitude of the magnetic self-field that the plasma beam 335 causes. When the magnetic self-field is large enough, the direction of the magnetic field at radial distances from the axis 315 within the plasma beam 335 reverses, giving rise to a FRC. (See FIGS. 3 and 5). It can be appreciated that, to maintain the radial distance of the circulating plasma beam 335 in the betatron orbit, it is necessary to increase the applied magnetic field from the outer coil 325 as the circulating plasma beam 335 increases in velocity. A control system is thus provided for maintaining an appropriate applied magnetic field, dictated by the current through the outer coil 325. Alternatively, a second outer coil may be used to provide the additional applied magnetic field that is required to maintain the radius of the plasma beam's orbit as it is accelerated. To increase the velocity of the circulating plasma beam 335 in its orbit, the betatron flux coil 320 is provided. Referring to FIG. 17, it can be appreciated that increasing a current through the betatron flux coil 320, by Ampere's Law, induces an azimuthal electric field, E, inside the chamber 310. The positively charged ions in the plasma beam 335 are accelerated by this induced electric field, leading to field reversal as described above. When ion beams 350, which are neutralized and polarized as described above, are added to the circulating plasma beam 335, the plasma beam 335 depolarizes the ion beams. For field reversal, the circulating plasma beam 335 is preferably accelerated to a rotational energy of about 100 eV, and preferably in a range of about 75 eV to 125 eV. To reach fusion relevant conditions, the circulating plasma beam 335 is preferably accelerated to about 200 keV and preferably to a range of about 100 keV to 3.3 MeV. FRC formation was successfully demonstrated utilizing the combined beam/betatron formation technique. The combined beam/betatron formation technique was performed experimentally in a chamber 1 m in diameter and 1.5 m in length using an externally applied magnetic field of up to 500 G, a magnetic field from the rotating plasma induced by the betatron flux coil 320 of up to 5 kG, and a vacuum of 1.2×10−5 torr. In the experiment, the background plasma had a density of 1013 cm−3 and the ion beam was a neutralized Hydrogen beam having a density of 1.2×1013 cm−3, a velocity of 2×107 cm/s, and a pulse length of around 20 μs (at half height). Field reversal was observed. Betatron Formation Technique Another preferred method of forming a FRC within the confinement system 300 is herein termed the betatron formation technique. This technique is based on driving the betatron induced current directly to accelerate a circulating plasma beam 335 using the betatron flux coil 320. A preferred embodiment of this technique uses the confinement system 300 depicted in FIG. 1, except that the injection of low energy ion beams is not necessary. As indicated, the main component in the betatron formation technique is the betatron flux coil 320 mounted in the center and along the axis of the chamber 310. Due to its separate parallel windings construction, the coil 320 exhibits very low inductance and, when coupled to an adequate power source, has a low LC time constant, which enables rapid ramp up of the current in the flux coil 320. Preferably, formation of the FRC commences by energizing the external field coils 325, 330. This provides an axial guide field as well as radial magnetic field components near the ends to axially confine the plasma injected into the chamber 310. Once sufficient magnetic field is established, the background plasma sources 345 are energized from their own power supplies. Plasma emanating from the guns streams along the axial guide field and spreads slightly due to its temperature. As the plasma reaches the mid-plane of the chamber 310, a continuous, axially extending, annular layer of cold, slowly moving plasma is established. At this point the betatron flux coil 320 is energized. The rapidly rising current in the coil 320 causes a fast changing axial flux in the coil's interior. By virtue of inductive effects this rapid increase in axial flux causes the generation of an azimuthal electric field E (see FIG. 18), which permeates the space around the flux coil. By Maxwell's equations, this electric field E is directly proportional to the change in strength of the magnetic flux inside the coil, i.e.: a faster betatron coil current ramp-up will lead to a stronger electric field. The inductively created electric field E couples to the charged particles in the plasma and causes a ponderomotive force, which accelerates the particles in the annular plasma layer. Electrons, by virtue of their smaller mass, are the first species to experience acceleration. The initial current formed by this process is, thus, primarily due to electrons. However, sufficient acceleration time (around hundreds of micro-seconds) will eventually also lead to ion current. Referring to FIG. 18, this electric field E accelerates the electrons and ions in opposite directions. Once both species reach their terminal velocities, current is carried about equally by ions and electrons. As noted above, the current carried by the rotating plasma gives rise to a self magnetic field. The creation of the actual FRC topology sets in when the self magnetic field created by the current in the plasma layer becomes comparable to the applied magnetic field from the external field coils 325, 330. At this point magnetic reconnection occurs and the open field lines of the initial externally produced magnetic field begin to close and form the FRC flux surfaces (see FIGS. 3 and 5). The base FRC established by this method exhibits modest magnetic field and particle energies that are typically not at reactor relevant operating parameters. However, the inductive electric acceleration field will persist, as long as the current in the betatron flux coil 320 continues to increase at a rapid rate. The effect of this process is that the energy and total magnetic field strength of the FRC continues to grow. The extent of this process is, thus, primarily limited by the flux coil power supply, as continued delivery of current requires a massive energy storage bank. However, it is, in principal, straightforward to accelerate the system to reactor relevant conditions. For field reversal, the circulating plasma beam 335 is preferably accelerated to a rotational energy of about 100 eV, and preferably in a range of about 75 eV to 125 eV. To reach fusion relevant conditions, the circulating plasma beam 335 is preferably accelerated to about 200 keV and preferably to a range of about 100 keV to 3.3 MeV. When ion beams are added to the circulating plasma beam 335, as described above, the plasma beam 335 depolarizes the ion beams. FRC formation utilizing the betatron formation technique was successfully demonstrated at the following parameter levels: Vacuum chamber dimensions: about 1 m diameter, 1.5 m length. Betatron coil radius of 10 cm. Plasma orbit radius of 20 cm. Mean external magnetic field produced in the vacuum chamber was up to 100 Gauss, with a ramp-up period of 150 μs and a mirror ratio of 2 to 1. (Source: Outer coils and betatron coils). The background plasma (substantially Hydrogen gas) was characterized by a mean density of about 1013 cm−3, kinetic temperature of less than 10 eV. The lifetime of the configuration was limited by the total energy stored in the experiment and generally was around 30 μs. The experiments proceeded by first injecting a background plasma layer by two sets of coaxial cable guns mounted in a circular fashion inside the chamber. Each collection of 8 guns was mounted on one of the two mirror coil assemblies. The guns were azimuthally spaced in an equidistant fashion and offset relative to the other set. This arrangement allowed for the guns to be fired simultaneously and thereby created an annular plasma layer. Upon establishment of this layer, the betatron flux coil was energized. Rising current in the betatron coil windings caused an increase in flux inside the coil, which gave rise to an azimuthal electric field curling around the betatron coil. Quick ramp-up and high current in the betatron flux coil produced a strong electric field, which accelerated the annular plasma layer and thereby induced a sizeable current. Sufficiently strong plasma current produced a magnetic self-field that altered the externally supplied field and caused the creation of the field reversed configuration. Detailed measurements with B-dot loops identified the extent, strength and duration of the FRC. An example of typical data is shown by the traces of B-dot probe signals in FIG. 19. The data curve A represents the absolute strength of the axial component of the magnetic field at the axial mid-plane (75 cm from either end plate) of the experimental chamber and at a radial position of 15 cm. The data curve B represents the absolute strength of the axial component of the magnetic field at the chamber axial mid-plane and at a radial position of 30 cm. The curve A data set, therefore, indicates magnetic field strength inside of the fuel plasma layer (between betatron coil and plasma) while the curve B data set depicts the magnetic field strength outside of the fuel plasma layer. The data clearly indicates that the inner magnetic field reverses orientation (is negative) between about 23 and 47 μs, while the outer field stays positive, i.e., does not reverse orientation. The time of reversal is limited by the ramp-up of current in the betatron coil. Once peak current is reached in the betatron coil, the induced current in the fuel plasma layer starts to decrease and the FRC rapidly decays. Up to now the lifetime of the FRC is limited by the energy that can be stored in the experiment. As with the injection and trapping experiments, the system can be upgraded to provide longer FRC lifetime and acceleration to reactor relevant parameters. Overall, this technique not only produces a compact FRC, but it is also robust and straightforward to implement. Most importantly, the base FRC created by this method can be easily accelerated to any desired level of rotational energy and magnetic field strength. This is crucial for fusion applications and classical confinement of high-energy fuel beams. Inductive Plasma Source The betatron and beam/betatron FRC formation techniques describe above, both rely on imparting energy to a background plasma via the flux coil 320. Analogous to a transformer, the flux coil performs the duties of the primary windings of the transformer, while the plasma acts as the secondary windings. For this inductive system to work efficiently, it is imperative that the plasma is a good conductor. Counter to typical conductors, such as metals, a plasma becomes less resistive and, thus, more conductive as its temperature increases. The temperature of plasma electrons, in particular, plays an important role and, to a large degree, determines dissipation, which is a function of electron-ion collisions. In essence, dissipation is due to resistance, which is caused by electron-ion collisions: the higher the collision frequency, the higher the resistivity. This is due to the collective phenomena in a plasma, where the coulomb collision cross-section is screened. The collision frequency (the rate at which successive collisions occur) is essentially a function of density, screened coulomb scattering cross-section and thermal (or average) velocity of the colliding/scattering charges, i.e.: νc=nσv. By definition v scales with T1/2, σ is proportional to v−4 or, thus, T−2. The collision frequency νc is, therefore, proportional to nT−3/2. The resistivity is related to the collision frequency by η=νcm/ne2. Hence, the resistivity is proportional to T−3/2 and, notably, independent of density—a direct result of the fact that even though the number of charge carriers increases with density, the number of scattering centers increases as well. Thus, higher temperature leads to higher plasma conductivity and less dissipative losses. To achieve better performance with regard to confinement in an FRC, a hot plasma is, therefore, highly desirable. In the case of the PEG system, enhanced electron temperature leads to improved FRC startup (the better a conductor the plasma becomes, the better the inductive coupling between the plasma and flux coil), better current sustainment (reduced plasma resistivity leads to less frictional/dissipative losses and hence less current loss) and higher magnetic field strength (the stronger the current, the more self-field). Adequate electron temperature during initial plasma formation and before the flux coil is engaged will lead to better coupling of the flux coil to the plasma (which advantageously tends to reduce the formation of azimuthal image currents in the chamber wall). This in turn will result in enhanced betatron acceleration (less resistivity leads to better inductive transfer of energy from flux coil to plasma) and plasma heating (some of the imparted directional energy as represented by the rotating current flow will thermalize and turn to random energy—ultimately leading to heating of the plasma by the flux coil), which will consequently increase the ion-electron collision time (due to higher temperature), reduce dissipation (less resistivity) and allow ultimately for the attainment of higher FRC fields (higher currents lead to stronger fields). To achieve better initial plasma temperature, an inductive plasma source is provided. As depicted in FIGS. 20A, 20B and 20C, the inductive plasma source 1010 is mountable within the chamber 310 about the end of the flux coil 320 and includes a single turn shock coil assembly 1030 that is preferably fed by a high voltage (about 5-15 kV) power source (not shown). Neutral gas, such as Hydrogen (or other appropriate gaseous fusion fuel), is introduced into the source 1010 through direct gas feeds via a Laval nozzle 1020. The gas flow is controlled preferably by sets of ultra fast puff valves to produce a clean shock front. Once the gas emanates from the nozzle 1020 and distributes itself over the surface of the coil windings 1040 of the shock coil 1030, the windings 1040 are energized. The ultra fast current and flux ramp-up in the low inductance shock coil 1030 leads to a very high electric field within the gas that causes breakdown, ionization and subsequent ejection of the formed plasma from the surface of the shock coil 1030 towards the center of the chamber 310. In a preferred embodiment, the shock coil 1030 comprises an annular disc shaped body 1032 bounded by an outer ring 1034 formed about its outer periphery and an annular hub 1036 formed about its inner periphery. The ring 1034 and hub 1036 extend axially beyond the surface of the body 1032 forming the edges of a open top annular channel 1035. The body 1032, ring 1034 and hub 1036 are preferably formed through unitary molded construction of an appropriate non-conductive material with good vacuum properties and low outgassing properties such as glass, plexiglass, pirex, quartz, ceramics or the like. A multi-sectioned shroud 1012 is preferably coupled to the ring 1034 of the shock coil 1030 to limit the produced plasma from drifting radially. Each section 1014 of the shroud 1012 includes a plurality of axially extending fingers 1016. The ends of each section 1014 include a mounting bracket 1015. The coil windings 1040 are preferably affixed to the face of the coil body 1032 in the channel 1035 using epoxy or some other appropriate adhesive. To obtain fast electro-magnetic characteristics of the shock coil 1030, it is important to keep its inductance as low as possible. This is achieved by using as few turns in the coil 1040 as possible, as well as building the coil 1040 up of multiple strands of wire 1042 that are wound in parallel. In an exemplary embodiment, the coil 1040 comprised 24 parallel strands of wire 1042, each of which executed one loop. The wires 1042 each begin at entry points 1044 that are located preferably about 15 degrees apart on the outer perimeter of the body 1032 and end after only one axis encircling turn at exit points 1046 on the inner radius of the body 1032. The coil windings 1040, therefore, cover the entire area between the inner and outer edges of channel 1035. Preferably, groups of strands 1042 are connected to the same capacitive storage bank. In general, power can be fed to all strands 1042 from the same capacitive storage bank or, as in an exemplary embodiment, 8 groups of 3 strands 1042 each are connected together and commonly fed by one of 2 separate capacitive storage banks. An annular disc-shaped nozzle body 1022 is coupled about its inner perimeter to the hub 1036 to form the Laval nozzle 1020. The surface 1024 of the nozzle body 1022 facing the hub 1036 has an expanding midsection profile defining an annular gas plenum 1025 between the surface 1024 and the face 1037 of the hub 1036. Adjacent the outer periphery of the nozzle body 1022, the surface 1024 has a contracting-to-expanding profile defining an azimuthally extending Laval-type nozzle outlet 1023 between the surface 1024 and the face 1037 of the hub 1036. Attached to the opposite side of the hub 1036 is a valve seat ring 1050 with several valve seats 1054 formed in the outer face of the ring 1050. The valve seats 1054 are aligned with gas feed channels 1052 formed through the hub 1036. In operation, neutral gas is feed through ultra fast puff valves in the valve seats 1054 to the gas channels 1052 extending through the hub 1036. Because of the constricting portion of the nozzle outlet 1023, the gas tends to feed into and fill the annular plenum 1025 prior to emanating from the nozzle 1020. Once the gas emanates from the nozzle 1020 and distributes itself over the surface of the coil windings 1040 of the shock coil 1030, the windings 1040 are energized. The ultra fast current and flux ramp-up in the low inductance shock coil 1030 leads to a very high electric field within the gas that causes breakdown, ionization and subsequent ejection of the formed plasma from the surface of the shock coil 1030 towards the center of the chamber 310. The current ramp-up is preferably well synchronized in all strands 1042 or groups of strands 1042 that are intended to be fired together. Another option that is possible and potentially advantageous, is to fire different groups of strands at different times. A delay can be deliberately instituted between engaging different groups of strands 1042 to fire different groups of strands at different times. When firing different groups of strands at different times it is important to group strands in a way so that the arrangement is azimuthally symmetric and provides sufficient coverage of the surface of the coil 1040 with current carrying wires 1042 at any given power pulse. In this fashion it is possible to create at least two consecutive but distinct plasma pulses. The delay between pulses is limited by how much neutral gas is available. In practice, it is possible to fire such pulses between about 5 and 600 micro-seconds apart. In practice, the input operating parameters are preferably as follows: Charging Voltage: about 10 to 25 kV split supply Current: up to about 50 kA total current through all windings combined Pulse/Rise Time: up to about 2 microseconds Gas Pressure: about −20 to 50 psi Plenum size: about 0.5 to 1 cm3 per valve—i.e.: about 4 to 8 cm3 total gas volume per shot In an exemplary embodiment, the input operating parameters were as follows: Charging Voltage: 12 to 17 kV split supply, i.e.: from −12 kV to +12 kV Current: 2 to 4.5 kA per group of 3 strands, i.e.: 16 to 36 kA total current through all windings combined Pulse/Rise Time: 1 to 1.5 microseconds Gas Pressure: −15 to 30 psi Plenum size: 0.5 to 1 cm3 per valve—i.e.: 4 to 8 cm3 total gas volume per shot The plasma created by this operational method of the inductive plasma source 1010 using the parameters noted above has the following advantageous characteristics: Density ˜4×1013 cm−3 Temperature ˜10-20 eV Annular scale ˜40-50 cm diameter Axial drift velocity ˜5-10 eV. Due to the shape and orientation of the source 1010, the shape of the emerging plasma is annular and has a diameter tending to equal the rotating plasma annulus of the to be formed FRC. In a PEG present system two such inductive plasma sources 1010 are preferably placed on either axial end of the chamber 310 and preferably fired in parallel. The two formed plasma distributions drift axially towards the center of the chamber 310 where they form the annular layer of plasma that is then accelerated by the flux coil 320 as described above.RF Drive for Ions and Electrons in FRC A RF current drive, called a rotomak, has been employed for FRCs in which the current is carried mainly by electrons. It involves a rotating radial magnetic field produced by two phased antennas. The electrons are magnetized and frozen to the rotating magnetic field lines. This maintains the current until Coulomb collisions of the ions with electrons cause the ions to be accelerated and reduce the current. The rotomak, however, is not suitable for maintaining the current indefinitely, but it has been successful for milliseconds. In the FRCs of the present system the current is mainly carried by ions that are in betatron orbits which would not be frozen to rotating magnetic field lines. The large orbit ions are important for stability and classical diffusion. Instead of antennas, electrodes are employed as in cyclotrons and the ions are driven by an electrostatic wave. The problem is completely electrostatic because the frequency of the RF is less than 10 Megacycles so that the wavelength (30 m) is much longer than any dimension of the plasma. Electrostatic fields can penetrate the FRC plasma much more easily than electromagnetic waves. The electrostatic wave produced by the electrodes is designed to travel at a speed that is close to the average azimuthal velocity of the ions, or of the electrons. If the wave travels faster than the average speed of the ions, it will accelerate them and thereby compensate for the drag due to the ion-electron collisions. Electrons, however, are accelerated by Coulomb collisions with the ions. In this case the wave must have a speed slower than the electron average velocity and the electrons will accelerate the wave. The average electron velocity is less than the average ion velocity so that the electrons must be driven at two different frequencies. The higher frequency will be for ions and energy is preferably supplied by the external circuit. For electrons, energy can be extracted at the lower frequency. Electrode Systems A quadrupole RF drive system is shown in FIGS. 21A and 21B. As depicted, the RF drive comprises a quadrupolar cyclotron 1110 located within the chamber 310 and having four elongate, azimuthally symmetrical electrodes 1112 with gaps 1114 there between. The quadrupole cyclotron 1110 preferably produces an electric potential wave that rotates in the same direction as the azimuthal velocity of ions, but at a greater velocity. Ions of appropriate speed can be trapped in this wave, and reflected periodically. This process increases the momentum and energy of the fuel ions and this increase is conveyed to the fuel ions that are not trapped by collisions. Fuel ions from the fuel plasma 335 may be replaced by injecting neutrals at any convenient velocity. An alternative and supplemental method to drive current is to augment the electrode system with additional magnetic field coils 1116 positioned about the flux coil 325 and quadrupole cyclotron 1110, and that are driven at half the frequency of the cyclotron electrodes 1112. The following discussion presented here, however, is dedicated to illustrate the electrode only version (without magnetic field coils 1116). In FIG. 21C electrodes are illustrated for two and four electrode configurations. The potential created by the electrodes with the indicated applied voltages are noted in FIG. 21C for vacuum in the space r<rb. The expressions are for the lowest harmonic. They are obtained by solving the Laplace equation ( 1 r ∂ ∂ r r ∂ ∂ r + 1 r 2 ∂ ∂ θ 2 ) Φ ( r , θ ; t ) = 0 ( 5 ) with appropriate boundary conditions. For example for the dipole cyclotron Φ ( r b , t ) = - V o cos ω t for 0 ≤ θ ≤ π = V o cos ω t for π ≤ θ ≤ 2 π Φ ( r , θ ; t ) is finite . ( 6 ) Since Φ(r,θ; t) is periodic in θ with a period 2π, it can be expanded in a Fourier series, i.e.: Φ ( r , θ ; t ) = ∑ n = - ∞ ∞ u n ( r , t ) ⅇ ⅈn θ ( 7 ) u n ( r , t ) = 1 2 π ∫ 0 2 π ⅆ θ ′ ⅇ - ⅈn θ ′ Φ ( r , θ ′ ; t ) ( 8 ) and un satisfies the equation ( 1 2 ∂ ∂ r r ∂ ∂ r + n 2 r 2 ) u n ( r , t ) = 0 u n ( r o , t ) = V o cos ω t in π ( ⅇ - ⅈn π - 1 ) = 0 if n = 2 , 4 … etc . u n ( 0 , t ) = 0 ( 9 ) Φ ( r , θ ; t ) = 4 V o cos ω t π ∑ l = 1 ∞ sin ( 2 l - 1 ) θ 2 l - 1 ( r r b ) 2 l - 1 . ( 10 ) The lowest harmonic is Φ 1 ( r , θ ; t ) = 2 V o π r r b [ sin ( ω t + θ ) - sin ( ω t - θ ) ] ( 11 ) Higher harmonics are Φ l ( r , θ ; t ) = 2 V o π ( r r b ) 2 l - 1 { sin [ ω t + ( 2 l - 1 ) θ ] - sin [ ω t - ( 2 l - 1 ) θ ] } ( 12 ) The wave speed in the azimuthal direction is {dot over (θ)}=±ω/(2l−1) so that the higher harmonics have a smaller phase velocity and amplitude. These comments apply to both cases in FIG. 21C. The frequency ω would be close to ωi the frequency of rotation of the ions in a rigid rotor equilibrium for the FRC. Thus {dot over (θ)}=ωi for l=1. For l=2{dot over (θ)}=ωi/3 and the wave amplitude would be substantially lower; it is thus a good approximation to consider only the lowest harmonic. Plasma Effect The response of the plasma can be described by a dielectric tensor. The electric field produces plasma currents which produce charge separation according to the charge conservation equation ∇ · J → + ∂ ρ ∂ t = 0 ( 13 ) where {right arrow over (J)} is current density and ρ is charge density. The appropriate equation is∇·{right arrow over (E)}=4πρ=4π{right arrow over (χ)}·{right arrow over (E)} (14)or∇·{right arrow over (ε)}·{right arrow over (E)}=−∇·{right arrow over (ε)}·∇Φ=0where =+4π is the dielectric tensor and χ is the polarizability. If only the contribution of the electrons is included the tensor is diagonal with one component ɛ ⊥ = 1 + 4 π nmc 2 B 2 ( 15 ) where n is the density and B is the FRC magnetic field. n and B vary rapidly with r and B=0 on a surface at r=ro within the plasma. The expression for ε⊥ is derived assuming electrons have a small gyroradius and the electric field changes slowly compared to Ω=eB/mc, the gyrofrequency. This approximation breaks down near the null surface. The characteristic orbits change from drift orbits to betatron orbits which have a much smaller response to the electric field, i.e. ε⊥≅1 near the null surface at r=ro. The ions mainly have betatron orbits and for the drift orbits the response to the electric field is small because the electric field changes at the rate ω≅ωi. The net result is that the Laplace equation is replaced by 1 r ∂ ∂ r r ∂ Φ ∂ r + 1 ɛ ⊥ ( r ) ⅆ ɛ ⊥ ⅆ r ∂ Φ ∂ r + 1 r 2 ∂ 2 Φ ∂ r 2 = 0 ( 16 ) which must be solved numerically. The additional term vanishes near r=ro. The potential for the lowest harmonic of the quadrupole case has the form Φ = V o F ( r ) 2 sin ( 2 θ - ω t ) ( 17 ) and a similar form for the dipole case. Waves traveling in the opposite direction to the ions (or electrons) will be neglected.Acceleration Due to Ions Trapped in an Electrostatic Wave We assume that ω=2ωi+Δω so that the wave {dot over (θ)}=ω/2=ωi+Δω/2 is a little faster than the ions. The standard rigid rotor distribution function is assumed for the ions f i ( x → , v → ) = ( m i 2 π T i ) 3 / 2 n i ( r ) exp { [ - m i 2 T i [ v r 2 + v z 2 + ( v θ - r ω i ) 2 ] ] } . ( 18 ) The reduced distribution function of interest is F i ( r , v θ ) = ( m i 2 π T i ) 1 / 2 exp [ - m i 2 T i ( v θ - r ω i ) 2 ] . The wave velocity of the electrostatic wave produced by the quadrupole cyclotron is νw=rω/2=rωi+Δνw. Ions moving faster than the wave reflect if v θ - v w < 2 e Φ o m i . ( 19 ) This increases the wave energy, i.e., ⅆ W + ⅆ t = ∑ i = 1 , 2 n i m i λ ∫ v θ = v w v θ = v w + 2 e Φ 0 m i ⅆ v θ F i ( r , v θ ) [ v θ 2 2 - ( 2 v w - v θ ) 2 2 ] ( v θ - v w ) . ( 20 ) Ions moving slower than the wave reflect if v w - v θ < 2 e Φ o m i . and the wave loses energy at the rate ⅆ W - ⅆ t = ∑ i = 1 , 2 n i m i λ ∫ v θ = v - 2 e Φ o m i v θ = v w ⅆ v θ F i ( r , v θ ) [ v θ 2 2 - ( 2 v w - v θ ) 2 2 ] ( v w - v θ ) . ( 21 ) The net results is simplified with the change of variable ν′θ=νθ−νw, i.e., ⅆ W ⅆ t = ⅆ W + ⅆ t - ⅆ W - ⅆ t = ∑ i = 1 , 2 2 n i m i v w λ ∫ 0 2 e Φ o m i ⅆ v θ ′ ( v θ ′ ) 2 [ F i ( v w + v θ ′ ) - F i ( v w - v θ ′ ) ] . ( 22 ) The approximation F i [ v w + _ v θ ] = F i ( v w ) + _ ∂ F i ∂ v θ ❘ v w v θ , ( 23 ) results in ⅆ W ⅆ t = ∑ i = 1 , 2 2 n i m i v w λ ( 2 e Φ o m i ) 2 ∂ F i ∂ v θ ❘ v o = v w . ( 24 ) This has a form similar to Landau damping, but it is not physically the same because Landau damping (growth) is a linear phenomena and this is clearly non-linear. Since ∂ F i ∂ v θ ❘ v w = ( m 2 π T i ) 1 / 2 m i T o ( v w - r ω o ) exp [ - m i 2 T i ( v w - r ω i ) 2 ] . ( 25 ) If νw=rωi there is no change in the wave energy. If ww>rωi or Δνw>0, the wave energy decreases; for Δνw<0 the wave energy increases. This is similar to the interpretation of Landau damping. In the first case Δνw>0, there are more ions going slower than the wave than faster. Therefore, the wave energy decreases. In the opposite case Δνw<0, the wave energy increases. The former case applies to maintaining the ion energy and momentum with a quadrupole cyclotron. This is current drive. The latter case provides the basis for a converter. Eqs. (22) and (24) can be used to evaluate the applicability to fusion reactor conditions. The power transferred to the ions when νwrωi=Δνw≅νi, the ion thermal velocity, is P = 2 π ∫ 0 r b ⅆ W ⅆ t r ⅆ r ,where dW/dt is determined by Eqs. (24) and (25). To simplify the integration Φo(r) is replaced by Φo(ro), the value at the peak density which is a lower bound of the wave amplitude. P = ( 2 π ) 3 / 2 ∑ i = 1 , 2 ( N i T i ) ω i [ 2 e i Φ o ( r o ) T i ] 2 ( 26 ) Ni is the line density of ions. i=1, 2 accommodates two types of ions which is usually the case in a reactor. Detailed calculations of F(r) indicate that the wave amplitude Φo(ro) is about a factor of 10 less than the maximum gap voltage which is 2Vo. This will determine the limitations of this method of RF drive. Vo will be limited by the maximum gap voltage that can be sustained which is probably about 10 kVolts for a 1 cm gap. Reactor Requirements For current drive a power P, is preferably transferred to the ions at frequency ωi and a power Pe is preferably transferred to the electrons at frequency ωe. This will compensate for the Coulomb interactions between electrons and ions, which reduces the ion velocity and increases the electron velocity. (In the absence of the power transfers, Coulomb collisions would lead to the same velocity for electrons and ions and no current). The average electric field to maintain the equilibrium of electrons and ions is given by2πr0Eθ=IR (27)where I = N e e 2 π ( ω i - ω e ) is the current/unit length and R = ( 2 π r 0 ) 2 m N e e 2 ( N 1 Z 1 m 1 N e t 1 e + N 2 Z 2 m 2 N e t 2 e ) is the resistance/unit length. Ne, N1, N2 are line densities of electrons and ions Ne=N1Z1+N2Z2 where Z1, Z2 are atomic numbers of the ions; t1e and t2e are momentum transfer times from ions to electrons. The average electric field is the same for ions or electrons because Ne≅N. for quasi-neutrality and the charge is opposite. The power that must be transferred to the ions isPi=2πr0IiθEθ (28)and the power that can be extracted from electrons isPe=−|2πr0Ieθ(Eθ (29)where Iiθ=Neeωi/2π and Ieθ=Neeωe/2π. For refueling with the RF drive the fuel may be replaced at any energy at rates given by the fusion times tF1=1/n1σν1 and tF2=1/n2σν2; n1 and n2 are plasma ion densities and σν are reactivities. The magnitude will, be seconds. The injected neutrals (to replace the fuel ions that burn and disappear) will ionize rapidly and accelerate due to Coulomb collisions up to the average ion velocity in a time of the order of milliseconds (for reactor densities of order 1015 cm−3). However this requires an addition to Eθ and an addition to transfer of power to maintain a steady state. The addition is δ 〈 E θ 〉 = V i θ - V b θ N e e 2 ( N 1 Z 1 m 1 t F 1 + N 2 Z 2 m 2 t F 2 ) ( 30 ) which will increase the required power transfer by about a factor of two (2). The power can be provided for current drive and refueling without exceeding the maximum gap voltage amplitude of 10 kVolts/cm. Considering that the frequency will be 1-10 Mega-Hertz and the magnetic field will be of order 100 kGauss no breakdown would be expected. The power that must be transferred for current drive and refueling is similar for any current drive method. However RF technology at 1-10 Mega-Hertz has been an established high-efficiency technology for many years. The method described that uses electrodes instead of antennas has a considerable advantage because the conditions for field penetration are much more relaxed than for electro-magnetic waves. Therefore this method would have advantages with respect to circulating power and efficiency. Fusion Significantly, these two techniques for forming a FRC inside of a containment system 300 described above, or the like, can result in plasmas having properties suitable for causing nuclear fusion therein. More particularly, the FRC formed by these methods can be accelerated to any desired level of rotational energy and magnetic field strength. This is crucial for fusion applications and classical confinement of high-energy fuel beams. In the confinement system 300, therefore, it becomes possible to trap and confine high-energy plasma beams for sufficient periods of time to cause a fusion reaction therewith. To accommodate fusion, the FRC formed by these methods is preferably accelerated to appropriate levels of rotational energy and magnetic field strength by betatron acceleration. Fusion, however, tends to require a particular set of physical conditions for any reaction to take place. In addition, to achieve efficient burn-up of the fuel and obtain a positive energy balance, the fuel has to be kept in this state substantially unchanged for prolonged periods of time. This is important, as high kinetic temperature and/or energy characterize a fusion relevant state. Creation of this state, therefore, requires sizeable input of energy, which can only be recovered if most of the fuel undergoes fusion. As a consequence, the confinement time of the fuel has to be longer than its burn time. This leads to a positive energy balance and consequently net energy output. A significant advantage of the present invention is that the confinement system and plasma described herein are capable of long confinement times, i.e., confinement times that exceed fuel burn times. A typical state for fusion is, thus, characterized by the following physical conditions (which tend to vary based on fuel and operating mode): Average ion temperature: in a range of about 30 to 230 keV and preferably in a range of about 80 keV to 230 keV Average electron temperature: in a range of about 30 to 100 keV and preferably in a range of about 80 to 100 keV Coherent energy of the fuel beams (injected ion beams and circulating plasma beam): in a range of about 100 keV to 3.3 MeV and preferably in a range of about 300 keV to 3.3 MeV. Total magnetic field: in a range of about 47.5 to 120 kG and preferably in a range of about 95 to 120 kG (with the externally applied field in a range of about 2.5 to 15 kG and preferably in a range of about 5 to 15 kG). Classical Confinement time: greater than the fuel burn time and preferably in a range of about 10 to 100 seconds. Fuel ion density: in a range of about 1014 to less than 1016 cm−3 and preferably in a range of about 1014 to 1015 cm−3. Total Fusion Power: preferably in a range of about 50 to 450 kW/cm (power per cm of chamber length) To accommodate the fusion state illustrated above, the FRC is preferably accelerated to a level of coherent rotational energy preferably in a range of about 100 keV to 3.3 MeV, and more preferably in a range of about 300 keV to 3.3 MeV, and a level of magnetic field strength preferably in a range of about 45 to 120 kG, and more preferably in a range of about 90 to 115 kG. At these levels, high energy ion beams, which are neutralized and polarized as described above, can be injected into the FRC and trapped to form a plasma beam layer wherein the plasma beam ions are magnetically confined and the plasma beam electrons are electrostatically confined. Preferably, the electron temperature is kept as low as practically possible to reduce the amount of bremsstrahlung radiation, which can, otherwise, lead to radiative energy losses. The electrostatic energy well of the present invention provides an effective means of accomplishing this. The ion temperature is preferably kept at a level that provides for efficient burn-up since the fusion cross-section is a function of ion temperature. High direct energy of the fuel ion beams is essential to provide classical transport as discussed in this application. It also minimizes the effects of instabilities on the fuel plasma. The magnetic field is consistent with the beam rotation energy. It is partially created by the plasma beam (self-field) and in turn provides the support and force to keep the plasma beam on the desired orbit. Fusion Products The fusion products are born in the power core predominantly near the null surface 86 from where they emerge by diffusion towards the separatrix 84 (see FIGS. 3 and 5). This is due to collisions with electrons (as collisions with ions do not change the center of mass and therefore do not cause them to change field lines). Because of their high kinetic energy (fusion product ions have much higher energy than the fuel ions), the fusion products can readily cross the separatrix 84. Once they are beyond the separatrix 84, they can leave along the open field lines 80 provided that they experience scattering from ion-ion collisions. Although this collisional process does not lead to diffusion, it can change the direction of the ion velocity vector such that it points parallel to the magnetic field. These open field lines 80 connect the FRC topology of the core with the uniform applied field provided outside the FRC topology. Product ions emerge on different field lines, which they follow with a distribution of energies. Advantageously, the product ions and charge-neutralizing electrons emerge in the form of rotating annular beams from both ends of the fuel plasma. For example for a 50 MW design of a p-B11 reaction, these beams will have a radius of about 50 centimeters and a thickness of about 10 centimeters. In the strong magnetic fields found outside the separatrix 84 (typically around 100 kG), the product ions have an associated distribution of gyro-radii that varies from a minimum value of about 1 cm to a maximum of around 3 cm for the most energetic product ions. Initially the product ions have longitudinal as well as rotational energy characterized by ½ M(vpar)2 and ½ M(vperp)2. vperp is the azimuthal velocity associated with rotation around a field line as the orbital center. Since the field lines spread out after leaving the vicinity of the FRC topology, the rotational energy tends to decrease while the total energy remains constant. This is a consequence of the adiabatic invariance of the magnetic moment of the product ions. It is well known in the art that charged particles orbiting in a magnetic field have a magnetic moment associated with their motion. In the case of particles moving along a slow changing magnetic field, there also exists an adiabatic invariant of the motion described by ½ M(vperp)2/B. The product ions orbiting around their respective field lines have a magnetic moment and such an adiabatic invariant associated with their motion. Since B decreases by a factor of about 10 (indicated by the spreading of the field lines), it follows that vperp will likewise decrease by about 3.2. Thus, by the time the product ions arrive at the uniform field region their rotational energy would be less than 5% of their total energy; in other words almost all the energy is in the longitudinal component. Energy Conversion The direct energy conversion system of the present invention comprises an inverse cyclotron converter (ICC) 420 shown in FIGS. 22A and 23A coupled to a (partially illustrated) power core 436 of a colliding beam fusion reactor (CBFR) 410 to form a plasma-electric power generation system 400. A second ICC (not shown) may be disposed symmetrically to the left of the CBFR 410. A magnetic cusp 486 is located between the CBFR 410 and the ICC 420 and is formed when the CBFR 410 and ICC 420 magnetic fields merge. Before describing the ICC 420 and its operation in detail, a review of a typical cyclotron accelerator is provided. In conventional cyclotron accelerators, energetic ions with velocities perpendicular to a magnetic field rotate in circles. The orbit radius of the energetic ions is determined by the magnetic field strength and their charge-to-mass ratio, and increases with energy. However, the rotation frequency of the ions is independent of their energy. This fact has been exploited in the design of cyclotron accelerators. FIG. 22B shows an end view of the inverse cyclotron converter in FIG. 22A. Referring to FIG. 24A, a conventional cyclotron accelerator 700 includes two mirror image C-shaped electrodes 710 forming mirror image D-shaped cavities placed in a homogenous magnetic field 720 having field lines perpendicular to the electrodes' plane of symmetry, i.e., the plane of the page. An oscillating electric potential is applied between the C-shaped electrodes (see FIG. 24B). Ions I are emitted from a source placed in the center of the cyclotron 700. The magnetic field 720 is adjusted so that the rotation frequency of the ions matches that of the electric potential and associated electric field. If an ion I crosses the gap 730 between the C-shaped electrodes 710 in the same direction as that of the electric field, it is accelerated. By accelerating the ion I, its energy and orbit radius increase. When the ion has traveled a half-circle arc (experiencing no increase in energy), it crosses the gap 730 again. Now the electric field between the C-shaped electrodes 710 has reversed direction. The ion I is again accelerated, and its energy is further increased. This process is repeated every time the ion crosses the gap 730 provided its rotation frequency continues to match that of the oscillating electric field (see FIG. 24C). If on the other hand a particle crosses the gap 730 when the electric field is in the opposite direction it will be decelerated and returned to the source at the center. Only particles with initial velocities perpendicular to the magnetic field 720 and that cross the gaps 730 in the proper phase of the oscillating electric field will be accelerated. Thus, proper phase matching is essential for acceleration. In principle, a cyclotron could be used to extract kinetic energy from a pencil beam of identical energetic ions. Deceleration of ions with a cyclotron, but without energy extraction has been observed for protons, as described by Bloch and Jeffries in Phys. Rev. 80, 305 (1950). The ions could be injected into the cavity such that they are brought into a decelerating phase relative to the oscillating field. All of the ions would then reverse the trajectory T of the accelerating ion shown in FIG. 24A. As the ions slow down due to interaction with the electric field, their kinetic energy is transformed into oscillating electric energy in the electric circuit of which the cyclotron is part. Direct conversion to electric energy would be achieved, tending to occur with very high efficiency. In practice, the ions of an ion beam would enter the cyclotron with all possible phases. Unless the varying phases are compensated for in the design of the cyclotron, half of the ions would be accelerated and the other half decelerated. As a result, the maximum conversion efficiency would effectively be 50%. Moreover the annular fusion product ion beams discussed above are of an unsuitable geometry for the conventional cyclotron. As discussed in greater detail below, the ICC of the present invention accommodates the annular character of the fusion product beams exiting the FRC of fusion reactor power core, and the random relative phase of the ions within the beam and the spread of their energies. Referring back to FIG. 22A, a portion of a power core 436 of the CBFR 410 is illustrated on the left side, wherein a plasma fuel core 435 is confined in a FRC 470 formed in part due to a magnetic field applied by outside field coils 425. The FRC 470 includes closed field lines 482, a separatrix 484 and open field lines 480, which, as noted above, determines the properties of the annular beam 437 of the fusion products. The open field lines 480 extend away from the power core 436 towards the magnetic cusp 486. As noted above, fusion products emerge from the power core 436 along open field lines 480 in the form of an annular beam 437 comprising energetic ions and charge neutralizing electrons. The geometry of the ICC 420 is like a hollow cylinder with a length of about five meters. Preferably, four or more equal, semi-cylindrical electrodes 494 with small, straight gaps 497 make up the cylinder surface. In operation, an oscillating potential is applied to the electrodes 494 in an alternating fashion. The electric field E within the converter has a quadrupole structure as indicated in the end view illustrated in FIG. 22B. The electric field E vanishes on the symmetry axis and increases linearly with the radius; the peak value is at the gap 497. In addition, the ICC 420 includes outside field coils 488 to form a uniform magnetic field within the ICC's hollow cylinder geometry. Because the current runs through the ICC field coils 488 in a direction opposite to the direction of the current running through the CBFR field coils 425, the field lines 496 in the ICC 420 run in a direction opposite to the direction of the open field lines 480 of the CBFR 410. At an end furthest from the power core 436 of the CBFR 410, the ICC 420 includes an ion collector 492. In between the CBFR 410 and the ICC 420 is a symmetric magnetic cusp 486 wherein the open field lines 480 of the CBFR 410 merge with the field lines 496 of the ICC 420. An annular shaped electron collector 490 is position about the magnetic cusp 486 and electrically coupled to the ion collector 498. As discussed below, the magnetic field of the magnetic cusps 486 converts the axial velocity of the beam 437 to a rotational velocity with high efficiency. FIG. 22C illustrates a typical ion orbit 422 within the converter 420. The CBFR 410 has a cylindrical symmetry. At its center is the fusion power core 436 with a fusion plasma core 435 contained in a FRC 470 magnetic field topology in which the fusion reactions take place. As noted, the product nuclei and charge-neutralizing electrons emerge as annular beams 437 from both ends of the fuel plasma 435. For example for a 50 MW design of a p-B11 reaction, these beams will have a radius of about 50 cm and a thickness of about 10 cm. The annular beam has a density n=107-108 cm3. For such a density, the magnetic cusp 486 separates the electrons and ions. The electrons follow the magnetic field lines to the electron collector 490 and the ions pass through the cusp 486 where the ion trajectories are modified to follow a substantially helical path along the length of the ICC 420. Energy is removed from the ions as they spiral past the electrodes 494 connected to a resonant circuit (not shown). The loss of perpendicular energy is greatest for the highest energy ions that initially circulate close to the electrodes 494, where the electric field is strongest. The ions arrive at the magnetic cusp 486 with the rotational energy approximately equal to the initial total energy, i.e., ½Mvp2≅½Mv02. There is a distribution of ion energies and ion initial radii r0 when the ions reach the magnetic cusp 486. However, the initial radii r0 tends to be approximately proportional to the initial velocity v0. The radial magnetic field and the radial beam velocity produce a Lorentz force in the azimuthal direction. The magnetic field at the cusp 486 does not change the particle energy but converts the initial axial velocity vP≅v0 to a residual axial velocity vz and an azimuthal velocity v⊥, where v02=vz2−v⊥2. The value of the azimuthal velocity v⊥ can be determined from the conservation of canonical momentum P θ = Mr 0 v ⊥ - qB 0 r 0 2 2 c = qB 0 r 0 2 2 c ( 31 ) A beam ion enters the left hand side of the cusp 486 with Bz=B0, vz=v0, v⊥=0 and r=r0. It emerges on the right hand side of the cusp 486 with r=r0, Bz=−B0, v⊥=qB0r0/Mc and v z = v 0 2 - v ⊥ 2 v z v 0 = 1 - ( r 0 Ω 0 v 0 ) 2 ( 32 ) where Ω 0 = qB 0 Mc is the cyclotron frequency. The rotation frequency of the ions is in a range of about 1-10 MHz, and preferably in a range of about 5-10 MHz, which is the frequency at which power generation takes place. In order for the ions to pass through the cusp 486, the effective ion gyro-radius must be greater than the width of the cusp 486 at the radius r0. It is quite feasible experimentally to reduce the axial velocity by a factor of 10 so that the residual axial energy will be reduced by a factor of 100. Then 99% of the ion energy will be converted to rotational energy. The ion beam has a distribution of values for v0 and r0. However, because r0 is proportional to v0 as previously indicated by the properties of the FRC based reactor, the conversion efficiency to rotational energy tends to be 99% for all ions. As depicted in FIG. 22B, the symmetrical electrode structure of the ICC 420 of the present invention preferably includes four electrodes 494. A tank circuit (not shown) is connected to the electrode structures 494 so that the instantaneous voltages and electric fields are as illustrated. The voltage and the tank circuit oscillate at a frequency of ω=Ω0. The azimuthal electric field E at the gaps 497 is illustrated in FIG. 22B and FIG. 25. FIG. 25 illustrates the electric field in the gaps 497 between electrodes 494 and the field an ion experiences as it rotates with angular velocity Ω0. It is apparent that in a complete revolution the particle will experience alternately acceleration and deceleration in an order determined by the initial phase. In addition to the azimuthal electric field Eθ there is also a radial electric field Er. The azimuthal field Eθ is maximum in the gaps 497 and decreases as the radius decreases. FIG. 22 assumes the particle rotates maintaining a constant radius. Because of the gradient in the electric field the deceleration will always dominate over the acceleration. The acceleration phase makes the ion radius increase so that when the ion next encounters a decelerating electric field the ion radius will be larger. The deceleration phase will dominate independent of the initial phase of the ion because the radial gradient of the azimuthal electric field Eθ is always positive. As a result, the energy conversion efficiency is not limited to 50% due to the initial phase problem associated with conventional cyclotrons. The electric field Er is also important. It also oscillates and produces a net effect in the radial direction that returns the beam trajectory to the original radius with zero velocity in the plane perpendicular to the axis as in FIG. 22C. The process by which ions are always decelerated is similar to the principle of strong focusing that is an essential feature of modern accelerators as described in U.S. Pat. No. 2,736,799. The combination of a positive (focusing) and negative lens (defocusing) is positive if the magnetic field has a positive gradient. A strong focusing quadrupole doublet lens is illustrated in FIG. 26. The first lens is focusing in the x-direction and defocusing in the y-direction. The second lens is similar with x and y properties interchanged. The magnetic field vanishes on the axis of symmetry and has a positive radial gradient. The net results for an ion beam passing through both lenses is focusing in all directions independent of the order of passage. Similar results have been reported for a beam passing through a resonant cavity containing a strong axial magnetic field and operating in the TE111 mode (see Yoshikawa et al.). This device is called a peniotron. In the TE111 mode the resonant cavity has standing waves in which the electric field has quadrupole symmetry. The results are qualitatively similar to some of the results described herein. There are quantitative differences in that the resonance cavity is much larger in size (10 meter length), and operates at a much higher frequency (155 MHz) and magnetic field (10 T). Energy extraction from the high frequency waves requires a rectenna. The energy spectrum of the beam reduces the efficiency of conversion. The existence of two kinds of ions is a more serious problem, but the efficiency of conversion is adequate for a D-He3 reactor that produces 15 MeV protons. A single particle orbit 422 for a particle within the ICC 420 is illustrated in FIG. 22C. This result was obtained by computer simulation and a similar result was obtained for the peniotron. An ion entering at some radius r0 spirals down the length of the ICC and after losing the initial rotational energy converges to a point on a circle of the same radius r0. The initial conditions are asymmetric; the final state reflects this asymmetry, but it is independent of the initial phase so that all particles are decelerated. The beam at the ion collector end of the ICC is again annular and of similar dimensions. The axial velocity would be reduced by a factor of 10 and the density correspondingly increased. For a single particle an extracting efficiency of 99% is feasible. However, various factors, such as perpendicular rotational energy of the annular beam before it enters the converter, may reduce this efficiency by about 5%. Electric power extraction would be at about 1-10 MHz and preferably about 5-10 MHz, with additional reduction in conversion efficiency due to power conditioning to connect to a power grid. As shown in FIGS. 23A and 23B, alternative embodiments of the electrode structures 494 in the ICC 420 may include two symmetrical semi-circular electrodes and/or tapered electrodes 494 that taper towards the ion collector 492. Adjustments to the ion dynamics inside the main magnetic field of the ICC 420 may be implemented using two auxiliary coil sets 500 and 510, as shown in FIGS. 27A and 27B. Both coil sets 500 and 510 involve adjacent conductors with oppositely directed currents, so the magnetic fields have a short range. A magnetic-field gradient, as schematically illustrated in FIG. 27A, will change the ion rotation frequency and phase. A multi-pole magnetic field, as schematically illustrated in FIG. 27B, will produce bunching, as in a linear accelerator. Reactor FIG. 28 illustrates a 100 MW reactor. The generator cut away illustrates a fusion power core region having superconducting coils to apply a uniform magnetic field and a flux coil for formation of a magnetic field with field-reversed topology. Adjacent opposing ends of the fusion power core region are ICC energy converters for direct conversion of the kinetic energy of the fusion products to electric power. The support equipment for such a reactor is illustrated in FIG. 29. Propulsion System Exploration of the solar system (and beyond) requires propulsion capabilities that far exceed the best available chemical or electric propulsion systems. For advanced propulsion applications, the present invention holds the most promise: design simplicity, high-thrust, high specific impulse, high specific power-density, low system mass, and fuels that produce little or no radio-activity. A plasma-thrust propulsion system, in accordance with the present invention, utilizes the high kinetic energy embedded in the fusion products as they are expelled axially out of the fusion plasma core. The system 800 is illustrated schematically in FIGS. 30 and 31. The system includes a FRC power core 836 colliding beam fusion reactor in which a fusion fuel core 835 is contained as described above. The reactor further comprises a magnetic field generator 825, a current coil (not shown) and ion beam injectors 840. An ICC direct-energy converter 820, as described above, is coupled to one end of the power core 836, and intercepts approximately half of the fusion product particles which emerge from both ends of the power core 836 in the form of annular beams 837. As described above, the ICC 820 decelerates them by an inverse cyclotron process, and converts their kinetic energy into electric energy. A magnetic nozzle 850 is positioned adjacent the other end of the power core 836 and directs the remaining fusion product particles into space as thrust T. The annular beam 837 of fusion products stream from one end of the fusion power core 836 along field lines 837 into the ICC 820 for energy conversion and from the other end of the power core 836 along field lines 837 out of the nozzle 850 for thrust T. Bremsstrahlung radiation is converted into electric energy by a thermoelectric-energy converter (TEC) 870. Bremsstrahlung energy that is not converted by the TEC 870 is passed to a Brayton-cycle heat engine 880. Waste heat is rejected to space. A power-control subsystem (810, see FIG. 32), monitors all sources and sinks of electric and heat energy to maintain system operation in the steady state and to provide an independent source of energy (i.e, fuel-cells, batteries, etc.) to initiate operation of the space craft and propulsion system from a non-operating state. Since the fusion products are charged a-particles, the system does not require the use of massive radiation and neutron shields and hence is characterized by significantly reduced system mass compared to other nuclear space propulsion systems. The performance of the plasma-thrust propulsion system 800 is characterized by the following kinetic parameters for a 100 MW p-B11 fusion core example having a design as depicted in FIG. 31: Specific Impulse, Isp 1.4 × 106 sThrust Power, PT50.8 MWThrust Power/Total Output Power, PT/Po 0.51Thrust, T28.1 NThrust/Total Output Power, T/Po281 mN/MW The system 800 exhibits a very high specific impulse, which allows for high terminal velocities of a space craft utilizing the plasma-thrust propulsion system. A key mission performance/limitation metric for all space vehicles is system mass. The principal mass components in the plasma-thrust propulsion system 800 are illustrated in FIGS. 31 and 32. The fusion core 835 requires approximately 50 MW of injected power for steady-state operation. The system generates approximately 77 MW of nuclear (particle) power, half of which is recovered in the direct-energy converter 820 with up to 90% efficiency. Thus, an additional 11.5 MW is needed to sustain the reactor, which is provided by the TEC 870 and Brayton-heat engine 880. The principal source of heat in the plasma-thruster propulsion system 880 is due to Bremsstrahlung radiation. The TEC 870 recovers approximately 20% of the radiation, or 4.6MW, transferring approximately 18.2 MW to the closed-cycle, Brayton-heat engine 880. The Brayton-heat engine 880 comprises a heat exchanger 860, turbo-alternator 884, compressor 882, and radiators 886, as shown in FIG. 31. The Brayton engine 880 supplies the remaining 7 MW of power needed to sustain the reactor, another 11 MW is dumped directly to space by means of radiators. A closed-cycle, a Brayton-heat engine is a mature and efficient option to convert excess heat rejected by the TEC 870. In Brayton engines the maximum-cycle temperature is constrained by material considerations, which limits the maximum thermodynamic-cycle efficiency. Based on a standard performance map for the Brayton engine, several design points can be extracted. Typical efficiencies can reach up to 60%. For the present case, 7 MW is needed to be recovered, hence, only a 40% efficiency in converting waste heat is acceptable and well within currently attainable limits of conventional Brayton engines. The component mass for the entire Brayton engine (less the heat radiators) is calculated based on specific-mass parameters typical of advanced industrial technologies, i.e. in the range of 3 kg/kWe. Turbomachines, including compressors, power turbines, and heat exchangers, are combined for a total subsystem mass of 18 MT. The radiator mass is estimated to be 6 MT, preferably using heat-pipe panels with state-of-the-art high thermal conductivity. Significant system weight also comes from the magnets 825 confining the plasma core 835. The superconducting magnetic coils 825 are preferably made of Nb3 Sn, which operates stably at 4.5K and at a field of 12.5-13.5 T. The cryogenic requirements for Nb3 Sn are less stringent than other materials considered. With a magnetic field requirement of 7 Tesla and a device length of approximately 7.5 meters, the coil needs about 1500 turns of wire carrying 56 kA of current. Using 0.5-cm radius wires, the total mass of this coil is about 3097 kg. The liquid helium cooling system is comprised of two pumps, one at each end of the main coil. The total mass of these pumps is approximately 60 kg. The outer structural shell is used to support the magnets and all internal components from outside. It is made of 0.01-m thick kevlar/carbon-carbon composite with a total mass of about 772 kg. The outermost layer is the insulation jacket to shield the interior from the large temperature variation in space is estimated at 643 kg. The total mass for the magnet subsystem 825 is, therefore, about 4.8 MT. At present, the ion injection system 840 most appropriate for space applications would be an induction linac or RFQ. Approximately 15 years ago an RFQ was flown on a scientific rocket and successfully demonstrated the use of high voltage power and the injection of ion beams into space. In a preferred embodiment, six injectors 840 distributed along the length of the CBFR, three for each species of ion. Each injector 840 is preferably a 30 beamlet RFQ with an overall dimension of 0.3 meters long and a 0.020-m radius. Each injector requires an ion source, preferably 0.02-meters long and 0.020-meters radius, that supplies ionized hydrogen or boron. One source is needed for each accelerator. Both the injector and the source are well within currently attainable limits; with design refinements for space their total mass, including the sources and the accelerators, should be about 60 kg. The cone-shaped ICC direct energy converter 820 is located at one end of the reactor 836, which is preferably made of stainless steel. With a base radius of 0.5 meters and a length of 2 meters, the ICC mass is approximately 1690 kg. An RF power supply 820 (inverter/converter) recovers the directed-ion flow, converting it into electric power. The power supply mass is about 30 kg. A storage battery 812 is used to start/re-start the CBFR. The stored capacity is about 30 MJ. Its mass is about 500 kg. Alternately, a fuel cell could also be used. Additional control units coordinate operation of all the components. The control-subsystem mass is estimated to be 30 kg. The total energy converter/starter subsystem mass is, therefore, estimated at about 2.25 MT. A magnetic nozzle 850 is located at the other end of the fusion core 835. The nozzle 850 focuses the fusion product stream as a directed particle flow. It is estimated that the mass of the magnetic nozzle and the ICC are about equal; since both are comprised of superconducting magnets and relatively low-mass, structural components. The TEC 870 recovers energy from the electromagnetic emissions of the fusion core. It is preferably a thin-film structure made of 0.02-cm thick boron-carbide/silicon-germanium, which has a mass density of about 5 g/cm3. The TEC 870 is located at the first wall and preferably completely lines the inner surface of the reactor core; the mass of the TEC 870 is estimated at about 400 kg. The radiant flux onto the TEC 870 is 1.2 MW/m2 and its peak operating temperature is assumed to be less than 1800° K. The total plasma-thruster propulsion system mass is thus estimated at about 33 MT. This defines the remaining mission-critical parameters for the presently discussed 100 MW unit: Total Mass/Total Power, MT/Po0.33 × 10−3 kg/WThrust/Mass, T/MT0.85 × 10−3 N/kg While the invention is susceptible to various modifications and alternative forms, a specific example thereof has been shown in the drawings and is herein described in detail. It should be understood, however, that the invention is not to be limited to the particular form disclosed, but to the contrary, the invention is to cover all modifications, equivalents, and alternatives falling within the spirit and scope of the appended claims. |
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claims | 1. A method for producing neutrons, the method comprising:producing a voltage of negative polarity of at least −100 keV on a surface of a deuterated or tritiated target in response to a temperature change of a pyroelectric crystal of less than about 40° C., the pyroelectric crystal having the deuterated or tritiated target coupled thereto;pulsing a deuterium ion source to produce a deuterium ion beam; anddirecting the ion beam onto the deuterated or tritiated target to make neutrons using at least one element selected from the group consisting of: a voltage of the pyroelectric crystal and a high gradient insulator (HGI) surrounding the pyroelectric crystal. 2. The method of claim 1, wherein the pyroelectric crystal is formed of a material selected from a group consisting of: lithium tantalite, lithium niobate, and barium strontiate. 3. The method of claim 1, further comprising a changing a temperature of the pyroelectric crystal using a thermal altering mechanism. 4. The method of claim 3, wherein the thermal altering mechanism includes at least one mechanism selected from the group consisting of: a chemical heating pack, a chemical cooling pack, a Peltier heater/cooler, a thermite composition, a resistive heating element, a dielectric fluid system, and a thermoelectric heater/cooler. 5. The method of claim 3, wherein the thermal altering mechanism raises or lowers a temperature of the pyroelectric crystal by about 10° C. to about 150° C. to produce a voltage of negative polarity on the surface of the deuterated or tritiated target of at least about −100 keV. 6. The method of claim 3, wherein the thermal altering mechanism raises or lowers a temperature of the pyroelectric crystal by less than about 40° C. to produce a voltage of negative polarity on the surface of the deuterated or tritiated target of at least about −100 keV. 7. The method of claim 1, wherein the element includes the high gradient insulator (HGI) surrounding the pyroelectric crystal, wherein the directing includes using an ion accelerating mechanism for accelerating the deuterium ion beam toward the deuterated or tritiated target. 8. The method of claim 1, wherein the ion source is deuterated such that the deuterium ion beam is produced when the ion source is pulsed. 9. The method of claim 1, wherein the deuterium ion source includes at least one source from the group consisting of: a cold cathode gated nanotip array, a nanotube ion source, and a spark source. 10. The method of claim 1, wherein the deuterated or tritiated target covers at least a portion of at least one side of the pyroelectric crystal. 11. The method of claim 10, wherein the deuterated or tritiated target has an inverted cone geometry with a focusing tip extending toward the deuterium ion source. 12. The method of claim 1, wherein the deuterated or tritiated target is positioned between the deuterium ion source and the pyroelectric crystal. 13. The method of claim 1, wherein the target is deuterated. 14. The method of claim 1, wherein the target is tritiated. 15. The method of claim 1, wherein directing the ion beam onto the deuterated or tritiated target includes accelerating the deuterium ion beam to the deuterated or tritiated target to produce a neutron beam. |
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description | The present invention relates generally to a passive filtration system for a nuclear reactor power plant and more specifically, to a passive filtration system in the fuel handling area of the nuclear reactor power plant. The generation of electric power by a nuclear reactor is accomplished by the nuclear fission of radioactive materials. Due to the volatility of the nuclear reaction, nuclear power plants are required by practice to be designed in such a manner that the health and safety of the public is assured. In conventional nuclear reactors used for generating electric power, the nuclear fuel becomes spent and is removed at periodic intervals from the nuclear reactor and replaced with fresh fuel. The spent fuel generates decay heat and remains radioactive after it has been removed from the nuclear reactor. Thus, a safe storage facility is provided to receive the spent fuel. In nuclear reactors, such as pressurized water reactors, a pool is provided as a storage pool for the spent fuel. The spent fuel pool is designed to contain a level of water such that the spent fuel is stored underwater. The spent fuel pool is typically constructed of concrete and is at least 40 feet deep. In addition to the level of the water being controlled and monitored, the quality of the water is also controlled and monitored to prevent fuel degradation when it is in the spent fuel pool. Further, the water in the spent fuel pool is continuously cooled to remove the heat which is produced by the spent fuel. In general, a nuclear power plant includes a spent fuel pool cooling system which is designed to remove decay heat generated by stored spent fuel from the water in the spent fuel pool. Removal of the decay heat maintains the spent fuel pool water temperature within acceptable regulatory limits. The spent fuel pool cooling system typically includes a spent fuel pool pump which circulates the high temperature water from within the spent fuel pool through a heat exchanger and then returns the cooled water to the spent fuel pool. In one embodiment, the spent fuel pool cooling system includes two mechanical trains of equipment. Each train includes one spent fuel pool pump, one spent fuel pool heat exchanger, one spent fuel pool demineralizer and one spent fuel pool filter. The two trains of equipment share common suction and discharge headers. In addition, the spent fuel pool cooling system includes the piping, valves and instrumentation necessary for system operation. In this embodiment, one train is continuously cooling and purifying the spent fuel pool while the other train is available for water transfers, in-containment refueling water storage tank purification, or alignment as a backup to the operating train of equipment. FIG. 1 shows a spent fuel pool cooling (SFPC) system 10 during its normal operation in accordance with the prior art. The SFPC 10 includes a spent fuel pool 15. The spent fuel pool 15 contains a level of water 16 which is at a high temperature as a result of the intense temperature of the spent fuel (not shown) that is transferred from the nuclear reactor (not shown) into the spent fuel pool 15. The SFPC system 10 includes trains A and B. Trains A and B are employed to cool the water in the spent fuel pool 15. As previously described, it is typical to operate either one of train A or train B to continuously cool and purify the spent fuel pool 15 while the other train is available as a back-up. Each of trains A and B include a SFPC pump 25, a heat exchanger 30, and a SFPC demineralizer and filter system 45. These trains share a common suction header 20 and a common discharge header 50. In each of trains A and B, water exits the spent fuel pool 15 through the suction header 20 and is pumped through the SFPC pump 25 to the SFPC heat exchanger 30. In the SFPC heat exchanger 30, a flow line 40 passes water from the component cooling water system (CCWS) (not shown) through the SFPC heat exchanger 30 and back to the CCWS. The heat from the water entering the SFPC heat exchanger 30 (from the spent fuel pool 15) is transferred to the water provided by the flow line 40 and is returned back to the CCWS through the flow line 40. Cooled water exits the SFPC heat exchanger 30 and passes through the SFPC demineralizer and filter system 45 positioned downstream of the SFPC heat exchanger 30. Purified, cooled water exits the demineralizer and filter system 45, is transported through the common discharge header 50, and is returned to the spent fuel pool 15. Recently, nuclear reactor manufacturers have offered passive plant designs, i.e., plants that will mitigate accident events in a nuclear reactor without operator intervention or off-site power. The Westinghouse Electric Company LLC offers the AP1000 passive plant design. The AP1000 design includes advanced passive safety features and extensive plant simplifications to enhance the safety, construction, operation, and maintenance of the plant. The AP1000 design emphasizes safety features that rely on natural forces. The safety systems in the AP1000 design use natural driving forces such as pressurized gas, gravity flow, natural circulation flow, and convection. The safety systems do not use active components (such as, pumps, fans or diesel generators) and are designed to function without safety grade support systems (such as, AC power, component cooling water, service water, and HVAC). The AP1000 fuel handling area is designed such that the primary means for fuel protection is provided by passive means and relies on the boiling of the spent fuel pool water inventory to remove decay heat. Thus, in extreme cases, the spent fuel pool can boil. Assuming a complete failure of the active spent fuel pool cooling system, spent fuel cooling can be provided by the heat capacity of the water in the spent fuel pool. Water make-up is provided to the spent fuel pool by a passive means to maintain the pool water level above the spent fuel while boiling of the pool water provides for the removal of decay heat. Boiling of the spent fuel pool water releases large quantities of steam into the fuel handling area. The steam mixes with the air in the fuel handling area and has to be released from this area to prevent a build-up of pressure. The steam/air mixture is released from the fuel handling area into the atmosphere. This can potentially result in the release of radioactive airborne contaminants into the atmosphere. Analysis has shown that minimal radiation doses that are well within acceptable limits may result from the onset of boiling. However, it is advantageous to provide a spent fuel filtration system and method for further reducing the radioactive doses that are released into the atmosphere from the onset of boiling of the spent fuel pool in the fuel handling area of a nuclear reactor. It is desired that the system and method be a passive mechanism which is simple to design and implement, and is effective to remove radioactive particulates in the event of a spent fuel pool boiling event in the nuclear reactor. In one aspect, the present invention provides a passive filtration system for a fuel handling area having a spent fuel pool in a nuclear reactor, to reduce a discharge into the atmosphere of particulates generated in a spent fuel pool boiling event. The passive filtration system includes a discharge path having a first end connected to the fuel handling area and a second end connected to the atmosphere; a vent mechanism positioned between the fuel handling area and the first end of the discharge path, the vent mechanism structured to release a steam and air mixture from the fuel handling area to the discharge path, the steam and air mixture includes the particulates; an air filtration unit located in the discharge path, the air filtration unit including at least one passive filter, the steam and air mixture forced through the at least one passive filter due to a differential pressure generated in the fuel handling area, the at least one passive filter structured to trap particulates from the steam and air mixture to produce a filtered steam and air mixture; and a second vent mechanism connected to the air filtration unit, the second vent mechanism structured to release the filtered steam and air mixture to the atmosphere. In an embodiment, the passive filtration system can further include at least one drain connected to the air filtration unit, the drain structured to return to the fuel handling area or other suitable discharge point condensate generated from the steam and air mixture in the air filtration unit. In a further embodiment, the passive filtration system can include two drains. In still a further embodiment, the passive filtration system can include one drain located forward of the filter and the other drain located behind the filter. In an embodiment, the first vent mechanism of the passive filtration system can include at least one temperature-actuated damper. In another embodiment, the second vent mechanism can include at least one fail open or gravity operated damper. In further embodiment, the first and second vent mechanisms can each include two dampers. In an embodiment, the steam and air mixture released from the first vent mechanism has a higher level of particulates as compared to the filtered steam and air mixture released from the second vent mechanism. In alternate embodiments, the nuclear reactor is a pressurized or boiling water reactor. In another embodiment, the passive filter includes a high efficiency particulate air filter. In still another embodiment, the particulates include radioactive particulates. In another aspect, the present invention provides, a method of filtering particulates from a steam and air mixture generated by a spent fuel pool boiling event in the fuel handling area of a nuclear reactor prior to discharge of the steam and air mixture to atmosphere. The method includes discharging the steam and air mixture from the fuel handling area through a venting mechanism; passing the steam and air mixture through a passive filter; trapping at least a portion of the particulates contained in steam and air mixture into the passive filter to produce a filtered steam and air mixture; and discharging the filtered steam and air mixture through a venting mechanism into the atmosphere. The discharging and the passing of the steam and air mixture employs a passive means comprising a differential pressure generated in the fuel handling area. The present invention relates to a passive filtration system and the use of at least one passive filter in the fuel handling area in a nuclear reactor, such as a pressurized water reactor, to reduce a release of particulates, such as radioactive particulates, into the atmosphere as a result of a spent fuel pool boiling event. In the nuclear reactor, a spent fuel pool is located in the fuel handling area. The spent fuel pool contains water and stores spent fuel removed from the nuclear reactor core. The spent fuel generates decay heat and remains radioactive after being removed from the nuclear reactor core and transferred into the spent fuel pool. Thus, a spent fuel pool cooling system is provided in nuclear reactors to remove decay heat and maintain the temperature of the water in the spent fuel pool at acceptable limits. Active and/or passive spent fuel pool cooling systems may be used. As previously described herein, FIG. 1 shows an example of an active spent fuel pool cooling system known in the art. Passive spent fuel pool cooling systems may be designed such that the water in the spent fuel pool boils to remove decay heat generated by the spent fuel. As a result of the boiling of the water in the spent fuel pool, large quantities of steam are generated in the fuel handling area. The steam mixes with the air in the fuel handling area. The steam and air mixture may contain particulates and contaminants, such as radioactive particulates and radioactive airborne contaminants. Further, the temperature and pressure in the fuel handling area increase as a result of boiling the water in the spent fuel pool. The mixture of steam and air is discharged from the fuel handling area, through a discharge path, and into the atmosphere to prevent the build-up of pressure in the fuel handling area. The release of the steam and air mixture can result in the release of airborne radioactive contaminants into the atmosphere. The passive filtration system of the present invention provides a means of filtering the steam and air mixture. A vent mechanism is positioned in the fuel handling area. The vent mechanism is structured to release the steam and air mixture into a discharge path which is connected to the fuel handling area. The vent mechanism can include at least one temperature-actuated damper. As the temperature increases, the at least one temperature-actuated damper opens to vent steam and air from the fuel handling area into the discharge path. In one embodiment, there are two temperature-actuated dampers such that one is available as a back-up. At least one passive filter can be positioned in the discharge path which extends from fuel handling area to the atmosphere. Thus, the steam and air which is vented through the temperature actuated damper(s) passes through the passive filter(s) prior to being discharged into the atmosphere. The steam and air mixture is forced through the passive filter(s) due to the pressure differential in the fuel handling area. The passive filter(s) is able to remove particulates and contaminants from the steam and air mixture generated in the fuel handling area as a result of a spent fuel pool boiling event. The particulates and contaminants can include radioactive particulates and airborne radioactive contaminants. Further, the passive filter(s) is effective in reducing the level of radioactive particulates and radioactive airborne contaminants that are discharged into the atmosphere. The passive filter(s) is available before, during and after a spent fuel pool boiling event. Analysis has found that the level of release of radioactive particulates is within acceptable limits provided by the United States Nuclear Regulatory Commission. However, the passive filter(s) of the present invention provides additional assurance that the release of radioactive particulates and contaminants is well within acceptable limits. The passive filter(s) for use in the present invention can include a wide variety of filters known in the art which are able to remove particulates and/or contaminants from steam, air or mixtures thereof, without an active means. In one embodiment, the filters are High Efficiency Particulate Air (HEPA) filters. Generally, HEPA filters are composed of a mat of randomly arranged fibers. The fibers can be composed of a variety of materials, such as but not limited to fiberglass. Typically, HEPA filters are operable to trap particles by having the particles adhere to the fibers or the particles being embedded into the fibers. In the present invention, the passive filter(s) provides a passive means for filtration of air to the atmosphere. The steam and air mixture is forced through the passive filter(s) by the differential pressure in the fuel handling area. Thus, there is no need for the use of an active means, such as a fan, to drive the steam and air mixture through the passive filter(s). The discharge path between the fuel handling area and the atmosphere can include various designs to incorporate the passive filter(s) and filtration path. In one embodiment, at least one passive filter is contained in a housing which is positioned in an air filtration unit that is located in the discharge path. The air filtration unit includes a vent mechanism which releases the filtered steam and air mixture into the atmosphere. The vent mechanism includes at least one fail-open or gravity operated discharge damper. The at least one fail-open or gravity operated discharge damper is positioned downstream of the passive filter(s). The number of discharge dampers can vary. It is typical to have more than one discharge damper for the purpose of redundancy. Thus, during normal operation, the fail-open or gravity operated discharge damper(s) is capable to isolate the passive filter(s) from the atmosphere. Further, the fail-open or gravity operated discharge damper(s) protects the passive filter(s) from damage when not in use (e.g., during normal plant operation of the nuclear reactor). In the present invention, the steam and air mixture which is released from the first vent mechanism has a higher level of particulates as compared to the filtered steam and air mixture that is released from the discharge damper(s). In an embodiment, the air filtration unit includes at least one water drain or a drain path capable to return condensed steam from the air filtration unit into the fuel handling area or other suitable discharge point to reduce the potential for an accidental release of condensate which may contain radioactivity. In one embodiment, the air filtration unit includes two drains. One drain is positioned forward of the passive filter(s) which is located in the air filtration unit and the other drain is position behind, e.g., after or to the rear of, the passive filter(s). During an emergency event, e.g., the spent fuel cooling system is not available and the spent fuel pool water heats up and boils to remove decay heat, the discharge damper(s) is open such that the passive filter(s) can receive steam and/or air from the fuel handling area and remove particulates, such as radioactive particulates, from the steam and air prior to its discharge from the discharge path into the atmosphere. During normal operation of the nuclear reactor, the discharge and/or temperature-operated damper(s) isolate the passive filter(s) and discharge/filtration path from the fuel handling area. FIG. 2 shows a passive spent fuel pool filtration system 100 in accordance with an embodiment of the present invention. The spent fuel pool filtration system 100 includes the spent fuel pool 15 and water level 16 as shown in FIG. 2. Further, the spent fuel pool filtration system 100 includes a fuel handling area 101 and a discharge path 115. A first end of the discharge path 115 is connected to the fuel handling area 101 and a second end of the discharge path 115 is connected to the atmosphere 155. In FIG. 2, the spent fuel pool 15 is positioned in a fuel handling area 101. The fuel handling area 101 includes dampers 105,110 which are temperature-activated and capable to open to release steam/air from the fuel handling area 101 to a discharge path 115. The dampers 105,110 are positioned between, e.g., at the interface of, the fuel handling area 101 and the first end of the discharge path 115. In the discharge path 115 is positioned an air filtration unit 125. The air filtration unit 125 includes a HEPA filter 130. In an embodiment, the air filtration unit 125 can include more than one HEPA filter 130. The steam/air in the discharge path 115 enters the air filtration unit 125 and passes through the HEPA filter 130. The HEPA filter is capable to remove particulates and contaminants from the steam/air mixture. The filtered steam/air mixture then exits the HEPA filter 130, passes through dampers 145,150, connected to the air filtration unit 125, and is discharged into the atmosphere 155. The dampers 145,150 are fail-open, and motorized or pneumatic; or are gravity operated. Further, the air filtration unit 125 includes drains 135,140. Any condensate, e.g., water, that condenses from the steam/air can be collected in the drains 135,140 and returned to the fuel handling area 101. The drain 135 is located upstream of the HEPA filter 130 and the drain 140 is located downstream, e.g., after, the HEPA filter 130. While the invention has been described in terms of various specific embodiments, those skilled in the art will recognize that the invention can be practiced with modifications within the spirit and scope of the appended claims. |
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050254642 | claims | 1. A support structure for spectral analysis comprising: a diffraction grating of equidistantly spaced wires arranged in parallel with each other; horizontal bars equidistantly spaced with respect to each other and aligned perpendicular to said grating wires in a plurality of 2MN-parallel rows, where M and N are integers and where N is the distance from one vertical support to the next support in a single row, measured in terms of the number of wire spacings lying between adjacent ones of said vertical supports; said horizontal bars bonded to said grating thereby structurally contributing to the support of said wires; vertical supports bonded to adjacent ones of said horizontal bars and aligned parallel to said wires; said vertical supports being equidistantly spaced relative to each other in each of said rows and having their positions relative to each vertical support in all of the other rows determined by a pseudo-random integer of the spacings between said wires. preparing a grating of equidistantly spaced wires; structurally supporting said wires with a matrix of horizontal bars and vertical supports arranged so as to form a plurality of horizontal rows; spacing the horizontal supports at regular intervals relative to each other; spacing the vertical bars in a regular pattern relative to each other in each row and in a pseudo-random pattern relative to vertical bars in each of the other rows; and attaching each of said vertical supports at the ends thereof to adjacent ones of said horizontal bars. (a) Creating an array of numbers j(m) for which j(m)=mod (m) i.e., : etc. for all m=0,1,2, . . . , [L] MN-1 where M is an integer; (b) Choose a random integer 0.ltoreq.n<[L]MN, and for the first row, K=j(n); (c) Create a new array from the old in which j(n) is missing, i.e., replace the old J(n), j(n+1), . . . , j([L]MN-2) with j(n+1), J(n+2) . . . , j([L]MN-1) so that the new (primed) array consists of j'(n-1)=j(n-1), j'(n)=j(n+1), . . . j'([L]MN-2)=j([L]MN-1); (d) Choose a random number 0.ltoreq.n', [L]MN-1; (e) For the second row, let k=j'(n'); (f) Create a new array; and (g) Repeat steps d, e, and f until the array is empty. 2. The support structure of claim 1 wherein said integer is a number from 0 to N-1. 3. A method of substantially eliminating the artifacts generated by the support structure matrix in a transmission grating diffraction plane comprising the steps of: 4. A method according to claim 3 wherein said step of pseudo-randomly spacing said parallel supports includes selecting a pseudo-random integer k, as a number from 0 to N-1, where N is the distance from one vertical support to the next measured in terms of the number of wire spacings lying between adjacent ones of said vertical supports. 5. A method according to claim 4, wherein said step to select k is performed by |
053655664 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS FIG. 1 shows a transillumination image of an examination subject, for example legs 1. Regions in which the radiation of a radiation transmitter 6 (shown in FIG. 2) would be incident directly on a radiation receiver (not shown) i.e., unattenuated regions, are identified with reference numerals 3, 4 and 5. Since the radiation directly incident on these regions 3, 4 and 5 results in an undesirable over-drive of the radiation receiver, such as an image intensifier, a radiation diaphragm 8 constructed in accordance with the invention is arranged in the beam path of the ray beam 7 of the radiation transmitter 6 between the radiation transmitter 6 and the examination subject, as shown in FIG. 2. This radiation diaphragm 8 includes first and second diaphragm lamellae 9 and 10 that are laterally adjustable into the ray beam 7 for occluding the first and third regions 3 and 5. A third diaphragm lamella 11 is inventively composed of elastic material and is arranged between the first and second diaphragm lamellae 9 and 10. This third diaphragm lamella 11 serves the purpose of occluding the radiation that would be incident in the second region 4. In order to be able to vary the effective width of the third diaphragm lamella 11, it is adjustable around an axis 12 aligned perpendicularly relative to the central ray of the ray beam 7. Preferably, the third diaphragm lamella 11 is also adjustable in the plane perpendicular to the central ray of the ray beam 7, so that its alignment can be matched better to the examination subject. In the embodiment of FIGS. 3, 4 and 5 adjustment means in the form of holders 15 and 16 are attached at opposite end faces 13 and 14 of the third diaphragm lamella 11, the diaphragm lamella 11 being adjustable around the axis 12 by means of holders 15 and 16, One holder 16 can be stationary while the other holder 15 is adjustable around the axis 12. Preferably, however, both holders 15 and 16 are adjustable around the axis 12 in the same direction or in opposite directions. FIGS. 4 and 5 show exemplary shapes the third diaphragm lamella can assume when the holders 15 and 16 are adjusted around the axis 12. For examinations wherein it is desired that the third diaphragm lamella 11 arcuately gate a region, the holders 15 and 16 are adjustable along the axis 12 and toward one another, as shown in FIG. 6. The radiation diaphragm of the invention can include a diaphragm lamella 17 shown in FIG. 7. The lamella 17 may be used in place of the lamella 11 and is likewise elastic, and has a mushroom-shaped cut-out, so that the ray beam 7 can be gated to correspond to the head of an examination subject. In order to enable the gating of the ray beam 7 corresponding to different head shapes, adjustment elements 19a and 20a of an adjustment system engage the fourth diaphragm lamella 17 at respective sides 19 and 20, aligned perpendicularly relative to one another. As shown in FIG. 8, the radiation diaphragm of the invention can include diaphragm lamella 21 formed by a plurality of sub-lamellae 22 through 29, mounted on an adjustment element 37 so that the alignment of the sub-lamellae 22 through 29 in the ray beam 7 is individually adjustable. FIG. 9 shows a diaphragm lamella 30 of the radiation diaphragm of the invention, formed by individual lamellae 31 through 33 that are aligned parallel to one another and are mounted on an adjustment element 38 so as to be adjustable relative to one another. It is thus possible to give the diaphragm lamella 30 different shapes. The adjustment system for the diaphragm lamellae 11, 17, 21 and 30 can have an electromechanical drive, for example an electric motor that engages the diaphragm lamellae 11, 17, 21 or 30 via a belt or a lever articulation guided in bearings for the adjustment thereof. The diaphragm lamellae 11, 17, 21 and 30 are preferably composed of an elastic material, for example rubber containing lead oxide, however, they can alternatively be composed of metal sheets arranged in a scale-like pattern. In accordance with the invention, the first and second diaphragm lamella 9 and 10 can likewise be flexible, such as composed of an elastic material, and can be rotatably and/or tiltably seated in the beam path by a correspondingly executed holder. The radiation diaphragm of the invention is preferably arranged optimally close to the radiation source 7, since it thus has a small, economical structure. The radiation diaphragm, however, must have dimensions such that the diaphragm lamellae have an adequate thickness for radiation absorption. The area of the diaphragm projection can be varied by varying the spacing relative to the focus of the radiation transmitter 6. The alignment of the diaphragm lamellae in the ray beam 7 can ensue in various ways. Due to the versatility of the elastic diaphragm lamellae, a user-friendly, automatic diaphragm alignment based on the content of a video system following the image intensifier can be implemented in addition to manual setting with visible auxiliary radiation. For automatic alignment of the diaphragm lamellae, only a brief-duration drive of the radiation transmitter 6 with low energy is then required, so that the radiation stress on the examination subject is low. In accordance with the invention, the diaphragm lamella of the radiation diaphragm can be executed as an elastic filter lamella which, for example, can assume the shapes shown in FIGS. 10 through 13. The filter lamella 33 is composed of an elastic plate engaged by holders 34 and 35 at at least two sides lying opposite one another for the alignment in the ray beam 7 of the radiation transmitter 6. The holders 34 and 35 are held in the adjustment system such that the desired alignment of the filter lamella 33 can be effected. An arcuate alignment of the filter lamella 33 is achieved as shown in FIG. 10, by adjusting the holders 34 and 35 in the direction to the central ray 36 of the ray beam 7 and/or obliquely relative thereto. The alignment of the filter lamella 33 shown in FIG. 11 is achieved by adjusting the holders 34 and 35 in different planes perpendicularly aligned relative to the central ray 36, and in the direction to the central ray 36. An alignment of the filter lamella 33 according to FIG. 12 is achieved by maintaining one holder, for example, the holder 34, stationary, and adjusting the holder 35 in a plane differing from the plane of the holder 34, and which is obliquely aligned relative to the central ray 36. In the alignment of the filter lamella 33 shown in FIG. 13, the holders 34 and 35 are arranged in the same plane but are obliquely aligned relative to the central ray 36. FIGS. 10 through 13 also show a diagram in which the radiation intensity I following the filter lamella is indicated. Holders of the adjustment system can likewise be provided at further sides of the filter lamella 33 lying opposite one another Although modifications and changes may be suggested by those skilled in the art, it is the intention of the inventor to embody within the patent warranted hereon all changes and modifications as reasonably and properly come within the scope of his contribution to the art. |
abstract | A method and an X-ray apparatus generate a projective X-ray representation of an examination object. Two projective images obtained from a phase contrast measurement are adapted to each other in respect of their representation format and a result image is generated by combining the adapted images. The result image allows extensive separation of different structures in the examination object that is used. |
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claims | 1. A method of multileaf collimator (MLC) leaf positioning in tracking-based adaptive radiotherapy, comprising:a. determining a radiotherapy beam pattern in real-time during a radiotherapy treatment beam delivery, wherein said radiotherapy beam pattern comprises transforming a treatment beam plan into radiotherapy beam coordinates;b. determining a dose discrepancy between said radiotherapy beam pattern and a deliverable MLC aperture in said real-time during said radiotherapy treatment beam delivery, wherein said dose discrepancy comprises a sum of an overdose cost and an underdose cost to a treatment volume;c minimizing said dose discrepancy in said real-time during said radiotherapy treatment beam delivery according to instantaneous target motion, wherein said dose discrepancy minimization provides a determined deliverable MLC aperture for said radiotherapy beam that configures said multileaf aperture to shape a radiotherapy beam that optimally accounts for changes in a patient's anatomy measured during said radiotherapy treatment beam delivery; andd. collapsing a plan aperture f and an estimated anatomical motion T onto a beam-eye-view plane, wherein an ideal motion-compensated aperture is determined by composing said plane aperture f with said collapsed beam-eye-motion plane according to g=f°T, wherein said map is represented with a binary function over a region of interest (ROI), wherein said ROI: Ω→{0,1} so that g ( x ) = { 1 x ∈ transformed plan aperture opening ; 0 else , wherein x=(x, y) denotes a beam element location in said BEV. 2. The method of claim 1, wherein said radiotherapy beam coordinates are based on a projection of i) a translation of at least part of said radiotherapy beam pattern, ii) a rotation of at least part of said radiotherapy beam pattern, and iii) a deformation of at least part of said radiotherapy beam pattern, or i), ii) or iii). 3. The method of claim 1, wherein said overdose cost comprises an integration of pixelwise overdose penalties, wherein said overdose penalties can be spatially variant. 4. The method of claim 1, wherein said underdose cost comprises an integration of pixelwise underdose penalties, wherein said underdose penalties can be spatially variant. 5. The method of claim 1, wherein at least one previously determined said dose discrepancy is used when determining a next said deliverable MLC aperture for said radiotherapy beam. 6. The method of claim 1, wherein a determination of said overdose cost and a determination of said underdose cost are based on a tissue type and a radio-sensitivity of said tissue. |
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summary | ||
abstract | An apparatus includes a beam deflection portion for deflecting the electron beam to change an irradiation position of the electron beam; a synchronization signal generation portion for generating a synchronization signal which is in synchronization with the rotation of the substrate; a controller for controlling the beam deflection portion on the basis of the synchronization signal in order to deflect the electron beam in a rotational radial direction of the substrate and in a rotational tangential direction of the substrate opposite to a rotational direction of the substrate, while drawing transition is performed from one circle to another circle; and a beam cutoff portion for cutting off the irradiation of the electron beam on the substrate, for a period during the electron beam is deflected in the rotational radial direction. |
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047770120 | description | FIG. 1 shows a reactor pressure vessel 1, for example of prestressed concrete, with a cylindrical cavity 2, wherein a high temperature reactor 3 and a number of steam generators (not shown) are installed. The core of the high temperature reactor 3 is formed by a pile 4 of spherical fuel elements 5, which are removed by means of a pebble removal installation 6 from the pile 4. The charging installation for the fuel elements 5 is not shown. Helium as the cooling gas flows from top to bottom through the pile 4. The pile 4 is surrounded on all sides by a graphite reflector which comprises a roof reflector 7, a bottom reflector 8 serving as the supporting floor and a cylindrical side reflector 9. In the roof reflector 7, passages 10 are provided for the absorber rods (not shown). They may be isserted directly into the pile 4. The bottom reflector 8, assembled of adjacently arranged graphite columns (as seen in FIG. 2), rests on round columns 14 supported by the bottom layers 11 of the high temperature reactor 3. A thermal bottom shield 12 adjoins the bottom layers 11 as the bottom plate. The free space between the round columns 14 constitutes the hot gas collector compartment of the high temperature reactor 3; several hot gas conduits 15 are connected with the hot gas collector department. They communicate with the steam generators. After cooling and compression, the helium is returned through lines 16, installed coaxially with the hot gas lines 15 to the reactor core. The side reflector 9 is surrounded by a similarly cylindrical thermal side shield 17, whereby an annular space 18 is provided between the two structural parts in communication with the lines 16. In the annular space 18, elastic supporting elements are arranged whereby the side reflector 9 is supported on the thermal side shield 17 (not shown). The annular space 18 is connected in flow with a cold gas space 19 bounded in the downward direction by the roof reflector 7 and upwardly by the thermal roof shield 20. The pebble removal installation consists of six ceramic removal tubes 21, adjoined outside the bottom layers 11 by a metal pebble removal tube 22 each. Each ceramic pebble removal tube 21 is surrounded by a boron shield 23, as seen in more detail in FIGS. 2 and 3. FIGS. 2 and 3 show one of the ceramic pebble removal tubes 21 in their immediate surroundings. The bottom reflector 8 consists of graphite columns 8a and 8b, which in turn, are assembled of individual graphite blocks. The uppermost of these graphite blocks have a plurality of cooling gas bores 24 which are in communication with the collector compartments 25 of the adjacent graphite blocks. By means of larger bores 26, each collector compartment 25 is connected with the hot gas collector compartment 13. The ceramic pebble removal tube 21 is surrounded directly by a ring of graphite columns 8a, extending through the hot gas collector compartment 13 and resting directly on the bottom layers 11. The bottom layers 11 are assembled of graphite blocks 11a, maintained in position in the immediate vicinity of the ceramic pebble removal tube 21 by means of dowels 27. The aforementioned boron shield 23 consists of a plurality of boron rods 28 and a row of solid boron plates 29. The boron rods 28 are arranged in vertical bores adapted to the rods and extending through the entire length of the graphite columns 8a. In this embodiment, two rows a and b of bores are provided; each of them arranged on a circle with the center axis of the pebble removal tube 21 as its center. The bores of the two rows a and b are staggered with respect to each other, so that the boron rods 28 are facing gaps. The boron rods 28 fill the entire length of the bores. The solid boron plates 29 are located in the area of the bottom layers 11. They are arranged directly outside and around the ceramic pebble removal tube 21 and are shaped so that they surround the pebble removal tube 21 completely in the circumferential direction and nearly completely in the axial direction. The boron shield 23 comprising the bodies 28 and 29 prevents any output by the fuel elements 5 in the ceramic pebble removal tubes 21, whereby the temperature in the metal pebble removal tube 22 adjacent to the removal tube 21 may be kept at an acceptable level. |
summary | ||
claims | 1. A recipe parameter management system for collecting, from a review system, data including recipe parameter setting values of recipe parameters when a defect review is conducted, and managing the data, comprising:a collecting unit for collecting from the review system, as recipe parameter setting history data, data including the recipe parameter setting values set to the review system when the defect review is conducted, numbers of trial reviews carried out until the recipe parameter setting values are set when the defect review is conducted, and defect images obtained when the defect review is conducted;a storage unit for storing therein the collected recipe parameter setting history data;a histogram generating unit for generating a histogram with respect to the recipe parameter setting values based on the recipe parameter setting values contained in the recipe parameter setting history data stored in the storage unit; anda display unit for displaying, for each recipe parameter, the histogram generated by the histogram generating unit and the numbers of trial reviews. 2. The recipe parameter management system according to claim 1, wherein the display unit extracts, from the storage unit, the defect images respectively corresponding to a minimum value, a central value, and a maximum value obtained from the histogram of the recipe parameter setting values and displays the defect images. 3. The recipe parameter management system according to claim 2, further comprising a setting unit for setting current setting values of the recipe parameters,wherein the display unit extracts, from the storage unit, defect images respectively corresponding to recipe parameter setting values nearest respectively to the current setting values set by the setting unit and displays those defect images. 4. The recipe parameter management system according to claim 1, wherein the display unit sorts the histogram of the recipe parameter setting values and the numbers of the trial reviews in a descending order of the numbers of the trial reviews and displays the histogram and the numbers of the trial reviews. 5. The recipe parameter management system according to claim 1, whereinthe display unit displays a filtering condition setting screen to set a condition to conduct filtering and display, filters the histogram of the recipe parameter setting values and the numbers of the trial reviews according to the filtering condition set via the filtering condition setting screen, and displays the histogram of the recipe parameter setting values and the numbers of the trial reviews. 6. The recipe parameter management system according to claim 1, wherein the recipe parameter setting history data includes at least one data item selected from a group consisting of a process step name, a recipe name, a day and time, a parameter type, a parameter name, a parameter value, an edit frequency, and an image data. 7. A recipe parameter management method for a recipe parameter management system for collecting, from a review system, data including recipe parameter setting values of recipe parameters when a defect review is conducted, and managing the data, the method including using a computer system to perform the following steps:collecting from the review system, as recipe parameter setting history data, data including the recipe parameter setting values set to the review system when the defect review is conducted, numbers of trial reviews carried out until the recipe parameter setting values are set when the defect review is conducted, and defect images obtained when the defect review is conducted;storing, in a predetermined storage, the recipe parameter setting history data collected;generating a histogram with respect to the recipe parameter setting values based on the recipe parameter setting values contained in the recipe parameter setting history data stored in the storage; anddisplaying, for each recipe parameter, the histogram generated and the numbers of trial reviews. 8. The recipe parameter management method according to claim 7, wherein the step of displaying further comprises the step of extracting, from the storage, the defect images respectively corresponding to a minimum value, a central value, and a maximum value obtained from the histogram of the recipe parameter setting values and displaying the defect images. 9. The recipe parameter management method according to claim 8, further comprising:a step of setting current setting values of the recipe parameters,wherein the step of displaying further comprises a step of extracting, from the storage, defect images respectively corresponding to recipe parameter setting values nearest respectively to the current setting values set by the step of setting current setting values of the recipe parameters and displaying those defect images. 10. The recipe parameter management method according to claim 7, wherein the step of displaying further comprises the step of sorting the histogram of the recipe parameter setting values and the numbers of the trial reviews in a descending order of the numbers of the trial reviews and displaying the histogram and the numbers of the trial reviews. 11. The recipe parameter management method according to claim 7, whereinthe step of displaying further comprises the steps of:displaying a filtering condition setting screen to set a condition to conduct filtering and display;filtering the histogram of the recipe parameter setting values and the numbers of the trial reviews according to the filtering condition set via the filtering condition setting screen; anddisplaying the histogram of the recipe parameter setting values and the numbers of the trial reviews. 12. The recipe parameter management method according to claim 7, wherein the recipe parameter setting history data includes at least one data item selected from a group consisting of a process step name, a recipe name, day and time, a parameter type, a parameter name, a parameter value, an edit frequency, and an image data. |
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040244059 | abstract | An X-ray eye shield useful in protecting eye tissue from radiation during dental radiography. The eye shield has a radiolucent frame and radiopaque lens cups which prevent passage of X-ray radiation. |
description | This application is based upon and claims the benefit of priority from prior Japanese Patent Application No. 2006-14881 filed on Jan. 24, 2006 in Japan, the entire contents of which are incorporated herein by reference. 1. Field of the Invention The present invention relates to a pattern area value calculating method, a method of calculating a proximity effect-corrected dose, a charged particle beam writing method, and a charged particle beam writing apparatus. For example, the present invention relates to a proximity effect correcting technique (to be described below) A pattern to be written is divided into predetermined unit sections (grids) The present invention relates to a proximity effect correcting technique which corrects a dose of an electron beam to be irradiated on each unit section in consideration of accumulated energy caused by back scattering of electrons. 2. Related Art A lithography technique which leads development of micropatterning of semiconductor devices is a very important process which uniquely generates a pattern in semiconductor manufacturing processes. In recent years, with high integration of an LSI, a circuit line width required for semiconductor devices progressively decreases year after year. In order to form a desired circuit pattern on the semiconductor devices, a high-definition original pattern (also called a reticle or a mask) is necessary. In this case, an electron beam writing technique has an essentially excellent resolution and is used in production of a high-definition original pattern. FIG. 11 is a conceptual diagram for explaining an operation of a conventional variable-shaped electron beam photolithography apparatus. A variable-shaped electron beam photolithography apparatus (electron beam (EB) writing apparatus) operates as follows. In a first aperture 410, a square, for example, rectangular opening 411 to shape an electron beam 330 is formed. In a second aperture 420, a variable-shaped opening 421 to shape the electron beam 330 having passed through the opening 411 formed in the first aperture 410 in a desired square shape is formed. The electron beam 330 irradiated from a charged particle source 430 and having passed through the opening 411 is deflected by a deflector. The electron beam 330 passes through a part of the variable-shaped opening 421 and is irradiated on a target object 340 placed on a stage. The stage continuously moves in one predetermined direction (for example, defined as an X direction) while irradiating the electron beam 330. More specifically, a square shape which can pass through both the opening 411 and the variable-shaped opening 421 is written in a writing region of the target object 340 placed on the stage. A scheme which causes an electron beam to pass through both the opening 411 and the variable-shaped opening 421 to form an arbitrary shape is called a variable shaped scheme. A pattern of a semiconductor integrated circuit is written on a resist material formed on the target object 340 by using an electron beam. In this case, the electron beam used in the pattern writing passes through the resist material and is incident on the target object 340. Then, back scattering occurs. A part of the electron beam is incident on the resist material again. As a result, the resist material is exposed in an area which is considerably larger than an incident part of the electron beam not to obtain a pattern having a desired line width. When patterns to be written approximate to each other due to micropatterning to increase the density, exposure of the resist material caused by back scattering occurs in a very wide range. Since this proximity effect is caused, correction must be performed. In general, when a pattern is written on the resist material on the substrate, a pattern to be written is divided into predetermined unit sections (to be referred to as grids or meshes). At the center of each unit section, accumulated energy caused by back scattering is calculated on the basis of an EID function. In consideration of the accumulated energy, a dose of an electron beam to be irradiated on each unit section is corrected. In relation to the proximity effect correction, a technique which calculates accumulated energy on the basis of an EID function by replacing the center point of the unit section with an area gravity point is disclosed in a reference (for example, see JP-A-9-186058). As described above, in the calculation of the proximity effect correction, a pattern to be written is divided into predetermined meshes, and accumulated energy caused by back scattering is calculated on the basis of an EID function at a center position of each mesh. However, an area value included in a mesh is regarded to be concentrated on the center of the mesh to estimate a back scattering energy distribution. For this reason, the position is different from an arrangement position of an actual pattern. As a consequence, the back scattering energy distribution has an error. Since the back scattering energy distribution with the error is used in calculation of a beam dose on the pattern, the error adversely affects the calculation. Therefore, the beam dose to be calculated also has an error. The conventional technique has the above problems. At the present or in the future, with an increase in degree of integration density of an LSI, highly accurate proximity effect correction is required. In this circumstance, an error caused by indetermination of the area position is a factor which decreases correction accuracy near a figure. The present invention has as its object to reduce an error of a back scattering energy distribution. In accordance with embodiment consistent with the present invention, there is provided a method for calculating area values of a pattern written by using a charged particle beam, including virtually dividing a pattern into a plurality of mesh-like first square regions surrounded by first grids defined at intervals of a predetermined size, virtually dividing the pattern into a plurality of mesh-like second square regions surrounded by second grids defined at intervals of the predetermined size, wherein the second grids being positionally deviated from the first grids by a half of the predetermined size, distributing an area value of a sub-pattern in each of the second square regions to a plurality of apexes of each of the second square regions such that a center-of-gravity position of the sub-pattern does not change, wherein the sub-pattern being a part of the pattern, and outputting the distributed area values as area values, for correcting a proximity effect, defined at the center position of each of the first square regions. Also, in accordance with embodiment consistent with the present invention, there is provided a proximity effect correcting method including virtually dividing a pattern which is written by using a charged particle beam into a plurality of mesh-like square regions surrounded by grids defined at intervals of a predetermined size, distributing a part of area value of a sub-pattern in each of the square regions to a center position of another square region such that a center-of-gravity position of the sub-pattern does not change, wherein the part of area value being defined by the center position of the another square region and the sub-pattern being a part of the pattern, and calculating an amount of proximity effect correction in each square region by use of an area value of each square region obtained by adding remaining area value which is not distributed to other square region and area value distributed from other square region to output the amount of proximity effect correction. Further, in accordance with embodiment consistent with the present invention, there is provided a method for writing a pattern using a charged particle beam, the method including, virtually dividing a pattern into a plurality of mesh-like square regions surrounded by grids defined at intervals of a predetermined size, distributing an area value of a sub-pattern in each of the square regions to positions where the distributed area values are defined by a center position of the square region and a center position of other square region, such that a center-of-gravity position of the sub-pattern in each of the square regions does not change, after the area values are distributed, calculating an exposure dose of the charged particle beam corrected with respect to proximity effect by using the area values defined by the center positions of the square regions, and writing the pattern on a target object at the exposure dose. Additionally, in accordance with embodiment consistent with the present invention, there is provided a charged particle beam writing apparatus for writing a pattern using a charged particle beam, including a dividing unit configured to virtually divide a pattern into a plurality of mesh-like first square regions surrounded by first grids defined at intervals of a predetermined size and a plurality of mesh-like second square regions surrounded by second grids defined at intervals of the predetermined size, wherein the second grids being positionally deviated from the first grids by a half of the predetermined size, a distributing unit configured to distribute an area value of a sub-pattern in each of the second square region to a plurality of apexes of each of the second square regions such that a center-of-gravity position of the sub-pattern in each of the second square region does not change, wherein the sub-pattern being a part of the pattern, a calculating unit configured to calculate an amount of proximity effect correction for correcting proximity effect in each of the first square regions by using area values distributed and a pattern writing unit configured to write the pattern on a target object at an exposure dose of the charged particle beam corrected with respect to proximity effect by using the amount of proximity effect correction. In respective embodiment, a configuration using an electron beam will be described below as an example of a charged particle beam. The charged particle beam is not limited to an electron beam. A beam such as an ion beam using other charged particles may be used. FIG. 1 is a diagram showing a main part of a flow chart in a first embodiment. In FIG. 1, a pattern area value calculating method performs a series of steps such as a check and mesh size calculating step (S102), a figure dividing mesh virtual dividing step (S104), a figure coordinate and figure size loading step (S106), a figure code loading step (S108), a mesh unit system converting step (S110), a figure dividing step (S112), an area, center of gravity, and moment calculating step (S114), an area value distributing step (S116), and an area value adding step (S118). In a proximity effect correcting method, a series of steps such as an area ratio calculating step (S120) and a beam dose calculating step (S122) is performed by using an area value obtained by the pattern area value calculating method. A charged particle beam writing method performs a pattern writing step (S124) by using a beam dose subjected to the proximity effect correction. FIG. 2 is a conceptual diagram showing an example of a main configuration of a writing apparatus according to the first embodiment. In FIG. 2, an electron beam writing apparatus is given as an example of a charged particle beam writing apparatus. A writing apparatus 100 serving as an example of an electron beam writing apparatus includes a pattern writing unit 150 and a control system. The writing apparatus 100 writes or “draws” a pattern onto a target object. The target object 101 includes a mask. The pattern writing unit 150 is equipped with an electron lens barrel 102 and a writing chamber 103. An electron gun assembly 201, a blanking (BLK) deflector 212, and a blanking (BLK) aperture 214 are arranged in the electron lens barrel 102. In the writing chamber 103, an XY stage 105 is arranged. The control system includes a deflecting amplifier 110, a deflection control circuit 112, a writing data generating circuit 120, and a stage control circuit 142. Arranged in the writing data generating circuit 120 are a proximity effect correcting unit 122, a shot data calculating unit 124, a shot data developing unit 126, an area processing calculating unit 130, and a magnetic disk device 128 serving as an example of a data storing device. The area processing calculating unit 130 has functions such as a dividing unit 132, an area calculating unit 134, a center-of-gravity calculating unit 136, a moment calculating unit 138, and a distributing unit 140. The magnetic disk device 128 has parameter data stored therein. The pattern data is input to the area processing calculating unit 130 from the magnetic disk device 128. Similarly, the pattern data is input from the magnetic disk device 128 to the shot data developing unit 126. In FIG. 2, constituent parts required to explain the first embodiment are described. For the writing apparatus 100, other necessary configurations are included as a matter of course. In the writing data generating circuit 120, all or some of the other parts than the magnetic disk device 128 may be constituted by a CPU serving as an example of a computer. In this case, the CPU executes the processes of the respective functions such as the proximity effect correcting unit 122, the shot data calculating unit 124, the shot data developing unit 126, and the area processing calculating unit 130. Alternatively, except for the proximity effect correcting unit 122, the shot data calculating unit 124, and the shot data developing unit 126, the area processing calculating unit 130 may be constituted by a CPU serving as an example of a computer. In this case, the CPU executes the processes of the functions such as the dividing unit 132, the area calculating unit 134, the center-of-gravity calculating unit 136, the moment calculating unit 138, and the distributing unit 140. However, the invention is not limited to the above configurations. All or some of the writing data generating circuit 120, the proximity effect correcting unit 122, the shot data calculating unit 124, the shot data developing unit 126, the area processing calculating unit 130, the dividing unit 132, the area calculating unit 134, the center-of-gravity calculating unit 136, the moment calculating unit 138, and the distributing unit 140 may be realized by hardware constituted by electric circuits. Alternatively, these units may be realized by a combination of hardware constituted by electric circuits and software, or may be realized by a combination of the hardware and firmware. The shot data calculating unit 124 in the writing data generating circuit 120 is connected to the deflection control circuit 112 through a bus (not shown). The proximity effect correcting unit 122 and the shot data developing unit 126 are connected to the shot data calculating unit 124 through a bus (not shown). The area processing calculating unit 130 is connected to the proximity effect correcting unit 122 through a bus (not shown). The stage control circuit 142 is connected to the shot data developing unit 126 through a bus (not shown). An electron beam 200 emitted from the electron gun assembly 201 is irradiated on a desired position of a target object 101 on the XY stage 105. The XY stage 105 is movably arranged. The XY stage 105 moves under the control of the stage control circuit 142. The electron beam 200 serves as an example of a charged particle beam. The stage control circuit 142 receives a shot density from the shot data developing unit 126 to calculate a stage speed of the XY stage 105 on the basis of the shot density. In this case, the electron beam 200 on the target object 101 is prevented from reaching the upper surface of the target object 101 when it is beam irradiation time at which the electron beam of a desired dose is incident on the target object 101. This is intended to prevent the electron beam 200 from being excessively irradiated on the target object 101. For example, the electron beam 200 is deflected by an electrostatic BLK deflector 212. The BLK aperture 214 cuts the electron beam 200. In this manner, the electron beam 200 is prevented from reaching the upper surface of the target object 101. A deflecting voltage of the BLK deflector 212 is controlled by the deflection control circuit 112 and the deflecting amplifier 110. In a beam-on (blanking-off) state, the electron beam 200 emitted from the electron gun assembly 201 travels a path indicated by a solid line in FIG. 2. In a beam-off (blanking-on) state, on the other hand, the electron beam 200 emitted from the electron gun assembly 201 travels on a path indicated by a dotted line in FIG. 2. The insides of the electron lens barrel 102 and the writing chamber 103 in which the XY stage 105 is arranged are evacuated by a vacuum pump (not shown), and a vacuum atmosphere has a pressure lower than the atmospheric pressure. In FIG. 2, constituent parts required to explain the first embodiment are described. However, the writing apparatus 100 may include, in addition to the configuration described above, the following configuration. That is, an illumination lens, a first aperture, a projection lens, a shaping deflector, a second aperture, an objective lens, an objective deflector, and the like may be arranged in the electron lens barrel 102. In a beam-on (blanking-off) state, the configuration is made such that the electron beam 200 emitted from the electron gun assembly 201 entirely illuminates a first aperture having a square, for example, rectangular opening through an illumination lens. First, the electron beam 200 is shaped in a square, for example, rectangular shape. The electron beam 200 of a first aperture image having passed through the first aperture is projected on a second aperture by a projection lens. A position of the first aperture image on the second aperture is controlled by a shaping deflector. In this manner, the beam shape and the beam size can be changed. The electron beam 200 of the second aperture image having passed through the second aperture is focused by an objective lens. The electron beam 200 is deflected by an objective deflector and irradiated on a desired position of the target object 101 on the X-Y stage 105. At this time, the XY stage 105 moves. With the configuration, a variable-shaped EB photolithography apparatus can be obtained. In step S102, as a check and mesh size calculating step, the writing data generating circuit 120 checks initial values such as mesh parameters N and m. In a proximity effect correcting process, a writing pattern which is written by using the electron beam 200 is divided into predetermined unit sections (to be referred to as grids or meshes) At a center position of each unit section, and accumulated energy caused by back scattering is calculated on the basis of an EID function. FIG. 3 is a diagram showing a part of a writing pattern divided like a mesh in the first embodiment. In FIG. 3, by way of example, a figure 40 and a figure 50 to be written in a writing pattern 10 are defined. The pattern 10 is virtually divided into a plurality of area meshes (first square regions) for a proximity effect correcting process. Each area mesh is defined as a region surrounded by area mesh grids 20 (solid line) written at intervals of a predetermined size. The area mesh is set at 2m/N (AU) expressed by using a minimum unit (to be referred to as an AU unit system) obtained when coordinates of a pattern for writing and a figure size are expressed as integer values as a predetermined size (mesh size) In general, several nm to several Å are often set per 1 AU. When 2m/N is set, division (will be necessary later) can be replaced with bit shift of an integer value. As a result, an amount of calculation can be reduced. For example, the values are given by 12≦m≦15 and 1≦N≦7. In the area value calculating method according to the first embodiment, by a method (to be described later), an area value of a pattern in area meshes defined by center positions of the area meshes is calculated. In the proximity effect correcting method, the area values defined at the center positions are used in proximity effect correction for electron beam writing. The writing data generating circuit 120 checks initial values such as mesh parameters N and m. The value 2m/N is calculated as a mesh size. In S104, as a figure dividing mesh virtual dividing step serving as an example of a virtual dividing step, the dividing unit 132 virtually divides the writing pattern 10 into a plurality of figure dividing meshes (second square regions). The figure dividing meshes are meshes having equal mesh sizes and obtained by deviating the area meshes with respect to mesh original positions in an x direction and a y direction by a half of a mesh size (½ mesh). Each figure dividing mesh, as shown in FIG. 3, is defined as a region surrounded by figure dividing mesh grids 30 (dot line) written at intervals of a mesh size equal to that of the area mesh. The mesh is defined by deviating the mesh original position by the ½ mesh to make it possible to set an intersecting point of the figure dividing mesh grids 30 as a center position of each area mesh. The intersecting point of the figure dividing mesh grids 30 is an apex of each figure dividing mesh. Therefore, the position of the apex of each figure dividing mesh can be set as the center position of each area mesh. In S106, as a figure coordinate and figure size loading step, the area processing calculating unit 130 loads pattern data from the magnetic disk device 128. The area processing calculating unit 130 loads figure coordinates and a figure size of the pattern 10 defined by the pattern data. In S108, as a figure code loading step, the area processing calculating unit 130 loads a figure code defined by the loaded figure coordinates. In FIG. 3, as an example, the figure 40 and the figure 50 which are based on the figure coordinates, the figure size, and the figure code are described. In S110, as a mesh unit system converting step, the area processing calculating unit 130 converts the coordinates and the figure size of the pattern 10 from the AU unit system into a mesh unit system. A conversion formula may be given by the following formula in which the values are divided by the mesh size:Conversion Formula: (coordinate, length) [mesh]=(coordinate, length) [AU]×N/2m In S112, as a figure dividing step, the dividing unit 132 divides the figure on a boundary between the figure dividing meshes. FIG. 4 is a diagram showing an example of a figure divided in figure dividing meshes according to the first embodiment. FIG. 4 shows, by way of example, a case in which the figure 40 is divided in a figure dividing mesh 32 and a figure dividing mesh 34. The figure 40 is a sub-pattern which is a part of pattern 10. As a result, in the figure dividing mesh 32, a figure 42 having an area S′ and serving as a part of the figure 40 having an area S is divided. Also the figure 42 is a sub-pattern which is apart of pattern 10. In the figure dividing mesh 34, a figure 44 having an area S″ and serving as a part of the figure 40 is divided. Also the figure 44 is a sub-pattern which is a part of pattern 10. As a matter of course, S=S′+S″ is satisfied. Division of the figure 50 will be omitted in the drawing and explanation. In S114, as an area, center of gravity, and moment calculating step, the area processing calculating unit 130 calculates the area, the center-of-gravity position, and the center-of-gravity moment of a figure on the basis of lengths of sides of the figures and figure coordinates. As an area value calculating step, the area calculating unit 134 calculates an area value of a figure on the basis of the lengths of the sides of the figures. FIG. 5 is a diagram showing an example of a figure dividing mesh obtained by dividing the figure in the first embodiment. FIG. 5 shows an appearance of the figure dividing mesh 32 as an example. When lengths of sides of the figure 42 divided in the figure dividing mesh 32 are L1 in the x direction and L2 in the y direction, the area S′ of the figure 42 can be calculated by multiplying L1 by L2. As a center-of-gravity position calculating step, the center-of-gravity calculating unit 136 calculates a center-of-gravity position of a figure on the basis of the lengths of the sides of the figures and the figure coordinates. In the example in FIG. 5, a lower left corner of the figure dividing mesh is set at an original point (0, 0). Coordinates of a figure original point of the figure 42 is given by (X1, y1). In this case, center-of-gravity position coordinates (gx1, gy1) of the figure 42 can be calculated by the following equation. Note that it is assumed that the following coordinate system is converted into a mesh unit system.gx1=x1+L1/2, gy1=y1+L2/2 As a center-of-gravity moment calculating step, the moment calculating unit 138 calculates a center-of-gravity moment of a figure on the basis of an area value of the figure and a center-of-gravity position of the figure. In the example in FIG. 5, a center-of-gravity moment (S′gx1, S′gy1) of the figure 42 can be calculated by the following equation:S′gx1=S′×gx1, S′gy1=S′×gy1 The steps S106 to S114 described above are looped with respect to all figures (repeated). In S116, as an area value dispersing step serving as a part of an area value distributing step, the distributing unit 140 distributes area values of patterns of each figure dividing mesh to a plurality of apexes of the figure dividing mesh. At this time, the area values are distributed to the apexes such that the center-of-gravity positions of the patterns in each figure dividing mesh do not change. The distributing unit 140 performs distribution such that area values of figures in each figure dividing mesh are distributed (dispersed) to a plurality of apexes of the figure dividing meshes by using the area values of the figures and the center-of-gravity moment of the figures in each figure dividing mesh. More specifically, the distribution is performed such that the area values are distributed (dispersed) to intersecting points of the figure dividing mesh grids 30. In other words, the area values of the patterns in each area mesh are distributed such that the area values are defined by the apexes of the figure dividing mesh at a center position of a certain area mesh and apexes of a figure dividing mesh at a center position of another area mesh. The area values are distributed such that center-of-gravity positions of the patterns in the area meshes are equal to each other. More specifically, some area values of the figures in the area meshes are distributed such that the center-of-gravity positions of the patterns are defined by center positions of another plurality of area meshes to be equal to each other. FIG. 6 is a diagram showing an example of distributed area values in the first embodiment. FIG. 6 shows, as an example, an appearance of the figure dividing mesh 32. Area values S′ of the figure 42 divided in the figure dividing mesh 32 are dispersed to four apexes (1 to 4) of the figure dividing mesh 32 as area values S′1 to S′4 and distributed. The area values dispersed are expressed by the following equations, respectively:S′1=S′−S′2−S′3−S′4 S′2=(S′gx1−S′gy1)/2+S′/4S′3=(S′gy1−S′gx1)/2+S′/4S′4=(S′gx1+S′gy1)/2−S′/4 According to these equations, area center-of-gravity position coordinates obtained when the area values S′1 to S′4 defined at the four apexes of the figure dividing mesh 32 can be made equal to center-of-gravity position coordinates (gx1, gy1) of the figure 42. An equation: S′=S′1+S′2+S′3+S′4 is satisfied as a matter of course. When a center-of-gravity moment is calculated such that a lower left corner of the figure dividing mesh is set as an original point (0, 0), a figure dividing mesh size expressed in the mesh unit system is 1. Therefore, a center-of-gravity moment of a sum of the areas arranged at the four apexes is given by:(1×S′2+1×S′4, 1×S′3+1×S′4).Therefore, it is understood that, when the equations S′1 to S′4 are assigned to the above equation, the resultant value is equal to the center-of-gravity moment calculated in FIG. 5. In S118, as an area value adding step serving as a part of the area value distributing step, the distributing unit 140 cumulatively adds area values of patterns in another figure dividing mesh when the area values are distributed to any one of the apexes of the corresponding figure dividing mesh. In the example in FIG. 4, a figure 44 serving as a part of the figure 40 is divided in the figure dividing mesh 32 and the figure dividing mesh 34. For this reason, of the area values of the figure 44 divided in the figure dividing mesh 34, area values distributed to the apexes of the figure dividing mesh 32 are added to each other, i.e., cumulatively added to each other. Distribution of the area values of the figure 44 divided in the figure dividing mesh 34 will be described later. The steps S116 to S118 described above are looped with respect to all figure dividing meshes (repeated). FIG. 7 is a diagram showing another example of the figure dividing mesh in which the figure in the first embodiment is divided. FIG. 7 shows, as an example, an appearance of the figure dividing mesh 34. As described above, as the area value calculating step, the area calculating unit 134 calculates an area value of a figure on the basis of lengths of sides of each figure. When the lengths of the sides of the figure 44 divided in the figure dividing mesh 34 are given by L1 in the x direction and L2 in the y direction, the area S″ of the figure 44 can be given by S″=L1×L2. As a center-of-gravity position calculating step, the center-of-gravity calculating unit 136 calculates a center-of-gravity position of a figure on the basis of lengths of sides of each figure and figure coordinates. In the example in FIG. 7, when coordinates of a figure original point of the figure 44 are given by (x2, y2), center-of-gravity position coordinates (gx2, gy2) of the figure 44 can be calculated by the following equation:gx2=x2+L1/2, gy2=y2+L2/2 As a center-of-gravity moment calculating step, the moment calculating unit 138 calculates a center-of-gravity moment of a figure on the basis of an area value of a figure and a center-of-gravity position of the figure. In the example in FIG. 7, a center-of-gravity moment (S″gx2, S″gy2) of the figure 44 can be calculated by the following equation:S″gy2=S″×gx2, S″gy2=S″×gy2 As an area value dispersing step (S116) serving as a part of the area value distributing step, the distributing unit 140 distributes area value of pattern in each figure dividing mesh to a plurality of apexes of the figure dividing mesh such that a center-of-gravity position of the pattern in the figure dividing mesh is not changed. The distributing unit 140 performs distribution such that the area value of the figures in the figure dividing mesh is distributed (dispersed) by using the area value of the figure in the figure dividing mesh and the center-of-gravity moments of the figure. In the distribution, as described above, the area value are distributed (dispersed) to a plurality of apexes of the figure dividing mesh, i.e., intersecting points of the figure dividing mesh grids 30. FIG. 8 is a diagram showing another example of the distributed area value in the first embodiment. FIG. 8 shows, as an example, an appearance of the figure dividing mesh 34. Area value S″ of the figure 44 divided in the figure dividing mesh 34 are dispersed to the four apexes (1 to 4) of the figure dividing mesh 34 as area values S″1 to S″4 and distributed. The area values dispersed are expressed by the following equations, respectively:S″1=S″−S″2−S″3−S″4 S″2=(S″gx2−S″gy2)/2+S″/4S″3=(S″gy2−S″gx2)/2+S″/4S″4=(S″gx2+S″gy2)/2−S″/4 According to these equations, area center-of-gravity position coordinates obtained when the area values S″1 to S″4 defined at the four apexes of the figure dividing mesh 34 can be made equal to center-of-gravity position coordinates (gx2, gy2) of the figure 44. An equation: S″=S″1+S″2+S″3+S″4 is satisfied as a matter of course. In this case, apex 1 of the four apexes of the figure dividing mesh 32 is also apex 3 of the figure dividing mesh 34. Similarly, apex 2 of the four apexes of the figure dividing mesh 32 is also apex 4 of the figure dividing mesh 34. Therefore, as an area value adding step (S118) serving as a part of the above-described area value distributing step, the distributing unit 140 cumulatively adds S″3 to dispersed S′1 with respect to apex 1 of the four apexes of the figure dividing mesh 32. Similarly, the distributing unit 140 cumulatively adds S″4 to dispersed S′2 with respect to apex 2 of the four apexes of the figure dividing mesh 32. FIGS. 9A and 9B are diagrams showing an example of distributed area values and a back scattering energy distribution in the first embodiment. In FIG. 9A, appearances of the figure dividing mesh 32 and the figure dividing mesh 34 are shown as an example. An area value S2 at apex 2 in FIG. 9A is a sum of an area value S′1 and an area value S″3. Similarly, an area value S5 at apex 5 in FIG. 9A is a sum of an area value S′2 and an area value S″4. When the area values at the respective apexes are cumulatively added to each other, area center-of-gravity position coordinates obtained when the area values S1 to S6 are synthesized with each other can be made equal to the center-of-gravity position coordinates (gx, gy) of the original figure 40 before the figure is not divided. The area values S1 to S6 are equal to area values defined with respect to six apexes of the figure dividing mesh 32 and the figure dividing mesh 34. The area value S of the figure 40 is given by S=S1+S2+S3+S4 as a matter of course. By using the area values (S1 to S6) in area meshes defined at the apexes of the figure dividing meshes, i.e., center positions of the area meshes, a back scattering energy distribution is calculated. In this case, as shown in FIG. 9B, a position of the calculated back scattering energy distribution can be made equal to a back scattering energy distribution of the figure 40. As described above, in the first embodiment, some area values of the patterns in each area mesh are distributed such that center-of-gravity positions of the patterns in the area mesh are defined by center positions of another plurality of area meshes to be equal to each other. In other words, the area values of the patterns in each figure dividing mesh are distributed to the plurality of apexes of the figure dividing mesh such that the center-of-gravity positions of the patterns in the figure dividing mesh do not change. In this manner, area values at the center positions of each area mesh can be obtained. Since the area values are distributed such that the center-of-gravity positions of the patterns in the figure dividing mesh do not change, the center-of-gravity positions obtained when the area values at the plurality of apexes of each figure dividing mesh are synchronized with each other do not change. As a result, the center-of-gravity positions obtained when the area values at the center positions of each area mesh are synchronized with each other also do not change. Accordingly, when a back scattering energy distribution is calculated by using the area values of the patterns in each area mesh defined at the center positions of the area mesh, deviations from the arrangement positions of the actual patterns can be canceled. For this reason, uncertainty of the area positions can be reduced. As a result, a position of the calculated back scattering energy distribution and a back scattering energy distribution of an actual figure can be made equal to each other, or an error of the back scattering energy distribution can be reduced. FIGS. 10A and 10B are diagrams showing an example of an appearance observed when area values of figures in an area mesh are not dispersed and a back scattering energy distribution. Description will be given with respect to a case in which area values of figures in an area mesh are not distributed in the same arrangement of figures as the arrangement of the figures shown in FIG. 3. As shown in FIG. 10A, when area values of the figure 40 in an area mesh 22 are not distributed, the center position of the area pattern 22 deviates from the center-of-gravity position coordinates (gx, gy) of the figure 40. For this reason, when the area values of the figure 40 are defined at the center position of the area mesh 22, the position of the back scattering energy distribution deviates from the arrangement position of the figure 40 if a back scattering energy distribution is calculated by using the area values. As a result, as shown in FIG. 10B, an error C is generated between the position of the calculated back scattering energy distribution and the back scattering energy distribution of the actual figure. An error D corresponding to the position of the figure 50 adversely affects writing of the figure 50. Therefore, as described in the first embodiment, the position of the calculated back scattering energy distribution is made equal to the back scattering energy distribution of an actual figure, or an error of the back scattering energy distribution is reduced to make it possible to eliminate or reduce the adverse affection. The distributing unit 140 outputs the area values to the proximity effect correcting unit 122. In S120, as an area ratio calculating step, the proximity effect correcting unit 122 receives area values held at grid intersecting points of the figure dividing mesh obtained by the pattern area value calculating method and calculates an area ratio in each area mesh. In S122, as an electron beam dose calculating step, the proximity effect correcting unit 122 calculates an amount of proximity effect correction in each area mesh depending on an area ratio in area meshes calculated by using area values held at grid intersecting points of the figure dividing meshes, i.e., the centers of the area meshes. The proximity effect correcting unit 122 outputs the amount of proximity effect correction to the shot data calculating unit 124. The shot data calculating unit 124 receives the amount of proximity effect correction from the proximity effect correcting unit 122. The shot data calculating unit 124 receives shot data developed by the shot data developing unit 126. The shot data calculating unit 124 calculates an exposure dose of electron beam obtained by performing proximity effect correction to the shot data. The back scattering energy distribution shown in FIG. 9B may be calculated such that an accumulated energy E generated by back scattering is calculated on the basis of an EID function. In consideration of the accumulated energy E, the dose of electron beam to be irradiated on each area mesh may be corrected to calculate an exposure of electron beam corrected with respect to proximity effect. In S124, as a pattern writing step, the pattern writing unit 150 writes the pattern 10 onto the target object 101 at the exposure dose of electron beam. The shot data calculating unit 124 outputs a signal to the deflection control circuit 112 such that the calculated dose of electron beam corrected with respect to the proximity effect is obtained. The deflection control circuit 112 irradiates (beam-on) the electron beam 200 on the target object 101 at the dose of electron beam corrected with respect to proximity effect through the deflecting amplifier 110. When it is beam irradiation time at which the dose of electron beam is obtained, a voltage is applied to the BLK deflector 212 such that the electron beam 200 collides with a plane of the BLK aperture 214 to deflect the electron beam 200 (beam-off). As described above, deviation from the arrangement position of the actual pattern can be eliminated. As a consequence, proximity effect correction in which a position of a calculated back scattering energy distribution is made equal to the back scattering energy distribution of an actual figure or an error of the back scattering energy distribution is reduced can be achieved. Therefore, a more accurate pattern can be written. The embodiment is described above with reference to the concrete examples. However, the present invention is not limited to the concrete examples. Parts such as an apparatus configuration or a control method which are not directly required to explain the present invention are omitted. However, a necessary apparatus configuration and a necessary control method can be appropriately selected and used. For example, although a control unit configuration for controlling the writing apparatus 100 is omitted, a necessary control unit configuration is appropriately selected and used, as a matter of course. All pattern area value calculating methods, proximity effect correcting methods, charged particle beam writing apparatuses, charge particle beam writing methods which include the elements of the present invention and which can be appropriately changed in design by a person skilled in the art are included in the spirit and scope of the invention. Additional advantages and modification will readily occur to those skilled in the art. Therefore, the invention in its broader aspects is not limited to the specific details and representative embodiments shown and described herein. Accordingly, various modifications may be made without departing from the spirit or scope of the general inventive concept as defined by the appended claims and their equivalents. |
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050009079 | summary | BACKGROUND OF THE INVENTION The present invention relates to pressurized water nuclear reactors having a water injection device for delivering a flow of emergency water stored in an external reservoir to the reactor vessel for temporarily compensating the egress of water upon failure of a pipe in the primary cooling circuit. PRIOR ART Reactors provided with such injection devices (FR-A-1 597 057) have been known for long. One of the defects of most prior art devices is that injection of cold emergency water, stored at atmospheric pressure in an external reservoir, causes a thermal shock when it enters the reactor vessel at a high operating temperature. This particularly detrimentally affects the portion of the ring of the vessel situated at the horizontal level of the core for, due to the high level of irradiation received by this portion, its mechanical characteristics may have decreased. It is necessary that the vessel resists the residual internal pressure upon a rupture of the primary circuit and occurrence of emergency injection. In present day reactors, attempts have been made to solve the problem by injecting water at one or more locations of the circuit which are as remote as possible from the vessel wall, so that the cold water mixes with hot water still present in the circuit before it contacts the side wall of the vessel. Another pressurized water nuclear reactor has been proposed in FR-A-2 314 919, which comprises a vessel which is closed by a cover, has cooling water input and output nozzles, and is connected to at least one conduit for the injection of pressurized water from an emergency reservoir; the duct opens into the vessel at the same level as or above the level of all nozzles; the reactor further comprises internals suspended from the vessel, having a core support plate and forming with the side wall of the vessel an annular downward water flow passage from the input or each inlet nozzle to a distribution space situated below the core support plate; a duct extending the conduit downward is located within the dividing structure (baffles and formers) of the internals equipment. In such a reactor, the injection function of the conduit is secondary: its main purpose is to reintroduce water delivered by a water recirculation circuit connected to the outlet nozzle. The duct is used for emergency injection based on the natural idea that, since the duct plays its function only during normal operation of the reactor, it is available for emergency injection should an accident occur. However, since the duct opens above the core support plate, the water reaches the fuel assemblies before it is appreciably heated, and may damage the fuel assemblies and particularly causes fuel rod sheath failure. A nuclear reactor is also known having conduits passing through the vessel and used solely for injecting emergency cooling water, whose output into the vessel has deflectors for directing the jet of incoming water toward the bottom of the vessel (FR-A-2 286 478). This solution does not overcome the thermal shock problem on the side ring of the vessel. On the contrary, it may be enhanced since the deflectors direct the stream of cold water along the vessel, directly toward the most sensitive portion thereof. SUMMARY OF THE INVENTION It is an object of the present invention to provide a nuclear reactor of the above-defined type which simultaneously allows injection of emergency cooling water directly into the vessel, consequently achieving maximum efficiency, and avoids causing a thermal shock on sensitive parts, such as the side ring of the vessel at the horizontal level of the core and the fuel assemblies. To this end, there is provided a reactor in which the duct is isolated thermally from the side wall of the vessel, at least at the same horizontal level as the core, and opens in the distribution space below the core support plate where the temperature of the cold water injected is rapidly increased by mixing with a large amount of water. In a particular embodiment, each duct is fixed rigidly to a casing belonging to the internal equipment and comprises end sealing means for substantially water-tight contact abutment against an outlet of the corresponding conduit; in a modification, each duct is fixed to the side wall of the vessel by means which hold it at a distance therefrom. |
description | This is the National Stage of PCT international application PCT/FR2019/050117, filed on Jan. 21, 2019 entitled “STORAGE BASKET FOR RADIOACTIVE MATERIALS, HAVING AN OPTIMISED SPACE REQUIREMENT AND HOUSINGS WITH MORE ACCURATE GEOMETRY”, which claims the priority of French Patent Application No. 1850638 filed Jan. 26, 2018, both of which are incorporated herein by reference in their entirety. The present invention relates to the field of transport and/or warehousing of radioactive materials. For example these can be assemblies of fresh or irradiated fuels, or radioactive waste. A basket, also called a storage “device” or “rack”, comprises a plurality of adjacent housings each able to receive radioactive materials. It is housed into the containment enclosure of a packaging, and designed to be able to simultaneously fulfil three essential functions, which will be briefly set out below. Indeed, there is first the function of thermal transfer of heat released by radioactive materials. Generally, aluminium or one of its alloys, due to its good thermal conduction properties, is used. The second function relates to neutron absorption, and the issue of maintaining the sub-criticality of the storage basket when the latter is loaded with fissile radioactive materials. This is made by using neutron absorbing materials, such as boron. Finally, the third essential function is concerned with the mechanical strength of the device. It is noted that the overall mechanical strength of the basket has to be compatible with regulatory safety requirements for transporting/warehousing radioactive materials, especially regarding tests called “free drop” tests. Many technical solutions are already known to make such baskets. For example, it has been contemplated to provide tubes arranged in parallel and each forming a housing for receiving radioactive materials. In this case, for example disclosed in document EP 1 212 755 A1, tubes are held to each other by transverse plates spaced apart from each other along the longitudinal central axis of the basket. To do so, each of these plates has tubes passing therethrough at through holes. However, a minimum material thickness has to be kept between two directly consecutive through holes within a same plate, in order to fulfil the aforesaid mechanical strength function. It generates a substantial transverse overall size for the basket. Furthermore, tubes generally extend over the whole height of the basket. It makes it difficult to obtain an accurate geometry of the housing, over the whole length of the latter. It is particularly complicated and expensive to reach the desired straightness over the whole length of the housing. The purpose of the invention is therefore to at least partially overcome the abovementioned drawbacks, relating to embodiments of prior art. To do so, one object of the invention is a storage basket for radioactive materials, the basket being to be arranged into a containment enclosure of a packaging for transporting and/or warehousing radioactive materials, the basket defining a plurality of housings each for receiving radioactive materials, the housings being parallel to each other and each extending along a housing axis parallel to a longitudinal central axis of the basket, the latter including: a transverse plate or a plurality of transverse plates distributed along the longitudinal central axis of the basket and arranged orthogonally to this axis, each plate having a plurality of holes passing therethrough; a plurality of housing tubes arranged parallel to the longitudinal central axis of the basket. According to the invention, the housing tubes are arranged alternately with the transverse plate(s) along the longitudinal central axis, so that the inner side surface of each housing is defined, successively along this axis, at least by the inner surface of a first housing tube, the inner surface of one of the holes of a first transverse plate, and the inner surface of a second housing tube. By means of this segmented design of housings along the longitudinal direction of the basket, it becomes easier to obtain the desired straightness over the whole length of these housings. Furthermore, since the holes of the transverse plates now form an integral part of the housings, these holes can thus be brought closer to each other while keeping the minimum material thickness required to fulfil the mechanical strength function. As a result the transverse overall size of the basket is advantageously decreased. The invention furthermore has at least any of the following optional characteristics, taken alone or in combination. The transverse plate(s) each comprise, at both their opposite faces, means for holding the housing tubes. These holding means preferably take the shape of recesses into which the ends of the housing tubes are inserted. The inner side surface of each housing has a circular, square, rectangular or hexagonal-shaped cross-section, or any other shape deemed to be appropriate by those skilled in the art. The transverse plate(s) each have a disc shape. Each housing tube is made of a steel preferably devoid of neutron absorbing elements, and each housing tube forms an inner tube surrounded by an outer tube made of an aluminium alloy preferably comprising neutron absorbing elements such as boron. Each inner tube axially projects from each of both opposite ends of the outer tube. The transverse plate(s) is (are) made of steel. The basket also includes a top plate and a bottom plate sandwiching between each other the alternating transverse plate(s) and housing tubes. The basket also includes tie rods each passing through the top plate, the bottom plate, as well as the transverse plate(s). Each housing is defined using a number N of transverse plate(s), number N being between 1 and 20. A ratio of the length of any of the housing tubes to the thickness of the plate or of any of the transverse plates is between 3 and 15. The length of the housing tubes is between 20 and 70 cm. Finally, one object of the invention is a packaging for transporting and/or warehousing radioactive materials, the packaging comprising a containment cavity delimited by a side body, a bottom and a lid, the packaging being fitted with a storage basket such as described above, arranged into the containment enclosure. Further advantages and characteristics of the invention will appear in the non-limiting detailed description hereinbelow. First with reference to FIG. 1, a packaging 100 for transporting and/or warehousing radioactive materials according to a preferred embodiment of the invention is represented. This packaging 100 includes a storage basket 1 in which radioactive materials 2 are to be stored, into housings 4 adjacent and parallel to a longitudinal central axis 6 of the basket. Also, each housing 4 has a housing axis 8 parallel to the central axis 6 of the basket 1, corresponding to the longitudinal central axis of the packaging. The packaging 100 defines a containment enclosure 12 in which the basket 1 is to be housed. Conventionally, the packaging is formed by an outer side body 14, a fixed bottom 16 and a removable lid 18 closing the enclosure 12. When the basket of the packaging is loaded with radioactive materials 2, the whole is conventionally called “package”. The storage basket 1 has a generally cylindrical-shaped axis 6, and with a circular cross-section complementary to that of the enclosure 12. The housings 4 also have a preferentially circular cylindrical-shaped cross-section, but which could alternatively be square, rectangular or even hexagonal, in order to accommodate the shape of the radioactive contents received in the housings 4. The aforesaid shape of the housings corresponds to their inner surfaces 4a, referenced 2 and 3 in the figures. By way of indication, it is noted that the number of housings 4 provided on the basket 1 is high, for example greater than 10. In the example represented in the figures, 32 housings are defined by the basket 1, each of these housings receiving radioactive materials 2 such as assemblies of fresh or irradiated fuels, or radioactive waste. The greatest width of each housing, referenced “LL” in FIG. 2 and corresponding to the internal diameter of each of these housings, is such that it has a low variable with respect to the external diameter of the basket referenced “DP”. More precisely, the ratio between dimensions LL and DP is lower than or equal to 0.2, which indeed reflects the presence of a high number of housings 4 within the basket. The basket 1 is made using various elements, among which at least one transverse plate 20, and preferably several transverse plates 20 spaced apart from each other along a longitudinal direction 22 of the basket, parallel to its longitudinal central axis 6. These transverse plates 20, also called slugs, are thus oriented transversally within the basket, that is arranged orthogonally to plane P and axes 6, 8. They are preferably made of steel, and each are generally of a disc shape perforated with a plurality of holes 24, the number of which corresponds to the total number of housings 4 within the basket. Between the plates 20, the basket 1 includes housing tubes 26 arranged parallel to the longitudinal central axis 6 of the basket. Much like the plates 20, the circular section tubes 26 are made of a steel, preferably devoid of neutron absorbing elements, such as a stainless steel. In this respect, it is indicated that by “neutron absorbing elements”, it is meant elements which have a cross-section area greater than 100 barns for thermal neutrons. However, in the embodiment represented, each housing tube 26 forms an internal tube surrounded by an external tube 26a made in turn of a material comprising neutron absorbing elements, such as an aluminium alloy comprising boron, gadolinium, hafnium, cadmium, or even indium. A very small radial clearance is provided between both coaxial tubes 26, 26a of each pair, since only one assembly clearance can indeed be kept. More precisely, two tubes 26 are associated with each hole 24 of each transverse plate 20, by being disposed on either side of the hole in question. One of the features of the invention lies indeed in forming housings 4 by associating housing tubes 26, and holes 24 passing through the plates 20. Since the transverse plates 20 and the housing tubes 26 are arranged alternately along the direction 22, the inner side surface 4a of each housing 4 is thus successively defined by the inner surface of a first housing tube 26, the inner surface of one of the holes of a first transverse plate 20, and the inner surface of a second housing tube 26, the inner surface of one of the holes of a second transverse plate 20, and so on. It is noted that the number of transverse plates is between 1 and 20, and preferentially between 5 and 15. Of course, this number depends on the height of the basket. The tubes 26 are then provided in a similar number, by having a length in the order of 20 to 70 cm. This length “L” depicted in FIG. 3 is greater than thickness “E” of the plates 20, since the ratio of the length L of any of the housing tubes 26, to the thickness E of any of the transverse plates 20, is between 3 and 15. The thickness E of the plate 20 is here considered in its current zone between the holes 24, at a distance from the periphery of the same which has a recess leading to a local reduction in the thickness. This recess forms a socket 30 which acts as holding means for the housing tube 26, an end of which is inserted into this socket. The external surface of the tube 26 is indeed guided by the side wall of the socket 30, whereas the bottom thereof forms a shoulder on which the tube 26 comes in axial abutment. Each hole 24 is thus fitted with two sockets 30 respectively provided on both opposite faces of the associated transverse plate 20, in order to hold both corresponding tubes 26. Preferably, only the steel housing tube 26 penetrates the socket 30, and not the associated external tube 26a. To do so, each tube 26 axially projects from each of both opposite ends of the external tube 26a, these ends resting in turn on the current part of the plate 20, having thickness E. Finally, it is noted that the basket 1 also comprises a top plate 32 and a bottom plate 34, sandwiching between each other the alternating transverse plates 20 and housing tubes 26. These plates 32, 34 also have holes for passing radioactive materials therethrough. They are connected to each other by tie rods 36 which pass therethrough, and which also pass through each transverse plate 20 at the periphery of the basket. By means of this design, the mechanical function is ensured by the tubes 26, plates 20 and tie rods 36 together, whereas the sub-criticality function is ensured by the external tubes 26a. Moreover, the segmented design of the housings 4 along the longitudinal direction 22 enables the desired straightness to be obtained, over the whole length thereof. And since the holes 24 of the plates 20 are an integral part of the housings, these holes can be brought closer to each other while keeping between each other a minimum material thickness in order to fulfil the mechanical strength function. As a result, the transverse overall size of the basket 1 is advantageously decreased. Of course, various modifications can be brought by those skilled in the art to the storage basket 1 and the packaging 100 just described, solely by way of indicating examples and within the scope defined by the appended claims. |
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052895122 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention is generally related to nuclear reactors and particularly to nuclear propulsion reactors. 2. General Background Desirable features of nuclear propulsion reactors for vehicles to be used in outer space are a high thrust-to-weight ratio and high exhaust temperature. Unfortunately, these are conflicting requirements where nuclear reactors are concerned. A high thrust-to-weight ratio tends to require low density materials while a high exhaust temperature tends to require high density materials. When these two features are sought in a single reactor core, the result is invariably a compromise in the core design that leads to a lower thrust-to-weight ratio and/or exhaust temperature than might otherwise be possible. In previous space nuclear propulsion reactor concepts the propellant gas is heated from its cryogenic inlet temperature to its final exhaust temperature by a single core region. This is the case for the ROVER/NERVA type engines that have been built and tested and known conceptual designs. This leaves a need for a nuclear propulsion reactor that provides a thrust-to-weight ratio and exhaust temperature that is more favorable than that of current designs. SUMMARY OF THE INVENTION The present invention addresses the above need in a straightforward manner. What is provided is a nuclear propulsion reactor that utilizes two stages for heating the propellant. Fuel elements in an annular first stage heat the propellant as it passes therethrough. The propellant exits the first stage into a plenum which then directs the propellant into a second stage. The second stage also contains fissionable material and is positioned in the axial region so that it is radially encompassed by the first stage. Fission reactions in the second stage are driven by leakage neutrons from the first stage. The second stage is formed from refractory (heat resistant) materials that are capable of withstanding higher exhaust temperatures. |
claims | 1. A nuclear fuel assembly comprising:A parallel array of elongated nuclear fuel elements supported between a lower nozzle and an upper nozzle and having an axial length along the elongated dimension of the nuclear fuel elements with a mid third region along the axial length;A plurality of substantially evenly spaced main support grids arranged in tandem along the axial length of the fuel elements, between the upper nozzle and the lower nozzle, at least partially enclosing an axial portion of the circumference of each fuel element within a support cell of the main support grids, with each support cell supporting only a single fuel element, to maintain a lateral spacing between fuel elements; andAt least one auxiliary grid positioned around the fuel elements in tandem with the main support grids at an elevation in the mid third region, the auxiliary grid comprising a plurality of support cells with one support cell for each fuel element, with each support cell supporting only a single fuel element, wherein the auxiliary grid is supported between two main support grids without any other auxiliary grids between the auxiliary grid and the adjacent main support grid and wherein the main support grids have a first fuel element support assembly and the auxiliary grid has a second fuel element support assembly and the first and second fuel element support assemblies are of a different design; andwherein the auxiliary grid support cells have walls that respectively at least partially enclose a portion of the circumference of the fuel elements along a portion of their axial length and the second fuel element support assembly is located on the walls of the auxiliary grid support cells and comprise a second set of dimples and/or springs that continuously contact and support the fuel elements and the main support grids' cells have walls that respectively at least partially enclose a portion of the circumference of the fuel elements along a portion of their axial length and the first fuel element support assembly is located on the walls of the main support grids' cells and comprise a first set of dimples and/or springs that continuously contact and support the fuel elements, wherein the second set of dimples and/or springs on the auxiliary support cells have a larger contact area with the fuel elements than the first set of dimples and/or springs on the walls of the main support grid cells. 2. The fuel assembly of claim 1 wherein the auxiliary grid is supported substantially midway between two main support grids. 3. The fuel assembly of claim 2 including a plurality of auxiliary grids positioned between some, but not all of the main support grids. 4. The fuel assembly of claim 3 wherein adjacent ones of the plurality of auxiliary grids share one main support grid between them. 5. The fuel assembly of claim 3 wherein the auxiliary grids are positioned along a mid span of the fuel elements within the mid third region. 6. The fuel assembly of claim 1 wherein the axial length of the walls of the auxiliary grid support cells is shorter than the corresponding walls of the main support grid cells. 7. The fuel assembly of claim 1 wherein the dimples and/or springs on the walls of the respective auxiliary grid support cells are coplanar along the same horizontal plane. 8. The fuel assembly of claim 1 wherein at least some of the main support grids have mixing vanes and at least some of the auxiliary grids do not have mixing vanes. 9. The fuel assembly of claim 1 wherein the auxiliary grid has an outer strap that extends around its circumference and includes upwardly extending guide tabs that are inwardly directed at an angle of less than 90 degrees with the strap in the direction of the adjacent fuel element, to prevent hang-up with adjacent fuel assemblies during removal or insertion into a reactor core. 10. A nuclear fuel assembly comprising:A parallel array of elongated nuclear fuel elements supported between a lower nozzle and an upper nozzle and having an axial length along the elongated dimension of the nuclear fuel elements with a mid third region along the axial length;A plurality of substantially evenly spaced main support grids arranged in tandem along the axial length of the fuel elements, between the upper nozzle and the lower nozzle, at least partially enclosing an axial portion of the circumference of each fuel element within a support cell of the main support grids, with each support cell supporting only a single fuel element, to maintain a lateral spacing between fuel elements; andAt least one auxiliary grid positioned around the fuel elements in tandem with and sandwiched between two of the plurality of main support grids at an elevation in the mid third region, the auxiliary grid comprising a plurality of support cells with one support cell for each fuel element, with each support cell supporting only a single fuel element, wherein the main support grids have a first fuel element support assembly and the auxiliary grid has a second fuel element support assembly and the first and second fuel element support assemblies are of a different design; andwherein the auxiliary grid support cells have walls that respectively at least partially enclose a portion of the circumference of the fuel elements along a portion of their axial length and the second fuel element support assembly is located on the walls of the auxiliary grid support cells and comprise a second set of dimples and/or springs that continuously contact and support the fuel elements and the main support grids' cells have walls that respectively at least partially enclose a portion of the circumference of the fuel elements along a portion of their axial length and the first fuel element support assembly is located on the walls of the main support grids' cells and comprise a first set of dimples and/or springs that continuously contact and support the fuel elements, wherein the second set of dimples and/or springs on the auxiliary support cells have a larger contact area with the fuel elements than the first set of dimples and/or springs on the walls of the main support grid cells and the full extent of the axial length of the walls of the auxiliary grid support cells is shorter than the corresponding walls of the main support grid cells. |
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claims | 1. A method of generating isotope products, the method comprising:selecting an irradiation target;placing the irradiation target in a target rod;installing the target rod in a water rod of a nuclear fuel assembly such that the irradiation target is surrounded by a fluid in the water rod;holding the target rod within the water rod with at least one securing device; andexposing the irradiation target to a neutron flux so as to substantially convert the irradiation target into isotope products,wherein the at least one securing device includes a washer affixed to the water rod at an axial position, the washer including a plurality of apertures, the target rod extending through one of the plurality of apertures. 2. The method of claim 1,wherein the placing the irradiation target includes forming the irradiation target into the target rod. 3. The method of claim 1, further comprising:harvesting the isotope products from the water rod. 4. The method of claim 1, wherein the placing the irradiation target includes positioning the irradiation target at a position within the water rod to achieve a desired neutronic or thermodynamic property of the nuclear fuel assembly. 5. The method of claim 1, wherein the selecting the irradiation target includes selecting a type and amount of irradiation target to achieve a desired activity of isotope product based on the properties of the irradiation target and amount and duration of the neutron flux to which the irradiation target is exposed. 6. The method of claim 1, wherein the exposing the irradiation target to neutron flux includes commencing power operation in a 100+ MWth reactor containing the fuel assembly. 7. A method of generating isotope products, the method comprising:placing an irradiation target in a target rod,installing the target rod in a water rod of a nuclear fuel assembly;holding the target rod within the water rod with at least one securing device such that the target rod is surrounded by a fluid in the water rod, the target rod installed at a position within the water rod to achieve a desired neutronic or thermodynamic property of the nuclear fuel assembly and a desired activity of isotope product, the installing based on the properties of the irradiation target and amount and duration of neutron flux to which the irradiation target will be exposed at the position; andexposing the irradiation target to the neutron flux at the position for the duration so as to convert the irradiation target into isotope products,wherein the at least one securing device includes a washer affixed to the water rod at an axial position, the washer including a plurality of apertures, the target rod extending through one of the plurality of apertures. 8. The method of claim 7, wherein the exposing the irradiation target to the neutron flux includes commencing power operation in a 100+ MWth reactor containing the fuel assembly. 9. The method of claim 7, wherein a plurality of irradiation targets are placed in the target rod, the plurality of irradiation targets not being fabricated of a same material. 10. A system for producing isotopes in a water rod of a fuel assembly, the system comprising:at least one target rod containing an irradiation target, the at least one target rod having a size that permits placement of the at least one target rod within the water rod, the at least one target rod having a configuration that permits the at least one target rod to be surrounded by a fluid in the water rod; andat least one securing device configured to hold the at least one target rod within the water rod during operation of a reactor containing the fuel assembly,wherein the at least one securing device includes a collar joined to the water rod at an axial position and extending radially into the water rod, the collar supporting the at least one target rod at the axial position. 11. The system of claim 10, wherein the collar and at least one target rod supported by the collar are joined. 12. The system of claim 10, wherein the at least one securing device further includes a bushing extending axially upward from the collar and joined to the collar, the bushing limiting radial movement of the at least one target rod within the water rod. 13. The system of claim 10, wherein the at least one securing device includes a washer affixed to the water rod at an axial position, the washer including a plurality of apertures, the at least one target rod extending through one of the plurality of apertures. 14. The system of claim 13, wherein the one of the plurality of apertures has a diameter substantially equal to that of the target rod extending therethrough, so as to frictionally join with and maintain the position of the target rod extending therethrough. 15. The system of claim 10, wherein the target rod has an outer wall that defines a cavity inside the target rod. 16. The system of claim 15, wherein one or more irradiation targets are positioned within the cavity. 17. The system of claim 10, wherein the target rod further includes a joining device configured to join to the water rod and hold the target rod stationary therein. 18. A nuclear fuel assembly, comprising:a plurality of fuel rods containing fissile material, the fuel rods extending in an axial direction;at least one water rod extending in the axial direction, the water rod being open-ended at ends of the fuel assembly so as to permit fluid to flow through the fuel assembly in the axial direction;at least one irradiation target positioned within the at least one water rod such that the at least one irradiation target is surrounded by the fluid in the water rod, the irradiation target substantially converting into an isotope product when exposed to neutron flux in the water rod;at least one target rod containing the irradiation target, the at least one target rod having a size that permits placement of the at least one target rod within the water rod; andat least one securing device configured to hold the at least one target rod within the water rod during operation of a reactor containing the fuel assembly,wherein the at least one securing device includes a collar joined to the water rod at an axial position and extending radially into the water rod, the collar supporting the at least one target rod at the axial position. 19. The fuel assembly of claim 18, wherein the at least one securing device further includes a bushing extending axially upward from the collar and joined to the collar, the bushing limiting radial movement of the at least one target rod within the water rod. 20. The fuel assembly of claim 18, wherein the at least one securing device includes a washer affixed to the water rod at an axial position, the washer including a plurality of apertures, the at least one target rod extending through one of the plurality of apertures. |
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description | This application is a national phase of International Application No. PCT/EP2006/065186 entitled “Package Serving To Accommodate A Case Containing Radioactive”, which was filed on Aug. 9, 2006, which was not published in English, and which claims priority of the French Patent Application No. 05 52499 filed Aug. 11, 2005. The present invention generally pertains to a cask intended to receive a canister containing radioactive material such as irradiated nuclear fuel assemblies, nuclear waste, etc. It also relates to a method to transfer said canister containing radioactive material, from a cask to a receiver housing, or conversely. By way of indication, if transfer is made from the cask to the receiver housing, the housing may for example be provided at a burial site ensuring the long-term storage of this type of canister. Finally, the invention also concerns an extraction/insertion system for said canister, whose function is to remove said canister from its associated cask and/or to insert the canister in this same cask. When a nuclear fuel assembly has been irradiated subsequent to its use in a nuclear plant, it may be placed in a sealed container called a canister before being placed in long-term storage, e.g. at a deep burial site. To ensure the transport of the canister towards the long-term storage site, the canister is first placed in a cask called a transfer cask to form an assembly called a waste package. It is then the entirety of the package which is transported to the burial site at which a transfer method is used to transfer this canister from the transfer cask to a receiver housing on the site. For this purpose, provision is generally made to align the cask holding the canister with the receiver housing, preferably horizontally, then to pull or push this canister in order to transfer it from its transfer cask to its associated receiver housing. This technique is particularly described in document U.S. Pat. No. 4,780,269, relating to a surface storage installation. The major drawback with said embodiment lies in the fact that when the canister is extracted from its cask, the outer side surface of this canister undergoes major friction against the inner surface of the cask delimiting the cask housing. This friction may lead to the tearing of particles from the outer side surface of the canister, exposing the canister to risks of corrosion which may be detrimental to the integrity of this canister during its long-term storage at the disposal site. Finally, it is to be noted that the phenomenon of particle stripping described above may also be observed at the time the canister is placed in its cask, when it is arranged horizontally, owing to similar friction which may occur against the inner surface of the cask housing. The purpose of the invention is therefore to propose a cask whose design can at least partly remedy the above-mentioned drawbacks of prior art embodiments. The purpose of the invention is also to present a package comprising said cask, and to propose a system to extract a cask containing radioactive material. Another purpose of the present invention is to propose an extraction/insertion system for said canister, whose function is to remove this canister from its associated cask and/or to insert the canister in this same cask. Finally, a further purpose of the present invention is to propose a transfer method for a canister containing radioactive material, from a cask to a receiver housing or conversely. To do so, the object of the invention is firstly a cask intended to receive a canister containing radioactive material, this cask having a inner surface delimiting a cask housing for this canister. According to the invention, it also comprises an extraction/insertion assembly for the canister bearing on the inner surface, this extraction/insertion assembly being designed so that the canister can be moved inside the cask housing along a longitudinal direction of the cask, in a carried position in which this canister is free of any contact with the inner surface. In addition, the canister extraction assembly is provided with a carriage, bearing against the inner surface, and with a canister support structure carried by the carriage, the carriage and support structure being designed so that they are able to assume a position in which they are drawn together in a radial direction of the cask, allowing contact between the canister and the inner surface, and a position in which they are drawn apart from each other in the radial direction of the cask, in which the canister borne by the support structure takes up its carried position. Finally, according to the invention, one of the two elements from among the carriage and support structure is provided with at least one guide ramp cooperating with a ramp follower provided on the other of the two elements, the ramp being made so that application of a relative translational movement between the carriage and the canister support structure in a longitudinal direction of the cask, causes the two elements to change over from the drawn-together position to the drawn-apart position, or conversely. One of the specificities of this cask, compared with those found in the prior art, therefore lies in the fact that it integrates a canister extraction/insertion assembly whose design allows this canister to be spaced away from the inner surface delimiting the cask housing, before starting to extract this canister. Therefore, during such extraction when the canister in contact with the assembly takes up its carried position, the lack of any contact and friction between this canister and the inner surface avoids the tearing away of particles on the outer side surface of the canister, and eliminates associated risks of corrosion. Similarly, the carried position can also be taken up by the canister in contact with the extraction/insertion assembly before it is inserted into the cask. This specificity allows the canister, when it is being inserted inside the cask, not to be in contact either with the inner surface of the cask before it has been fully inserted inside its cask housing. As arises from the foregoing, it is to be understood that the extraction/insertion assembly can be used either to ensure removal of the canister from the cask housing, or to ensure insertion of this canister inside this same cask, or to ensure both above-cited application. In each of these two procedures, when the canister lies fully outside its cask housing, the assembly carrying the cask may also take up an outside position in which it is fully extracted from a cask body defining the inner surface. On the other hand, when the canister lies fully inside the cask housing, as is the case in particular during its transport, the assembly is in a retracted position in which it lies fully within the cask body. Additionally, as indicated above, the canister extraction assembly is provided with a carriage bearing against the inner surface and with a canister support structure carried by the carriage, the carriage and support structure being designed so that they can assume a position in which they are drawn close to each other in a radial direction of the cask, allowing contact between the canister and the inner surface, and a position in which they are drawn apart in the radial direction of the cask in which the canister carried by the support structure takes up its carried position. Therefore, it will be understood that design of the assembly allows it to be positioned at any point around the inner surface and against it, even if a preferred position consists of making provision for it to rest on a lower end portion of this inner surface, when the cask is arranged horizontally or slightly at an angle to the horizontal, as is usually the case for the transfer of the canister to a receiver housing. This preferred positioning allows the canister to rest simply under gravity on the support structure, when the extraction assembly lies in its drawn-apart radial position bringing this canister to its carried position. Evidently, other positions of the extraction assembly within the cask, requiring the provision of securing means to secure the canister to the structure support so that this canister is maintained in its carried position, can be envisaged without departing from the scope of the invention. Also, as specified above, one of the two elements from among the carriage and support structure is provided with at least one guide ramp cooperating with a ramp follower provided on the other of the two elements, the ramp being designed so that application of a relative translational movement between the carriage and the canister support structure, in a longitudinal direction of the cask, causes the changing of the two elements from the drawn-together position to the drawn-apart position, or conversely. Therefore, one advantage provided by said configuration is that the relative movement required to obtain the changeover from one position to another must be made in a direction that is identical to the direction of extraction of the canister from its cask. Therefore, it then advantageously becomes possible to use the same mobilizing means to ensure insertion of the canister in its cask, the extraction of the canister out of its cask e.g. to implement a transfer method towards a deep burial site, and also to ensure the changeover from the drawn-together position to the drawn-apart position, and conversely. Preferably the carriage used is a travelling carriage rolling on the inner surface of the cask. However, it could evidently be a rail-mounted or slipper pad carriage without departing from the scope of the invention. In parallel, it is preferably provided that the above-mentioned ramp follower is a roller. Again preferably, the canister extraction assembly is arranged in a cavity opened towards the cask housing, and defined by the inner surface of the cask. Here again, it could alternatively be provided that the open cavity is made on the outer side surface of the canister and not on the inner surface of the cask, or this cavity could be defined jointly by the inner surface of the cask and by the outer side surface of the canister. Finally, if the cavity is defined by the inner surface, provision is preferably made for the cask to have two canister supports partly defining the inner surface and being spaced at an angle around a longitudinal axis of the cask, so as partly to delimit this open cavity between them. Naturally other configurations may be envisaged, such as one in which the entirety of the inner surface is provided on a single-piece body. Another object of the present invention concerns a package of radioactive material, comprising a cask such as described above and a canister containing the radioactive material and arranged inside the cask housing. A further object of the invention relates to an extraction/insertion system for a canister containing radioactive material, this canister being intended to be extracted/inserted in a cask housing delimited by an inner surface of a cask. According to the invention, the system comprises a canister extraction/insertion assembly such as described above and whose design is such that the application of a relative translational movement between the carriage and the canister support structure in the longitudinal direction of the cask, causes a changeover of the two elements from the drawn-together position to the drawn-apart position, or conversely, this extraction system also comprising mobilizing means in a longitudinal direction of the cask connected to one of the two elements, and retractable abutment means cooperating with the other of the two elements. Evidently, in this system in which the mobilizing means are preferably connected to the carriage of the assembly, the retractable abutment means are piloted so that the utilisation of these mobilizing means alternately cause displacement of the entirety of the assembly in a longitudinal direction of the cask, and a relative movement between the two elements so that they over from the drawn-together position to the drawn-apart position, or conversely. By connecting the mobilizing means to the carriage and by causing the retractable abutment means to cooperate with the support structure, it is advantageously observed that during simultaneous locking of this structure and longitudinal movement of the carriage, the canister support structure is then moved solely in radial direction, which does not induce any relative movement between the canister and this support structure. Any friction between these two, elements is therefore advantageously prevented. It is noted that this system is preferably intended to be used for implementation of a method to transfer a canister containing radioactive material from a cask towards a receiver housing, such as the method which is described below and which is also a subject of the invention. Nonetheless, it may also be intended for implementation of a canister transfer method in which this canister is to be extracted from a receiver housing for insertion inside a cask, without departing from the scope of the invention. Also, it is noted that the system may be such that its canister extraction/insertion assembly forms an integral part of the cask, as is the case for the cask subject of the invention. However, alternatively, this assembly could be an integral part of the means defining the receiver housing, or of any other means such as a motorized vehicle. The method of the invention therefore relates to a transfer method for a canister containing radioactive material, from a first to a second entity from among the group consisting of a cask such as described above and a receiver housing delimited by an inner surface. It comprises the following successive steps consisting of: bringing the canister located inside the first entity to a carried position in which this canister is devoid of any contact with the inner surface associated with this first entity; setting in movement the extraction/insertion assembly carrying the canister so as to cause this extraction assembly and the canister to enter inside the second entity. Therefore, in the preferred case in which the first entity is a cask, the second stated step then consists of mobilizing the extraction/insertion assembly carrying the canister, in a direction of extraction of lying in the longitudinal direction of the cask, so as to cause the extraction assembly and the canister to enter inside the receiver housing. Preferably, if the method is intended to ensure transfer from the cask housing to a receiver housing, it also comprises the following successive steps consisting of: bringing the canister to a position in which it is deposited inside the receiver housing wherein this canister is devoid of any contact with the extraction assembly; setting in movement the extraction assembly in a direction opposite to the extraction direction, so as to re-insert this extraction assembly inside the cask. Preferably, the canister extraction assembly used belongs to the extraction system presented above, which is used to implement this transfer method. Therefore, provision can be made so that the step consisting of bringing the canister to its carried position inside the cask is performed by carrying out the following successive operations: connecting means to mobilize the extraction system to one of the two elements, either the carriage or the support structure, of the extraction assembly; actuating a first abutment belonging to the retractable abutment means, so as to bring this first abutment from a retracted position to an abutment position, allowing the locking in translation of the other of the two elements in the extraction direction lying in the longitudinal direction of the cask; actuating the mobilizing means in the direction of extraction to cause displacement of the carriage and support structure from the drawn-together position to the drawn-apart position; and actuating the first abutment so as to bring it from the abutment position to the retracted position. Naturally, it would also have been possible to provide a type of guide ramp requiring actuation of the mobilizing means in the opposite direction to the extraction direction in the longitudinal direction of the cask, in order to cause changeover from the drawn-together position to the drawn-apart position, without departing from the scope of the invention. Still preferably, the step consisting of bringing the canister to a deposited position inside the receiver housing is conducted by implementing the following successive operations: actuating a second abutment belonging to the retractable abutment means so as to bring this second abutment from a retracted position to an abutment position, allowing the locking in translation of the other of the two elements in the opposite direction to the extraction direction; actuating mobilizing means in the opposite direction in order to cause changeover of the carriage and support structure from the drawn-apart position to the drawn-together position; and actuating the second abutment so as to bring it from the abutment position to the retracted position. Other advantages and characteristics of the invention will become apparent on reading the detailed, non-limiting description given below. With reference firstly to FIGS. 1 and 2a, a cask is shown intended to receive a canister (not shown) containing radioactive material, this cask being in the form of a preferred embodiment of the present invention. The cask 1 globally comprises a hollow cask body 2 of cylindrical shape and defining a cask housing 4, a lid 6 closing the housing 4, and two covers 8 respectively arranged at the two ends of the cask body 2. The above-mentioned elements are of conventional design known to those skilled in the art, and therefore allow the housing of a canister containing irradiated nuclear fuel assemblies for example and/or nuclear waste. They therefore ensure the usual functions of neutron protection, protection against gamma radiation, and mechanical resistance. One of the particularities of the present invention consists firstly of providing that the cask housing 4 is delimited by a cask inner surface 10 whose section is not circular, contrary to cask sections found in the prior art intended to receive cylindrical canisters of circular section. As can be seen more clearly in FIG. 2a, the inner surface 10 is jointly delimited firstly by an inner cylindrical wall with circular section 15 of the body 2, arranged along a longitudinal axis 12 of the cask parallel to a longitudinal direction 14 of this same cask, and secondly by two support surfaces 16 respectively belonging to two canister supports 18 mounted fixedly on the above-mentioned inner wall 15. The two supports 18, therefore partly defining the inner surface 10, are spaced at an angle around the longitudinal axis 12. Therefore, when the cask 1 lies in a substantially horizontal position such as shown FIG. 2a and corresponding to the position taken up during transfer of the canister towards a receiver housing, the two supports 18 arranged symmetrically with respect to a median vertical plane of the cask, together delimit a cavity 20 open towards the cask housing 4 and closed downwardly by a lower portion of the inner wall 15. More precisely, it is noted that the open cavity 20 is partly delimited by the two lower sides 19 of the two canister supports 18 preferably extending substantially over the entire length of the body 2, and that the two support surfaces 16 of these canister supports 18 are cylindrical portions of circular section arranged along one same longitudinal axis 24, preferably separate from the longitudinal axis 12 of the cask 1. Therefore, it is to be understood that the inner surface 10 is designed to delimit both the cask housing 4 and the open cavity 20 leading into it. Nevertheless, it is specified that this inner surface 10 could be obtained in another manner, other than by adding canister supports on an inner wall 15 of circular section of body 2, without departing from the scope of the invention. The open cavity 20, which is therefore preferably located in the lower part of the body 2 when the cask is in canister transfer position, is intended to receive a canister extraction/insertion assembly 36 such as the one which will now be described with reference to FIG. 3. In this figure, it can be seen that the canister extraction/insertion assembly 36, hereinafter called canister extraction assembly, globally consists firstly of a carriage 38 intended to bear against the inner surface 10, and more precisely against the lower portion of the inner wall 15 which delimits the open cavity 20, and secondly of a canister support structure 40 carried by this carriage 38. The structure 40 therefore has an upper part of upwardly incurved shape as is clearly visible FIG. 3, so that it is able to hold this canister when it is solely carried by this same support structure 40. The longitudinal direction 14 of the cask is shown so as to indicate the positioning of the assembly 36 inside the cask 1, this assembly 36 preferably extending substantially over the entire length of the cask body 2, in cavity 20. Also, it can be seen that the carriage 38 is preferably a carriage able to run over the lower portion of the inner wall 15 which delimits the open cavity 20. The above-mentioned carriage 38 and structure 40 are designed so that they can take up a drawn-together position in a radial direction of the cask 1, schematically illustrated by arrow 42 in FIG. 3, and a drawn-apart position in this same direction 42, orthogonal to direction 14. In FIG. 4a partly showing the assembly 36 in the same configuration as in FIG. 3, i.e. when the carriage 38 and structure 40 assume their drawn-together position, it can be seen that the support structure 40 is provided with several guide ramps 48 (only one being shown in this FIG. 4a), whilst the carriage 38 is equipped with a ramp follower 50 associated with each of the ramps 48 and in the form of a roller. In this respect it can be seen FIG. 3 that the assembly 36 is effectively provided with several ramp/roller assemblies, preferably distributed on each side of this assembly. In the drawn-together position, for each ramp/roller assembly, the roller 50 bears on an upper rear end of the ramp 48, whose geometric shape is such that it descends in the longitudinal direction 14, in the direction of canister extraction schematically illustrated by arrow 52. Evidently, the notion of <<descent>> is to be considered when the extraction assembly 36 rests in the open cavity 50 located in a lower part of the body 2 of cask 1 arranged horizontally, such as is shown FIG. 2a. With said geometry, the ramp 48 is therefore designed so that the application of a relative translational movement between the carriage 38 and the canister support structure 40, in direction 14 of the cask, causes the changeover of these two elements 38, 40 from the drawn-together position to the drawn-apart position. As can be seen FIG. 4b, the application of said relative movement intended to move the carriage 38 in the direction of extraction 52 relative to the canister support structure, results in moving the roller 50 inside its associated ramp 48, until this roller 50 reaches a lower front end of this ramp 48. It is therefore the particular geometry of the ramp 48, through which the roller 50 passes, which allows the automatic causing of a relative radial movement drawing apart the two elements 38, 40, subsequent to mere application of a relative movement in the longitudinal direction 14 of the cask. Naturally, the above-mentioned notions <<front>> and <<rear>> can be respectively likened to so-called notions of <<towards the opening>> and <<towards the bottom part>> of the cask housing 4, when the cask 1 has been positioned horizontally for extraction of the canister, a position in which direction 14 is substantially parallel to the horizontal. Also, the application of a relative movement intended to move the carriage 38 in the direction of extraction 52 relative to the structure 40 results in moving the roller 50 inside its associated ramp 48 until the roller 50 meets up with the upper front end of this ramp 48. This then ensures changeover of the assembly 36 from the drawn-apart position to the drawn-together position. In this respect, it is specified that the upper front end and the lower rear end may each be provided with a notch into which the roller can enter 50 for the purpose of firmly maintaining the drawn-together and drawn-apart positions. In FIG. 2b, a package 60 can be seen comprising the cask 1 provided with the canister extraction assembly 36, and a canister 62 sealingly containing radioactive material and preferably being made in copper. This canister 62 is preferably cylindrical and of circular section. In this FIG. 2b, the extraction assembly 36 resting in its lower open cavity 20, is shown in the drawn-together position in which its two elements 38, 40 are sufficiently drawn close to one another in the radial direction 42 so that the canister 62 is able to rest under gravity on surfaces 16 of the two supports 18, without being in contact with the support structure 40 carried by the carriage 38 itself equipped with wheels 66 in contact with the inner wall 15 partly delimiting cavity 20. In addition, an upper part of this canister 62 is spaced away from the inner wall 15 of the body 2, and rests in a so-called deposited position in its cask housing 4, in which it is solely in contact with the lower portions of the surfaces 16. On the other hand, as can be seen FIG. 5, when the extraction assembly 36 takes up its drawn-apart radial position, its support structure 40 lifts up the canister 62 by entering inside the housing 4, which causes this canister 62 to take up a so-called carried position in which it is no longer in contact at all with the inner surface 10, and hence no longer in contact with the supports 18 from which the canister 62 has been drawn away. The canister 62 is then solely held by the support structure 40, owing to gravity and the incurved shape of this structure preventing this canister from escaping sideways. Consequently, to ensure extraction of the canister 62 out from the cask without damaging its outer side surface, all that is required is to set assembly 36 in movement in direction 14, and more particularly in the direction of extraction 52. Therefore the canister 62 placed in movement advantageously does not undergo any friction since it remains fixed with respect to the structure 40 which lifts it, and it is the wheels 66 of the carriage 38 which move along the inner surface 10 in the longitudinal direction 14. In this respect, the invention also relates to a transfer method for said canister 62 containing radioactive material, from a cask 1 towards a receiver housing, or conversely, this receiver housing possibly being provided at a burial site for the long-term storage of this type of canister. FIGS. 6a to 6f shows different successive steps of a transfer method according one preferred embodiment of the present invention, whose implementation is preferably ensured by a canister extraction/insertion system 70 also subject of the invention. This embodiment concerns a preferred non-limiting case in which the method consists of ensuring transfer of a canister from a cask 1 to a receiver housing. For this reason, the system 70 will be called an extraction system 70 in the remainder hereof. As can be seen FIG. 6a, this system 70 globally comprises the canister extraction assembly 36 already described and intended to equip cask 1, mobilizing means 73 in direction 14 connected to the carriage 38, and retractable abutment means 78 cooperating with the support structure 40. Globally, as will be seen below in the description of the method, it is to be understood that this system 70 is designed and piloted so that it can generate movement of the canister 62 in direction 14, and its radial displacement, intended for example to cause it to change from its deposited position inside housing 4 to its carried position. With reference therefore to FIG. 6a, it can be seen that the method to transfer canister 62 consists firstly of conducting a series of preparatory operations such as opening the cask 1 by removing its upper cover 8 and lid 6, placing this cask 1 on a docking cylinder 80 positioned in the continuation of the receiver housing 82 into which the canister 62 is to be transferred, and connecting the mobilizing means 73 to the extraction assembly 36 e.g. by passing a telescopic arm 84, oriented in direction 14 and forming an integral part of means 73, through the bottom part of the cask body 2. Before starting the transfer operations, it is therefore ensured that the longitudinal axis of the cask merges with a longitudinal axis 86 of the receiver housing 82, which is delimited by inner surface 88 whose shape is preferably identical to the shape of the inner surface 10 of the cask 1, and which therefore also delimits a lower cavity 90 opened towards the receiver housing 82. As already mentioned above, it is also ensured that the open cavity 20 is located at a lower part of the cask 1 positioned horizontally or slightly at an angle with respect to the horizontal. When the cask 1 is correctly positioned relative to the receiver housing 82, a first step consists of bringing the canister 62 to its carried position inside the cask housing 4, i.e. to cause the extraction assembly 36 to change over from the drawn-together radial position to the drawn-apart radial position. For this purpose, once the mobilizing means 73 are mechanically joined to the carriage 36, a first abutment 74 is actuated belonging to the retractable abutment means 78, so as to bring this first abutment 74 from a retracted position to an abutting position such as shown FIG. 6a. In this abutting position the first abutment 74 crossing cask 1 or the docking cylinder 80 comes to abut a front end of the support structure 40, thereby allowing the latter to be locked in translation in the direction of extraction 52. The mobilizing means 73 can then be actuated in the direction of extraction 52 in order to cause relative movement between the support structure 40 locked longitudinally in translation and the carriage 38 directly driven by these means 73, for the purpose already described of causing these elements 38, 40 to change over from the drawn-together position to the drawn-apart position shown FIG. 6b. During this operation, it is noted that the support structure 40 advantageously only undergoes a radial translational movement, which prevents friction against the outer side surface of the canister. Then, once contact has been released between the canister 62 and the inner surface 10, the first abutment 74 is actuated so as to return it to its retracted position in which it releases the structure 40 which is then no longer locked in translation in the direction of extraction 52. By way of indication, it is noted that this abutment 74 can for example be pivot-mounted on the cask body 2, and be piloted manually or automated fashion. The following step consists of setting in movement the extraction assembly 36, lifting the canister 62, in the direction of extraction 52 and using mobilizing means 73, so as to cause this assembly 36 and the canister 62 to enter inside the receiver housing 82, as shown FIG. 6c. At the time of entering, it is observed that the assembly 36 comes to insert itself in the lower open cavity 90 delimited by inner surface 88 and located in the continuation of lower open cavity 20 of the cask. Once the canister 62 has been inserted to a sufficient depth inside housing 82, the step is conducted of bringing this canister to its deposited position inside the receiver housing 82, in which this canister is free of any contact with the extraction assembly 36 but in which it rests on the inner surface 88, preferably on two canister supports 94 (only one being illustrated), identical to supports 18 and located in the continuation thereof. To do so, a second abutment 76 is actuated belonging to the retractable abutment means 78, so as to bring this second abutment 76 from a retracted position to an abutment position such as shown FIG. 6d. In this abutment position, the second abutment 76 crossing the docking cylinder 80 comes to abut a rear end of the support structure 40, therefore ensuring the locking in translation of this structure in an opposite direction 98 to the direction of extraction 52. The mobilizing means 73 can then be actuated in the opposite direction 98 to cause relative movement between the structure 40 locked in translation and the carriage 38 directly driven by these means 73, for the purpose of causing these elements 38, 40 to change over from the drawn-apart position to the drawn-together position shown FIG. 6e. Next, once contact has been released between the canister 62 and the support structure 40, the second abutment 76 is actuated to return it to its retracted position in which it releases the structure 40 which is then no longer locked in translation in the opposite direction 98. By way of indication, it is noted that this abutment 76 may for example be pivot-mounted on the docking cylinder 80 or on the body delimiting the receiver housing 82, and can be automatically or manually piloted. Finally, to complete this transfer method, the extraction assembly 36 is again set in movement in the opposite direction 98 so that it can be re-inserted inside the cask 1, without the canister 62, in its associated open cavity as can be clearly seen FIG. 6f. Thereafter the mobilizing means 73 can be uncoupled from the carriage 38 and the cask separated from the docking cylinder 80, so that this cylinder can again be used for another canister-transfer. Naturally, it is to be understood that if the method of the invention is intended to ensure the transfer of a canister from a receiver housing towards a cask 1, the steps to be conducted are implemented in reverse order to the order just described. Evidently, various modification may be made by persons skilled in the art to cask 1, package 60, extraction system 70 and the transfer method which have just been described solely as non-limiting examples. |
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052271285 | abstract | A reactor core removable fuel assembly includes upper and lower tie plates having pluralities of fuel rods and hollow control rods extending therebetween. The lower tie plate includes a lower manifold therein joined in flow communication with a reservoir containing a neutron absorbing control liquid, with the reservoir being removable from the reactor core together with the fuel assembly. The control liquid is selectively pumped from the reservoir through the lower manifold and into the control rods for selectively varying the level of the control liquid therein for controlling reactivity. |
abstract | A method for filtration of harmful gas effluents from an industrial installation including the steps of providing a gas effluent from an industrial installation, the gas effluent including a mixture of gases; filtering the harmful, elements from the gas effluent by membrane separation through a plurality of membranes, the membrane separation being achieved by sifting, sorption and/or diffusion, each membrane being adapted for filtering a specific harmful element; sorting the filtered harmful elements and storing them in separate storage reservoirs, and discharging the processed gas effluent to the atmosphere. |
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abstract | The present invention provides a method and an apparatus for producing a two-dimensional patterned beam, e.g. a two-dimensional patterned and focused ion beam, for fabricating a nano-structure on a substrate with the precursor gas. In comparison with the conventional focused ion beam that is applied for fabricating a dot-like nano-structure the method is more simplified and easy to be achieved. |
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051951134 | summary | BACKGROUND OF THE INVENTION The present invention relates to an X-ray exposure apparatus, and particularly, to one capable of effectively adjusting an alignment of an axis of radiated X-ray beams of the X-ray exposure apparatus with an optical axis of an optical alignment system of the X-ray exposure apparatus and a method of positioning the X-ray exposure apparatus. In a known technology, a light exposure device has been generally utilized for forming a large scale integrated circuit (LSI) pattern, but recently, the formation of patterns by light exposure device has approached its limit in resolution for the requirement of the formation of extremely fine LSI patterns. In view of this limit, recently, an X-ray exposure apparatus capable of forming a fine LSI pattern in comparison with the light exposure device are being searched and developed. The X-ray exposure device requires an X-ray source provided with high luminance and, in this viewpoint, an attention has been paid to synchrotrons as X-ray sources. A conventional X-ray exposure apparatus provided with a synchrotron as the X-ray source will be described hereunder. X-ray beams generated from the synchrotron are reflected by a reflecting mirror and then enter a chamber of an exposure apparatus body of the exposure apparatus. The interior of the chamber is of a helium atmosphere for preventing the X-ray beams from being attenuated, and in the inside of the chamber are accommodated a mask stage holding an X-ray mask that is movable, a wafer stage holding a semiconductor wafer to be movable and an alignment optical system for detecting a positional offset between patterns described on the X-ray mask and the semiconductor wafer. According to this structure of the X-ray exposure apparatus, X-ray beams entering the chamber of the X-ray exposure apparatus body are irradiated on the X-ray mask, and then irradiated on the surface of the semiconductor wafer to thereby expose a circuit pattern of the X-ray mask onto the semiconductor wafer. The exposure apparatus body is mounted on an oscillation removing table which is set on a floor through an elastic member such as air spring, thus effectively preventing the exposure apparatus body from being oscillated or affected by an external shock or the like from the floor. The X-ray beams obtained by a synchrotron orbit radiation (SOR) constitute horizontal linear beams irradiated in a horizontal direction. In order to enlarge an irradiated area for carrying out an exposure transfer of the circuit pattern of the X-ray mask onto the semiconductor wafer, an X-ray reflecting mirror is arranged between the synchrotron and the exposure apparatus body, and the X-ray beams are swinged by swinging the reflecting mirror. Furthermore, in order to transfer the circuit pattern on the X-ray mask onto the semiconductor wafer in a precisely overlapped manner, it is required that the alignment optical system and an exposure optical system for the transfer be stably constructed with a predetermined performance. This requirement, in one example, is achieved by assembling, for adjustment with mechanically high precision, the mask stage, the wafer stage and the alignment optical system as the exposure apparatus body. For the reason described above, in the X-ray exposure apparatus utilizing the SOR, the adjustment of the axes of the X-ray beams and the alignment optical system are required to have high precision. In other words, in order to carry out transferring, in a precisely overlapped state, the circuit pattern on the X-ray mask to the semiconductor wafer on which a resist is coated, it is necessary to adjust the positions of the X-ray mask and the semiconductor wafer so as not to have a positional offset between a position of a mask pattern and a position of a pattern to be formed on the semiconductor wafer. Namely, the precise coincidence of the optical axis of the X-ray beams with an optical axis, as measurement reference, of the alignment optical system is required. In the case of no coincidence, a positional offset is caused in an amount such as shown by the following equation, EQU .delta.=G.times..DELTA..theta. in which G is a gap between the X-ray mask and the semiconductor wafer, and .DELTA..theta. is an angular difference between the axis of the beam and the axis of the alignment optical system, and .delta. is an amount of positional offset. As described above, the X-ray beams are reflected by the reflecting mirror and then enter the exposure apparatus body, so that the X-ray beams are not ordinarily parallel to the floor on which the X-ray exposure apparatus is settled and has an inclination of several angles. On the contrary, since the exposure apparatus body is precisely assembled with a horizontal plane as the reference level, the optical axis of the alignment optical system as the measurement reference is made substantially horizontal. For this reason, in the prior art, in order to make the axis of the X-ray beams coincident with the optical axis of the alignment optical system as the measurement reference, there is provided a method wherein two parallel X-ray reflecting mirrors are arranged with a predetermined distance therebetween between the synchrotron and the exposure apparatus body to thereby reflect the X-ray beams twice by the two reflecting mirrors to obtain horizontal light beams. In another method, an angle of the X-ray reflecting mirror is generally adjusted in accordance with the arranged position and the inclination of the exposure apparatus body. In the former prior art method, however, since the X-ray beams are twice reflected by the two X-ray reflecting mirrors, the X-ray beams are largely attenuated. However, in the latter prior art method, since the inclination of the X-ray reflecting mirror is offset with respect to the optimum inclination according to the location or inclination of the exposure apparatus body, the X-ray beams cannot be effectively utilized, and hence the function of the X-ray exposure apparatus utilizing the SOR cannot be adequately attained. SUMMARY OF THE INVENTION An object of the present invention is to substantially eliminate defects or drawbacks encountered in the prior art and to provide an X-ray exposure apparatus and a method of positioning the same, capable of easily achieving the coincidence of an axis of an X-ray beam generated with an optical axis of an alignment optical system of an exposure apparatus body by utilizing one reflecting mirror with an optimum reflecting angle being maintained. This and other objects can be achieved according to the present invention, in one aspect, by providing an X-ray exposure apparatus comprising an X-ray beam generating source such as a synchrotron, an exposure apparatus body disposed independently from the X-ray generating source, the exposure apparatus body including an outer casing defining an exposure chamber, an alignment optical system accommodated in the exposure chamber, an oscillation removing member disposed on a base for supporting the exposure apparatus body to be swingable thereto, a raising mechanism mounted on the oscillation removing member for raising the exposure apparatus body in a floating manner and adjusting a height position and an inclination thereof so that an optical axis of the alignment optical system substantially coincides with an axis of an X-ray beam generated from the X-ray source, and a securing mechanism for securing the exposure apparatus body after the adjustment in position and inclination thereof. In a preferred embodiment, the raising mechanism comprises an air cushion device secured to the oscillation removing member and expandable vertically so as to raise the exposure apparatus body in a floating manner, and the air cushion device comprises a plurality of air cushions disposed at predetermined portions on the oscillation removing member, the air cushions being adjustable in a desirable expanded height individually by adjusting an amount of air to be supplied thereinto. In another aspect, there is provided a method of positioning an X-ray exposure apparatus comprising an X-ray beam generating source and an exposure apparatus body supported on an oscillation removing member and including an alignment optical system, the method comprising the steps of raising the exposure apparatus body in a floating manner by a raising mechanism, adjusting a height position and an inclination of the exposure apparatus body by controlling the raising mechanism so that an axis of an X-ray beam generated from the X-ray generating source substantially coincides with an optical axis, as a measurement reference, of the alignment optical system, and securing the exposure apparatus body to the height position and the inclination adjusted by a securing mechanism. According to the present invention described above, the exposure apparatus body is set independently from the X-ray beam generating source and the exposure apparatus body is located to be swingable with respect to the base such as the floor through the elastic member. Thereafter, the exposure apparatus body is raised by the raising mechanism in a floating manner and the height position and the inclination of the exposure apparatus body are adjusted so that the axis of the X-ray beam generated from the X-ray source substantially coincides with the optical axis as measurement reference of the alignment optical system. The exposure apparatus body is then secured to this adjusted position by means of the securing mechanism. |
abstract | A storage area network (SAN) management application generates device allocation reports displaying foundation variables, device specific parameters, and computed, derived fields for different types of storage arrays, without burdening the allocation report with extraneous parameters through the use of a layout indicative of the information included on the report, providing a streamlined and seamless allocation report. The SAN management application defines a layout indicative of the foundation variables, device attributes, and derived fields requested in an allocation report. The user selected layout indicates the requested allocation parameters for a report, indicative of the foundation variable, device attributes, and derived fields, and also indicates the device usage metrics for computing the derived fields from the foundation variables and device attributes. In this manner, a SAN operator requests an allocation report indicative of only the information sought, and need not correlate multiple reports or manually synthesize report output for determining derived fields. |
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claims | 1. An apparatus for detecting properties of a sample, the apparatus comprising:an electron beam generator adapted to produce an electron beam and direct the electron beam at a desired point on the sample, the sample thereby emitting characteristic x-rays at takeoff angles,a diffractor adapted to receive and deflect the x-rays,means for preserving, at least in part, information indicative of the takeoff angles of the x-rays,a position sensitive detector adapted to receive the deflected x-rays and detect the information indicative of the takeoff angles of the x-rays and generate signals that are characteristic of the received x-rays, andan analyzer adapted to receive the signals from the detector and determine the properties of the sample based at least in part on the information indicative of the takeoff angles of the x-rays. 2. The apparatus of claim 1, wherein the means for preserving the information indicative of the takeoff angles of the x-rays comprises the diffractor having a cylindrically curved surface with an axis of curvature in a substantially horizontal plane, the diffractor adapted to receive the x-rays in a substantially vertical plane, and thereby deflect the x-rays along a substantially vertical plane, at angles that are dependent at least in part on the takeoff angles of the x-rays. 3. The apparatus of claim 1, wherein the means for preserving the information indicative of the takeoff angles of the x-rays comprises:a collimator adapted to receive and parallelize the x-rays and convert the takeoff angles of the x-rays to positional differences between the parallelized x-rays,the diffractor having a flat surface to preserve at least in part the positional differences between the parallelized x-rays as they are deflected, andthe detector adapted to sense in at least one dimension and report the positional differences between the parallelized x-rays. 4. The apparatus of claim 3, wherein the collimator includes a plurality of spaced and nested parabolic surfaces, where each of the parabolic surfaces has a common focal point, and the focal point coincides with the desired point on the sample where the electron beam is directed, and each of the parabolic surfaces receives x-rays within a given range of takeoff angles. 5. The apparatus of claim 1, further comprising a filter disposed between the sample and the detector, the filter adapted to substantially permit transmission of the x-rays while substantially prohibiting transmission of at least one of other energy, particles, and backscattered electrons from the sample to the detector. 6. The apparatus of claim 1, wherein the diffractor is one of a crystalline flat surface diffractor, a multilayer flat surface diffractor, a crystalline curved surface diffractor, a multilayer curved surface diffractor, a flat grating, and a curved grating. 7. The apparatus of claim 1, wherein the detector consists of at least one of a two dimensional array of detector elements, a charge coupled device containing a two dimensional array of pixels, a linear array of semiconductor detectors, a position sensitive proportional counter, and a multi-wire proportional counter. 8. The apparatus of claim 1, wherein the detector also detects positions of the x-rays, and the analyzer determines the properties of the sample based at least in part on the positional differences between the x-rays, the positions of the x-rays, and a number of x-rays impinging the detector at a given x-ray position per unit time. 9. The apparatus of claim 1, wherein the properties detected by the apparatus include elemental composition of the sample and thickness of the sample. 10. An apparatus for detecting properties of a sample, the apparatus comprising:an electron beam generator adapted to produce an electron beam and direct the electron beam at a desired point on the sample, the sample thereby emitting characteristic x-rays,a collimator adapted to receive and parallelize the x-rays and convert the takeoff angles of the x-rays to positional differences between the parallelized x-rays,a diffractor having a flat surface and adapted to receive and deflect the x-rays, while preserving, at least in part, the positional differences between the x-rays,a detector adapted to receive the x-rays from the diffractor and generate signals that are characteristic of the received x-rays, where the diffractor and the detector are mounted in a fixed relationship on a common rotatable stage that is rotatable relative to at least one of the electron beam generator, the sample, and the collimator, andan analyzer adapted to receive the signals from the detector and determine the properties of the sample based at least in part on the positional differences between the x-rays. 11. The apparatus of claim 10, further comprising a filter disposed between the sample and the detector, the filter adapted to substantially permit transmission of the x-rays while substantially prohibiting transmission of at least one of other energy, particles, and backscattered electrons from the sample to the detector. 12. The apparatus of claim 10, wherein the collimator includes a plurality of spaced and nested parabolic surfaces, where each of the parabolic surfaces has a common focal point, and the focal point coincides with the desired point on the sample where the electron beam is directed, and each of the parabolic surfaces receives x-rays within a given range of takeoff angles. 13. The apparatus of claim 10, wherein the detector is a position sensitive detector consisting of at least one of a two dimensional array detector elements, a charge coupled device containing a two dimensional array of pixels, a position sensitive proportional counter, and a multi-wire proportional counter, and the detector also detects positions of the x-rays, and the analyzer determines the properties of the sample based at least in part on the positional differences between the x-rays, the positions of the x-rays, and a number of x-rays impinging the detector at a given x-ray position per unit time. 14. The apparatus of claim 10, wherein the properties detected by the apparatus include elemental composition of the sample and thickness of the sample. 15. An apparatus for detecting properties of a sample, the apparatus comprising:an electron beam generator adapted to produce an electron beam and direct the electron beam at a desired point on the sample, the sample thereby emitting characteristic x-rays at takeoff angles,a first curved surface diffractor adapted to receive and deflect the x-rays that are received at low takeoff angles,a second curved surface diffractor adapted to receive and deflect the x-rays that are received at high takeoff angles,a first detector adapted to receive the x-rays from the first diffractor, and generate signals that are characteristic of the received x-rays,a second detector adapted to receive the x-rays from the second diffractor, and generate signals that are characteristic of the received x-rays, andan analyzer adapted to receive the signals from the first detector and the second detector and determine the properties of the sample, based at least in part on differences between the signals received from the first detector and the second detector. 16. The apparatus of claim 15, further comprising a filter disposed between the sample and the first and second detectors, the filter adapted to substantially permit transmission of the x-rays while substantially prohibiting transmission of at least one of other energy, particles, and backscattered electrons from the sample to the first and second detectors. 17. The apparatus of claim 15, wherein the first and second detectors are position sensitive detectors each consisting of at least one of a two dimensional array of detector elements, a charge coupled device containing a two dimensional array of pixels, a linear array of semiconductor detectors, a position sensitive proportional counter, and a multi-wire proportional counter, and the detectors also detects positions of the x-rays, and the analyzer determines the properties of the sample based at least in part on the positional differences between the x-rays, the positions of the x-rays, and a number of x-rays impinging the detectors at a given x-ray position per unit time. 18. An apparatus for detecting properties of a sample, the apparatus comprising:an electron beam generator adapted to produce an electron beam and direct the electron beam at a desired point on the sample, the sample thereby emitting characteristic x-rays at takeoff angles,a curved surface diffractor adapted to receive the x-rays within a desired solid angle of from about ten millisteradians to about fifty millisteradians, and deflect the x-rays,a detector adapted to receive the deflected x-rays and generate signals that are characteristic of the received x-rays, andan analyzer adapted to receive the signals from the detector and determine the properties of the sample. 19. The apparatus of claim 18, wherein the diffractor is adapted to receive the x-rays within the desired solid angle by at least one of enlarging a receiving surface area of the diffractor so as to receive the desired solid angle, and bringing the diffractor closer in proximity to the sample so as to receive the desired solid angle. 20. The apparatus of claim 18, further comprising a filter disposed between the sample and the detector, the filter adapted to substantially permit transmission of the x-rays while substantially prohibiting transmission of at least one of other energy, particles, and backscattered electrons from the sample to the detector. |
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summary | ||
claims | 1. A power module assembly comprising:a reactor vessel containing a reactor core surrounded by a primary coolant;a containment vessel submerged in a containment cooling pool, wherein the containment vessel is configured to prohibit a release of the primary coolant outside of the containment vessel;one or more inlets submerged in the containment cooling pool, wherein the one or more inlets are configured to draw a secondary coolant from the containment cooling pool during an emergency operation, wherein the emergency operation comprises a loss of secondary coolant flow;a heat exchanger in contact with the primary coolant, wherein the heat exchanger is configured to remove heat generated by the reactor core, and wherein the heat is removed by circulating the secondary coolant from the containment cooling pool through the heat exchanger via natural circulation; andone or more outlets submerged in the containment cooling pool, wherein the one or more outlets are configured to vent the secondary coolant into the containment cooling pool, and wherein the primary coolant remains completely retained within the containment vessel during the emergency operation. 2. The power module assembly according to claim 1, wherein the natural circulation comprises a circulation path of the secondary coolant entering the one or more inlets, passing through the heat exchanger, exiting the one or more outlets, and re-entering the one or more inlets. 3. The power module assembly according to claim 1, wherein the secondary coolant is vented into the containment cooling pool by the one or more outlets after passing through the heat exchanger. 4. The power module assembly according to claim 1, wherein the secondary coolant circulates through the heat exchanger by the natural circulation due to both a temperature difference between the secondary coolant and the primary coolant and a difference in elevation between the one or more inlets and the one or more outlets. 5. The power module assembly according to claim 1, wherein the one or more outlets are positioned at an elevation of the containment cooling pool that is above an elevation of the one or more inlets, and wherein the secondary coolant circulates through the primary coolant by the natural circulation due to an elevation difference between the one or more inlets and the one or more outlets. 6. The power module assembly according to claim 1, further comprising:an inlet line configured to deliver the secondary coolant from the one or more inlets to the heat exchanger; andan outlet line connecting the heat exchanger to the one or more outlets. 7. The power module assembly according to claim 5, wherein the natural circulation occurs as a result of a temperature change of the secondary coolant that is drawn and vented into the containment cooling pool, and wherein the natural circulation further occurs as a result of the elevation difference. 8. The power module assembly according to claim 6, wherein the secondary coolant is circulated from the containment cooling pool through the heat exchanger without any pump. 9. The power module assembly according to claim 6, further comprising an accumulator tank configured to inject auxiliary coolant into the inlet line when a loss of secondary coolant flow is detected, wherein the accumulator tank provides the auxiliary coolant to the heat exchanger until the natural circulation of the secondary coolant is established. 10. A power module assembly comprising:a reactor vessel containing a reactor core surrounded by a primary coolant;containment means configured to prohibit a release of the primary coolant into a surrounding containment cooling pool;means for removing heat from the reactor vessel, wherein the means for removing heat is at least partially surrounded by the primary coolant;means for drawing emergency feedwater from the containment cooling pool when a loss of feedwater condition is detected;means for venting the emergency feedwater into the containment cooling pool while the containment means continues to prohibit the release of the primary coolant into the containment cooling pool, wherein the means for venting is submerged in the containment cooling pool; andmeans for circulating the emergency feedwater from the containment cooling pool through the means for removing heat and back to the containment cooling pool, wherein the emergency feedwater is circulated through natural circulation. 11. The power module assembly according to claim 10, wherein the means for drawing is submerged in the containment cooling pool, and wherein the means for drawing is positioned at an elevation in the containment cooling pool that is lower than an elevation of the means for venting. 12. The power module assembly according to claim 11, wherein the natural circulation is due to an elevation difference between the means for drawing and the means for venting, and wherein the natural circulation is further due to a difference in density of the emergency feedwater that is vented from the means for venting and the emergency feedwater that is drawn from the containment cooling pool. 13. The power module assembly according to claim 10, wherein the means for circulating does not comprise a pump. 14. The power module assembly according to claim 10, wherein the means for drawing does not comprise a pump. 15. The power module assembly according to claim 1, wherein the containment vessel is internally dry before the emergency operation. 16. The power module assembly according to claim 10, wherein the containment means is internally dry before the loss of feedwater condition. 17. The power module assembly according to claim 10, wherein the emergency feedwater circulates through the means for removing heat by the natural circulation due to both a temperature difference between the emergency feedwater and the primary coolant and a difference in elevation between the means for drawing and the means for venting. 18. The power module assembly according to claim 10, wherein the means for venting are positioned at an elevation of the containment cooling pool that is above an elevation of the means for drawing, and wherein the emergency feedwater circulates through the primary coolant by the natural circulation due to an elevation difference between the means for drawing and the means for venting. |
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abstract | This disclosure relates to reactor control, and more particularly to a control rod drive mechanism and a reactor control system. The control rod drive mechanism includes a lifting-lowering assembly, a mounting assembly and a release assembly. The mounting assembly is configured to mount a control rod. The lifting-lowering assembly includes a fixing component, a scissor-type lifting-lowering mechanism and a lifting-lowering component. An end of the scissor-type lifting-lowering mechanism is connected to the fixing component, and the other end is connected to the lifting-lowering component. The scissor-type lifting-lowering mechanism is configured to drive the lifting-lowering component to move close to or away from the fixing component. The release assembly is movably arranged on the lifting-lowering component, and is detachably connected to the mounting assembly. The release assembly is configured to move relative to the lifting-lowering component when power is off to release the mounting assembly. |
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042960746 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to the removal of stainless steel, zirconium or zirconium alloy cladding materials from a metallic element selected from the group consisting of uranium, thorium and mixtures thereof. The present invention is particularly applicable to the selective destructive removal of cladding materials from nuclear fuel elements containing fissile or fertile fuels such as uranium, thorium and mixtures thereof. 2. Prior Art There are numerous types of nuclear power fuel elements. The present invention is particularly applicable to those nuclear fuel elements of the solid type which comprise a body or core of thermal neutron fissionable uranium, thorium or mixtures thereof which may be present in an elemental state or alloyed with zirconium, niobium or other low cross section materials which are clad in a low cross section corrosion resistant material such as stainless steel, zirconium, or zirconium alloys. Nuclear power fuel elements generally contain two types of nuclear fuel material, both of which are valuable. It is essential that the fuel element contain a fissionable nuclear fuel material such as uranium isotopes U 233 or U 235. Fuel elements also contain nuclear fuel materials that are not originally fissionable, but which can be converted to fissionable material and are, therefore, said to be fertile or potential nuclear fuel materials. For example, U 238 is a fertile material often present in fuel elements in considerable amounts. In some instances as much as 99.3% of the uranium content may be present in the form of U 238 in the case of an unenriched element. During the course of the use of the element in a power reactor, the fissionable material such as U 233 and U 235 releases neutrons. Some of the neutrons are trapped by the fertile but unfissionable U 238 present in the element and the U 238 eventually becomes Pu 239 which is fissionable. In the same way, thorium which is a fertile but unfissionable material, absorbs neutrons to become U 233 which is fissionable and useful as nuclear fuel material. Fuel elements of the solid type, with which the present invention is particularly applicable, deteriorate due to radiation damage long before the useful content of the fissionable material is used. At the same time, radioactive fission products accumulate in the fuel element. Some are gases and others are solid; however, each is objectionable in reducing the efficiency of the reactor as a whole and each exert some part in the destruction or disintegration of the fuel element. More particularly, many of the fission products have a high neutron capture cross section thus reducing the total amounts of neutrons available for production of thermal energy. In addition, the gaseous fission products build up pressure within the cladding material which can result in permanent structural damage to the elements and possibly to the reactor. Since these deleterious effects occur at a time when only a small fraction of the fissile values have been burned by the fission process and since the unburned fuel is too valuable to be wasted, it advantageously is reprocessed to render it fit for reuse. None of the heretofore known methods for recovering fuel and fertile uranium or thorium from such elements has been completely satisfactory. One method, for recovering unburned fissile and fertile fuel values from solid neutron irradiated fuel elements, involves dissolution of the cladding and the fuel followed by a liquid-liquid solvent extraction process in which an aqueous nitrate feed solution containing said values is selectively extracted by contact with an organic aqueous immiscible extractant. An example of a solvent extraction process for recovering uranium values, for example, is found in U.S. Pat. No. 2,848,300. A major disadvantage of aqueous dissolution of cladding, however, is that large aqueous feed volumes containing dissolved metals must be carried through the solvent extraction process. This in turn leads to a large radioactive waste volume requiring expensive waste storage and handling. In addition, the solutions generally are highly corrosive and have a high chloride content. Removal of chloride from the aqueous feed must be accomplished prior to solvent extraction for recovery of the thorium or uranium. In an attempt to reduce the volume of high level radioactive waste pollution, various other methods have been proposed, such as separately dissolving the cladding material in concentrated sulfuric acid thus making the fuel core available for ready dissolution in a nitric acid solution. However, a cladding material such as stainless steel is relatively passive in sulfuric acid and even when it does react, there is a high probability that cross contamination between the decladding solution and the core solution will result, thus further complicating the problem of recovering the fuel. U.S. Pat. No. 2,827,405 suggests a method of desheathing fuel rods of uranium metal bars by puncturing the sheath to expose the uranium core at a plurality of points. The rod then is reacted with steam at an elevated temperature to oxidize the uranium and break the bond between the sheath and the uranium. The fuel is recovered as an oxide requiring expensive processing to convert it back to a metal. Another method suggested in U.S. Pat. No. 2,962,371 comprises reacting the element at an elevated temperature with essentially pure anhydrous hydrogen for a time sufficient to hydride the cladding so that it falls from the core. This invention however, is concerned with zirconium-clad fuel elements although it is suggested that it is also applicable to elements that are clad in alloys of zirconium. Another process for recovering the core of a zirconium-clad fuel element is disclosed in U.S. Pat. No. 3,007,769. The process comprises immersing the clad element in a substantially neutral solution of ammonium fluoride to effect the dissolution of the zirconium and separate the neutron fissionable material values from the solution. U.S. Pat. No. 3,089,751 suggests a process for the selective separation of uranium from ferritic stainless steels. In accordance with the process disclosed therein, a nuclear fuel element consisting of a core of uranium clad in a ferritic stainless steel is heated to a temperature in the range of 850.degree. C. to 1050.degree. C. for a period of time sufficient to render the cladding susceptible to intergranular corrosion. The heated element is then cooled rapidly to a temperature range of 850.degree. C. to 615.degree. C. and then to about room temperature. The cooled element then is contacted with an aqueous nitrate solution to selectively and quantitatively dissolve the uranium from the core. Gas phase processes for effecting the dissolution of fuel or the cladding material are disclosed in U.S. Pat. Nos. 3,149,909; 3,156,526 and 3,343,924. The problem of handling and containing gaseous fuel, however, is even greater than that for liquid phase processes. U.S. Pat. No. 3,929,961 suggests a method of treating a nuclear fuel element enclosed in a stainless steel metal sheath which comprises disposing the fuel element with a portion thereof in an induction coil, subjecting the induction coil to a radio frequency magnetic field to induce local induction heating of the metal sheath sufficient to raise the temperature of the portion of the sheath within the coil to its melting temperature and effect local melting therein. The fuel element is moved axially relative to the induction coil with continued heating to rupture the metal sheath. The fuel values are subsequently recovered by dissolution. Thus it is seen that the prior art processes either convert the fuel to an oxide or at some point require liquid or gas phase processing with all of the problems associated therewith. SUMMARY OF THE INVENTION The present invention provides a method of treating an assembly comprising an element selected from the group consisting of uranium, thorium and mixtures thereof encased in a cladding of stainless steel or a zirconium alloy to separate the selected element from the cladding. In accordance with the present method, the assembly is subjected to a scoring or perforating step to expose the selected element. Thereafter, the assembly is exposed to hydrogen at a pressure of from about 0.5 to 2.0 atmospheres (360 to 1400 torr) and a temperature of 450.degree. C. to 680.degree. C. to form a hydride of the element. The hydride, having a greater volume than the elemental metal, expands, rupturing the cladding material. Thereafter, the temperature is further increased to a range of from about 700.degree. C. to 900.degree. C. to decompose the hydride back to the element. The dehydriding results in the element being in the form of friable particulates such that after at least one and preferably after about three successive hydriding-dehydriding steps, the selected element is readily recoverable from the cladding material, utilizing conventional mechanical separation techniques such as sieving or the like. In a particularly preferred embodiment of the invention, during the hydriding step, the temperature is cycled between about 500.degree. C. and 650.degree. C. to enhance the completeness of the hydriding and maximize the removal or evolution of any volatile compounds contained within the assembly. The present invention is particularly applicable to the treatment of irradiated fuel elements for the recovery of fissionable and fertile values therefrom. |
summary | ||
claims | 1. A method of implanting ions into a substrate while changing a relative positional relation between an ion beam and the substrate, wherein a first ion implantation process in which a uniform dose amount distribution is formed within the substrate and a second ion implantation process in which a non-uniform dose amount distribution is formed within the substrate are performed in a predetermined order, and a cross-sectional size of an ion beam that is irradiated on the substrate during the second ion implantation process is set smaller than a cross-sectional size of an ion beam that is irradiated on the substrate during the first ion implantation process. 2. The method of implanting the ions according to claim 1, wherein, when processing a plurality of substrates using the first and second ion implantation processes, one of the ion implantation processes is successively performed on the substrates, and thereafter, the other ion implantation process is successively performed on the substrates. |
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claims | 1. A radiation protection shield for protecting medical personnel from radiation being applied to a patient positioned on a table, the shield comprising:a frame having a plurality of vertical supports that divide a radiation-source side of the frame from a user side of the frame;a primary screen including a radiation-resistant material, the primary screen connected to said frame;a secondary screen slidably connected to the primary screen so that the secondary screen is moveable along a plane defined by the plurality of vertical supports to one of a retracted configuration and an extended configuration; andan upper flange extending (i) across at least a portion of the frame and (ii) from a top region of the secondary screen and towards the radiation-source side of the frame, the upper flange effective in reduction of radiation scatter;wherein the upper flange effectuates reduction of radiation scatter in both of the first retracted configuration and the extended configuration; the shield further comprising a shelf having (i) a shelf outer end on a side of the shelf adjacent to an outer end of the frame and (ii) a shelf inner end on a side of the shelf opposite to the shelf outer end, the shelf having a varying width that is wider at the shelf outer end than the shelf inner end. 2. The radiation protection shield as set forth in claim 1, wherein the primary screen includes a lead-impregnated acrylic. 3. The radiation protection shield as set forth in claim 1 further comprising:a plurality of rollers attached to a bottom of the frame to allow a lower shield to move along a floor that the shield is positioned on during use of the radiation protection shield. 4. The radiation protection shield as set forth in claim 3, wherein said rollers are lockable to restrict the rollers from moving thereby limiting movement of the lower shield. 5. The radiation protection shield as set forth in claim 1, wherein the secondary screen includes a lead-impregnated acrylic. 6. The radiation protection shield as set forth in claim 1, wherein:the primary screen and secondary screen are curved forming an inner concave surface on the radiation-source side of the frame and an outer convex surface opposite the concave surface on the user side of the frame; andthe inner surface is positioned adjacent a radiation source during use of the system. 7. The radiation protection shield as set forth in claim 1 further comprising:a hand rail connected to said primary screen to facilitate moving of a lower shield. 8. The radiation protection shield as set forth in claim 1 further comprising:a toe flange extending downward from the primary screen. 9. The radiation protection shield as set forth in claim 1 further comprising:a kick plate positioned adjacent a lower edge of the shield for protecting the primary screen and for medical personnel to use to move a lower shield. 10. The radiation protection shield as set forth in claim 1 further comprising:an upper flange extending from a top region of the shield to form a right angle with the primary screen. 11. The radiation protection shield as set forth in claim 1, further comprising:an upper flange extending from a top region of the shield to form a right angle with the secondary screen. 12. The radiation protection shield as set forth in claim 11 further comprising:an upper shelf extending from the top region of the shield adjacent said upper flange to form a right angle with the secondary screen. 13. The radiation protection shield as set forth in claim 12, wherein said upper shelf extends farther from the secondary screen than the upper flange extends from the secondary screen. 14. The radiation protection shield as set forth in claim 12, wherein the upper shelf is positioned adjacent an end of said shield. 15. The radiation protection shield as set forth in claim 14 further comprising:an upper shelf extending from a top region of the shield. 16. The radiation protection shield as set forth in claim 15, wherein said shelf includes a lead-impregnated acrylic. 17. The radiation protection shield as set forth in claim 1,wherein,the secondary screen is connected to the plurality of vertical supports; andthe frame includes sleeves receiving the plurality of vertical supports. 18. The radiation protection shield as set forth in claim 17 further comprising:springs attached to the frame adjacent said sleeves for providing an upward force to the secondary screen by way of said rods. 19. The radiation protection shield as set forth in claim 17 further comprising:a locking mechanism attached to the frame adjacent said sleeves for selectively locking the rods in place in the sleeves thereby locking the secondary screen in place with respect to the primary screen. 20. A radiation protection shield for protecting medical personnel from radiation being applied to a patient positioned on a table, the shield comprising:a frame having a radiation-source side and a user side;multiple screens telescopically connected to each other in a vertical direction along the frame so that an overall height of the shield can be selectively adjusted during use of the shield,an upper flange extending (i) across a first portion of the frame and (ii) from a top region of the secondary screen and towards the radiation-source side of the frame, the upper flange effective in reduction of radiation scatter, anda shelf extending (i) across a second portion of the frame and (ii) from the top region of the secondary screen and towards the radiation-source side of the frame, the shelf having (i) a shelf outer end on a side of the shelf adjacent to an outer end of the frame and (ii) a shelf inner end on a side of the shelf opposite to the shelf outer end, the shelf having a varying width that is wider at the shelf outer end than the shelf inner end. 21. A radiation protection shield as set forth in claim 1, further comprising:a toe drape angled toward the radiation-source side of the frame. |
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050698630 | summary | CROSS-REFERENCE TO RELATED APPLICATIONS None. BACKGROUND OF THE INVENTION The present invention relates to nuclear power plants and more particularly to fuel transfer systems employed therein during refueling to transfer fuel assemblies between the containment building and the storage pool. When a nuclear reactor is shut down for refueling, fully or partially spent fuel assemblies are lifted from the reactor core in the containment building and moved through a pool of water to a fuel transfer system that transfers the assemblies usually one or two at a time to the auxiliary building where other apparatus takes the assemblies and deposits them in the pool storage area. New fuel assemblies or partially spent assemblies are carried by the fuel transfer system from the storage pool area to the containment building where they are placed in the reactor core. Generally, the fuel transfer operation takes place under water to limit radiation exposure. Nonetheless, it is desirable that the fuel transfer system be economic in manufacture and both effective and reliable in operation to provide the required fuel assembly transport performance. Typically, a stainless steel tube having a diameter between 20 inches and 36 inches provides a horizontal transfer path between the containment building and the spent fuel storage area. A transfer car may operate on a pair of spaced rails over the transfer path. A basket is provided on the car to carry the fuel assembly during transfer. The basket may be an end pivot type in which case the basket is turned on its end to an upright position above the car for loading and unloading a fuel assembly at each end of the car travel. In this case, the space below the car and between the rails can be occupied with system structure since the basket does not pass through the undercar space when it is turned to the vertical position. The basket may also be a center pivot type and this is normally the preferred scheme since the fuel assemblies can be upended at both ends of car travel with reduced loading on the upending mechanism. However, the center pivot basket does require undercar clearance space for the basket half that moves under the car when the basket is turned to the vertical position. Since the car must operate in both directions along the transfer path, it is necessary that the drive system for the car provide driving force in either of the two directions. Further, it is desirable that the drive system be structured so that it is reliable for underwater operation. One drive system architecture is the bilateral type and it involves placing a drive unit on the containment side of the containment wall to provide pulling force that directs the car away from the storage area and toward the containment building. Another drive unit located near the storage area provides pulling force that directs the car away from the containment building and toward the storage area. The fact that a drive unit must be located on opposite sides of the containment wall in this scheme is disadvantageous from a service and maintenance standpoint since a cable hookup must be provided from the drive unit to the car from the containment side before the fuel transfer operation can be started after a plant shutdown. Another drive system architecture is the unilateral type and it involves apparatus advantageously located only on the storage area side of the containment wall. The drive system is organized so that it provides drive force for directing the car in either direction over the fuel transfer path. In one prior art bilateral scheme, a fixed chain is welded to the bottom of the car midway between the rails and it is driven in either direction by a sprocket which in turn is driven by an underwater air motor. Another prior bilateral scheme involves a continuous chain that runs along the center of the track. It is linked to the car and directly driven by a drive shaft of an underwater motor or indirectly driven through a coupling by a drive shaft of an electric motor above the water level. Again, the basket is end pivoted. A variation on this scheme involves a pair of continuous chains located outside the rails so that a center pivoted basket may be employed. A prior unilateral scheme involves use of a fixed drum in the storage area and a drum on the car interconnected by cable. The car drum is coupled through a sprocket that engages pins on the rails to propel the car in one direction or the other. When the cable is pulled in one direction, the car drum is turned in one direction to propel the car in one direction along the track. When the cable is pulled in the opposite direction, the car is pulled in the opposite track direction and the sprocket rewinds the cable on the car drum. A center pivoted basket is used, but this scheme suffers from unreliability from a number of sources including the pin and sprocket drive arrangement. In all of these schemes, underwater limit switches are normally required for system operation to enable the car to be brought to a controlled stop at its ends of travel. System reliability is accordingly adversely affected because the underwater limit switches are prone to leak over time. The present invention is directed to a fuel transfer system having its drive system located on one side of the containment wall, preferably externally of the containment building. The drive is preferably structured for a center-pivoted car basket and otherwise for fuel transfer operation with significantly improved reliability. SUMMARY OF THE INVENTION A fuel transfer system provides for moving fuel assemblies along a track running between the auxiliary building side and the containment building side of a containment wall in a nuclear power plant. The track extends through a transfer tube within the containment wall. The system comprises a car having wheels for movement along spaced rails of the track and a carrying basket for one or more fuel assemblies. Winch means are located on the auxiliary building side of the containment wall and above the water level existing over the track during refueling operations to drive the car along the track. First cable means and second cable means extend substantially vertically downward from the winch means to the track level. First sheave means direct the first and second cable means substantially in the horizontal direction along the track. Means are provided for securing the first cable means to the car so that winch pulling force on the first cable means drives the car away from the containment building. Second sheave means are located near the containment end of the transfer tube. The second cable means extend substantially horizontally along the track from the first sheave means to the second sheave means where it is redirected to extend substantially horizontally in the reverse direction along the track. Means are provided for securing the second cable means to the car so that winch pulling force on the second cable means drives the car toward the containment building. The winch means are operated to pull the cable means so as to move the car selectively between one end position in which the car is within the auxiliary building for fuel assembly loading and unloading and the other end position in which the car is principally located in the containment building with at least a cable securance portion of the car located over the track within the transfer tube and to the auxiliary building side of the second sheave means. Both of the cable securing means are located on the cable securance car portion. |
043127045 | abstract | Grill gates respectively formed by oppositely moving, interfitting rod combs with gaps between the rods which are smaller than the particle diameter are so arranged so that each comb may be individually moved to the open position and shut again without allowing any of the even sized particles of the bulk material to pass through the other comb while testing the readiness of the comb operating system. Grooves of circular arc cross-section fitting the size of the spheres of a high-temperature reactor are provided in the upstream sides of the wall produced by the interfitting combs to enable close packing of the spheres against the closure wall in an ordered arrangement that allows either comb to be withdrawn and reinserted without breakage of the spherical elements. The grooves are oriented in the length dimension of the comb rods, and one set of comb rods provides the groove bottoms and the adjoining rods of the other set the cooperating groove flanks. The outer rods of one of the combs carries a more or less semi-circular plate fitting the conduit wall through which plate slots are provided for passage of the rods of the other comb. |
053234278 | claims | 1. A nuclear reactor containment arrangement, comprising: a reactor vessel having a peripheral wall and a horizontally extending annular flange thereon; a containment wall spaced apart from and surrounding the peripheral wall of the reactor vessel and defining an annular gap therebetween; and an annular ring seal extending across the annular gap and providing a water-tight seal therebetween, characterized by: a reactor vessel having a peripheral wall and a horizontally extending annular flange thereon; a containment wall spaced apart from, and surrounding the peripheral wall of the reactor vessel and defining an annular gap therebetween; and a annular ring seal extending across the annular gap and providing a water-tight seal therebetween, characterized by: an annular main seal plate characterized by inner and outer edges spaced radially apart, spaced apart surfaces extending between the inner and outer edges, the main seal plate extending substantially horizontally across a portion of the annular gap; inner connection means for supporting the main seal plate on the annular flange with accommodation for radial thermal expansion and contraction of the reactor vessel, and providing a water-tight seal between the main seal plate and the annular flange; and outer connection means for supporting the main seal plate by the containment wall with a water-tight seal and with accommodation for axial thermal expansion and contraction and lateral movement of the reactor vessel; the outer connection means including a first Belleville plate connected to the main seal plate, a second Belleville plate connected to the containment wall, the first and second Belleville plates accommodating for axial thermal expansion and contraction of the reactor vessel, and means connected between the first and second Belleville plates accommodating for lateral movement of the reactor vessel. 2. The containment arrangement of claim 1, characterized in that the inner flexible sealing means comprises a sealing member, the sealing member being characterized by an annular horizontal portion sealingly affixed to the upper surface of the annular flange and a cylindrical portion extending upwards from the horizontal portion and sealingly affixed to the main seal plate proximate the inner edge of the main seal plate. 3. The containment arrangement of claim 2, characterized in that the outer sealing means comprises an attachment plate sealingly affixed to the containment wall, and the second annular Belleville plate is sealingly affixed at its outer edge to the attachment plate. 4. The containment arrangement of claim 3, characterized in that the containment wall comprises vertically extending upper and lower portions, the upper portion having a greater diameter than the lower portion, and a ledge therebetween, in that the attachment plate comprises a cylindrical portion extending down from the ledge along the lower portion of the containment wall and an annular portion positioned on the ledge, and in that the outer edge of the second annular Belleville plate is sealingly affixed to the cylindrical portion of the attachment plate. 5. A nuclear reactor containment arrangement, comprising: 6. The containment arrangement of claim 5, characterized in that the inner connection means comprises a cylindrical support affixed to an upper surface of the annular flange and extending upward therefrom and upon which the main seal plate rests, and flexible sealing means radially inside the cylindrical support and extending between the main seal plate and the annular flange for providing a watertight seal therebetween. 7. The containment arrangement of claim 5, characterized in that the means accommodating for lateral movement of the reactor vessel comprises a corrugated cylinder characterized by axially spaced apart first and second edges and radial corrugations extending longitudinally between the first and second edges. 8. The containment arrangement of claim 7, characterized in that the outer connection means further includes an annular protrusion extending from one of the surfaces of the main seal plate proximate the outer edge of the main seal plate and providing a surface for attachment of the first Belleville plate. 9. The containment arrangement of claim 7, characterized in that the inner connection means comprises a cylindrical support affixed to an upper surface of the annular flange and extending upward therefrom and which supports the main seal plate, and flexible sealing means proximate the cylindrical support and extending between the main seal plate and the annular flange for providing a water-tight seal therebetween. 10. The containment arrangement of claim 9, characterized in that the flexible sealing means comprises a sealing member characterized by an annular horizontal portion sealingly affixed to the annular flange and a cylindrical portion extending upwards from the horizontal portion and sealingly affixed to the main seal plate. |
abstract | In one characterization, the present invention relates to a radiation-shielding assembly for holding a container having a radioactive material disposed therein. The assembly may, at least in one regard, be referred to as an elution shield and/or a dispensing shield. The assembly includes a body at least partially defining a cavity. There is at least one opening through the body into the cavity. The assembly may include a cap that at least generally hinders escape of radiation from the assembly through the opening. The cap may be releasably attached to the body in one orientation and may establish non-attached engagement with the body in another orientation. The assembly may include an adjustable spacer system for adapting the assembly for use with containers having different heights. |
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043371671 | claims | 1. A container for radioactive nuclear waste materials which are ultimately to be buried underground, said container being composed of a native nickel-iron alloy produced under thermodynamically stable conditions and possessing a composition in the range exhibited by the natural materials awaruite and josephinite. 2. A container as recited in claim 1 wherein said alloy is selected from the group consisting of awaruite and josephinite. 3. A container as recited in claim 2 wherein said alloy is awaruite. 4. A container as recited in claim 2 wherein said alloy is josephinite. 5. A container as recited in claim 1 wherein the nickel content of said alloy is in the range 60-90 percent and the iron content of said alloy is in the range 10-40 percent. 6. A container as recited in claim 5 wherein said alloy also contains up to 5 percent cobalt. 7. A container as recited in claim 6 wherein said alloy also contains up to 5 percent copper. 8. A container as recited in claim 5 wherein said alloy also contains up to 5 percent copper. 9. A container as recited in claim 1 wherein said alloy is composed of the stoichiometric alloy phase Ni.sub.3 Fe. 10. A method of containing radioactive nuclear waste materials over extended periods of time, said method comprising the steps of: (a) encapsulating the waste materials in a container composed of a native nickel-iron alloy produced under thermodynamically stable conditions and possessing a composition in the range exhibited by the natural materials awaruite and josephinite and (b) burying the container underground in an impervious, stable rock formation. (1) encapsulating the waste materials in a container composed of a nickel-iron alloy having the properties of those natural minerals produced under thermodynamically stable conditions within serpentinite-type rocks and possessing a composition in the range exhibited by the mineral awaruite and (2) burying the container underground in an impervious stable rock formation. 11. A container for radioactive nuclear waste materials which are ultimately to be buried underground, said container being composed of a nickel-iron alloy having the properties of those natural minerals produced under thermodynamically stable conditions within serpentinite-type rocks and possessing a composition in the range exhibited by the mineral awaruite. 12. A container as recited in claim 11 wherein said alloy is awaruite. 13. A container as recited in claim 11 wherein the nickel content of said alloy is in the range 60-90 percent and the iron content of said alloy is in the range 10-40 percent. 14. A container as recited in claim 13 wherein said alloy also contains up to 5 percent cobalt. 15. A container as recited in claim 14 wherein said alloy also contains up to 5 percent copper. 16. A container as recited in claim 13 wherein said alloy also contains up to 5 percent copper. 17. A container as recited in claim 11 wherein said alloy is composed of the stoichiometric alloy phase Ni.sub.3 Fe. 18. A method of containing radioactive nuclear waste materials over extended periods of time, said method comprising the steps of: |
055685285 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates generally to compensation of a rod position indication system of a nuclear reactor vessel which system includes a linear variable differential transformer (LVDT) having a generally linear output that is representative of the position of a control rod and, more particularly, to compensating such a rod position indication system for any non-linearity in the LVDT output. 2. Description of the Related Art In a commercial nuclear reactor, heat, from which steam and ultimately electricity are generated, is produced by fissioning of a fissible material such as enriched uranium. This fissible material, or nuclear fuel, is typically contained within a nuclear core made up of a multiplicity of fuel rods supported in a plurality of nuclear fuel assemblies, coextensively arranged in a spaced parallel array. Movable control rods are dispersed throughout the core to control the fission process. The control rods generally comprise a plurality of elongated rods containing neutron absorbing materials which fit in longitudinal openings defined in the fuel assemblies and among the fuel rods by guide thimbles of the fuel assemblies. The guide thimbles thus guide the control rods during their movement into and out of the core. Inserting a control rod into the core adds more absorber material and, hence, decreases the nuclear reaction; conversely, withdrawing a control rod removes absorber material and, hence, increases a nuclear reaction and thereby the power output of the core. The nuclear reactor core and the control rods are positioned within and supported by a reactor vessel through which a reactor coolant flows. The control rods are supported in cluster assemblies moved into and from the nuclear core by control rod drive mechanisms which, in turn, are mounted by an upper internals arrangement located within the nuclear reactor vessel above the nuclear core. Typically, a reactor pressure vessel is pressurized to a relatively high internal pressure. The control rod drive mechanisms operate within the same pressure environment that exists within the reactor pressure vessel. Hence, the control rod drive mechanisms are housed within pressure housings of the upper internals arrangement which are tubular extensions of the reactor pressure vessel. One of the more commonly used types of control rod drive mechanisms is referred to as a "magnetic jack." With this type of mechanism, the control rods are jacked into and from the nuclear core in a series of motions each involving moving the control rod a discrete incremental distance or "step;" hence, such movement is commonly referred to as stepping of the control rods. There are typically 231 steps between the fully withdrawn position and the fully inserted position of the control rods. For example, 0 steps indicate the fully inserted position, and 231 steps indicate the fully withdrawn position. This type of mechanism is illustrated and described in U.S. Pat. Nos. to Frisch (3,158,766) and Dewesse (3,992,255) which are assigned to the assignee of the present invention. This magnetic jack type of control rod drive mechanism includes three electromagnetic coils and armatures or plungers which are operated to raise and lower a drive rod shaft and thereby the control rod cluster assembly. The three coils are mounted about and outside of the pressure housing. Two of the coils actuate respective plungers of movable and stationary grippers contained within the housing. The third coil actuates a lift plunger connected to the movable gripper. Actuation of the movable and stationary plungers, in turn, operate sets of circumferentially spaced latches which grip the drive rod shaft having multiple axially-spaced circumferential grooves. The stationary gripper latches are actuated to hold the drive shaft in a desired axial position. The movable gripper latches are actuated to raise and lower the drive rod shaft. Each jacking or stepping movement of the control rod drive mechanism moves the drive rod shaft 5/8 inch (1.58 cm) The jacking or stepping movement is thus accomplished by the operation of the three sets of axially spaced electromagnetic coils to actuate the corresponding stationary, movable and lift plungers so as to alternately and sequentially grip, move and release the control rod drive shaft of the respective mechanism. A number of indicators have been used in the past to determine control rod position. One such indicator is an analog indicator. This analog indicator includes a plurality of layered, wound coils concentrically arranged in a stack and supported by a nonmagnetic stainless steel tubular substructure that is slid over a nonmagnetic rod travel housing. The coils are arranged alternately as primary and secondary coils, with all the primary coils connected in series and all the secondary coils connected in series. The coils form, in effect, a long linear voltage transformer distributed over the height of the travel housing such that the coupling from primary to secondary is affected by the extent to which the magnetic drive rod penetrates the coil stack. Rod position is determined by applying a constant sinusoidal excitation current to the primary and measuring the voltage induced across the secondary. The magnitude of the induced secondary voltage corresponds to rod position. The secondary output voltage varies substantially linearly with the position of control rod, as is well known in the art. This secondary voltage is processed by instrumentation, which is well known in the art, and displayed on a control panel. Although the present device for detecting control rod position is satisfactory, it is not without drawbacks. When the control rods are near the fully withdrawn position (i.e., 231 steps), the output of the transformer varies substantially non-linearly with the position of the control rods, which obviously creates an error in the indication of the position of the control rod. Consequently, a need exists for a method and apparatus for compensating a rod position indication system for non-linearity when the control rods are substantially withdrawn. SUMMARY OF THE INVENTION The present invention provides an improvement designed to satisfy the aforementioned needs. Particularly, the present invention is directed to a method for compensating a rod position indication system for non-linearity comprising the steps of (a) applying an excitation current to a primary of a sensor for inducing a generally linear voltage that is representative of the position of a control rod on a secondary of the sensor, wherein the excitation current includes a first frequency sufficient to provide the generally linear output; and (b) modifying the excitation current to include a second frequency sufficient for providing the generally linear output when the first frequency is insufficient to provide the generally linear output. It is an object of the present invention to provide a method for compensating a rod position indication system for non-linearity. It is a feature of the present invention to provide a current source having a 60 cycle frequency for exciting the sensor over a predetermined range of rod positions, and having a 20 cycle frequency for exciting the sensor over the remaining range of control rod positions. |
abstract | The invention comprises a segmented rolling floor apparatus and method of use thereof, such as for use in a charged particle cancer therapy system. The segmented rolling floor comprises a first spool and a second spool, attached to opposite ends of the rolling floor, which cooperatively wind and unwind the rolling floor. The segmented rolling floor circumferentially surrounds a nozzle system penetrating through an aperture in the segmented rolling floor, where the nozzle system is used to deliver charged particles, from an accelerator, to a tumor of a patient. The rolling floor and nozzle systems move at respective rates maintaining the nozzle system in the aperture allowing for a safe/walkable floor while allowing treatment of the tumor as a gantry rotates the nozzle system and delivers protons to the tumor from positions above and below the floor. |
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abstract | An electron beam apparatus equipped with a height detection system includes an electron beam unit emitting an electron beam to the specimen, and a height detection system for detecting height of the specimen which is set on a table. The height detection system includes an illumination system configured to direct first and second beams of light through a mask with a multi-slit pattern to a surface of the specimen at substantially opposite azimuth angles and at substantially equal angles of incidence, first and second detectors which respectively detect first and second multi-slit images of the first and second beams reflected from the specimen and generate output signals thereof, and a device which receives the output signals and generates a comparison signal which is responsive to the height of the specimen. An objective lens of the electron beam unit is controlled in accordance with the comparison signal. |
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description | The present invention will be described in detail with reference to a few specific embodiments thereof as illustrated in the accompanying drawings. In the following description, numerous specific details are set forth in order to provide a thorough understanding of the present invention. It will be apparent, however, to one skilled in the art, that the present invention may be practiced without some or all of these specific details. In other instances, well known process steps have not been described in detail in order to not unnecessarily obscure the present invention. In general terms, X-ray micrography includes analysis of the X-ray region of the electromagnetic spectrum that is associated with a sample so as to gain information regarding the sample. The X-ray region of the electromagnetic spectrum includes frequencies that range from 1.0xc3x971017 Hz to 1.0xc3x971021 Hz. X-ray micrography is performed by bombarding a specimen with electrically charged particles which have sufficient energy to cause X-ray photons to be emitted from the specimen. By counting the emitted photons at one or more energy levels, the composition and thickness of conductive layers in a semiconductor device may be determined. The composition of a material may be determined since the specific energy levels of X-ray photons emitted from such material are related to the material""s composition. For example, the thickness of a conductive layer may be determined by taking the integral of the number of X-ray photons around the characteristic energy emission levels. The integral taken is directly proportional to the thickness of the conductive layer. The actual thickness of a metal layer is then determined by applying a calibration factor, which is specific to each type of material being measured, to the count data. Procedures for determining composition and thickness of layers within a film stack are described further below with reference to FIGS. 7 through 10. FIGS. 1, 2A, and 2B illustrate generally how X-ray photons may be caused to emanate from a specimen in accordance with one implementation of the present invention. FIG. 1 illustrates a charged particle 10, such as an electron, colliding with the surface of a specimen 12. The charged particle may collide with an atom of the specimen 12 and thereby causes an X-ray photon 14 to be emitted from the specimen 12. FIGS. 2A and 2B illustrate this collision event at the atomic level. FIG. 2A shows an atom 16 of the specimen 12. The atom 16 has a nucleus 18 surrounded by electrons 20 at different discrete distances from nucleus 18 called electron shells. A given electron shell has a binding energy level equal to the amount of energy required to remove an electron from the electron shell. The binding energy level of an electron shell is inversely proportional to the distance of the electron shell from the nucleus. The innermost electron shell of an atom is called the K shell, and has the highest binding energy associated with it. In FIG. 2A, K-shell electron 22 is located in K shell 24. The two shells beyond the K shell are the L and M shell, the M shell being the farthest away from the nucleus 18. FIG. 2A also shows the charged particle 10 impacting atom 16 within the specimen 12. If the energy level of the particle 10 is greater than the binding energy level of a K-shell 24, the entire energy of the particle 10 is absorbed by atom 16 and one of the electrons in K shell 24 is ejected from the atom 16. As depicted in FIG. 2A, K-shell electron 22 is ejected from atom 16 after particle 10 is absorbed by atom 16. With a vacancy in K shell 24, atom 16 is energetic and unstable. The most probable stabilization mechanism is the filling of the vacancy in K shell 24 by an electron located in an electron shell with a lower binding energy level. As shown in FIG. 2B, an L-shell electron 26 in L shell 28, which is farther from nucleus 18 than K shell 24, may fill the vacancy in K shell 24. As L-shell electron 26 fills the vacancy in K shell 24, atom 16 may simultaneously emit an X-ray photon 14 with energy (NKxe2x88x92NL), where NK and NL are the binding energy levels of K and L shell, respectively. With a vacancy now in L shell 28, ionized atom 16 is more stable and less energetic. The above described X-ray emission theory may be utilized to determine the composition and thickness of films within a film stack since X-ray emissions from the film stack depend on the film stack""s composition and thickness values. That is, each material type has an associated atomic shell arrangement. Additionally, the binding energy level for each shell varies with material type. For example, a first material type emits X-rays at energy levels that differ from energy levels of a second material type. In other words, each material type has associated X-ray energy peaks at which X-rays are expected to be emitted. Thus, an unknown material""s composition may be determined by comparing the unknown material""s X-ray energy peaks with a known material""s energy peaks. When a match is found, the unknown material is identified as having the same composition as the matching material. FIG. 3 illustrates a system utilizing an electron beam induced X-ray microanalysis test system according to one embodiment of the present invention. The system represented in FIG. 3 includes a beam generator 400, which directs a charged particle beam at the specimen 330. The typical spot size of the system is approximately 10 microns in diameter. However, the spot size may range between 1-100 microns in diameter. The specimen 330 is a multi-layered semiconductor wafer for which layer thickness and composition measurements are desired. X-ray detectors 500 are positioned above the specimen 330 in order to collect the X-ray photons emitted from the specimen 330. Each of the X-ray detectors are coupled with an analysis unit 320. The analysis unit 320 can be configured to analyze the data collected by the X-ray detectors 500 and to generate useful information concerning the individual layers. The analysis unit 320 may take the form of any suitable processing or computing system, such as a workstation. The beam generator 400 may be any suitable device that directs charged particles towards a specimen, and which, in turn, causes X-rays to emanate from the sample under test. The generator 400 is capable of projecting the charged particles with sufficient energy to penetrate at least two layers of the film stack on a integrated circuit. By way of example, a conductive layer and a liner layer may be two of the film stack layers penetrated by the beam generator 400. Preferably, the particles penetrate substantially through the entire conductive and liner layers so as to cause X-rays to emanate from the entire width of the respective layers. As a result, X-ray measurements from the entire thickness of the penetrated layers may then be taken. By way of example, the beam generator 400 may take the form of a scanning electron microscope (SEM). FIG. 4 is a diagrammatic representation of a typical scanning electron microscope (SEM) system 400. As shown, the SEM system 400 includes an electron beam generator (402 through 416) that generates and directs an electron beam 401 substantially toward an area of interest on a specimen 424. The SEM system 400 also includes a detector 426 arranged to detect charged particles 405 (secondary electrons and/or backscattered electrons) emitted from the specimen 424. As shown, the electron beam generator includes an electron source unit 402, an alignment octupole 406, an electrostatic predeflector 408, a variable aperture 410, a wien filter 414, and a magnetic objective lens 416. The source unit 402 may be implemented in any suitable form for generating and emitting electrons. For example, the source unit 402 may be in the form of a filament that is heated such that electrons within the filament are excited and emitted from the filament. The octupole 406 is configured to align the beam after a particular gun lens voltage is selected. In other words, the beam may have to be moved such that it is realigned with respect to the aperture 410. The aperture 410 forms a hole through which the beam is directed. The lower quadrupole 408 may be included to compensate for mechanical alignment discrepancies. That is, the lower quadrupole 408 is used to adjust the alignment of the beam with respect to any misaligned through-holes of the SEM through which the beam must travel. The magnetic objective lens 416 provides a mechanism for accelerating the beam towards the sample. Finally, the Wien filter 414 deflects secondary electrons towards the detector 426. The specimen or sample 330 may take a variety of forms for which certain measurements are desired. Specifically, the electron beam induced X-ray microanalysis test system may be used to measure the film characteristics of various thin film devices such as a semiconductor wafer or a magnetic recording head. In the illustrated embodiment, the specimen 330 is a semiconductor wafer which contains alternating conductive 340, dielectric 360, and liner 350 layers formed on a substrate of semiconductor material. The wafer 330 includes multiple conductive layers 340 which couple devices (e.g., transistors and capacitors) within a semiconductor device. Portions of each conductive layer 340 are typically coupled to an adjacent conductive layer portion 340 through a connection path called a xe2x80x9cplugxe2x80x9d 370. Typically, the conductive layer is typically about 10,000 Angstroms in thickness. Each conductive layer 340 is also separated by a dielectric material layer 360. The dielectric material, for example silicon dioxide, electrically insulates each conductive layer 340 to prevent unwanted short circuits. At the beginning of the fabrication process, the conductive layers have initial thicknesses of 1000 Angstroms. During the fabrication process, additional materials are added to form the conductive layers and therefore result in conductive layers having thickness of approximately 10,000 Angstroms. In some implementations of the test system, the initial and final thicknesses of the conductive layers will be measured. Since conductors, such as Cu, may easily diffuse into the adjacent dielectric layers 360, a thin liner layer 350 is formed between the conductive layers 340 and the dielectric layers 360. The liner layers 350 prevent each conductive layer 340 from diffusing into the dielectric layer and shorting with an adjacent conductive layer. Such a short may be detrimental to the proper operation of the semiconductor device. Any suitable liner layer material may be utilized to prevent a conductive layer from diffusing into an adjacent conductive layer. For example, a liner layer of Tantalum or Tantalum-Nitride may be used. The liner layer may have any suitable thickness, such as 300 Angstroms. The formation of the conductive and liner layers are monitored closely since they affect semiconductor device operation. The microanalysis system of the present invention is advantageously capable of measuring at least one conductive layer 340 and at least one liner layer 350 of a semiconductor wafer. Any suitable detector for measuring X-rays at specific energy levels may be utilized. FIG. 5 illustrates a cross-sectional view of a wavelength dispersive system (WDS) X-ray detector in accordance with one embodiment of the present invention. Each X-ray detector 500 includes a housing 530 having an aperture 535. Although preferred, the housing and aperture are optional and are not required for practicing the techniques of the present invention. An electron beam 545 is directed to a focus point 550 on a thin film device 555 (i.e., a semiconductor wafer). The electron beam 545 causes photons 540 to emanate from the focus point 550. The aperture 535 permits a limited amount of photons 540 to enter each detector 500. Upon entering the detector 500, each photon travels along a path to a concave reflective surface 510. The reflective surface 510 directs a portion of photons to a sensor 520. The reflective surface 510 is designed and positioned so that only photons with a specific energy level are directed to the sensor 520. The reflective surface 510 may be positioned to direct only photons with an energy level characteristic of a certain material to facilitate a film characterization process. By detecting photons of only a specific energy level, detector 500 is capable of obtaining high signal to noise ratios. It should be noted that the reflective surface may be a Bragg reflector or a crystal capable of directing photons towards the sensor. A cross-sectional view of an alternative and preferred embodiment of a WDS X-ray detector 500xe2x80x2 is illustrated in FIG. 6. Detector 500xe2x80x2 has a collimator 560 that captures the photons 540 emanating from the focus point 550, and then through its reflective surfaces causes the photons 540 to travel in substantially parallel paths. The collimator 560 is generally made from metal foil material. The photons then reflect off of a substantially flat reflective surface 565 such that the photons 540 continue in parallel paths towards the sensor 520. Similarly with detector 500, the reflective surface 565 in detector 500xe2x80x2 may also be Bragg reflector or a crystal. A common device which contains the general elements of the detector 500 and 500xe2x80x2 is a Wavelength Dispersive System (WDS). By utilizing multiple WDS detectors, one or more photon peaks may be detected for each type of material that is expected to be present within the measured film stack of the specimen. That is, characteristic emission levels for one or more types of material in the film stack may be measured. One or more individual detectors may also be dedicated to detect the various characteristic emission levels for each type of material. For example, two WDS detectors may be dedicated for detecting two peaks associated with a copper material. As described earlier each material has emission levels characteristic of photons released due to an electron falling from each of the K, L, or M shells. By using multiple WDS detectors, the test system is able to obtain information for each of a multiple number of film layers. Another type of detector, an Energy Dispersive System (EDS), collects photons in a wide spectrum of energies. EDS are capable of collecting a greater range of signals. As a result however, EDS detectors also collect photons having energies surrounding the characteristic photon energies. This causes EDS detectors to have lower signal to noise ratios. The test system of the illustrated embodiment is capable of obtaining measurements having precision within 0.5% accuracy. Film stack thickness measurements may also be made within two seconds if the electron beam current is increased to approximately 1.0xc3x9710xe2x88x925 Amps. Thus, this test system allows for both accurate testing at a high throughput rate. FIG. 7 illustrates a flow diagram representing one particular implementation of a method for measuring film stack characteristics on a semiconductor wafer. Initially, electron beam 400 parameters are set in operation 610. Parameters such as the electron beam current and voltage are set so that the electron beam has a landing energy sufficient for the beam to penetrate at least one conductive layer and one liner layer. Landing energy is defined as the energy at which the beam hits the specimen. In addition to the beam current and voltage, the scan pattern of the electron beam may affect the charge distribution of on the specimen 330, and thus, affect how the information is collected during the testing procedure. Charts, such as the ones illustrated in FIGS. 8A and 8B, may be used to determine the energy levels at which to set the electron beam 400 so that the desired layers are penetrated by the electron beam. These charts are well known and widely available to those skilled in the art. FIG. 8A shows the depths to which an electron beam will penetrate as the electron beam energy level varies. FIG. 8B shows how the electron beam penetration depth varies with an element""s atomic number while at a constant electron beam energy level. Typical electron beam energy levels range from 15 to 30 keV. After the beam parameters are set for the specific specimen, each of the WDS X-ray detectors 500 are set to detect the photons which will emanate from the layers within the specimen in operation 620. That is, the detectors 500 are set to detect photons having specific X-ray energy levels. Ideally, one WDS 500 is set to detect photons having energy levels corresponding to the characteristic X-ray energy emission levels of each material in the film stack to be analyzed. Multiple detectors 500 may also be set to detect X-ray energies of a single material in order to detect each of the multiple characteristic emission levels of that specific material. Each emission level for a single material represents photons emitted as a result of an electron falling from a different electron shell (i.e., K, L, M, etc.). A WDS is set for a specific emission level by orienting a reflective surface 510 at a specific angle such that only X-ray photons centered around the desired energy level are directed at a sensor 520. The sensor 520 will then be able to collect photons for analysis purposes. The information collected by the detectors may be transmitted to processing system 320 which is connected to each detector 500. Processing system 320 will then perform the desired analysis. After the electron beam 400 and the X-ray detectors 500 are properly set, the electron beam probe 400 is activated so that an electron beam impinges upon the semiconductor wafer 330 in operation 630. The electron beam penetrates at least one conductive layer and one liner layer, thereby, causing X-ray photons to emanate from the penetrated layers. X-ray photons emanating from each layer will be used to gain information about each of the respective layers. In operation 640, these X-rays are detected by the X-ray detectors 500. The number of X-rays at each characteristic emission energy level are counted. The number of X-rays counted is designated as the count for that particular emission level. FIGS. 9 and 10 show illustrative charts of the counts for the emission levels for a copper (Cu) K-shell emission and a tantalum (Ta) layer L-shell emission, respectively, wherein the Cu layer is formed on top of the Ta layer. The counts for each layer are illustrated for varying thickness values of the Cu layer. In FIG. 9, it can be seen that as the thickness of the Cu layer increases, so to does the count. In contrast, in FIG. 10, the count for Ta decreases as the thickness of the Cu layer increases since fewer X-ray photons are created as the Cu layer increases. Fewer X-ray photons for the Ta layer are created as the Cu layer increases in thickness because the electron beam penetrates to the Ta layer to a lesser degree as the Cu layer increases in thickness. Data from the collected X-rays is saved by the processing system 320 in operation 650. The data is then analyzed in operation 660. Any suitable analysis technique may be used for determining characteristics of the film stack based on X-ray emission at specific energy levels. For example, a regression routine may be used. The procedure for measuring film stack characteristics 600 is completed when the conductive layer characteristics are obtained through the regression routine. Alternative methods of analyzing the collected data may also be used. For example, a series of calibration measurements are made, then measurement points between the calibration points are obtained through interpolation techniques. FIG. 11 illustrates a flow diagram representing the procedure 1000 for determining particular film stack characteristic values. Generally, the procedure 1000 is a regressive technique where certain computational operations are repeated until a desired result is obtained. The operations include generating predicted data (the counts) regarding the film stack layers by inputting estimated film stack characteristics, such as thickness and composition values, into equations which model X-ray emission from the film stack layers. In other words, X-ray emission results are calculated for a film stack having the input estimated film stack characteristics. Various estimated film stack values are repeatedly inputted into the modeling equations until the resulting predicted data values closely match the data actually obtained from the X-ray micrography system. Any suitable regression technique may be utilized to repeatedly generate x-ray emission predictions for a particular film stack model and match such predictions to the measured raw data. For example, the well known Monte Carlo regression technique may be utilized. Monte Carlo techniques are useful for solving computationally complex equations, such as the above described equations. Another example of a regression technique that may be used is the Phi-Rho-Z model. Several software applications are available that automatically model films stacks with the Phi-Rho-Z model. For example, the Stratagem, available from Samx Guyan Court, in France, or the Citzaf, available from NIST (National Institute of Standard and Technology), software packages may be used. Parameters that are relevant to x-ray emission are input into the software, along with the measured raw data. The software then repeatedly uses the model equations to produce predicted data until the predicted data matches the raw data. For this specific software application, the parameters include the raw data and starting estimated values for film compositions, film thickness, landing energy of the electron beam, beam current, take-off angle, and detector efficiency. When the predicted data values (e.g., output from the film stack model) closely match the raw data values, it is believed that the estimated film stack characteristics are a good estimate of the actual film stack characteristics. The accuracy of this method, is of course, limited by the ability of the modeling equations to accurately predict the characteristics of the actual specimen. As shown in FIG. 11, raw data is initially obtained from the X-ray micrography system, for example, in operation 1010. In operation 1020, a first set of estimated film stack characteristics values are selected to be inputted into the modeling equations. In operation 1030, the equations are solved according to the estimated characteristic values in order to generate a set of predicted data values. In operation 1040, the raw data and the predicted data are compared. If the difference between the respective values is less than a certain margin of error, then the estimated film stack characteristic values are an acceptable estimate of the actual film stack characteristics. However, if the difference between the respective values are greater than the margin of error, then a new set of estimated film stack characteristics are selected. The regression operations (i.e., 1020, 1030 and 1040) continue until the difference in data values is less than the estimated margin of error, which is represented in operation 1050. A margin of error should be approximately equal to or less than 0.5% for acceptable estimates of film stack characteristics. While this invention has been described in terms of several preferred embodiments, there are alteration, permutations, and equivalents which fall within the scope of this invention. It should also be noted that there are many alternative ways of implementing the methods and apparatuses of the present invention. It is therefore intended that the following appended claims be interpreted as including all such alterations, permutations, and equivalents as fall within the true spirit and scope of the present invention. |
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description | The invention lies in the nuclear technology field and pertains, more specifically, to a containment vessel in a nuclear power installation. My commonly assigned earlier German patent DE 198 09 000 C1 (corresponding to my copending application Ser. No. 09/655,091) described an innovative structural and safety concept for a boiling water reactor. In the boiling water reactor disclosed in that document, the reactor pressure vessel is disposed centrally in a containment. In addition to the reactor pressure vessel, a closed condensation chamber and a flood basin arranged above it are provided for emergency cooling of the boiling water reactor. The flood basin is open toward a central region, in which the reactor pressure vessel is arranged, and forms a pressure chamber together with the latter. A so-called building condenser is arranged above the flood basin, i.e. in the upper region of the pressure chamber or containment. The building condenser is in communication with a cooling liquid from a cooling basin arranged above the containment and is used to dissipate the heat from the pressure chamber. The efficiency of the building condenser reacts sensitively to the presence of non-condensable gases, such as nitrogen or hydrogen. The latter may be formed in particular in the event of extreme emergencies. This is because the non-condensable gases reduce the ability of the building condenser to dissipate heat from any steam which may be present in the pressure chamber into the cooling basin. On account of its low relative density, hydrogen accumulates in the upper region of the pressure chamber, so that a high concentration of non-condensable gases may be present in particular in the vicinity of the building condenser, leading to an increase in pressure in the containment. To dissipate the heat from the pressure chamber in the event of an emergency, i.e. to dissipate non-condensable gases from the pressure chamber, there are known concepts in which the pressure chamber is connected to a condensation chamber via condensation tubes. The steam which is present in the pressure chamber in the event of an emergency, together with the non-condensable gases, passes via these condensation tubes into the condensation chamber. Since the condensation tubes are generally immersed in the cooling liquid in the condensation chamber to a depth of several meters, the steam condenses and only the entrained fractions of the non-condensable gases remain in the condensation chamber. A system of that type is known, for example, from my earlier German patent DE 198 09 000 C1. The containment described in that document has a condensation chamber, a pressure chamber and a building condenser arranged in the upper region of the pressure chamber, with a diverter tube also being provided, flow-connecting the upper region of the pressure chamber to the condensation chamber in order to divert the non-condensable gases out of the upper region of the pressure chamber into the condensation chamber in a targeted and direct manner. Conventional condensation tubes substantially comprise a vertically running tube, the upper end of which is connected to the pressure chamber and the lower end of which is immersed in a cooling liquid in the condensation chamber. The condensation tubes generally have a diameter of approximately 400 to 600 mm and at their lower end are cut off substantially perpendicular to the tube axis. With this conventional design, in particular in the event of large leak cross sections, high loads are imposed on the base and the side walls of the condensation chamber as a result of water being thrown up during the initial overflow of air or nitrogen and as a result of the phenomenon known as chugging toward the end of the overflow phase. During chugging, the pressure amplitudes may amount to several bar, and consequently the pressure loads caused by chugging may be the determining factor in the building structure of the containment. It is accordingly an object of the invention to provide a containment of a nuclear power plant which overcomes the above-mentioned disadvantages of the heretofore-known devices and methods of this general type and which is further developed such that the pressure loads on the base and the walls of the condensation chamber in the event of an emergency are reduced to a considerable degree. With the foregoing and other objects in view there is provided, in accordance with the invention, a containment of a nuclear power plant, comprising: a containment structure having formed therein a pressure chamber and a condensation chamber with a base; a substantially vertical condensation tube having an upper end communicating with the pressure chamber and a lower end immersed in a cooling liquid in the condensation chamber; the lower end of the condensation tube being formed with an elbow and an outlet nozzle; the elbow having an elbow angle causing a lower end of the elbow to be immersed obliquely in the cooling liquid in the condensation chamber; and the outlet nozzle having an outlet opening substantially shielded with respect to the base of the condensation chamber. In other words, the lower end of the vertically running condensation tube has an elbow and an outlet nozzle, the elbow having an elbow angle which is such that the lower end of the elbow is immersed obliquely in the cooling liquid in the condensation chamber, and the outlet nozzle having an outlet opening which is substantially shielded with respect to the base of the condensation chamber. In this context, the invention is based on the consideration that while water is being thrown up during the initial overflow of air or nitrogen, considerably lower pressure loads on the base and walls of the condensation chamber are likely, since the air which emerges, on account of the substantially horizontal flow out of the outlet nozzle which is designed in accordance with the invention, is distributed over a significantly larger area. In the event of chugging, i.e. in the event of only low mass flows of steam flowing out and steam bubbles being formed in the condensation chamber, the dynamic pressure loads on the condensation chamber walls are considerably lower, since the outflow area from the outlet nozzle is predominantly closed off by the cooling liquid, whereas in the standard embodiment the entire cross section of the tube has always been uncovered. Tests carried out by the inventors have confirmed that the pressure loads on the base and walls of the condensation chamber are significantly reduced compared to conventional designs of the condensation tubes. In accordance with one preferred embodiment of the invention, the outlet nozzle is formed by a tube section, of which the side that faces the base of the condensation chamber is longer than the side that is remote from the base of the condensation chamber, so that the local mixing zone of steam and water, in which the highest pressure transients are formed as a result of the bubble collapsing during chugging, is shielded from the base of the condensation chamber. Moreover, the elbow angle of the elbow of the condensation tube is preferably between approximately 70° and approximately 85°, preferably approximately 82°, so that the lower end of the elbow is immersed in the cooling liquid in the condensation chamber in such a manner that it is inclined obliquely downward. In accordance with a further configuration of the invention, a significant part of the condensation tube is embedded in the wall of the condensation chamber. The condensation chamber wall can in this way absorb all the forces which occur in the condensation tube and ensure additional protection in the event of possible breaking of a condensation tube. Moreover, this avoids the in some cases highly complex holding structures for the condensation tubes, which are customarily arranged freely in the condensation chamber. The advantages which are achieved by the invention consist in particular in the fact that a completely new outlet geometry of the condensation tube is provided, leading to significantly more favorable properties with regard to the pressure loads which occur on the base and walls of the condensation chamber. The fact that the condensation tube with the specially designed outlet nozzle is immersed obliquely in the cooling liquid in the condensation chamber gives rise to a substantially horizontal outflow over a significantly larger area, and the outflow area is substantially closed off by the cooling liquid. In this way, in the event of an emergency in the boiling water reactor, significantly lower pressure loads on the walls and the base of the condensation chamber are present both during the initial phase when water is thrown up and during the chugging phase. Other features which are considered as characteristic for the invention are set forth in the appended claims. Although the invention is illustrated and described herein as embodied in a containment of a nuclear power plant, it is nevertheless not intended to be limited to the details shown, since various modifications and structural changes may be made therein without departing from the spirit of the invention and within the scope and range of equivalents of the claims. The construction and method of operation of the invention, however, together with additional objects and advantages thereof will be best understood from the following description of specific embodiments when read in connection with the accompanying drawings. Referring now to the figures of the drawing in detail and first, particularly, to FIGS. 1 and 2 thereof, there is shown the structure of a containment in accordance with a preferred exemplary embodiment of the invention. The pressure load on the base of the condensation chamber when using a condensation tube in accordance with the invention will be compared with the pressure load when using a conventional condensation tube on the basis of FIGS. 3A and 3B. In accordance with FIG. 1, a reactor pressure vessel 12 is disposed centrally in a closed containment 10. A condensation chamber 14 and a flood basin 16 arranged above it are provided in the containment 10, laterally next to the reactor pressure vessel 12. The flood basin 16 is open at the top toward the interior of the containment 10. The interior is also referred to as the pressure chamber 18, which forms a common pressure space with the flood basin 16. The condensation chamber 14 and the flood basin 16 are each partly filled with a cooling liquid 20, in particular water, to a filling level 22. The maximum filling level 22 in the flood basin 16 is determined by the upper end of an overflow pipe 24. The overflow pipe 24 connects the flood basin 16 to the condensation chamber 14 and opens out into the cooling liquid 20 of the condensation chamber 14, so that in the event of the maximum filling level 22 being exceeded the cooling liquid 20 flows out of the flood basin 16 into the condensation chamber 14. The flood basin 16 is also connected, via a flood line 26, to the reactor pressure vessel 12. In the event of an emergency, the flood basin 16 can supply the pressure vessel 12 with sufficient cooling liquid 20. The condensation chamber 14 is substantially closed off from the pressure chamber 18. It is only in communication with the pressure chamber 18 via one or more condensation tubes 28. The condensation tube 28 is immersed in the cooling liquid 20 in the condensation chamber 14, so that there is no exchange of gases between the condensation chamber 14 and the pressure chamber 18. The condensation tube 28 is normally closed off by a water column 30 in the condensation tube 28; only in the event of an emergency, when the pressure rises in the pressure chamber 18, does steam flow into the condensation chamber 14 via the condensation tube 28 in order to be condensed. The precise structure and functioning of the condensation tube 28 are explained in more detail below with reference to FIG. 2. In the left-hand half of FIG. 1, a building condenser 32 is arranged in the upper region of the containment 10 and therefore in the upper region of the pressure chamber 18. The building condenser 32 is designed as a heat exchanger with heat exchanger tubes and is flow-connected to a cooling basin 34, which is arranged outside the containment 10 on top of its cover 36. The building condenser 32 takes up the heat from its surroundings inside the containment 10 and transmits it to the cooling basin 34, with the result that heat can be dissipated from the containment 10 into the external surroundings. In the event of an emergency, for example in the event of a steam line in the containment 10 breaking, with the associated escape of steam, or in the event of a loss of coolant, the temperature and pressure in the containment 10 rise. Various emergency cooling devices, of which only the building condenser 32 and the flood basin 16 with associated flood line 26 are shown in FIG. 1, it is ensured that the emergency final pressure in the containment 10 does not exceed a permissible threshold. This is achieved primarily by cooling and condensing the steam. In this context, an important role is played by the building condenser 32, by means of which heat can be dissipated from the containment 10 to the outside. During an emergency, under certain circumstances non-condensable gases, such as for example hydrogen or inert gases such as air or nitrogen, will be released and accumulate in the upper region of the containment 10, i.e. in the upper region of the pressure chamber 18. The non-condensable gases collect in the upper region of the pressure chamber and lead to an increase in the pressure in the containment 10. When a certain pressure is reached in the pressure chamber 18, the steam together with the non-condensable gases, through the condensation tube 28, can overcome the pressure of the water column 30 in the condensation tube 28 and flow into the condensation chamber 24. The entrained steam is cooled and condensed in the condensation chamber 14, while the non-condensable gases remain in the condensation chamber 14. In principle, the non-condensable gases impair the efficiency of the building condenser 32 by significantly reducing the heat transfer capacity of the building condenser 32. When non-condensable gases are present, the building condenser 32 can only dissipate significantly less heat per unit time and area from the steam to the cooling basin 34 than when the non-condensable gases are absent. Since these non-condensable gases are diverted out of the vicinity of the building condenser 32 by the condensation tube 28, the building condenser 32 can be designed for saturated steam. Therefore, it does not need large and specially designed heat transfer surfaces, which if non-condensable gases were present would be absolutely imperative in order to enable sufficient heat to be dissipated. Therefore, the building condenser 32 can be of simpler, more compact and therefore more advantageous design. The structure and functioning of the condensation tube 28 which leads into the condensation chamber 14 in accordance with the present invention will now be described in more detail on the basis of the diagrammatic illustration present in FIG. 2. In the event of an emergency, with the associated increased pressure in the pressure chamber 18, steam flows out of the pressure chamber 18, together with the non-condensable gases, through the condensation tube 28 and into the condensation chamber 14. As illustrated in the graph presented in FIG. 3B, in the process, in the case of a conventional condensation tube, i.e. a substantially vertically running condensation tube with a lower end which is cut off perpendicular to the tube axis, pressures of up to 2 bar on the base and walls of the condensation chamber 14 occur when the water is thrown up during the initial overflow of air, and pressures of up to 10 bar on the base and walls of the condensation chamber 14 occur during the phenomenon known as chugging, i.e. the formation of steam bubbles in the condensation chamber 14 toward the end of the overflow phase. To reduce these high pressure loads on the walls and base of the condensation chamber 14, the condensation chamber 28 of the containment 10 according to the invention is constructed as follows. The condensation tube 28 has a substantially vertically running main section 28a, at the upper end of which there is provided an inlet opening 28b inside the pressure chamber 18. The lower end of the vertical section 28a of the condensation tube 28 is adjoined by an elbow 28c. The elbow 28c is substantially a curved tube section with an elbow angle 28e of preferably between approximately 70° and 85°, particularly preferably of approximately 82°. The condensation tube 28 projects, by means of this elbow 28c, into the cooling liquid, below the filling level of the cooling liquid 20 in the condensation chamber 14, with a slight downward inclination. An outlet nozzle 28d is provided at the lower end of the elbow 28c. In the exemplary embodiment shown, the outlet nozzle 28d is made from a straight piece of tube, the length of which on the side facing the base of the condensation chamber 14 is considerably longer than on the side remote from the base. This particular design of the condensation tube 28 with the elbow 28c and the special outlet nozzle 28d means that in the event of an emergency the pressure loads on the base and walls of the condensation chamber 14 are likely to be significantly lower both during the initial throwing-up of water and during the subsequent chugging. This is also confirmed by tests, the results of which are illustrated in the graph presented in FIG. 3A. The pressure loads which occur are in a range below approximately 1 bar throughout the entire time, i.e. are significantly lower than the pressure loads of initially at most 2 bar and up to 10 bar toward the end in the case of the conventional condensation tube (cf. FIG. 3B). Unlike conventional containments 10, the condensation tube 28 is also not held in the condensation chamber by means of suitable holding structures. Instead, a significant part of the condensation tube 28, in particular the vertically running main section 28a and a large part of the elbow 28c, is embedded in the concrete wall of the condensation chamber. As a result, the wall 38 of the condensation chamber 14 absorbs all the forces which occur in the condensation tube and offers additional protection against the possibility of a condensation tube 28 fracturing. Overall, the invention provides a containment 10 of a nuclear power plant which has a condensation tube which significantly reduces the pressure loads on the base and walls of the condensation chamber compared to a conventional condensation tube. This increases the safety of the containment and reduces the demands imposed on the building structure of the containment. Although the present invention has been described above on the basis of a preferred exemplary embodiment, it will be understood by those of skill in the pertinent art that various modifications to this embodiment can be performed while still remaining within the scope of protection of the present invention as defined by the appended claims. In particular, the design of the outlet nozzle is not limited to the form of a straight section of tube with long sides of different lengths described above. The only crucial factor in designing the outlet nozzle is the outlet effect of the media flowing through the condensation tubes which is brought about by the outlet nozzle. Furthermore, the condensation tubes comprising the vertical main section, the elbow and the outlet nozzle may be in both single-part form and composed of a plurality of separately manufactured components which are subsequently tightly connected to one another. |
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description | The present disclosure generally relates to nuclear fuel pellet inspection. More specifically, the present embodiments are directed towards X-ray methods and apparatuses for detecting physical defects in nuclear fuel pellets. In a typical configuration, fuel for nuclear power plants is often in the form of cylindrical pellets stacked within a cladding tube. The cylindrical pellets typically contain the nuclear fuel material, and the cladding is often composed of a relatively inert substance. Together, these form a nuclear fuel rod which, during operation of the power plant, may be used to drive selected processes for power generation. One example would be steam generation that is used to power a turbine. The manufacturing of such fuel rods is typically performed by stacking the cylindrical pellets inside the cladding tube and welding end plugs onto the end of the cladding tubes. Nuclear fuel reliability is an issue throughout the nuclear industry and is a key objective in many fuel rod manufacturing processes. Nuclear fuel pellet defects can lead to a variety of fuel performance deficiencies, among other problems. For example, pellets missing a portion of their cylindrical surface (e.g., a pit or chip), cracks, and other defects may lead to pellet-cladding interaction (PCI) in which the fuel pellet contacts the cladding resulting in stress, which may be accompanied by corrosive compounds that attack the cladding and can result in cladding rupture. Such fuel rod failures can allow highly radioactive contaminants to be dispersed in the reactor coolant. Such events can lead to higher than desired radiation exposure to power plant operators and to unplanned shutdowns of the reactor to remove and replace the failed fuel. Such unplanned outages can lead to significant financial and/or capital losses to both the energy provider and their customers. The present disclosure is generally directed to partially or substantially automated procedures for inspecting nuclear fuel pellets using X-ray radiation. In one embodiment, a system for inspecting nuclear fuel rods is provided. Among other features, the system may generally include a first X-ray source configured to produce X-ray radiation at one or more energies, the first X-ray source being disposed on one side of a moveable track configured to transport one or more nuclear fuel rods. A first X-ray detector having a first scintillator and a first diode array is also provided. The first X-ray detector is generally configured to detect at least a fraction of the X-ray radiation produced by the first X-ray source and produce a first digital X-ray image of one or more nuclear fuel rods. The first X-ray detector may be disposed on the opposite side of the moveable track from the first X-ray source. A tangible, machine-readable medium is also provided, which is generally configured to store one or more computer-implemented algorithms which, when executed by a processor, facilitate the detection of defects in one or more nuclear fuel rods using X-ray images. In another embodiment, a method for inspecting nuclear fuel rods is provided. Among other features, the method may generally include providing a nuclear fuel rod to an inspection area, the nuclear fuel rod containing nuclear fuel pellets disposed within the nuclear fuel rod along a long axis. The nuclear fuel rod is irradiated with a first portion of X-ray radiation produced by an X-ray source. At least a fraction of the first portion of the X-ray radiation is detected with a detector generally containing a scintillator and a diode panel to produce a first digital X-ray image of the nuclear fuel rod. The nuclear fuel rod is then rotated at a first angle about the long axis using a nuclear fuel rod handling mechanism, followed by irradiation with a second portion of X-ray irradiation produced by the X-ray source. At least a fraction of the second portion of the X-ray radiation is detected, again with the detector, to produce a second digital X-ray image of the nuclear fuel rod. The first and second digital X-ray images are inspected for defects within the nuclear fuel pellets. In a further embodiment, an imaging masking assembly is provided. The imaging masking assembly may generally include a first monolithic structure configured to be disposed proximate one side of a nuclear fuel rod having a long axis, the monolithic structure extending past the nuclear fuel rod in a direction parallel to the long axis and parallel to an X-ray beam produced by an X-ray source. The assembly may also include a nuclear fuel rod handling assembly configured to maintain the position of the nuclear fuel rod proximate the monolithic structure and configured to positionally displace the nuclear fuel rod during imaging. In a typical nuclear fuel rod manufacturing configuration, nuclear fuel pellets are inspected via visual methods during the production process. However, oftentimes several process steps must be completed subsequent to this inspection which may cause damage to the fuel pellets (among other defects). For example, because of the tight tolerances between the outer diameter of the fuel pellet and the inner diameter of the cladding rod, fuel pellets may be damaged upon insertion into the cladding. Once loaded into the cladding, handling of the fuel rods may also result in unwanted movement of the pellets within the rods, as the pellets may impact one another. Such impacts may lead to undetected defects within the fuel rods. Accordingly, it is now recognized that substantial automation of the pellet/fuel rod inspection process using X-ray radiation as well as faster and more accurate performance of the same may be achieved by implementing the methods and apparatuses described herein. As such, the present disclosure is directed towards X-ray methods and apparatuses for detecting defects in nuclear fuel pellets. In particular, the present disclosure utilizes X-ray sources and X-ray detectors to produce X-ray images of nuclear fuel rods (and the pellets contained inside). The X-ray sources may be configured to produce X-ray radiation at one or more energies, and the detectors may be configured to detect the same. According to an aspect of the present disclosure, the detectors may generally include a scintillator and a diode array for the production of digital X-ray images in a substantially real-time fashion. In a further embodiment, a fuel rod handling mechanism may be provided such that a number of digital X-ray images of the fuel rods may be obtained from a variety of poses to increase the probability of detecting any defects which may be present in the fuel rods or fuel pellets. According to another aspect, a processor or processing circuitry for executing stored routines (such as routines stored on a tangible, machine-readable medium) may be provided. The processor may be configured to execute one or more algorithms for the recognition, detection, and flagging of possible pellet defects. In an embodiment of this aspect, the processor may also be configured to execute algorithms to correct any image distortion resulting from random events, such as X-ray scattering arising from X-ray-pellet interactions. In certain of the embodiments generally described above, a masking fixture may also be provided to increase the image sharpness and contrast at the edges of the fuel pellets. Further, the masking structure may serve to protect the detector from unattenuated X-ray radiation that has not passed through the fuel rods. Moving now to the figures and referring initially to FIG. 1, a block diagram illustration of an embodiment of a nuclear fuel rod manufacturing process 2 is provided, which generally includes three sections, a rod assembly section 4, a rod inspection section 6, and a bundle assembly section 8. FIG. 1 also depicts the general flow of rods through the process 2. For example, fuel rods 10 are assembled in rod assembly 4, which leads to rod inspection 6. In rod inspection 6, the number of rods 10 may be reduced to a set of inspected rods 12 (i.e., rods that have passed an inspection process). Once the rods 12 have been initially inspected, they may be passed to the bundle assembly area where the inspected set of rods 12 may be assembled into rod bundles 14. It should be noted that while the present embodiments described are directed towards the implementation of the methods and systems according to the present disclosure into one of these sections, it is well within the scope of the present disclosure to implement the approaches disclosed herein into these or other areas of a fuel rod manufacturing facility. Such implementation is within the skills of a person of ordinary skill in the art and would be a matter of routine optimization. For example, one possible implementation of the approaches described herein may be applied to substantially real-time inspection of individual nuclear fuel pellets before loading into fuel rods. In the depicted embodiment, the process 2 begins at the rod assembly section 4. In one embodiment of rod assembly 4, hollow, long cylindrical cladding rods 16 are filled with nuclear pellets 18. However, there may be different possible configurations of a fuel rod 20, such that the pellets 18 contained within may have different patterns of placement. In some embodiments, the cladding rods 16 may be several inches to several feet long and have an internal diameter ranging from a few inches to less than an inch. According to present embodiments, the cladding rod 16 is made of a material capable of substantially isolating a nuclear fuel. For example, the cladding 16 may be a zirconium-based alloy or stainless steel, and the fuel pellets 18 may be uranium oxide (UO2) or a similar nuclear fuel. In some embodiments, the fuel pellets 18 may be doped with a few percent (e.g., between about 0.5 wt % and about 10 wt %) of gadolinium oxide (Gd2O3, gadolinia) in order to absorb neutrons and allow the fuel rods 20 to exhibit more desirable reactivity characteristics. Accordingly, other neutron-absorbing materials are also contemplated. Rod assembly 4 may include receiving fuel pellets 18 from a pellet tray, for example, using a properly configured pellet and/or rod handling mechanism. The pellets 18 are loaded into the cladding rod 16, and in some embodiments, a spring 22 is included to bias the pellets 18 against one of two end plugs 24 of the fuel rod 20. In other embodiments, this step may be performed as a separate step. The spring 22 is generally chosen and configured to bias the position of the fuel pellets 18 against an inner end surface of the fuel rod (end plug 24) to substantially limit translational movement of the pellets 18 along a long axis 26 of the fuel rod 20. Once all of the pellets 18 have been loaded into the cladding 16 in rod assembly 4, the fuel rod 20 then undergoes a number of procedures dedicated to sealing the rod 20 and inspecting the cladding 16 for defects, generally referred to as rod inspection 6. Sealing the rod 20 may include any number of procedures known in the art, such as resistance welding and/or tungsten inert gas (TIG) welding, and is generally performed to seal both end plugs 24 onto the cladding 16. In a typical configuration, rod assembly 4 and/or rod inspection 6 may include a number of stations, including, for example, a rod type detect station, a spring check station, an evacuation/low pressure backfill station, a seam weld station, a cool-down station, a high pressure weld station, a helium leak detection station, a parallelism gage station, a ring gage station, an ultrasonic (UT) microscope station, and an unload station. It should be understood that either rod loading 4 or rod inspection 6 may include any one or a combination of these stations. In rod inspection 6, the rods 20 are checked for any leaks of the inert gas which was used to partially backfill the rod 20 before welding on the end plugs 24. The welding seams between the cladding rod 16 and the end caps 24 may also be checked to ensure a flush, circular contact. Lastly, the rods 20, and more particularly the cladding 16, are checked for any defects. Upon inspection of the cladding 16, the rods 10 are then either accepted, rejected, or flagged for further inspection based at least in part on the integrity of the cladding 16. The rejected rods will go to a holding area, while the accepted rods 12 will be transported to the bundle assembly section 8. Any calibration and verification rods may also go to the holding area. The bundle assembly area 8 generally contains areas of scanning including passive scanning and active scanning, as well as areas where the rods 20 are actually assembled into bundles, further tested, and finally stored until delivery to a customer. According to the present disclosure, the rod assembly 4, rod inspection 6, and/or bundle assembly 8 sections may be potential locations for implementing X-ray scanning of the fuel rods 20 into the process flow. For example, in some embodiments, X-ray inspection of the welding resulting from TIG welding processes may be used to inspect the resultant weld for integrity and quality. Accordingly, such X-ray equipment may be properly modified according to the present disclosure to allow the facile implementation of the apparatuses and methods described herein. It should be noted that in a typical manufacturing setting, tens, hundreds, or even thousands of individual rods 20 may be produced and processed every day or every few days. Accordingly, using conventional X-ray inspection processes, only a fraction of the rods 20 being produced may be inspected due to the time constraints associated with conventional radiographic inspections. For example, conventional X-ray inspection of individual rods 20 may require a total time of five or more minutes per rod 20, including the time taken to acquire the image, process the image data, and inspect the image for pellet defects. Conversely, according to the approaches described herein, it may be possible to implement direct radiography in the imaging of individual fuel rods 20, such that substantially real-time imaging and inspection of all rods produced in a manufacturing facility may be performed without negatively impacting the throughput of the facility. Such approaches may be implemented using a number of possible configurations, one embodiment of which is depicted in FIG. 2, which is a block diagram of one embodiment of a system 50 for the X-ray inspection of nuclear fuel rods, such as an individual rod 20. Of course, in some embodiments, the rods inspected may be assembled rods 10, initially inspected rods 12, or rod bundles 14. In the illustrated embodiment, the system 50 is a direct radiography system designed to acquire two- or three-dimensional digital radiography data about nuclear fuel rods and, in some embodiments, to process such data to determine the presence of any pellet defects (quality excursions) in a substantially real-time fashion (e.g., in less than five minutes, less than three minutes, less than a minute, less than thirty seconds, less than ten seconds, less than a second, and so forth). While the present discussion is directed towards the implementation of digital or direct radiography (DR) to determine pellet defects, the use of computed radiography (CR) and/or computed tomography (CT) is also contemplated, and is merely a matter of design choice. Further, the implementation of computed radiography and/or computed tomography in place of direct radiography could be accomplished with routine experimentation and optimization. In the embodiment illustrated in FIG. 2, the system 50 generally includes an X-ray source 52 positioned adjacent to a collimator 54. The X-ray source 52 may be an X-ray tube or a solid-state X-ray emitter, and may be configured to emit X-ray radiation at one or more energies. In the embodiment depicted, the collimator 54 is configured to permit a stream of radiation 56 to pass into a region, such as an area where one or more fuel rods 20 are present. In certain embodiments, the stream of radiation 56 may have an energy sufficient to penetrate the nuclear fuel rod 20. For example, the stream of radiation 56 may have energies ranging from about 50 kV to about 1 MV (e.g., about 200 kV, 250 kV, 300 kV, 380 kV, 420 kV, 450 kV, and above, or any ranges in between). Further, the particular energy used for the stream of radiation 56 may be chosen depending on the nature of information that is to be collected. For example, lower energies (e.g., between 200 and 450 kV) may provide surface information about the pellets 18 within the fuel rods 20 (e.g., pits and/or chips), while higher energies (e.g., above about 420 kV) may provide non-surface information (i.e., cracks, etc.) about the pellets 18. During operation, a portion of the radiation 58 passes through or around the rod 20 (or bundle of rods 14) and impacts a detector array 60. According to the present disclosure, a masking structure 62 may be provided to substantially limit the amount of radiation 56 that bypasses the fuel rod 20, and to increase the quality of the image that results from the detection of the portion of radiation 58. For example, as depicted, the masking structure 62 extends in the direction of X-ray propagation beyond the fuel rod 20. In such a configuration, the masking structure 62 may substantially limit scattering events that result from the interaction of the stream of radiation 56 and the fuel rod 20 (including the cladding 16 and pellets 18). In one implementation the masking structure 62 is configured to be disposed proximate the fuel rod 20 on one or both sides, though not directly in between the rod 20 and the source 52 or detector 60. The masking structure 62 may, in some embodiments, extend out in other directions as well, such as perpendicular to the general beam direction, to protect the detector 60 from unmitigated X-ray radiation. These and other features regarding the masking structure 62 are described in more detail hereinbelow. During operation, upon the X-ray radiation 58 passing through the rod 20 and past the masking structure 62, detector elements of the array produce electrical signals that represent the intensity of the incident X-ray beam (fraction or portion of radiation 58). For example, the illustrated detector 60 includes a scintillator 64 and a photodiode array 66. The scintillator 64 is configured such that the X-ray radiation 58 is converted from ionizing radiation to light (i.e., visible light). The light produced by the scintillator 64 then interacts with the photodiode array 66 to produce an electric current, such as via a photovoltaic effect. In order to produce the photovoltaic effect, the diode array 66 may be constructed from a material exhibiting photovoltaic properties. For example, in one embodiment, the diode array 66 may be an amorphous silicon flat panel with a coupled field effect transistor (FET). In some embodiments, the scintillator 64 may be configured to absorb a substantial amount of the fraction of radiation 58, which may have energies between about 50 kV and 1 MV. In one embodiment, the scintillator 64 may be constructed from thallium-doped cesium iodide (CsI:Tl) or other scintillation material that does not exhibit appreciable hygroscopicity. In such an embodiment, the CsI:Tl may be grown directly onto the diode array 66 (amorphous silicon) and then hermetically sealed with a cover plate to reduce the effects of hygroscopicity. As such, it may be possible to tailor the thickness of the scintillator 64 (the scintillating layer) to desired heights. As an example, the CsI:Tl may be grown to between about 0.050 mm to about 4 mm (e.g., about 0.40 mm, 1.0 mm, 1.5 mm, 2.0 mm) in thickness. In some embodiments, when choosing a thickness for the scintillator 64, an operator may consider a number of factors. For example, thicknesses of up to about 10 mm may be achieved by growing CsI:Tl needles onto an amorphous silicon panel. While such thickness may impart greater signal-to-noise ratios than would otherwise be achieved using shorter needles, greater scintillator thickness (e.g., over 4.0 mm) may lead to reduced contrast around the edges of the pellets 18 and thus, lowered spatial resolution of pellet defects. However, there may be embodiments or situations where such lengths are desirable and an increase in contrast resolution achieved with a thicker scintillator may outweigh the decrease in its spatial resolution for the detection of these defects. In another embodiment, the scintillator 64 may be constructed from polycrystalline terbium-doped gadolinium oxysulfide (Gd2O2S:Tb) phosphor sheets. In such an embodiment, particles of Gd2O2S:Tb are mixed with a binder and adhered to a polymeric (e.g., Mylar) support. When Gd2O2S:Tb phosphors are used, the particles with binder may form sheets having a thickness between about 0.050 mm to about 2 mm (e.g., about 1 mm). In certain embodiments where a phosphor is used, it may be desirable to use, in lieu of or in combination with Gd2O2S:Tb, terbium-doped yttrium oxysulfide (Y2O2S:Tb), silver-doped zinc sulfide-cadmium sulfide alloy (ZnSCdS:Ag), or similar phosphor, or any combination thereof. Other phosphors may include those employing copper-doped zinc sulfide (ZnS:Cu), copper-doped zinc-cadmium sulfide (ZnCdS:Cu) and similar phosphors using aluminum dopants, to name a few. A choice of any related phosphor known in the art is merely one of design and should be readily apparent to those of skill in the art and is considered to be within the scope of this disclosure. In a further embodiment, it may be desirable to use fiber optic scintillating glass to construct the scintillator 64. In a typical configuration of fiber optic glass, a core fiber is the scintillating material (e.g. a terbium-doped heavy silicate glass of moderate effective atomic number and density), which is surrounded by a cladding glass that allows light to propagate down the fiber toward one or more photodiode arrays 66. In such embodiments, the fiber/cladding is close-packed into a faceplate of many millions of fibers (e.g., between about 1 and about 20 million), with a total thickness (length of cladding glass) of between about 1 mm and about 10 mm (e.g., about 2 mm). The fiber optic scintillating glass faceplate may also include extramural absorbing material, either surrounding the cladding, or placed in a statistical manner as individual fibers throughout the faceplate. Both methods may be used to block stray light that exits fibers from reaching the diode structure below. Thus, the scintillator 64 may be constructed from a number of materials designed to emit light to the photodiode array 66 upon absorbing ionizing radiation. The scintillator 64, as mentioned, has a thickness of between about 200 microns to about 10 millimeters, with the thickness being chosen based on a number of factors including the type and material of the fuel rod 20, geometry of the defect, the energy of the emitted X-ray radiation 56, and so forth. The scintillator 64 may have tens, hundreds, thousands, or even millions of rows of needles and/or detector elements. In one embodiment, by decreasing the size and increasing the number of the needles in the scintillator 64, the resolution of a resultant image may be increased. In embodiments where the scintillator 64 has a high resolution, the resolution of a resultant digital X-ray image may be limited by the resolution of the photodiode array 66. For example, as mentioned, the photodiode array 66 may be a flat panel of amorphous silicon. In one embodiment, the amorphous silicon panel may have a resolution of about 1024 by 1024 over a 16-inch by 16-inch area. In another embodiment, it may have a resolution of about 2400×3000 over a 9.4-inch by 11.8-inch. The process of generation of light by the scintillator 64, its transfer to the diode array 66, and the generation of a digital image by the array 66 may occur within a fraction of a second, e.g., in near-real time. Therefore, the direct radiography configuration according to present embodiments may allow substantially real-time imaging of the rod 20, with imaging speeds of up to about 30 Hz being attainable. As a result, imaging and inspection of a rod 20 may be performed in a matter of seconds. For example, inspection of a rod 20 may occur in under 3 seconds, 5 seconds, 10 seconds, 30 seconds, 1 minute, 3 minutes, 5 minutes and so forth. Accordingly, the implementation of direct radiography into a manufacturing process, such as the manufacturing process 2, may permit all or a majority of the individual rod 20 that are produced to be inspected, rather than inspecting on a selective, random or subset basis. In operation, once the photodiode array 66 generates electrical signals corresponding to the numbers of X-rays that have passed through the rod 20, the signal is passed to a system controller 68, and more particularly, to an image data acquisition system 70 that is configured to receive and/or process the signals generated by the detector 60. The system controller 68 is configured to control the X-ray source 52, and furnishes both power and control signals for rod inspection procedures. Moreover, the detector 60 is coupled to the system controller 68, which commands acquisition of the signals generated in the detector 60. The system controller 68 may also execute various signal processing and filtration functions, such as for initial adjustment of dynamic ranges, interleaving of digital image data, and so forth. In general operation, system controller 68 commands operation of the imaging system 50 to execute inspection protocols and, in some embodiments, to process acquired data. In the present context, system controller 68 also includes signal processing circuitry, typically based upon a general purpose or application-specific digital computer, associated memory circuitry for storing programs and routines executed by the computer (such as programs and routines for implementing the embodiments described herein), as well as configuration parameters and image data, interface circuits, and so forth. In the embodiment illustrated in FIG. 2, the system controller 68 contains a motor controller 72, which is configured to control a rotational subsystem 74 and a translational subsystem 76 of a rod positioning system 78. The rod positioning system 78 is configured to allow an operator or controller to move the rod 20 translationally along its long axis 26 and/or rotationally about its long axis 26. Such positioning may be useful in detecting pellet defects, as certain defects may only be imaged and thus visible from one pose of the rod 20 with respect to the X-ray source 52 and the detector 60. Further, the rod positioning system 78 may be useful in some aberration correction methods (e.g., shift imaging) performed by the system controller 68, which is described, along with other implementations of the present disclosure, hereinbelow. As depicted, the rod positioning system 78 may perform the translation and/or rotation by clasping on to one or both ends of the fuel rod 20. The rotational subsystem 74 allows the rod positioning system 78 to then rotate the rod 20 about its long axis 26, either in a smooth, continuous rotation or in a step-and-shoot manner, where the rod 20 is rotated at an angle, an image is taken, and the rod 20 is then rotated again for subsequent image capture. Accordingly, the rotation performed by the rod positioning system may be between about 90° and about 1° (e.g., about 90, 55, 45, 22.5, 11.25 degrees). In certain embodiments, the rod positioning system may also perform small rotations of less than 10°, for example less than about 9, 8, 7, 6, 5, 4, 3, 2, or 1 degree. Similarly, the translational subsystem 76 may allow the rod positioning system 78 to perform a small translation along the long axis 26 of the rod 20, the translation being only to a small extent (i.e., between about 1 micron and several centimeters). To this end, the motor controller 72 is also configured to control a translational subsystem 80 of a moveable track 82, along which the rod 20 (or a bundle of rods 20) rests. The moveable track 82 may be configured to move in small (e.g., about 1 foot) or in large (e.g., about 15 foot) increments, or in a continuous fashion depending upon the throughput of the system 50 and particular application-specific design choices. For example, the moveable track 82 may translate the rod 20 along its long axis 26 after a required number of images have been taken of a particular section of the rod 20 (e.g., a section of a few pellets). In such an embodiment, the moveable track 82 would translate the rod 20 such that a new section within the same rod 20 (e.g., a new section of a few pellets) would be imaged by the X-ray source 52 and detector 60. Further, the moveable track 82 may serve to transport the rod 20 (or rod bundle(s)) out of the area defining the inspection system 50. The system controller 68 may coordinate the timing of the actuation of the rod positioning system 78 and the moveable track 82 with the X-ray source 52 and the detector 60. As such, the system controller 68 may control the X-ray source 52 with an X-ray controller 84. Particularly, the X-ray controller 84 may be configured to provide power and timing signals to the X-ray source 52. For example, in one embodiment, the X-ray controller 84 may provide the proper power to the X-ray source 52 to probe the rod 20 for surface defects. In addition, the X-ray controller 84 may provide a different amount of power to the X-ray source, such that the rod 20 is probed for non-surface defects, such as cracks in the pellets 18, and the like. The system controller 68, in the embodiment illustrated, also includes processing circuitry 86. The data collected by the image data acquisition system 70 may be transmitted to the processing circuitry 68 for subsequent X-ray image manipulation and inspection. The processing circuitry 86 may include (or may communicate with) a memory 88 that can store data processed by the processing circuitry 86 or data to be processed (such as digital X-ray images produced by the imaging of the rod 20) by the processing circuitry 86. It should be noted that any type of computer accessible memory device capable of storing the desired amount of data and/or code may be utilized by the imaging system 50. Moreover, the memory 88 may include one or more memory devices, such as magnetic, solid state, or optical devices, of similar or different types, which may be local and/or remote to the system 50. The memory 88 may store data, processing parameters, and/or computer programs (e.g., image recognizing and manipulating algorithms) having one or more routines for performing the processes described herein. For example, in one embodiment the memory 88 may store X-ray inspection software. The software may contain code configured to acquire, report, review, and archive information about the rod 20. For example, during operation, the code may allow the processing circuitry 86 to interact with the inspection equipment (X-ray source 52 and detector 60) to collect digital information. The code may contain a database of relevant inspection implementations and/or protocols and may be configured to control the inspection equipment. The software may also accept acquired data and removable media, such as CD and DVD, as well as provide application tools for analysis, enhancement, measurement, and storage of received data. An example of such software is the Rhythm Software Suite produced by GE Inspection Technologies. The processing circuitry 86 may be configured to control features enabled by the system controller 68, e.g., inspection operations and rod movement. For example, the processing circuitry 86 may be configured to receive commands and inspection parameters from an operator via an operator interface 90 typically equipped with, for example, a keyboard, mouse and/or other input devices. An operator may thereby control the system 50 via the input devices. A display 92 coupled to the operator interface 90 may be utilized to observe digital X-ray images of the rod 20 (and thus the pellets 18). Additionally, an image may be printed by a printer 94, which may be coupled to the operator interface 90. In some embodiments, one or more operator interfaces 90 may be linked to the system 50 for outputting system parameters, requesting inspection, viewing images, and so forth. In general, displays, printers, workstations, and similar devices supplied within the system 50 may be local to the data acquisition components, or may be remote from these components, such as elsewhere within a manufacturing facility, or in an entirely different location, linked to the image acquisition system via one or more configurable networks, such as the Internet, virtual private networks, and the like. The processing circuitry 86 may also be coupled to a picture archiving and communications system (PACS) 96. Image data generated or processed by the processing circuitry 86 may be transmitted to and stored at the PACS 96 for subsequent processing or review. It should be noted that PACS 96 might be coupled to a remote client 98, or to an internal or external network, so that others at different locations may gain access to the image data (e.g., a customer who purchases the rods 20). Moving now to FIG. 3, an embodiment of a masking fixture 62 is illustrated from a head-on view. FIG. 3 also illustrates the general configuration of rods 20 and their position with respect to the masking fixture 62 during imaging. For example, FIG. 3 may be considered as being viewed from a perspective perpendicular to the X-ray beam path 106 from of the X-ray source 52. For example, as depicted, beam path 106 is the general direction of X-ray propagation of X-rays 56 and portion of X-rays 58. Further depicted in the illustrated embodiment are multiple rods 20 disposed within the masking fixture 62. According to the present disclosure, it may be desirable to image as many rods 20 as possible at once, with limiting factors including the size of the detector 60, the size of the masking fixture 62, as well as the handling capacity of the rod positioning system 78 and the moveable track 82, among others. Looking down long axis 26, the rod 20 according to present embodiments contains the cladding 16 and fuel pellets 18 having a pellet surface 108. The masking fixture 62, as depicted, is disposed proximate the rods 20, such that substantially no space or only a small space 110 exists between the fixture 62 and one of the rods 20. The spacing between the rod 20 and the masking fixture 62 may be sufficient, in certain embodiments, to allow the rod 20 to freely rotate and translate about and along the long axis 26 with respect to the masking fixture 62. In one embodiment according to the approaches described herein, the masking fixture 62 may curve around the rod 20, such that the cladding 16 of the rod 20, or more specifically a surface of the cladding 16 having a tangential relationship to beam path 106, is substantially masked, allowing only the pellets 18 and surface of the pellets 108 to be imaged. Additionally, a tolerance 112 between the pellets 18 and the cladding 16 may allow a higher contrast with regard to the surface of the pellets 18, such that a an inspection routine may more efficiently identify any pellet defects and/or abnormalities. To substantially limit the amount of radiation that passes by the rods 20, it may be desirable for the masking assembly 62 to be constructed from one or more high Z elements. In one embodiment, the masking assembly 62 is a monolithic structure made of tungsten (W). High Z elements are those elements with a large atomic number, and for the purposes of the present disclosure, high Z elements consist of those elements with an atomic number greater than or equal to 74. In one embodiment the masking assembly 62 contains little or substantially no other materials which may impede the ability of the masking assembly 62 to absorb X-ray radiation. Therefore, the masking assembly 62, in one embodiment is constructed only from elements such as tungsten and high Z-value elements, with the only other materials present in the masking fixture 62 being impurities from the purposely included elements. In another embodiment, it may be desirable to construct the masking fixture 62 such that a lighter weight material (e.g., a polymer) is encapsulated within an outer shell of a high-Z material, such as tungsten. In such an embodiment, the masking fixture may retain its X-ray absorption properties while having a lighter weight than would otherwise be achieved. A perspective view of the masking fixture 62 during operation of the system 50 is illustrated in FIG. 4. In accordance with the present disclosure, the illustrated embodiment depicts the masking fixture 62 as being disposed between the X-ray source 52 and the detector 60. Further depicted is a representation of the radiation 56 which emanates from the X-ray source 52 and out of the beam collimator 54. A portion of the X-ray radiation 56 passes through the rod 20 (radiation 58), and another portion of the X-ray radiation 56 strikes the masking assembly 62. The masking assembly 62, in the depicted embodiment, includes two monolithic structures placed on either side of the rod 20, and the two monolithic structures extend past the rod 20 towards the X-ray source 52 and the detector 60. Such a configuration where the masking structure 62 extends past the rod 20 may substantially limit the amount of scattered radiation that reaches the detector 60 as a result of non-absorptive interactions between the rod 20 and the incident radiation 56. Moving now to FIG. 5, an embodiment of a system 114 employing more than one X-ray source 52 and more than one detector 60 is illustrated. In the depicted embodiment, the system 114 includes an X-ray source-detector combination 116 having the X-ray source 52 (e.g., a first X-ray source 118), the collimator 54 (e.g., a first collimator 120) which controls an amount of radiation 122 that is incident onto the rod 20, and the masking fixture 62 (e.g., a first masking fixture 124) which is configured to protect the detector 60 (e.g., a first detector 126) and limit X-ray scattering off of the rod 20. Further depicted in FIG. 5 are sections of the rod 20, generally denoted using numeral 128, which are imaged one at a time. For example, the first collimator 120 may control the width of the radiation beam 122 to cover a first section 130 of the rod 120. Further, the first detector 126 may have a width slightly larger than the width of the first section 130 (and most subsequent sections) as a result of the X-ray radiation 122 having a fan or cone-shaped beam. For example, the width may be between about ⅛ inch and 5 inches greater in width than the sections 128. As such, the overall features contained within the rod 20 (e.g., the pellets 18) may be inspected by dividing the rod into the sections 128 and imaging each section 128 in sequence. It should be noted that the masking fixture 62, as well as the imaging equipment may be fixed in place, with the rod 20 being moved translationally along the moveable track 82 and rotated using the rod positioning system 78, both of which are described with regard to FIG. 2. In one embodiment, for example, the moveable track 82 may move the rod 20 along its long axis 26 by a length substantially equal to that of the width of sections 128. In some embodiments, by adding more than one source-detector combination, the throughput of inspection of the fuel rods 20 may be increased, as fewer sections 128 may need to be imaged by a single source-detector combination. For example, a typical fuel rod length is about 14 feet. In some embodiments, the fuel rod 20 may be divided equally over the length of 14 feet, with each section 128 being on the order of the width of the detector 60 (e.g., about 16 inches, with the detector being slightly larger). Thus, the rod 20 may be divided into about 11 sections 128 having about a 16-inch width with a smaller 4-inch section 132 remaining, for a total of about 12 sections. Thus, one X-ray source-detector combination would have to perform 12 different inspection sequences. However, by using more than one properly-spaced source-detector combination, the number of required inspection sequences may decrease. In some embodiments it may be desirable to have two X-ray source-detector combinations side-by-side, such that two consecutive sections 128 may be imaged substantially simultaneously. Three combinations which are side-by-side may image three consecutive sections 128, and so on. While multiple source-detector combinations may be useful, the capital cost of equipment as well as the floor space needed for the equipment that accompanies each combination (e.g., chillers and power supplies) may also be a consideration. Further, the imaging equipment as well as the accompanying chillers, power supplies, and so forth may limit the ability to place the combinations directly adjacent to one another (e.g., within less than 16 inches of one another). As such, according to the present disclosure, the first X-ray source-detector combination 116 may be accompanied by one or more additional X-ray source-detector combinations (e.g., a second X-ray source-detector combination 134), with an integer number of approximately one section 128 length(s) in between each combination. As illustrated, the additional X-ray source-detector combinations contain similar equipment and configurations to that of the first X-ray source-detector combination 116, and generally includes the X-ray source 52 (e.g., a second X-ray source 136) and the beam collimator 54 (e.g., a second beam collimator 138) that is configured to control the shape and area of an X-ray beam 140 during operation. The masking fixture 62 (e.g., a second masking fixture 142) is provided which is configured to substantially limit scattering and bypass radiation, and the detector 60 (e.g., a second detector 144) is provided to produce digital X-ray images of a second section 146 being imaged. According to the present disclosure, including multiple X-ray source-detector combinations (116, 134) within the pellet inspection system 114 may yield a number of possible configurations. For example, the first X-ray source-detector combination 116 may be configured to perform tangential imaging (imaging of surface defects) of the pellets 18 within the fuel rod 20, while the second X-ray source-detector combination 134 may be configured to perform non-tangential imaging (imaging of deeper defects, such as cracks) of the pellets 18 within the fuel rod 20, and/or vice-versa. In one embodiment, such configurations may be accomplished by configuring the first X-ray source 118 to generate X-ray radiation 122 having an energy less than about 450 kV. As such, the second X-ray source 136 may be configured to generate X-ray radiation 140 having an energy greater than about 450 kV. Accordingly, it may be possible to configure system 114 such that one X-ray source-detector combination performs a certain type of scan (e.g., tangential), while the other performs a different type (e.g., non-tangential). Either combination 118, 136 may perform either mode of scanning, the choice of which may be purely one of design and throughput. In another embodiment, the system 114 may be configured such that the number of inspection sequences is decreased by performing non-tangential and tangential imaging together using a single X-ray source-detector combination (118 or 134), with each combination (118 or 134) performing inspection sequences on different sections 128. In such an embodiment, the first combination 116 may perform an inspection sequence on section 130, followed by an inspection sequence on section 148 (the adjacent section). Accordingly, the second combination 134 may perform an inspection sequence on section 146, followed by an inspection on section 150. Such movement is depicted in FIG. 5 as dashed boxes representative of the positions of the detectors 126, 144 at a second position 152. Of course, to avoid dual-inspection of section 146, the rod 20 may be moved translationally (by the moveable track 82, for example) over a length substantially equal to three times the width of one of the sections 128, such that in a third sequence, the first combination 116 inspects section 154 and the second combination 134 inspects section 156 at a third position 158. Other configurations may be possible using multiple X-ray source-detector combinations, and are considered to be within the scope of this disclosure. The systems described above may be used singularly or in combination with one another using the flow chart illustrated within FIG. 6, which is a flow-chart depicting an embodiment of a method 170 for inspecting fuel pellets 18 in accordance with the present disclosure. Method 170 may be performed by a system controller, such as the system controller 68, or a human operator, or a combination of both, as will be apparent from the following description. Method 170 includes, among other features, providing (block 172) the rod 20 (or bundles of rods 20) to an inspection area (such as the inspection systems 50, 114). The rod or rods 20 may be provided, for example, by the moveable track 82 and/or the rod positioning system 78. In some embodiments, only a portion of the rod 20 (or bundle of rods 20) may be provided to the inspection area, such as that described in FIG. 5. In one embodiment, for example in an automated routine, the system controller 68 may direct the moveable track 82 to provide the rod 20 or the section 128 of the rod 20 to the inspection area, such that the system controller 68 may begin an imaging sequence. In another embodiment, a human operator may, for example, engage the moveable track 82 to perform the same action. Once the rod 20 has entered into the inspection area, a human operator or the system controller 68 will engage the system 50, 114 to begin a subset of imaging (block 174). As illustrated, the set of imaging steps generally includes a number of steps that are used to eventually generate a digital X-ray image 176 of the rod 20. The imaging steps may begin with irradiation (block 178) of the rod 20. The irradiation may be performed by the X-ray source 52 where the system controller 68 may direct the X-ray controller 84 to provide a certain power level to the X-ray source 52. The power level provided may be selected for tangential or non-tangential imaging of the rod 20, and as mentioned may be anywhere between 50 kV and up to 10 MV (e.g., between about 120 kV and 420 kV, or between about 450 kV and 1 MV). Once the rod 20 has been irradiated (block 178), the radiation that passes through the rod 20 (e.g., radiation 58) is detected by the detector 60 (block 180). For example, in the embodiments described where the detector 60 includes the scintillator 64 and the diode array 66, the radiation that passes through the rod 20 may be absorbed by the scintillator 64, where the scintillator generates light that is passed to the diode array 66. The diode array 66 then converts, through a photovoltaic effect, the light from the scintillator 64 into electrical signals, which may be amplified and digitized by a transistor, such as a field-effect transistor disposed onto the diode array 66, which generates the digital X-ray image data 176. To complete the imaging steps, the digital X-ray image 176 may be directed to the image data acquisition system 70 and eventually to the processing circuitry 86 for processing the image data to detect pellet defects (block 182). For example, the processing circuitry 86 may execute code stored in memory 88 directed to the recognition of pellet abnormalities and similar features. The X-ray controller 84, having directed the X-ray source 52 to emit X-ray radiation at a discrete energy level (or range), may provide the processor 86 with such information, such that the code is executed in a manner where the defect detection is focused on surface defects of the pellets 18 or on non-surface defects of the pellets 18. Further, the processor 86 may execute (block 182) code configured to correct any abnormalities within the digital X-ray image 176. For example, the digital X-ray image 176 may have one or more bright spots or areas that result from X-ray scattering due to interactions of the incident X-ray beam with the rod 20 or other features. In an embodiment, the system controller 68, if the processor 86 detects such abnormalities in the digital image 176, may start a shift imaging routine 184. In such an embodiment, the system controller 68 may direct the rod positioning system 78 to perform a positional micro-displacement step 186. To perform positional micro-displacement, the rod positioning system 78 may move the rod 20 by rotating the rod 20 about its long axis 26 at an angle that is substantially smaller than that used for imaging the entire circumference of the pellets 18, which is described hereinbelow. Alternatively or additionally, the rod positioning system 78 may translate the rod 20 along its long axis 26 by a length on the order of a few microns to a millimeter (e.g., about 1 microns to about 1 millimeter). The system controller 68 may then direct the system 50 to perform subsequent imaging steps 174 (block 188) to produce a digital image 190 which is slightly shifted from the digital X-ray image 176. In some embodiments, the processor 86 may then execute code stored in memory 88 that is configured to correlate each image with the other (e.g., correct for the micro-displacement via registration of the two images) and divide (block 192) the digital image 176 by the digital image 190, or vice-versa. In other embodiments, no digital shift of pixels may be necessary. Such embodiments may arise when the positional micro-displacement 186 is on the order of the width of a pellet defect, which may allow scatter and background to be subtracted and/or normalized, resulting in the acquisition of defect information. The division may be performed such that each pixel of one image is divided by the same pixel in another image. The quotient is represented as a digital representation 194, which may be a digital map of the digital image 176, where areas (pixels) are represented as a number. For those pixels (or sets of pixels) that have a value that is not one, this may represent a defect in a pellet 18 or a scattering event. Thus, the shift imaging sequence 184 may be useful for defect detection in addition to image correction (block 196). In one embodiment of imaging correction and/or flagging, the processor 86 may be able to distinguish defects in the pellets 18 from scattering events by pattern recognition. For example, the memory 88 may contain a database of common shapes or artifacts observed when the pellet 18 exhibits a defect. The processor 86 may compare an area of similar non-unitary numbers within the digital representation 194 against the database. If the area matches a database shape, then the digital image 176 may be flagged for the presence of a possible defect. Conversely, if the area does not match a database shape, or if the area is on the order of a single pixel, the processor 86 may correct the area using, for example, an anti-aliasing algorithm. This process may be performed over any area within the digital representation 194, which may result in the correction and/or flagging of the digital image. Once processing and optional shift imaging are complete, the rod 20 may be rotated about its long axis 26 at an angle (block 198) in order to image a different portion of the pellets 18 (block 200), the imaging generally including imaging steps 174. Imaging of the different portion or section of the pellets 18 may be performed during the inspection process, in some embodiments, using the rod positioning system 78. For example, the system controller 68 may direct the rod positioning system 78 to rotate the rod 20 at a given angle. The angle of rotation may be selected for a number of reasons, including the desired time allotment for imaging a section of the rod 20, the mechanical limitations of the rod positioning system 78, the requirements of the processing circuitry 86 that may be used to process image data, and the energy that is used to image the pellets 18 to name a few. For example, in some embodiments, the angle may be between 11.25° and 90° (e.g., about 11.25°, 22.5°, 45°, or 90°), such that the number of times step 198 is performed is directly dependent upon the number of rotations necessary at the chosen angle to eventually perform a full 180° rotation. In one embodiment, the angle may be about 11.25°, such that 16 rotations may be necessary to complete a full 180° rotation. In some embodiments, it may be desirable to have a relatively small angle of rotation (e.g., 11.25° or 22.5°) in order to increase the probability of detecting a defect in the pellets 18. In another embodiment, the rod positioning system 78 may continuously rotate the rod 20 about its long axis. In such an embodiment, the system 50, 114 may operate on a substantially real-time basis, with digital images being captured at a rate of about 30 Hz. After performing rotation and imaging, a query 200 is performed to ascertain whether full 180° rotation and imaging has been completed. If the full 180° imaging has not been completed, the method 170 may cycle back to irradiation of the rod (block 178). In embodiments where full 180° rotation and imaging has been completed, the rod 20 is then translated (block 202) along its long axis 26 by the moveable track 82, such that a new section of the rod 20, having non-imaged pellets are imaged by the system 50 (block 174). Translation of the rod 20 may include a movement of anywhere from a few inches to several feet, as mentioned. In one embodiment, the length that the rod 20 is moved may be substantially equal to the width of the area of observation between the X-ray source 52 and the detector 60. As with rotation followed by imaging, a query 204 is performed to ascertain whether the full length of the rod 20 has been imaged. If the full length of the rod 20 has not been imaged, the method 170 may cycle back to irradiation of the rod (block 178). Once full imaging of the length of the rod 20 is complete, the entire rod 20 may be processed (block 206). The processing may involve processing all of the acquired X-ray image data for the sections, as well as for the entire rod 20. In certain embodiments, due to the ability to acquire substantially real-time images of the rod 20, it may be desirable to perform a three-dimensional reconstruction of the rod 20 or sections of the rod 20, such that a digital representation of the rod 20 may be stored in a database and/or reviewed by the processor 86 or human operator. In one embodiment, it may be desirable to perform such three-dimensional reconstruction on rods that contain potentially defective pellets. In some embodiments, the processor 86 may flag (block 208) certain sections of the rod 20 or the entire rod 20, such that rods 20 that have acceptable sections may be re-worked (sent to a re-working station) and completely disassembled to salvage pellets 18 with no defects. Rods 20 that are not flagged or rejected may be sent on for further storage until they are packed and shipped to a customer. This written description uses examples to disclose the invention, including the best mode, and also to enable any person skilled in the art to practice the invention, including making and using any devices or systems and performing any incorporated methods. The patentable scope of the invention 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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06240153& | claims | 1. An apparatus for cleaning a nuclear reactor stud comprising: a housing having a sealable compartment for enclosing the reactor stud during cleaning; said sealable compartment having a base, a top cover, and side walls; a port in the top cover of said sealable compartment for lowering the reactor stud into said sealable compartment; a port covering for the port for sealing the sealable compartment during cleaning of the reactor stud; a turntable mounted in said sealable compartment for mounting the reactor stud with the longitudinal axis of the stud in a substantially vertical position and for rotating the reactor stud about the longitudinal axis; a cleaning mechanism support within said sealable compartment; said cleaning mechanism support being movable from a first position for entry of the reactor stud onto the turntable to a second position for contact of said cleaning brush mechanism with the reactor stud; cleaning brush mechanisms rotatably mounted on said cleaning mechanism in said compartment for contacting and cleaning the reactor stud; said cleaning brush mechanisms comprising brushes which are cylindrically shaped, are rotatable around their longitudinal cylindrical axis, and are horizontally moveable between a first position out of contact with said reactor stud and a second position in contact with said reactor stud; and said reactor stud has a threaded portion along its length and said cylindrical brushes are of a length corresponding to the length of the threaded portion of the reactor stud. a housing having at least two sealable compartments for enclosing reactor studs during cleaning; said sealable compartments each having a base, a top cover, and side walls; a port in the top cover of each said sealable compartment for lowering a reactor stud into said sealable compartment; a port covering for the port in the top cover of each said sealable compartment during cleaning of the reactor stud; a turntable mounted in each said sealable compartment for mounting a reactor stud with the longitudinal axis of the stud in a substantially vertical position and for rotating the reactor stud about the longitudinal axis; a cleaning mechanism support within said sealable compartment; a cleaning mechanism comprising brushes rotatably mounted on said cleaning mechanism support for contacting and cleaning a reactor stud; a drive mechanism mounted on the top cover of the sealed compartment for rotating said brushes of said cleaning mechanism; a cleaning fluid circulation system comprising a sump, an application nozzle mounted in each said compartment for directing cleaning fluid on a reactor stud a pump for delivering cleaning agent from the sump to each nozzle, a drain for collecting the cleaning agent from the reactor stud and returning the cleaning agent to the sump, and a filter for removing contaminants from the cleaning agent. said sealable compartments having a base, a top cover, and side walls; a port in the top cover of each said sealable compartment for lowering the reactor stud into said sealable compartment; a port covering for the port in the top cover of each said sealable compartment during cleaning of the reactor stud; a turntable mounted in each of said sealable compartments for mounting the reactor stud with the longitudinal axis of the stud in a substantially vertical position and for rotating the reactor stud about the longitudinal axis; a cleaning mechanism support within said sealable compartment; a cleaning mechanism comprising brushes rotatably mounted on said cleaning mechanism support for contacting and cleaning the reactor stud; a brush drive motor for rotating said brushes of said cleaning mechanism; a drive mechanism for rotating said brushes on said cleaning mechanism support; and a cleaning fluid circulation system, said system comprising a sump, an application nozzle for applying cleaning agent to the reactor stud, a pump for delivering cleaning agent from the sump to the nozzle, a drain for collecting the cleaning agent from the reactor stud and returning the cleaning agent to the sump, and a filter for removing contaminants from the cleaning agent. 2. The apparatus of claim 1, wherein said housing comprises at least two sealable compartments for enclosing reactor studs during cleaning. 3. The apparatus of claim 1, wherein said sealable compartment further comprises a viewing port for observing the interior of said compartment. 4. The apparatus of claim 1, wherein said sealable compartment further comprises a sealable access door. 5. The apparatus of claim 1 further comprising a cleaning fluid circulation system comprising a sump, an application nozzle for directing cleaning fluid on the reactor stud, a pump for delivering cleaning agent from the sump to each nozzle, a drain for collecting the cleaning agent from the reactor stud and returning the cleaning agent to the sump, and a filter for removing contaminants from the cleaning agent. 6. An apparatus for cleaning two or more nuclear reactor studs comprising: 7. The apparatus of claim 6, wherein each said sealable compartment further comprises a viewing port for observing a reactor stud. 8. The apparatus of claim 6, wherein each said sealable compartment further comprises a sealable access door. 9. The apparatus of claim 6, wherein said cleaning fluid circulation system further comprises an adjustable spray wand for delivering cleaning agent to the reactor stud. 10. The apparatus of claim 6, wherein said drive mechanism rotates said cleaning mechanism in a direction opposite that of the rotation of said turntable. 11. The apparatus of claim 6 further comprising retractable stabilizing supports for the housing. 12. The apparatus of claim 6, wherein said brushes are cylindrically shaped and are rotatable around the longitudinal axis of the cylinder. 13. The apparatus of claim 12, wherein the reactor stud has a threaded portion along its length and said cylindrical brushes are of a length corresponding to the length of threaded portion of the reactor stud. 14. An apparatus for cleaning a nuclear reactor stud comprising: a housing having at least two sealable compartments for enclosing the reactor stud during cleaning; 15. The apparatus of claim 14, wherein said sealable compartment further comprises a viewing port for observing the reactor stud. 16. The apparatus of claim 14, wherein said sealable compartment further comprises a sealable access door. 17. The apparatus of claim 14, wherein said cleaning fluid circulation system further comprises an adjustable spray wand for delivering cleaning agent to the reactor stud. 18. The apparatus of claim 14, wherein said drive mechanism is rotatably mounted over the top cover of said sealed compartment. 19. The apparatus of claim 14, wherein said drive mechanism rotates said cleaning mechanism in direction opposite that of the rotation of said turntable. 20. The apparatus of claim 14 further comprising retractable stabilizing supports for the housing. 21. The apparatus of claim 14, wherein said cleaning brushes are cylindrically shaped and are rotatable around the longitudinal axis of the cylinder. 22. The apparatus of claim 21, wherein the reactor stud has a threaded portion along its length and said cylindrical brushes are of a length corresponding to the length of the threaded portion of the reactor stud. |
063109344 | claims | 1. An X-ray projection exposure apparatus operating in a vacuum or a reduced-pressure environment, said apparatus comprising: a mask chuck for holding a reflection X-ray mask having a mask pattern thereon in the vacuum or in the reduced pressure environment, a void being formed between the mask and said mask chuck; a wafer chuck for holding a wafer onto which the mask pattern is transferred; an X-ray illuminating system for illuminating the reflection X-ray mask, held by said mask chuck, with X-rays; an X-ray projection optical system for projecting the mask pattern of the reflection X-ray mask onto the wafer held by said wafer chuck with a predetermined magnification; and supply means for supplying the void formed between the mask and said mask chuck with a cooling gas for cooling the mask. a chuck base fixing the mask chuck; and a temperature control mechanism for controlling the temperature of said chuck base. a mask chuck for holding a reflection X-ray mask having a mask pattern thereon in the vacuum or in the reduced pressure environment, a void being formed between the mask and said mask chuck; an X-ray illuminating system for illuminating the reflection X-ray mask, held by said mask chuck, with X-rays; and supply means for supplying the void formed between the mask and the mask chuck with a cooling gas for cooling the mask. a chuck base fixing the mask chuck; and a temperature control mechanism for controlling a temperature of said chuck base. 2. An apparatus according to claim 1, further comprising recovering means for recovering the cooling gas. 3. An apparatus according to claim 1, wherein said mask chuck comprises a plurality of projections for supporting the reflection X-ray mask. 4. An apparatus according to claim 3, wherein a ratio of an area of contact between distal ends of said plurality of projections and the mask to the entire area of the mask is at most 10%. 5. An apparatus according to claim 3, wherein a plurality of voids are formed between said plurality of projections. 6. An apparatus according to claim 1, further comprising a plurality of pin-shaped projections formed on said mask chuck. 7. An apparatus according to claim 1, wherein said mask chuck comprises a static electricity generating mechanism for generating static electricity for attracting and holding the reflection X-ray mask by an electrostatic force. 8. An apparatus according to claim 7, further comprising a detection mechanism for detecting an attraction force with which the reflection X-ray mask is held as a result of the electrostatic force generated by said static electricity generating mechanism. 9. An apparatus according to claim 1, further comprising: 10. An apparatus according to claim 9, wherein said chuck base comprises one of a ceramic material and a glass material. 11. An apparatus according to claim 1, further comprising a grounded earth pawl provided at least at a side of said mask chuck for supporting the reflection X-ray mask. 12. An apparatus according to claim 1, further comprising a reflection X-ray mask having an X-ray reflecting multilayer film, and a mask pattern on said reflection X-ray mask made of an absorbing member formed on the X-ray reflecting multilayer film. 13. An apparatus according to claim 1, wherein the X-rays include vacuum-ultraviolet rays or soft X-rays. 14. An apparatus according to claim 1, further comprising discharging means for discharging the cooling gas. 15. An X-ray projection exposure apparatus operating in a vacuum or a reduced-pressure environment, said apparatus comprising: 16. An apparatus according to claim 15, further comprising recovering means for recovering the cooling gas. 17. An apparatus according to claim 15, wherein said mask chuck comprises a plurality of projections for supporting the reflection X-ray mask. 18. An apparatus according to claim 17, wherein a ratio of an area of contact between distal ends of said plurality of projections and the mask to the entire area of the mask is at most 10%. 19. An apparatus according to claim 17, wherein a plurality of voids are formed between said plurality of projections. 20. An apparatus according to claim 15, further comprising a plurality of pin-shaped projections formed on said mask chuck. 21. An apparatus according to claim 15, wherein said mask chuck comprises a static electricity generating mechanism for generating static electricity for attracting and holding the reflection X-ray mask by an electrostatic force. 22. An apparatus according to claim 21, further comprising a detection mechanism for detecting an attraction force with which the reflection X-ray mask is held as a result of the electrostatic force generated by said static electricity generating mechanism. 23. An apparatus according to claim 15, further comprising: 24. An apparatus according to claim 23, wherein said chuck base comprises one of a ceramic material and a glass material. 25. An apparatus according to claim 15, further comprising a grounded earth pawl provided at least at a side of said mask chuck for supporting the reflection X-ray mask. 26. An apparatus according to claim 15, further comprising a reflection X-ray mask having an X-ray reflecting multilayer film, and a mask pattern on said reflection X-ray mask made of an absorbing member formed on the X-ray reflecting multilayer film. 27. An apparatus according to claim 15, wherein the X-rays include vacuum-ultraviolet rays or soft X-rays. 28. An apparatus according to claim 15, further comprising discharging means for discharging the cooling gas. |
050874128 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The invention is generally related to nuclear reactors and in particular to thermal power reactors for use in outer space. 2. General Background Nuclear reactors designed for use in outer space may be classified according to the energy distribution of the neutrons in the core. This energy distribution can be tailored according to the amount and type of moderator (material that reduces the energy level of neutrons) and reflector (material that returns or reflects neutrons to the core region) used in and around the core. The following three classifications are generally used. First, a fast reactor is one in which little or no moderator is used and the average neutron energy is close to that at which the fission neutrons are born. The first successful space reactor, SNAP-8, and the SP-100 currently under development are examples of this type. These are usually liquid metal cooled reactors and are characterized by relatively high specific fuel mass (Kg/Kw). Second, an intermediate reactor is one in which the average neutron energy at which fission occurs is in the range from a few electron volts (ev) to a few thousand electron volts (Kev). An example of this is the NERVA--type propulsion reactor which is moderated partly by the graphite matrix of the fuel elements and partly by separate columns of zirconium hydride. These reactors are intended for short term operation at very high power and are relatively massive compared to more recent conceptual designs. Third, a thermal reactor is one in which the average neutron energy at which fission occurs is less than one electron volt. At this energy level, the fission cross sections of the important fissile materials become very large and the fissile loading is reduced relative to that required in the first two reactor types. Because of the large fission cross sections thermal reactors require substantial quantities of an efficient moderator between and around the fuel elements. The relatively small amount of fissile material required in a thermal reactor provides important advantages over fast and intermediate reactors. A number of missions in outer space are not presently feasible because of the mass of the propulsion system and/or the on-board power system. In the case of a nuclear system with a solid moderator and high power density, adequate cooling of the moderator imposes a severe mass penalty. Compliance with safety requirements also imposes additional mass penalties. A further problem limiting some missions is degradation of the reactor moderator caused by radiation damage. It can be seen from the above that a need exists for reactors used in space applications with enhanced safety, low specific mass, and capability of extended high power operation without radiation damage to the moderator. SUMMARY OF THE INVENTION The present invention solves the above problems in the form of a liquid moderated and/or reflected reactor. In a gas-cooled nuclear reactor, the fuel elements are surrounded by a liquid moderator which shifts the neutron energy spectrum into a range of high fission probability. This permits a sustained chain reaction and power production with minimum reactor fuel and mass. Heat removal from the moderator/reflector is accomplished by forced circulation of the liquid moderator/reflector over an array of heat conducting fins projecting into the liquid from the fuel element housing thimbles. The heat from the liquid is conducted through the fins to the cooler reactor coolant gas on the inside of the thimbles whereas in conventional pressurized water reactors the heat flow is into the liquid rather than from the liquid. Unlike conventional water moderated reactors, the liquid moderator/reflector is not utilized in the energy conversion process. |
description | 1. Field of the Invention The present invention is generally related to methods of improving nuclear reactor performance during core operation, and more particularly related to methods of improving reactor performance and of operating a core so as to increase a scram effectiveness. 2. Description of the Related Art FIG. 1 is a schematic diagram of a simplified boiling water reactor (BWR); FIG. 2 is a cross-sectional diagram depicting a conventional arrangement of multiple fuel rod bundles within a core of a BWR. A BWR generates power from a controlled nuclear fission reaction. As shown in FIG. 1, a simplified BWR includes a reactor chamber 101 that contains a nuclear fuel core and water. Generated steam may be transferred through pipe 102 to turbine 103, where electric power is generated, then water returns to the core through pipe 104. As shown in FIG. 2, the core 201 is made of approximately five hundred (500) bundles 202 of fuel rods arranged in a given manner within the reactor core. FIG. 3 is a schematic cross-sectional diagram of a conventional arrangement of fuel rods within a single fuel bundle. As shown in FIG. 3, each bundle 301 may contain roughly about one hundred (100) fuel rods 302. Water in the core surrounds the rods. Heat generated by a nuclear reaction is transferred from the rods to the water circulating through the core, boiling some of the water. The heat generated in the core may be controlled to maintain safe and efficient reactor operations. In a BWR, there are basically three modes of heat transfer to be considered in defining thermal limits for the reactor: (i) Nucleate boiling, (ii) transition boiling and (iii) filn boiling. “Nucleate boiling” is a desired efficient mode of heat transfer in which the BWR is designed to operate. “Transition boiling” is manifested by an unstable fuel rod cladding surface temperature which rises suddenly as steam blanketing of the heat transfer surface on the rod occurs. The fuel rod cladding surface temperature then drops to the nucleate boiling temperature as the steam blanket is swept away by the coolant flow, and then rises again. At still higher fuel rod/bundle operating powers, “film boiling” occurs, which results in higher fuel rod cladding surface temperatures. The cladding surface temperature in film boiling, and possibly the temperature peaks in transition boiling, may reach values which could cause weakening of the cladding and/or accelerated corrosion. Fuel rod overheating may be generally defined as the onset of the transition from nucleate boiling to film boiling. The conventional basis for reactor core and fuel rod design is defined such that some “margin,” accommodating various design and operational “uncertainties,” is maintained between the most limiting operating condition and the transition boiling condition, at all times for the life of the core. The onset of transition boiling can be predicted by a correlation to the steam quality at which boiling transition occurs, as which may be referred to as “critical quality.” Steam quality can be readily measured and is generally a function of a given, measured distance above the boiling boundary (boiling length) for any given mass flow rate, power level, pressure and bundle flow geometry, among other factors. A “critical power” may be defined as that bundle power which would produce the critical quality of steam. Accordingly, a “critical power ratio” (CPR) may be defined as the ratio of the critical power to the bundle operating power at the reactor condition of interest. CPR may be descriptive of the relationship between normal operating conditions and conditions which produce a boiling transition. Conventionally, CPR is used to rate reactor design and operation in an effort to assure a safe and efficient operation of the reactor, the CPR is kept above a given value for each fuel assembly in the core. Reactor operating limits may be conventionally defined in terms of the most limiting fuel bundle assembly in the core, which may be defined as the “minimum critical power ratio” (MCPR). Reactor operating limits are typically stated in terms of MCPR. In nuclear power generation engineering principles, it is widely recognized that there is a possibility, however small, that the occurrence of a reactor transient event, combined with the various “uncertainties” and tolerances inherent in reactor design and operation, may cause transition boiling to occur locally at a fuel rod for some given period of time. Accordingly, MCPR operating limits are conventionally set in accordance with a United States Nuclear Regulatory Commission (USNRC) design basis requirement that transients caused by a single operator error or a single equipment malfunction shall be limited such that, taking into consideration uncertainties in the core operating state, more than 99.9% of the fuel rods may be expected to avoid boiling transition during that error or malfunction. A safety limit minimum critical power ratio (SLMCPR) is defined under current USNRC requirements as the MCPR where no more than 0.1% of the fuel rods are subject to boiling transition (also known as NRSBT for Number of Rods Subject to Boiling Transition). The corresponding operating limit MCPR (OLMCPR) describes the core operating conditions such that the MCPR is not lower than the SLMCPR to a certain statistical confidence. During operation of a reactor core, of a BWR, for example, nuclear power production may be controlled in part by control rods. Generally, the control rods may be moved to a deeper position in the core to reduce reactivity in the reactor, or moved further out from the core center or bottom to increase reactivity in the reactor. A scram operation involves the rapid insertion of substantial negative reactivity, usually via spring or hydraulic-assisted injection, of all control rods in the core to a fully inserted position. A reactor scram reduces the fission process within the core to thereby reduce power production. A reactor scram may be initiated automatically by a reactor protection system or manually by a reactor operator, for example. A scram is generally least effective when control rods are either fully inserted in the core or fully withdrawn from the core. In the case where the control rods are fully inserted in the core, the scram target condition has already been met; thus, initiating a scram will not change the reactivity of the core. Namely, this is because all rods are already fully inserted. In the case where the control rods are fully withdrawn from the core, a rate at which the scram reduces reactivity in the core is lowest, since the control rods traverse the longest possible distance (i.e., the distance between full withdrawal and insertion). In other words, once initiated, the scram requires a longer period of time with fully withdrawn control rods. The reactivity in the core is generally lowest at the end of operating cycle (EOC), which may be the period prior to a planned maintenance outage for the reactor. For this reason, control rods in conventional reactors are typically fully removed from the core at EOC, so as to attain the highest available level of reactivity. However, a reactor scram may be required during the EOC. A scram initiated at EOC may be less effective because the control rods may be fully withdrawn. For this reason, the OLMCPR at EOC may be set at a higher level, due to the decreased rate of power reduction during a scram at EOC. An example embodiment of the present invention is directed to a method of improving nuclear reactor performance, including implementing an operational solution for the nuclear reactor using at least one control rod criteria in order to increase scram effectiveness during at least a portion of an operating cycle for the nuclear reactor. For example, the at least one control rod criteria may include a consideration of a partial insertion of control rods during the portion of the operating cycle. In an effort to place the example embodiments of the present invention into context, a general example method for determining control rod insertion during an operating cycle will be described, prior to describing example methods of simulation. General Method of Selecting Control Rods for Partial Insertion FIG. 12 is a flow chart illustrating a process for determining control rod insertion during an operating cycle of a nuclear reactor, according to an example embodiment of the present invention. For example, the process of FIG. 12 may be applied at an end of operating cycle (EOC). As such, the following example embodiment given with respect to FIG. 12 illustrates the process as applied at the EOC. However, it is understood that example embodiments of the present invention are not limited to employment at the EOC, but rather may further be employed at any portion of an operating cycle. Referring to FIG. 12, a user (e.g., a core designer) selects (S1400) at least one control rod to consider for partial insertion in the core at EOC. The user may employ well-known methodologies in making the control rod selection, as is evident to one of ordinary skill in the art. The set of possible control rods for selection may be all control rods in the reactor. For example, control rod selection may be based on information in an operational plan. The operational plan may typically set forth the control rod movements during sequences for a next cycle of operation, for example. In one example, the user selects the control rods which are removed latest from the core (e.g., in a last control rod sequence) according to the operational plan. The latest removed control rods are typically the control rods which may absorb the most reactivity at EOC. Thus, the control rods which may absorb the most reactivity at the EOC may be selected. The user may select a partial-degree of insertion (S1405) at which to simulate the selected control rods. The user may employ experience and/or well-known methods to select the degree of insertion for which to simulate the control rods selected as S1400, as is evident to one of ordinary skill in the art. The selected control rods from S1400 may be simulated (S1410) at the selected degree of insertion (S1405) for the duration of an operation cycle (e.g., including EOC). Example methods of performing this simulation (e.g., trial and error, direct calculation, 3D modeling—SLMCPR Addition, 3D modeling—0.1% NRSBT, etc.) will be described in detail below. The simulation results may include a consideration of scram operation as well as a consideration of normal core operation at EOC. The simulation result from S1410 may be compared (S1415) with desired performance criteria. An example of desired performance criteria may be an operating limit minimum core power ratio (OLMCPR), although the example embodiments may use other desired performance criteria such as peak fuel centerline temperature, as is known to one of ordinary skill in the art. The susceptibility to boiling transition during the transient may be quantified statistically as either (1) the probability that a single rod in the core is susceptible to boiling transition or (2) the expected fraction of total rods in the core susceptible to boiling transition. Such a statistical relationship is possible because each individual trial value of NRSBT has been determined by summing the probabilities that individual fuel rods have CPR values less than 1.0 during the transient. In an example, the nominal value for each NRSBT distribution may be associated with the distribution of initial rod CPR values for all fuel rods in the core. By this process, a relationship can be established between the minimal initial MCPR value for all fuel rods in the core, and the probability and confidence level that the fuel rods will be susceptible to boiling transition during the transient. The minimal initial MCPR value for the core, when using the probability and confidence level established by the USNRC design basis requirement for the number of rods not susceptible to boiling transition during the AOO transient, is by definition the minimum Operating Limit MCPR required to demonstrate compliance with the USNRC. Alternatively, at S1415, the user (e.g., a core designer) may employ his or her experience to determine whether the simulation results from S1410 indicate an acceptable or improved solution at EOC. Based on the evaluation in S1415, the user determines (S1420) whether to continue the above-described process with the consideration of other parameters (e.g., other control rods and/or degrees of insertion). If the user determines the simulated solution is acceptable (output of S1420 is “Yes”), the process ends and the resultant partial insertion of the selected control rods may be implemented in the sequence designated in the operational plan (e.g., at a next cycle of operation). Alternatively, if the desired performance criteria is not met (output of S1420 is “No”) and/or the user wishes to evaluate simulations with other parameters, the process proceeds back to S1400 and repeats with a different degree of insertion (S1405) and/or with a different selection of control rods (S1400). Examples of simulating core operation will now be described. The examples described hereafter relate to a trial and error based method, a direct calculation, three dimensional (3D) modeling with SLMCPR addition and 3D modeling based on 0.1% NRSBT methods. While the simulation methods described below are directed to matching the NRSBT to 0.1%, it is understood that 0.1% is merely a safety standard set by the USNRC. Thus, alternatively, if the safety standard was based on another metric (e.g., as in Europe), similar methods may be employed to satisfy the other metric. In one example, the partial rod insertion may be tested during an actual operation of the nuclear reactor. Thus, referring to FIG. 12, after the user selects at least one control rod (S1400) and a given degree of insertion (S1405) for the selected control rod(s), the user may “simulate” the solution via actual implementation during reactor operation. Sensors in the reactor may store data associated with the core operation, which may be evaluated by the core designer at S1415. FIG. 4 is a graph showing the determination of the NRSBT according to an ideal process. In this direct calculation example, the OLMCPR may be calculated directly, so that for the limiting anticipated operational occurrence (AOO), less than 0.1% of the rods in the core would be expected to experience boiling transition. This approach is described in U.S. Pat. No. 5,912,933 to Shaug et al, for example. As shown in FIG. 4, there is a histogram 400 of rod CPR values 401 versus number of rods 402 at the specific CPR value. While the CPR value is usually associated with a fuel bundle, it actually refers to the limiting rod in a bundle. Each rod in the bundle has a CPR value that is determined by the local power distribution and relative position of the rod within the bundle (R-factor). The lowest CPR of any one rod in the bundle is used to characterize the CPR for the entire bundle. The CPR 401 for a given rod has an associated probability distribution function (PDF) which reflects the uncertainties in its determination. The PDF may be determined experimentally and is shown as an Experimental Critical Power Ratio (ECPR) distribution 410. Thus, if a nominal CPR value (411) is 1.0, then the PDF 410 of probable actual CPR values range from 0.90 to 1.10. The variability in the rod CPR values is due to uncertainties in the initial rod condition, i.e., uncertainties in the measurements of parameters at the reactor operating state (core power) and in the modeling of derived parameters (power distribution). In order to take the effect of a transient event on the CPR values into account, a safety margin may be introduced to CPR values by shifting the acceptable nominal CPR value 405 for the lowest rod CPR to a larger CPR value, i.e., 1.25. The ECPR histogram distribution 403 for the lowest CPR rod is thus shifted such that the entire CPR histogram is above a CPR value of 1.20, and well above a CPR value of 1.0. Moreover, the nominal CPR values 407 for rods other than the lowest CPR rod are above the nominal CPR value, e.g., 1.25, of the lowest CPR rod. During a transient in rod operation, the histogram 407 of rod CPRs shifts to the left to lower CPR values, resulting in the histogram 408. With this shift, the “nominal” CPR value 406 during the transient is at the point, e.g., 1.05, where the minimum CPR value is reached during the transient. The limiting rod will have an associated PDF 404, which includes both the uncertainties in the initial rod conditions and “transient uncertainties.” The maximum change in critical power ratio during the transient (“transient ΔCPR 409”) includes uncertainties in the modeling of the transient and/or uncertainties in both the physical models and plant parameters. FIG. 5 is a flow chart illustrating a sequence of processing steps executable by a data processing system for performing an evaluation of OLMCPR using the ideal process. FIG. 5 is described in detail in commonly-assigned U.S. Pat. No. 6,111,572 to Bolger et al., entitled “Determination of Operating Limit Minimum Critical Power Ratio”, the entire relevant portion of which is incorporated by reference herein and described below. In an example embodiment, this transient ΔCPR 409 and associated OLMCPR may be generated as shown in FIG. 5, and described as follows. Step 1: Assume a set of base core operating conditions using the parameters to run the plant that generates a core MCPR equal to the OLMCPR, as shown by block 501. Step 2: Using the parameters, such as core power, core flow, core pressure, etc., that predict a general bundle CPR set forth in block 506, determine the MCPR for each bundle in the core, as shown by block 502. Step 3: Using parameters, such as rod placement within each bundle and rod power, which change each bundle CPR into individual rod CPR values set forth in block 507, determine the MCPR for each rod in the core, as shown by block 503. Step 4: Using the ECPR probability distribution, generated by Equations 1 and 2, set forth in block 508, determine the percentage of NRSBT in the core by summing the probabilities of each rod in the core that is subject to boiling transition, as shown by block 504. This summation may be shown by Equation 3. ECPR = ( Critical Power Predicted by Correlation ) ( Measured Critical Power ) Equation 1 P i = P ( z i ) = 1 2 π ∫ z i ∞ ⅇ 1 2 u 2 ⅆ u Equation 2 NRSBT ( % ) = 100 N rod × ∑ i = 1 N rod [ P i × ( 1 Rod ) ] Equation 3 where zi indicates a rod operating at MCPR(i), u indicates (MCPR(i)-mean ECPR)/(ECPR standard deviation), Pi and P(zi) indicate a probability that a rod i may experience boiling transition and Nrod indicates a total number of rods. Step 5: Vary the parameters set forth in blocks 506 and 507 for a set number of Monte Carlo statistical trials, as shown by block 505. The Monte Carlo process is well-known in the art and is a general method of collecting data to be used in a simulation. The Monte Carlo method provides approximate solutions to a variety of mathematical problems by performing statistical sampling experiments (e.g., on a computer). The method typically applies to problems with no probabilistic content as well as to those with inherent probabilistic structure. Among all numerical methods that rely on N-point evaluations in M-dimensional space to produce an approximate solution, the Monte Carlo method may have an absolute error of estimate that decreases as N superscript −1/2 whereas, in the absence of exploitable special structure all others have errors that decrease as N superscript −1/M at best. Compile the statistics from all the trials from steps 2 through 4 to generate a probability distribution of NRSBT. Step 6: Compare the value of NRSBT percentage to 0.1%, as shown in block 509. If the percentage is greater than 0.1%, reset the core parameters to different initial conditions in order to comply with the USNRC regulations, as shown in block 510. Similar to Step 1 and block 501, the new initial conditions are assumed to generate an OLMCPR. The determination of NRSBT restarts and loops until the value of NRSBT is equal to 0.1%. Similarly, if the percentage is less than 0.1%, the core parameters are reset to increase the value of NRSBT in order to operate the core more efficiently or with fewer effluents. Step 7: If the percentage of NRSBT equals 0.1%, the assumed value of OLMCPR, which equals core MCPR, complies with the USNRC regulations, as shown by block 511. Accordingly, the operating core conditions are set as the assumed parameters. While the above-described example assumes that the OLMCPR must meet the 0.1% standard, it is understood that, alternatively, the above-described example may be applied to any safety criteria. FIG. 6 is a graph showing the linear addition of ΔCPR to the SLMCPR to determine operating limit minimum core power ratio (OLMCPR), which is the currently approved process. FIG. 6 is described in detail in commonly-assigned U.S. Pat. No. 6,111,572 to Bolger et al., entitled “Determination of Operating Limit Minimum Critical Power Ratio”, the entire relevant portion of which is incorporated by reference herein and described below. In the 3D modeling process with SLMCPR addition for determining simulation results, the OLMCPR determination is divided into two primary steps, as shown by FIG. 6. Using a process similar to the above-described direct calculation, first the SLMCPR is determined so that less than 0.1% of the rods in the core will be expected to experience boiling transition at this value. In other words, 99.9% of the fuel rods in the core will be expected to avoid boiling transition if the MCPR in the core is greater than SLMCPR. Second, the OLMCPR is then established by summing the maximum change in MCPR (as shown by an error factor ΔCPR95/95) expected from the most limiting transient event and the SLMCPR. Since FIG. 6 is somewhat similar to the FIG. 4; thus, only a brief description of its components follows for purposes of brevity. Histogram 600 shows the number of rods at a specific CPR value 602 versus the corresponding CPR value 601. The histogram 608 results with the lowest CPR rod 607 at a value of, e.g., 1.05, which equals the SLMCPR 603. Limiting rod distribution 606 shows the uncertainty in determination of the limiting CPR rod 607. Similar to the above described direct calculation, the SLMCPR 603 is determined when the percentage of NRSBT is equal to 0.1%. However, unlike the above-described direct calculation, the ID modeling process is unable to fully predict and measure certain parameters, such as the power distribution within each bundle and the power distribution along each rod. Thus, the uncertainties in calculating the SLMCPR do not allow equating the OLMCPR with the SLMCPR. Accordingly, the error factor, ΔCPR95/95 605, is linearly added to the SLMCPR 603 to determine the OLMCPR 609. ΔCPR.sub.95/95 605 conservatively corrects for limitations in the calculation of the SLMCPR 603. FIG. 7 is a flow chart illustrating a sequence of processing steps executable by a data processing system for performing a evaluation of the OLMCPR using the currently approved process. FIG. 7 is described in detail in commonly-assigned U.S. Pat. No. 6,111,572 to Bolger et al., entitled “Determination of Operating Limit Minimum Critical Power Ratio”, the entire relevant portion of which is incorporated by reference herein and described below. Using the ID modeling process, the OLMCPR 609 is generated as shown in FIG. 7, and described as follows: Step 1: Assume a set of base core operating conditions using the parameters to run the plant generates a core MCPR equal to the SLMCPR, as shown by block 701. Step 2: Using the parameters, such as core power, core flow, core pressure, bundle power, etc., that predict a general bundle CPR set forth in block 706, determine the MCPR for each bundle in the core as shown by block 702. This process step may have large uncertainties in predicting the bundle power, potentially biasing the calculations. Step 3: Using parameters, such as rod placement within each bundle and rod power, which change each bundle CPR into individual rod CPR values set forth in block 707, determine the MCPR for each rod in the core, as shown by block 703. Individual rod power may be difficult to measure; combining that uncertainty with bundle power distribution uncertainty serves to increase the uncertainty in practical calculations of the SLMCPR. Step 4: Using the ECPR probability distribution set forth in block 708, generated by Equations 1 and 2 shown above, determine the percentage of NRSBT in the core by summing the probabilities of each rod in the core that is subject to boiling transition, as shown by block 704. This summation may be performed using Equation 3 from above. Step 5: Vary the parameters set forth in blocks 706 and 707 for a set number of Monte Carlo statistical trials, as shown by block 705. Compile the statistics from all the trials from steps 2 through 4 to generate a probability distribution of NRSBT. Step 6: Compare the value of percentage of NRSBT to 0.1%, as shown in block 709. If the percentage is greater than 0.1%, reset the core parameters to different initial conditions in order to comply with the USNRC regulations, as shown in block 710. Similar to Step 1 and block 701, the new initial conditions are assumed to generate the SLMCPR. The determination of NRSBT loops until the value of NRSBT is equal to 0.1%. Similarly, if the percentage is less than 0.1%, the core parameters are reset to increase the value of NRSBT in order to operate the core more efficiently. Step 7: If the percentage of NRSBT equals 0.1%, the assumed value of SLMCPR, which equals core MCPR, is the limit at which the core may operate, as shown by block 711. Step 8: Since this process includes relatively uncertain simulations in steps 2 and 3, as shown by blocks 702 and 703, the change in CPR is evaluated at a 95% confidence interval, ΔCPR95/95. The OLMCPR equals the linear addition of the SLMCPR to the ΔCPR95/95. The resulting value of the OLMCPR complies with the USNRC regulations. In this example of a 3D modeling process to attain simulation results, a generic bias may be calculated for a change in critical power ratio during a transient event (ΔCPR/ICPR) and a resulting Probability Distribution Function (PDF) may be used to predict a more accurate OLMCPR without first calculating a SLMCPR. From a large number of experimental trials that take many factors into account, a PDF for a transient referred to as ΔCPR/ICPR is created and the standard deviation in ΔCPR/ICPR is determined for each transient event. A nominal ΔCPR/ICPR for the transient event starting from nominal initial conditions is also determined. Histograms of individual rod CPR values for the minimum point in the transient are created by drawing random values of initial CPR and transient ΔCPR/ICPR uncertainty. The initial critical power ratios (ICPR) are converted, or translated, to MCPRs by a common random value of ΔCPR/ICPR. From the MCPR values, the percentage of NRSBT is calculated for each trial. If the percentage of NRSBT is greater than 0.1%, initial operating conditions are changed and the process is repeated until the NRSBT is equal to 0.1%. The NRSBT distribution histogram is analyzed using statistical methods to determine a “central tendency” of the distribution. Typically the mean or median is used as a statistic to quantify central tendency. The value of this statistic is defined here as the nominal value. In the discussions that follow, examples are given where the mean value is chosen as the nominal value although the present invention is not limited to this choice. Use of the median value or the value of some other statistic for central tendency as the nominal value is also contemplated as part of the example embodiments of the present invention. The uncertainty in the nominal value of the statistic that is used to quantify central tendency is expressed in terms of a “confidence interval” for the nominal value. A confidence interval is defined such that there is a specified probability (usually of 50% or greater) that the interval contains the nominal value. For example, a 95% probability that the interval bounds the mean, defines a 95% confidence interval for the mean. The specified probability used to establish this confidence interval is called the “level of confidence” or confidence level. In accordance with one example, the present invention may include a system including a data processing apparatus programmed to execute specific routines for simulating BWR core operating conditions and for calculating and statistically demonstrating the OLMCPR of a reactor in accordance with the improved method of the present invention as described in detail below. FIG. 8 shows a block diagram of an example data processing system, contemplated for performing the multi-dimensional simulation of reactor core transient response and for the direct evaluation of OLMCPR for a BWR reactor core in accordance with the example embodiments of the present invention. The system may include a central processing unit 801 (CPU), a storage memory 802, user interfacing I/O devices 803 and, optionally, one or more displays 804. Storage memory 802 may include a database (not shown) of reactor plant state information, parameter values and routines for implementing multi-dimensional simulations of core operating conditions and evaluating OLMCPR in accordance with the example method of the present invention as described herein below. For example storage memory 802 may include any well-known memory (e.g., a Read Only Memory (ROM), Random Access Memory (RAM), etc.) A statistical study may be performed for each type of AOO, for each class of BWR plant type, and for each fuel type, for example to determine the generic transient bias and uncertainty in the ΔCPR/ICPR. Enough trials (on the order of one hundred) are made starting with the nominal conditions, using random variations in the model and plant parameters. Uncertainties in initial conditions that contribute to the ΔCPR/ICPR (e.g., core power) are also included in the perturbations. The data are utilized to determine bias and standard deviation on the transient ΔCPR/ICPR. FIG. 9 is a flow chart illustrating a sequence of processing steps used in calculating the OLMCPR using the generic uncertainty in ΔCPR/ICPR. FIG. 9 is described in detail in commonly-assigned U.S. Pat. No. 6,111,572 to Bolger et al., entitled “Determination of Operating Limit Minimum Critical Power Ratio”, the entire relevant portion of which is incorporated by reference herein and described below. A flow chart for an example process of the present invention is shown in FIG. 9. Block 909 remains unvaried throughout the calculation of the OLMCPR, and the ΔCPR/ICPR for individual transient events for each reactor type and fuel type must be determined before the process is used. FIG. 10 is a graph showing the determination of generic uncertainty in ΔCPR/ICPR using the present invention. FIG. 10 is described in detail in commonly-assigned U.S. Pat. No. 6,111,572 to Bolger et al., entitled “Determination of Operating Limit Minimum Critical Power Ratio”, the entire relevant portion of which is incorporated by reference herein and described below. FIG. 10 shows the resulting graph of ΔCPR/ICPR for one specific type of AOO. Histogram 1000 shows the number of trials 1002 with a resulting CPR 1001 for each rod versus the corresponding CPR 1001 values. The PDF 1003 represents the distribution of CPR before the transient event. Each CPR value then changes according to individual ΔCPR/ICPR 1006 values. The aggregate of the transient CPR values yields the PDF 1004 during the transient event. The nominal ΔCPR/ICPR 1005 is defined to be the difference in the nominal CPR value of the PDF 1003 and the nominal CPR value of the PDF 1004. The calculation of the OLMCPR may be as follows. Step 1: Assume a set of base core operating conditions using the parameters to run the plant generates a core MCPR equal to the OLMCPR as shown by block 901. Step 2: Using the parameters, such as core power, core flow, core pressure, bundle power and others, that predict a general bundle CPR set forth in block 907, determine the ICPR for each bundle in the core, as shown by block 902. Step 3: Using parameters, such as rod placement within each bundle and rod power distribution, that change each bundle CPR into individual rod CPR values set forth in block 908, determine the ICPR for each rod in the core, as shown by block 903. Step 4: Using a randomly drawn individual ΔCPR/ICPR 1006 value from the graph of the appropriate transient represented in FIG. 10, MCPR values are projected for corresponding values of ICPR according to Equation 4. FIG. 11 is a graph showing the determination of the NRSBT using the generic uncertainty of ΔCPR/ICPR. FIG. 11 is described in detail in commonly-assigned U.S. Pat. No. 6,111,572 to Bolger et al., entitled “Determination of Operating Limit Minimum Critical Power Ratio”, the entire relevant portion of which is incorporated by reference herein and described below. In FIG. 11, this process is represented by Shift 1109. Histogram 1100 shows the number of rods at a specific CPR value 1102 versus the corresponding CPR value 1101. The histogram 1107 is translated to histogram 1108 during the transient using a randomly selected ΔCPR/ICPR 1006 value. Lowest CPR value 1105 becomes lowest CPR value 1106, and lowest CPR rod PDF 1103 becomes lowest CPR rod 1104. MCPR i = ICPR i ( 1 - ( Δ CPR ICPR ) 1 ) Equation 4 Step 5: Using the ECPR probability distribution shown as PDF 1104 and set forth in block 910, determine the percentage of NRSBT in the core by summing the probabilities of each rod in the core that is subject to boiling transition as shown by block 905. This summation is performed using Equation 3, shown above. Step 6: Vary the parameters set forth in blocks 907 and 908 for a set number of Monte Carlo statistical trials as shown by block 906. Compile the statistics from all the trials from steps 2 through 5 to generate a probability distribution of NRSBT. Step 7: Compare the value of percentage of NRSBT to 0.1% as shown in block 911. If the percentage is greater than 0.1%, reset the core parameters to different initial conditions in order to comply with the USNRC regulations as shown in block 912. Similar to Step 1 and block 901, the new initial conditions are assumed to generate the OLMCPR. The determination of NRSBT restarts and runs until the value of NRSBT is equal to 0.1%. Similarly, if the percentage is less than 0.1%, the core parameters are reset to increase the value of NRSBT in order to operate the core more efficiently or to reduce effluents. Step 8: If the percentage of NRSBT equals 0.1%, the assumed value of OLMCPR, which equals core MCPR, complies with the USNRC regulations as shown by block 913. Accordingly, the operating core conditions are set as the assumed parameters. Example embodiments of the present invention being thus described, it will be obvious that the same may be varied in many ways. For example, while the simulation described with respect to S1410 of FIG. 12 has been described as being one of trial and error, direct calculation and 3D modeling methods, it is evident to one of ordinary skill in the art that any well-known simulation method may be employed to evaluate proposed solutions in consideration of rod placement during an operating cycle of a nuclear reactor (e.g., at the EOC). Further, while acceptable solutions have been described as being in accordance with the 0.1% OLMPCR standard, it is understood that the simulation results (S1410) may be evaluated (S1415) and deemed acceptable (S1420) based on any desired performance criteria. Further, while above described as being applied at the EOC, it is understood that other example embodiments of the present invention may be directed to any portion of an operating cycle for a nuclear reactor. Such variations are not to be regarded as a departure from the spirit and scope of the example embodiments of the present invention, and all such modifications as would be obvious to one skilled in the art are intended to be included within the scope of the following claims. |
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description | The present invention relates to a filter for separating particles in a coolant fluid in a nuclear reactor, of the type comprising channels for circulation of coolant fluid through the filter, at least one channel extending along a channel centerline and comprising an upstream section, a downstream section and an intermediate section extending between the upstream section and the downstream section and being enlarged relative to the upstream section and the downstream section. U.S. Pat. No. 6,876,713 A1 discloses a filter of the above-mentioned type in which the channels are elongated transversely to the flow direction and separating members are provided in the form of elongated circular cylinders each extending transversely to the flow direction in the intermediate section of one of the channels to define in the channel at least one passage bent in the flow direction to catch particles elongated in the flow direction. However, elongated particles extending transversely to the fluid flow direction might flow without being caught by the separating members. An object of the invention is to provide a filter allowing catching particles more efficiently along with limiting the fluid flow resistance of the filter. To this end, the invention provides a filter of the above-mentioned type, comprising at least one separating member defining inside the intermediate section of the at least one channel an annular passage whose axis is substantially coaxial to the channel in the intermediate section. In other embodiments, the filter comprises one or several of the following features, taken in isolation or in any technical feasible combination: the filter comprises centering elements for maintaining the or each separating member spaced from the inner surface of the corresponding intermediate section; the filter comprises spherical separating members; the filter comprises separating members having an elliptical cross section; the filter comprises separating members having a double cone shape; the filter comprises at least one pair of separating members disposed inside the same intermediate section to define a pair of coaxial annular passages whose axis is substantially coaxial to the channel centerline in the intermediate section a first separating member being tubular and defining an outer annular fluid flow passage with the inner surface of the intermediate section, the second separating member being disposed inside the first separating member and defining therein an inner annular passage; the filter comprises at least one separating member having at least one fluid flow hole passing through the separating member; the filter comprises at least a first set of channel and a second set of channels, the intermediate sections of the channels of the first set being offset relative to the intermediate sections of the channels of the second set along the filter main flow direction of the coolant fluid through the filter; the channels are arranged in a pattern such that each channel of the first set is surrounded by channels of the second set; the filter comprises a one-piece filtering plate having the channels extending there through, each channel being defined by a duct extending through the filtering plate and having a narrow section and a enlarged section, and a tubular insert inserted inside the enlarged section; the filter comprises a filtering plate comprising at least two stacked parts, each channel extending through the different parts and each separating member being disposed between two of plate-like parts; the filtering plate comprises a lower part, an upper part and an intermediate part interposed between the upper part and the lower part, each separating member being disposed between the intermediate part and one of the upper and lower parts; The invention also provides to a nuclear fuel assembly lower nozzle defining a filter as defined above. In an embodiment, the nuclear fuel assembly comprises a bundle of fuel rods and an armature for supporting the fuel rods, the armature comprising a lower nozzle and an upper nozzle, the fuel rods extending between the nozzles, wherein the lower nozzle defines a filter as defined above. The nuclear fuel assembly 2 of FIG. 1 comprises a bundle of nuclear fuel rods 4 and an armature 6 for supporting the fuel rods 4. The fuel assembly 2 is elongated along a longitudinal axis L extending vertically when the fuel assembly 2 is disposed inside a nuclear reactor. In the following, the terms “upper” and “lower” refer to the position of the fuel assembly 2 in a nuclear reactor. Each fuel rod 4 comprises a tubular cladding, pellets of nuclear fuel stacked inside the cladding and caps closing the ends of the cladding. The armature 6 comprises a lower nozzle 8, an upper nozzle 10, a plurality of guide tubes 12 and a plurality of spacer grids 14. The lower nozzle 8 and the upper nozzle 10 are spaced one from the other along axis L. The guide tubes 12 connect the lower nozzle 8 and the upper nozzle 10 together. The guide tubes 12 extend parallel to axis L and maintain a predetermined spacing between the nozzles 8, 10. Each guide tube 12 opens upwards through the upper nozzle 10 for allowing insertion of a control rod into the guide tube 12. The spacer grids 14 are distributed along the guide tubes 12 between the nozzles 8, 10 and connected to the guide tubes 12. The fuel rods 4 extend parallel to axis L between the nozzles 8, 10 and through the spacer grids 14. The spacer grids 14 support the fuel rods 4 transversely and longitudinally relative to axis L. The lower nozzle 8 defines a filter for filtering coolant fluid flowing upwardly along the fuel assembly 2. The lower nozzle 8 comprises a filtering plate 16 and a plurality of feet 18 extending downwardly from the filtering plate 16. As illustrated on FIG. 1, the feet 18 rests on the core support plate 20 of a nuclear reactor and the filtering plate 16 extends horizontally above at least one water inlet 22 of said core support plate 20 adapted for allowing water to flow upwardly out from the inlet 22 and through the filtering plate 16 in a main flow direction illustrated by arrow F and substantially parallel to the longitudinal axis L The filtering plate 16 comprises a plurality of analogous fluid flow channels 24 for allowing water to flow through the filtering plate 16, only one channel 24 being visible on the partial sectional views of FIG. 2 and FIG. 3. FIG. 2 is a side sectional view and FIG. 3 is a top sectional view along on FIG. 2. The channel 24 extends along a channel centerline A through the filtering plate 16 from a lower face 26 facing downwardly to an opposed upper face 28 facing upwardly. The channel 24 comprises along its channel centerline A an inlet upstream section 30, an outlet downstream section 32 and an intermediate section 34 extending between the upstream section 30 and the downstream section 32. In the embodiment illustrated on FIG. 2, the centerline A is straight and substantially parallel to the longitudinal axis L of the fuel assembly 2. The downstream section 32 and the upstream section 30 have substantially the same cross-section. The intermediate section 34 is enlarged transversely to the channel centerline A relative to each of the upstream section 30 and the downstream section 32. The upstream section 30 and the downstream section 32 each have a cylindrical shape of circular cross-section, and the intermediate section 34 is spherical, the inner diameter of the intermediate section 34 being greater than the inner diameter of each of the upstream section 30 and the downstream section 32. The filtering plate 16 comprises a separating member 36 disposed inside the intermediate section 34 of the channel 24 so as to define therein an annular passage 38 whose axis is parallel and namely coaxial to the channel centerline A in the intermediate section 34. The annular passage 38 is bent along channel centerline A. To this end, the separating member 36 has a dimension transverse to the channel centerline A superior to that of each of the upstream section 30 and the downstream section 32. The separating member 36 is spherical and has an outer diameter that is inferior to the inner diameter of the intermediate section 34 but superior to the inner diameter of each of the upstream section 30 and the downstream section 32. The filtering plate 16 comprises centering elements 40 (FIG. 3) disposed between the separating member 36 and the inner surface of the intermediate section 34 to maintain the spacing between the separating member 36 and the inner surface of the intermediate section 34 and define the annular passage 38. The centering elements 40 are provided in the form of bosses distributed on the outer surface of the separating member 36. Optionally or alternatively, centering bosses are distributed on the inner surface of the intermediate section 34. In use a flow of coolant fluid flows through the channel 24 mainly parallel to the channel centerline A. The coolant fluid might carry debris which can potentially damage the fuel rods 4 (FIG. 1). The annular passage 38 allows an important flow of coolant fluid while having locally small transverse dimensions to efficiently catch debris. The bending of the annular passage 38 along the channel centerline A enables to efficiently catch elongated debris extending parallel to the channel centerline A. The annular passage 38 coaxial to the channel centerline A is curved in a plane perpendicular to the channel centerline A (FIG. 3), which enhances filtering by enabling to catch elongated particles extending in any direction inclined relative to the channel centerline A. The annular passage 38 having an axis parallel to channel centerline A in the intermediate section 34, and thus to the main direction of the channel flow in the intermediate section 34, enables to limit the flow resistance through the channel 24. FIGS. 4 and 5 are enlarged sectional views corresponding to FIGS. 2 and 3 respectively. FIG. 5 is a sectional view along V-V on FIG. 4. As illustrated on FIGS. 4 and 5, the filtering plate 16 comprises a plurality of channels 24 extending parallel to each other through the filtering plate 16, one separating member 36 being disposed inside the intermediate section 34 of each channel 24. The filtering plate 16 comprises a first set S1 of channels 24 and a second set S2 of channels 24, the channels 24 of the first set S1 having their intermediate sections 34 offset relative to that of the channels 24 of the second set S2 along the direction perpendicular to the plane of the filtering plate 16. Each channel 24 of the first set S1 is adjacent to channels 24 of the second set S2 with a reduced distance between the channel centerlines A, by providing intermediate sections 34 overlapping in view from the main flow direction F (see FIG. 5). As a result, a great number of channels 24 is provided in a limited area and the fluid flow resistance of the lower nozzle 8 is reduced. The channels 24 of the first set S1 and the channels 24 of the second set S2 are staggered in rows (FIGS. 4 and 5) to provide a very compact arrangement. The intermediate section 34 and the separating member 36 disposed therein can vary in shape as it will appear from the following description of alternative embodiments illustrated on FIGS. 6 to 9. In the alternative embodiment of FIG. 6, the separating member 36 has an elliptical cross-section in a plane parallel to the channel centerline A (FIG. 6). The intermediate section 34 has a corresponding shape. As a result, the curvature radius of the annular passage 38 along the channel centerline A is made small to increase the debris catching capacity. In the alternative embodiment of FIG. 7, the separating member 36 has a double cone shape formed of two coaxial conical portions 42 of circular base pointing in opposite direction. The inner surface of the intermediate section 34 has a corresponding shape and is formed of two coaxial conical segments 44 of circular base converging in opposite directions. As a result, the annular passage 38 is angled whereby debris catching can be increased. In the alternative embodiment of FIG. 8, a pair of separating members 36, 46 is disposed inside the intermediate section 34 to define two co-axial annular passages 38, 48. The first separating member 36 is tubular and defines an outer annular passage 38 with the inner surface of the intermediate section 34. The second separating member 46 is disposed inside the first separating member 36 and defines therein an inner annular passage 48. The inner annular passage 48 is bent along the channel centerline A and its axis is coaxial to the channel centerline A. As a result, the fluid flow resistance is reduced while filtering efficiency is maintained. In the alternative embodiment of FIG. 9, the second separating member 46 is provided with fluid flow holes 50 extending through the second separating member 46 for reducing fluid flow resistance. The holes 50 are inclined relative to the channel centerline A in the intermediate section 34. Two intersecting holes 50 are illustrated on FIG. 9. FIG. 10 is a sectional view along X-X on FIG. 9 and illustrates the centering elements 40 provided between the first separating member 36 and the inner surface of the intermediate section 34, and second centering elements 51 provided between the second separating member 46 and the first separating member 36 to maintain the annular space between the inner surface of the first separating member 36 and the outer surface of the second separating member 46. The embodiments of FIGS. 2 and 6 to 9 may be combined. It is for example possible to provide holes 50 extending through a separating member 36 such as that of FIGS. 2, 6 and 7. FIGS. 11 to 15 illustrate successive steps of a method for manufacturing a filtering plate 16 as that of the lower nozzle 8 of FIGS. 1 to 5. In a first step illustrated on FIG. 11 a solid plate 16 is provided. In a second step illustrated on FIG. 12, a plurality of through ducts 52 of circular cross-section is machined through the plate 16. Each duct 52 extends between the lower face 26 and the upper face 28 of the plate 16. In a third step illustrated on FIG. 13, a section of each duct 52 is enlarged, e.g. by machining the plate 16, whereby each duct 52 subsequently has an enlarged section 54 and a narrow section 56. The enlarged section 54 of each duct 52 of a first set S1 extends from the upper face 28 of the filtering plate 16, and the enlarged section 54 of each duct 52 of a second set S2 extends from the lower face 26 of the filtering plate 16. In a fourth step illustrated on FIG. 14, separating members 36 are introduced in the enlarged section 54 of each duct 52. In a fifth step illustrated on FIG. 15, a tubular insert 58 is introduced in the enlarged section 54 of each duct 52. The tubular insert 58 has a bore 60 of smaller cross-section than the separating member 36. Each channel 24 is thus defined in a respective duct 52, one of the downstream section 32 and the upstream section 30 being defined by the bore 60 of the tubular insert 58 and the other by the narrow section 56 of the duct 52. The intermediate section 34 of the channel 24 is defined inside the enlarged section 54. Each separating member 36 is retained in the corresponding channel 24 by the tubular insert 58. The ducts 52 of the first set S1 and the ducts 52 of the second set S2 enable to obtain two sets of channels 24 having offset intermediate sections 34 with a transverse spacing between channels 24 of the first set S1 and channels 24 of the second set S2 inferior to the inner diameter of the intermediate sections 34 as explained with reference to FIGS. 4 and 5. FIGS. 16 and 17 illustrate another method of manufacturing a filtering plate 16 as that of the lower nozzle 8 of FIGS. 1 to 5. As illustrated on FIGS. 16 and 17, the filtering plate 16 comprises superimposed plate-like parts 62, 64, 66, each channel 24 extending through the different parts 62, 64, 66 and each separating member 36 being disposed between two of plate-like parts 62, 64, 66. The plate-like parts 62, 64, 66 have aligned holes 68 of different diameters formed therein so as to define channels 24 upon stacking the parts 62, 64, 66. The intermediate section 34 of each channel 24 is defined at the junction of two adjacent parts 62, 64, 66, by enlarged sections of the holes 68. In the illustrated embodiment, the filtering plate 16 comprises a lower part 62, an upper part 64 and an intermediate part 66. The intermediate sections 34 (FIG. 17) of the channels 24 of the first set S1 are formed at the junction between the upper part 64 and the intermediate part 66, and the intermediate sections 34 of the channels 24 of the second set S2 are formed at the junction between the intermediate part 66 and the lower part 62. The aligned holes 68 of the superimposed parts 62, 64, 66 are arranged to form the first set S1 of channels 24 and the second set S2 of channels 24 with offset intermediate section 34. One separating member 36 is disposed inside each channel 24 of the first set S1 between the upper part 64 and the intermediate part 66, and one separating member 36 is disposed inside each channel 24 of the second set S2 between the intermediate part 66 and the lower part 62. This arrangement allows obtaining two sets S1, S2 of channels 24 having offset intermediate sections 34 with a transverse spacing between the channel centerlines A inferior to the inner diameter of the intermediate sections 34. This method allows providing more than two sets S1, S2 of channels 24 by increasing the number of intermediate sections 34 forming the filtering plate 16. In the illustrated embodiments, the channels extend along straight channel centerlines A (or channel axes) which are substantially parallel to the fuel assembly axis L in order to limit the flow resistance. In alternative embodiments, channels extend along straight channel centerlines which are inclined at an angle comprised between 0 and 45° relative to the fuel assembly axis L. The invention is not to be limited to channels extending along straight channel centerlines. A channel may also extend along a broken or curved centerline. Hence, in a general manner, the separating member disposed inside the intermediate section is arranged to define therein an annular passage whose axis is substantially coaxial to the channel centerline in the intermediate section. The axis of the annular passage is thus substantially parallel to the main direction of the coolant fluid flow through the intermediate section of the channel, whereby particles elongated in a direction inclined relative to the channel centerline in the intermediate section are efficiently caught. Beside, providing an annular passage bent along the channel centerline allows efficiently catching particles elongated parallel to the channel centerline in the intermediate section. Moreover, the fluid flow resistance is kept low since the axis of the annular passage is substantially coaxial to the channel centerline in the intermediate section. The invention is applicable to lower nozzles of fuel assembly for Light Water Reactors (LWR) such as Boiling Water Reactors (BWR) or Pressurized Water Reactors (PWR), and more generally to any filter. |
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claims | 1. An apparatus for manufacturing a disc-shaped object, comprising:an electron beam irradiation apparatus including a first rotational unit provided in an openable/closable shield container and accommodating a disc-shaped object rotationally driven, and an electron beam irradiation unit for irradiating the surface of said on-rotating disc-shaped object with electron beams;a chamber including a second rotational unit capable of accommodating said disc-shaped object and an exchange chamber that is air-tightly closable and openable/closable independently of said shield container; anda rotational unit for exchanging said first and second rotational units with each other by rotating said first rotational unit in said shield container and said second rotational unit in said exchange chamber,wherein said electron beam irradiation unit and said disc-shaped object are relatively moved when irradiating said on-rotating disc-shaped object with the electron beams. 2. An apparatus for manufacturing a disc-shaped object according to claim 1, wherein a width of said electron beam irradiation unit in a direction orthogonal to a rotating direction of said irradiation target object within a rotating plane of said disc-shaped object, is smaller than a radius of said disc-shaped object. 3. An apparatus for manufacturing a disc-shaped object according to claim 1, wherein a rotating speed of said disc-shaped object is changed corresponding to a position of the irradiation by said electron beam irradiation unit over said disc-shaped object. 4. An apparatus for manufacturing a disc-shaped object according to claim 3, wherein said first rotational unit and said second rotational unit are so constructed as to be capable of revolving, and said first rotational unit irradiates the surface of said on-rotating disc-shaped object with the electron beams from said electron beam irradiation unit. 5. An apparatus for manufacturing a disc-shaped object according to claim 3, wherein the rotating speed of said disc-shaped object is decreased when said electron beam irradiation unit irradiates an outer periphery side of said disc-shaped object with the electron beams and is increased when irradiating an inner periphery side with the electron beams. 6. An apparatus for manufacturing a disc-shaped object according to claim 1, wherein a moving velocity of said electron beam irradiation unit is changed corresponding to the position of the irradiation by said electron beam irradiation unit over said disc-shaped object. 7. An apparatus for manufacturing a disc-shaped object according to claim 6, wherein the moving velocity of said electron beam irradiation unit is decreased when said electron beam irradiation unit irradiates the outer periphery side of said disc-shaped object with the electron beams and is increased when irradiating the inner periphery side with the electron beams. 8. An apparatus for manufacturing a disc-shaped object according to claim 1, wherein said electron beam irradiation unit comprises an irradiation window of a single electron beam irradiation tube. |
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description | The present patent document claims the benefit of the filing date under 35 U.S.C. § 119(e) of Provisional U.S. Patent Application Ser. No. 62/382,098, filed Aug. 31, 2016, which is hereby incorporated by reference. The present embodiments relate to single photon emission computed tomography (SPECT). In particular, the effects of scatter are reduced in SPECT. In SPECT, a radioactive substance is administered to a patient. An imaging detector detects, through a collimator, some of the γ-radiation emitted from the patient. The detected emissions are tomographically reconstructed to generate an image object of locations of the emissions in a patient. Due to the gamma radiation scattering, some of the detections are of scatter rather than direct or primary photon emissions. This scattering degrades the SPECT image, so is to be reduced. The collimator typically has parallel holes to filter out photons that are not from particular directions. The scattering process is typically Compton scattering, which relates a change in direction with a reduction of energy as described by the Compton scattering formula. Scattered photons can fulfill the directionality condition as imposed the specific collimator, and it may even fulfill the energy acceptance conditions. Typically the energy window must be set large enough to allow for all primary photons to enter, however a large fraction of all detected counts are scattered photons. From simulations, the scatter fraction in the photo peak may be as high as 50% or more. Scatter may be a problem for collimators with non-parallel holes (non-parallel-hole collimators). Non-parallel-hole collimators are usually used to achieve better tradeoff between sensitivity and resolution for imaging certain organs (e.g., heart, thyroid and brain). The non-parallel-hole collimators greatly increase the difficulty in accurate modeling of the image formation process, including scatter. Thus, the quantitative accuracy of SPECT imaging using non-parallel-hole collimators is usually poor. To improve the image formation process, energy-window based scatter correction has been used for non-parallel-hole collimators. The scatter at energy windows adjacent to but not at an energy window for the primary photons is measured. The scatter at the energy window for the primary photons is estimated by linear interpolation from the scatter in the adjacent windows. This estimate is used in image formation to reduce the contribution of scatter. This-energy window type of scatter correction may be inaccurate or rely on assumption in interpolation that are not true. By way of introduction, the preferred embodiments described below include methods, systems, and non-transitory computer readable media for model-based scatter correction in SPECT with a non-parallel-hole collimator. Model-based scatter correction uses scatter kernels based on simulation to model the scatter for a given system and patient. For non-parallel-hole collimators, the measured sensitivity and measured vector maps are used in the modeling of scatter. The measured sensitivity is used to normalize the scatter kernels simulated for a parallel-hole collimator rather than attempting to simulate scatter with the complicated arrangement of holes. The measured vector maps are used to accurately project the model-based scatter sources into a data or emissions space. In a first aspect, a method is provided for model-based scatter correction in a SPECT system. A SPECT detector having a non-parallel-hole collimator detects emissions from a patient. An image object is generated from the emissions. The image object is forward projected to a first data model in data space. A model-based scatter source is formed from convolution of the image object with scatter kernels. The scatter kernels are kernels from a simulation of scatter with a parallel hole collimator normalized by sensitivity as a function of location measured for the non-parallel-hole collimator. A model of scatter is determined from a vector map measured from the non-parallel-hole collimator and the model-based scatter source. The first data model is combined with the model of scatter into a projection data model. A SPECT image of the patient is displayed and is a function of the projection data model. In a second aspect, a method is provided for model-based scatter correction in a SPECT system. Scatter kernels simulated with a parallel-hole collimator are normalized by a measured sensitivity for a non-parallel-hole collimator. Scatter is corrected in SPECT reconstruction with the normalized scatter kernels. A SPECT image is generated for a patient from the SPECT reconstruction as corrected for scatter. In a third aspect, a system includes a non-parallel-hole collimator and a detector for detecting emissions from a patient. The detector is adjacent to the non-parallel-hole collimator. A reconstruction processor is configured to reconstruct a three-dimensional activity distribution of the emissions from the patient. The reconstruction includes model-based scatter correction accounting for the non-parallel-hole collimator. A display is configured to display an image of the three-dimensional activity distribution. The present invention is defined by the following claims, and nothing in this section should be taken as a limitation on those claims. Further aspects and advantages of the invention are discussed below in conjunction with the preferred embodiments and may be later claimed independently or in combination. Iterative SPECT reconstruction uses model-based scatter correction for non-parallel-hole collimators instead of attempting to simulate scatter for each patient-SPECT system combination and instead of using energy-window-based scatter correction. The model-based scatter correction is altered to account for the non-parallel-hole collimator. The alteration accurately and practically account for the effects of non-parallel-hole collimators on scatter components by separating the modeling into two steps. In a first step, the effects of non-parallel-hole collimators on scatter kernels are eliminated by simulating an ideal parallel-hole collimator and normalizing the kernels resulting from simulation by sensitivity of the non-parallel-hole collimator. In a second step, the effects of non-parallel-hole collimators on scatter components during the forward radiation transfer process are accounted for using a combination of a model-based scatter source generated from the normalized scatter kernels and vector maps of the non-parallel-hole collimator. The model of scatter separates the image formation model for scattered photons from unscattered photons. By improving the accuracy in the image formation model using model-based scatter correction, accurate quantification for SPECT imaging may be provided using non-parallel-hole collimators. For example, coronary flow reserve (CFR) may provide additional information for better diagnosis of heart disease. Accurate quantitative imaging is a prerequisite for measuring CFR. SPECT systems using non-parallel-hole collimators (e.g., IQSPECT) may achieve better trade-off between sensitivity and resolution compared to parallel-hole collimators. The quantitative accuracy provided by model-based scatter correction with the non-parallel-hole collimator may allow for accurate CFR. Another important application may be I-123 DatScan for diagnosing Parkinson's disease. Using non-parallel-hole collimators with model-based scatter correction may provide higher resolution and better quantitative accuracy, which is beneficial to differentiating different types of Parkinson's disease. Other applications may benefit from the improved accuracy due to model-based scatter correction with non-parallel-hole collimators. FIG. 1 shows a SPECT system 10. Any now known or later developed SPECT system 10 may be used. The SPECT system 10 detects emissions due to radioactive decay in a patient. The SPECT system 10 may provide qualitative or quantitative imaging. The SPECT system 10 implements the method of FIG. 3, the method of FIG. 4, or another method. Model-based scatter correction is used with the non-parallel-hole collimator 20 for imaging a patient. The SPECT system 10 includes a reconstruction processor 12, a memory 14, a display 16, a detector 18, and the non-parallel-hole collimator 20. The processor 12, memory 14, and/or display 16 are part of the imaging system with the detector 18 or are separate (e.g., a computer or workstation). Additional, different, or fewer components may be provided. For example, the system 10 is a computer without the detector 18 and the non-parallel-hole collimator 20. As another example, user input, patient bed, x-ray scanner, computed tomography (CT) scanner, or other SPECT imaging-related devices are provided. Other parts of the system 10 may include power supplies, communications systems, and user interface systems. The non-parallel-hole collimator 20 is lead or other material for blocking gamma radiation resulting from emissions in the patient. A slab of material extending over the surface of the detector 18 is provided. The surface facing the patient of the non-parallel-hole collimator 20 is flat over a two-dimensional area, but may have surface curvature in other embodiments. Square, rectangular, hexagonal, circular, or other surface area shapes may be provided. The non-parallel-hole collimator 20 is of any uniform or varying thickness, may contact or be spaced from the detector 18, and is positioned so that photons arriving at the detector 18 pass through the non-parallel-hole collimator 20. A plurality of holes is in the non-parallel-hole collimator 20. The holes are circular, but may be hexagonal or other shapes. Fan-beam shapes, elongated shapes, or other hole shapes may be used. The holes are cylindrical channels or other shapes with parallel walls, but conical or other non-parallel wall shapes for the channels may be used. Uniform or non-uniform spacing of holes is provided across the non-parallel-hole collimator. The holes extend through the non-parallel-hole collimator 20, providing channels through which photon from the emissions may pass to reach the detector 18. Any photons arriving at a hole at a generally non-parallel angle to the hole are blocked from reaching the detector. “Generally” is used to account for a range of acceptance angles (e.g., +/−2 degrees) due to the hole size. Any photons arriving at a hole at a generally parallel angle to the hole pass through and arrive at the detector 18. The holes are non-parallel. Any range of deviation from parallel may be used, such as +/−5, 10, 15, or 25 degrees. Any variation as a function of locations across the non-parallel-hole collimator 20 may be used. For example, the non-parallel-hole collimator 20 is a multi-focal collimator where two or more focal regions or one focal region and a generally non-focused (e.g., infinite focus) is also provided. Some of the holes may be parallel or closer to parallel than other holes and corresponding channels. For example, the holes around the edges of the collimator are parallel and at least some of the holes inward from the edges are non-parallel with the holes and channels on the edges. This may allow a greater concentration of emissions from one or more regions to be counted. FIG. 2 shows an example non-parallel-hole collimator 20 in cross-section. Sample lines of response are shown extending along the channels through the holes and into a patient space. The lines of response are non-parallel to each other due to the non-parallel-holes of the non-parallel-hole collimator 20. The lines of response inward from the two edges of the cross-section angle towards a region spaced from the non-parallel-hole collimator 20. The lines of response on the edges and at the center are parallel to each other. Any variation in angles across one or more dimensions of the surface of the non-parallel-hole collimator 20 may be used. Other arrangements of lines of response may be provided. The detector 18 is a gamma camera connected with a gantry. The gamma camera is a planar photon detector, such as having crystals or scintillators with photomultiplier tubes, SiPM, or another optical detector. Direct (e.g., CZT) or indirect (e.g., with scintillation crystals) detectors may be used. The gantry rotates the gamma camera about the patient. Other structures of detectors may be used. The detector 18 is adjacent to the non-parallel-hole collimator 20. The non-parallel-hole collimator 20 rests against or contacts the detector 18. Spacers or standoffs may connect the non-parallel-hole collimator 20 to the detector 18. Alternatively, the non-parallel-hole collimator 20 is positioned against or away from the detector 18 by any connection (e.g., direct or indirect) but sufficiently adjacent that photons that arrive at the detection surface of the detector 18 (e.g., arrive at the scintillation crystals) from the patient pass through the non-parallel-hole collimator 20. The SPECT system 10, using the detector 18, detects emissions from the patient 22 for measuring uptake or physiological function. During scanning of a patient, the detector 18 detects emission events. The emissions occur from any location in a finite source (i.e., the patient). The radiotracer in the patient migrates to, connects with, or otherwise concentrates at specific types of tissue or locations associated with specific biochemical reactions. Thus, a greater number of emissions occur from locations of that type of tissue or reaction. As a gamma camera moved to different locations relative to the patient, the detector 18 detects emission events at different positions or angles relative to the patient, forming lines of response for the events. The patient bed may move to define a field of view relative to the patient. The reconstruction processor 12 is a general processor, digital signal processor, graphics processing unit, application specific integrated circuit, field programmable gate array, digital circuit, analog circuit, combinations thereof, or other now known or later developed device for reconstructing an image object from detected emissions. The reconstruction processor 12 is a single device, a plurality of devices, or a network. For more than one device, parallel or sequential division of processing may be used. Different devices making up the reconstruction processor 12 may perform different functions, such as one processor (e.g., application specific integrated circuit or field programmable gate array) for reconstructing the object and another (e.g., graphics processing unit) for rendering an image from the reconstructed image object. In one embodiment, the reconstruction processor 12 is a control processor or other processor of SPECT system 10. In other embodiments, the reconstruction processor 12 is part of a separate workstation or computer. The reconstruction processor 12 operates pursuant to stored instructions to perform various acts described herein, such as reconstructing of act 32, forming the model-based scatter source of act 36, correcting for scatter of act 40, and generating an image of act 46. The reconstruction processor 12 is configured by software, firmware, and/or hardware to reconstruct a volume or object (e.g., three-dimensional activity distribution) from emissions. Any reconstruction may be used to estimate the activity concentration or distribution of the tracer in the patient. The reconstruction processor 12 accesses the detected emission events from the memory 14, from the detector 18, or buffers to reconstruct. The detected emissions are used to reconstruct the distribution of the radioisotope in three dimensions. For reconstruction, forward and backward projection are used iteratively until a merit function indicates completion of the reconstruction (i.e., a best or sufficient match of the image object to the detected emissions). The forward projection uses system information to project a currently estimated activity distribution in object or image space to a data space corresponding to detected emissions along lines of response. For a first instance, an initial distribution of activity in the image or object space is used or estimated from the detected emissions. Residuals between the forward projection and the detected emissions are determined. The backward projection transforms the residuals from the data space to the image or object space. By repeating the forward and backward projections, a series of distributions of activities result. Once a stop criterion or criteria are met in the reconstruction, then the resulting activity distribution is used for SPECT imaging. As part of reconstruction, the reconstruction processor 12 accounts for the non-parallel-hole collimator 20 with model-based scatter correction. The scatter correction is performed as part of the forward projection from the image or object space to the data space. The scatter correction model is used in applying the image formation process to the activity distribution. The resulting projection data model has reduced scatter. For scatter correction, a scatter response function (SRF) is combined with the activity distribution of the patient to form a model-based scatter source. The SRF is represented by scatter kernels. The reconstruction processor 12 convolves the scatter kernels with the activity distribution to create the model-based scatter source. The scatter kernels for the given SPECT system 10 are used. These system-specific scatter kernels are stored in the memory 14 and were created based on simulation of scatter with a parallel-hole collimator. The kernels from the simulation do not account for the non-parallel-hole collimator 20, but the simulation is more efficiently performed with an ideal parallel-hole collimator. To make the kernels specific to a given SPECT system 10, the simulated kernels are normalized by sensitivity of the non-parallel-hole collimator 20. The sensitivity as a function of location on the surface of the non-parallel-hole collimator 20 is measured. This measured sensitivity is used to normalize across the kernels, adapting the scatter kernels and resulting SRF to the specific non-parallel-hole collimator 20. The reconstruction processor 12 is further configured to combine the model-based scatter source generated from the scatter kernels and vector maps measured for the non-parallel-hole collimator 20. For forward radiation transfer of the scatter model to the data space, the vector maps for the non-parallel-hole collimator 20 are used. These vector maps define the acceptance angles and/or lines of response along which the forward projections occur. The model-based scatter source is used to model detection of scatter by the SPECT system 10. This detection is modeled as the forward radiation transfer to create the model of scatter in the data space. The reconstruction processor 12 then combines the forward projection of the activity distribution with the model of scatter to reduce the scatter in the resulting projection data model. The reconstruction processor 12 generates one or more images based on the reconstruction. Any given image represents the emissions from the patient. The image shows the spatial distribution, such as with a multi-planar reconstruction or a volume rendering. For quantitative imaging, the image represents accurate measures (e.g., in Bq/ml) of the activity concentration. Alternatively or additionally, the image shows a quantity or quantities (e.g., alphanumeric) representing the activity concentration or specific uptake values for one or more locations or regions. The display 16 is a CRT, LCD, plasma screen, projector, printer, or other output device for showing an image. The display 16 displays an image of the reconstructed functional volume. The image is of or part of the three-dimensional activity distribution. A three-dimensional rendering, multi-planar reconstruction, cross-section or other type of image of the activity distribution in the patient is displayed. The memory 14 is a buffer, cache, RAM, removable media, hard drive, magnetic, optical, database, or other now known or later developed memory. The memory 14 is a single device or group of two or more devices. The memory 14 is part of the SPECT system 10 or a remote workstation or database, such as a PACS memory. The detected emission events, counts, location, or other detection information are stored in the memory 14. The memory 14 may store data at different stages of processing, such as a data model, residuals, image or object representation (i.e., activity distribution), filtered data, thresholded data, results from morphological processing, masks, scatter kernels, vector maps, model-based scatter source, model of scatter, or other data. Projection operators, transposed operators, a measure of completeness of reconstruction, merit function data, the reconstructed object, system matrix, thresholds, results of calculations, an image to be displayed, an already displayed image, or other data may be stored. The data is stored in any format. The memory 14 is additionally or alternatively a non-transitory computer readable storage medium with processing instructions. The memory 14 stores data representing instructions executable by the programmed processor 12. The instructions for implementing the processes, methods, and/or techniques discussed herein are provided on non-transitory computer-readable storage media or memories, such as a cache, buffer, RAM, removable media, hard drive, or other computer readable storage media. Computer readable storage media include various types of volatile and nonvolatile storage media. The functions, acts or tasks illustrated in the figures or described herein are executed in response to one or more sets of instructions stored in or on computer readable storage media. The functions, acts or tasks are independent of the particular type of instructions set, storage media, processor or processing strategy and may be performed by software, hardware, integrated circuits, firmware, micro code and the like, operating alone or in combination. Likewise, processing strategies may include multiprocessing, multitasking, parallel processing and the like. In one embodiment, the instructions are stored on a removable media device for reading by local or remote systems. In other embodiments, the instructions are stored in a remote location for transfer through a computer network or over telephone lines. In yet other embodiments, the instructions are stored within a given computer, CPU, GPU, or system. FIG. 3 show one embodiment of a method for model-based scatter correction in a SPECT system. The model-based scatter correction is performed for a non-parallel-hole collimator. The scatter correction is specific to the non-parallel-hole collimator. Scatter kernels are adapted to account for the non-parallel-hole collimator, allowing use of simulation of scatter with a simulated parallel-hole collimator. The measured vector maps for the non-parallel-hole collimator are further used to adapt the model-based scatter correction to the non-parallel-hole collimator. FIG. 4 shows another representation of the method. The methods of FIGS. 3 and 4 implement model-based scatter correction with the image forming process accounting for the non-parallel-hole collimator. As part of iterative reconstruction, the methods generate the image object from the emissions, forward project the image object, and combine the model of scatter with the forward projection of the image object. Once reconstructed with scatter reduction based on the model of scatter, a SPECT image resulting from image object from the iterative reconstruction is displayed. In this reconstruction process, the model of scatter is applied as part of transforming in one or multiple (e.g., all) instances (e.g., iterations) from the image object to the projection data model. FIGS. 3 and 4 are directed to acts in this transformation. The method is implemented by a SPECT system, such as the SPECT system of FIG. 1. The SPECT system includes a non-parallel-hole collimator, so the modeling of scatter accounts for this non-parallel-hole collimator in the image formation process. A detector, such as a gamma camera with the non-parallel-hole collimator, detects emissions in act 30, a reconstruction processor performs acts 32-44, and a display, memory interface, or network interface is used for act 46. Other devices may be used to perform any of the acts, such as act 38 being performed by a separate processor and the resulting scatter kernels stored in memory are used by the reconstruction processor in act 36. The acts are performed in the order shown (e.g., top to bottom or numerical for FIG. 3 or along the arrows for FIG. 4) or other orders. Additional, different, or fewer acts may be performed. For example, act 30 is not provided where the detected emissions are stored or transferred from memory. As another example, act 38 is not performed where the stored scatter kernels of the model of scatter are previously normalized. In other examples, an act for backward projection is provided as part of the reconstruction of act 32. In act 30, emissions from a patient are detected. The activity concentration in a patient having received a radiotracer is determined as part of reconstruction. After ingesting or injecting the radiotracer into the patient, the patient is positioned relative to a detector, and/or the detector is positioned relative to the patient. For SPECT, the detector may be rotated or moved relative to the patient, allowing detection of emissions from different angles and/or locations in the patient. Emissions from the radiotracer within the patient are detected over time. A non-parallel-hole collimator in front of the detector limits the direction of photons detected by the detector, so each detected emission is associated with an energy and line of response (e.g., a cone of possible locations from which the emission occurred). Any multi-focal or other arrangement of non-parallel holes may be used to limit the directions from which emissions are detected. For example, the emissions are detected along lines of response that are more parallel with each other at the edges of the planar array of the gamma camera than other lines of response more inward. The angles of the holes of the collimator determine the lines of response. Plane, fan beams, or cones of response with spatial variance in orientation may be used. In act 32, a reconstruction processor reconstructs an image object from acquired projection data. Computed tomography implements reconstruction to determine a spatial distribution (i.e., activity distribution) of emissions from the detected lines of response. The projection data represents the detected emissions. The quantity or amount of uptake for each location (e.g., voxel) may be estimated as part of the reconstruction. The SPECT system may estimate the activity concentration of an injected radiopharmaceutical or tracer for different locations in a patient volume. For an initial iteration in the reconstruction, a default or other image object may be used. An “object” or “image object” is an activity distribution in an object space (also referred to as image space) and is a reconstruction of the data set D measured in a data space. The object space is the space in which the result of the image reconstruction is defined and which corresponds to the volume that was imaged using the SPECT imaging system (e.g., the input object, such as a patient, provided to the nuclear imaging system). The image object may be a three-dimensional (3D) image object or may have any other dimensionality, e.g., for N-dimensional imaging. Any now known or later developed reconstruction methods may be used, such as based on Maximum Likelihood Expectation Maximization (ML-EM), Ordered Subset Expectation Maximization (OSEM), penalized weighted least squares (PWLS), Maximum A Posteriori (MAP), multi-modal reconstruction, non-negative least squares (NNLS), or another approach. Different processes for dealing with motion or other sources of distortion may be used for a same method. The reconstruction is iterative. The image reconstruction processor uses a system matrix H (or projection operators) to describe the properties of the SPECT system and uses an iteratively improved data model to calculate the image object based on the data set D of detected emissions. The iterative reconstruction forward projects a current estimate of the object or image (e.g., object or image space) to projection or data space using the system matrix or forward projectors representing the detection (e.g., image process) in act 34. The forward projection multiplies the system matrix or projection operators with the current volume to emulate the detection by the nuclear imaging system. The reconstruction includes projection operators (i.e., forward projector) that incorporate the effects of the detector on the photons (i.e., collimation and detection process) for a patient and isotope. The forward projector contains a model of the imaging formation process specific to the detector and/or imaging system. The image formation model includes the interaction of photons with patients (e.g., attenuation and scatter), the collimation-detection process (e.g., collimator detector response including collimator geometric response, septal penetration and scatter, partial deposition in crystal and detector intrinsic resolution), and related radionuclide properties (e.g., emission abundances). The system matrix or projection operators are the mathematical representation of the projection from the object space to the projection space (e.g., forward projector). In some nuclear imaging systems, such as SPECT for small animal imaging, the system matrix is stored and used directly in each iteration to calculate the projection data model from a current estimate of the activity distribution. In most clinical nuclear imaging systems, due to the large dimension of the system matrix, the system matrix is not stored. Instead, a series of mathematical projection operators, collectively called the forward projector, are performed in each iteration. The projection operators mathematically provide multiplication by the system matrix. In one representation for SPECT, the forward projection is an application of the system matrix H to an object in object space. Projecting an estimated image Iα (where a represents a measurement angle) into data space results in a data model Mi of that estimated image: M i = ∑ α H i α I α Representing the system matrix as a product of operators yieldsH=HxΘ□−H2ΘH1 Other representations may be used. The forward projection of act 34 may be limited to emissions at particular energies. An energy window associated with emissions of the radionucleotide used for the patient is used. The energy window is set around the photon energy of the non-scatter emissions from the radionucleotide. Emissions at energies outside this primary photon window are likely scatter or from other sources, so are not to be used. Emissions within the energy window are more likely from the radionucleotide, but may include first, second, third or other orders of scatter from any source. Since the detected emissions are in a projection space (e.g., generally known location in two-dimensions but not three), the forward projection of the current volume is compared to the detected or measured emissions. This comparison is tested for accuracy with a merit function (e.g., ML-EM, NNLS, or Mighell chi square). If sufficiently accurate and/or having no further increase in accuracy, the iteration ceases and the current image object is output as the reconstructed image object. If the merit function indicates insufficient or improving accuracy, a difference or residual between the forward projection and the detected emissions is backward projected. This backward projection provides a gradient or change for the image object. The direction and step size is determined for the change and is applied to update the image object. The process is then repeated for another iteration of the reconstruction. Once complete, an image object I, which may be an N-dimensional image object (typically N=3 in medical imaging applications), may then be displayed on a display using a volume rendering or other imaging technique. The scatter model is applied in parallel with the forward projection operation of act 34, so the scatter in the object space is transformed into the data space for combination with the forward projection resulting from act 34. The data model is adjusted or includes correction for scatter prior to testing with the merit function. For use on a SPECT system, the scatter correction is desired to be an accurate and practical method to account for the effects of non-parallel-hole collimators on scatter components. Energy window-based scatter correction may be inaccurate. Simulation of scatter with the non-parallel-hole collimator for each system may be impractical. Instead, model-based scatter correction is used. To account for the non-parallel-hole collimator, the model-based scatter correction is separated in two ways. Act 38 represent one way, and act 42 represents another way. In act 36, the reconstruction processor forms a model-based scatter source. This model-based scatter source is specific to the non-parallel-hole collimator and the patient, so uses the non-parallel-hole collimator specific scatter kernels 39 and the patient-specific activity distribution 32 as inputs. To form the model-based scatter source, the scatter kernels are convolved with the activity distribution 32 (i.e., image object). A library of points in material are simulated, such as with Monte Carlo simulation. If the material is known or assumed, the appropriate kernels are convolved given the material present. Another method is brute force computation of Monte Carlo, which is very compute intensive. The scatter kernels 39 represent the scatter response function of the detection with the non-parallel-hole collimator. In act 37, the interaction of scatter resulting from different sources with a detector and collimator are simulated. Any type of simulation may be used. Monte Carlo or other stochastic simulation may be used. To keep the processing burden reasonable given stochastic simulation, the collimator in the simulation is treated as an ideal parallel-hole collimator. Non-ideal may be used for the parallel-hole collimator, such as to account for conical lines of response due to the finite size of the holes. The simulation is performed for all systems of a given type, such as all SPECT systems using a same combination of collimator and detector. The simulation is for that combination, such as based at least in part of the size, shape, and/or material characteristics of the collimator and detector. The simulation is not performed by the SPECT system 10, but by a computer, workstation, or server. Alternatively, the SPECT system 10 performs the simulation. The results of the simulation are scatter kernels for the parallel-hole collimator and detector combination. The scatter kernels model the common physics in the image formation process for scatter. The simulation is for a given radiotracer. The simulation provides for the source or sources to emit at the energy level for the primary photons. An energy window may or may not be used in the simulation. In alternative embodiments, the simulation provides scatter kernels for different energy levels. To adapt the scatter kernels to the non-parallel-hole collimator, the scatter kernels from the simulation are normalized by sensitivity of the non-parallel-hole collimator. This normalization adjusts the scatter kernels based on the sensitivity. The sensitivity of the non-parallel-hole collimator varies across the non-parallel-hole collimator at act 38. Due to the variation in the angles, distribution (e.g., non-uniform), size, and/or shape of the holes, the sensitivity is different at different locations. This spatial variation in sensitivity is applied to the scatter kernels. The scatter kernels represent the image formation process, so respond to the spatial variation of sensitivity. The sensitivity is measured for the non-parallel-hole collimator. Each non-parallel-hole collimator may have a different spatial distribution of sensitivity, even if of a same design. Manufacturing tolerances, damage, variation in installation, and/or another aspect may result in the sensitivity for a given non-parallel-hole collimator being different than the same type or part non-parallel-hole collimator in another SPECT system. The normalization by the measured sensitivity provides non-parallel-hole collimator specific scatter kernels for the one non-parallel-hole collimator. In alternative embodiments, the sensitivity is by class or type, such as providing an average sensitivity profile or by location for the part or type of non-parallel-hole collimator. This normalization by the measured sensitivity for this type provides non-parallel-hole collimator specific scatter kernels for the type. In this way, the effects of non-parallel-hole collimators are eliminated from the scatter kernels for modeling the source of scatter. The sensitivity is measured at the energy level for the primary photons. Different sensitivities are provided for different energy levels or energy windows. The normalization is by the sensitivities for the primary photons or non-scattered energies from the radiotracer emissions. The normalization is performed with the simulation 37 or at another time. The SPECT system may receive the normalized scatter kernels 39 for use with any number of patients. The scatter kernels 39 may be updated at a calibration of the SPECT system or are maintained. The SPECT system may not create the non-parallel-hole collimator specific scatter kernels 39. In other embodiments, the SPECT system generates the non-parallel-hole collimator specific scatter kernels 39 as needed or as part of generating the model-based scatter source in act 36. In act 40, the model-based scatter source is used to correct for scatter in SPECT reconstruction. The model-based scatter source generated by the normalized scatter kernels is combined with vector maps 41 measured for the non-parallel-hole collimator. The combination occurs as a part of the forward radiation transfer. Scatter from the model-based scatter source is forward projected. The combination creates a model of scatter in act 42. The combination of the model-based scatter source generated from scatter kernels and vector maps 41 of the non-parallel-hole collimator is performed by ray tracing in image formation. The vector maps 41 and the model-based scatter source are convolved with a point spread function as part of the forward radiation transfer. The emissions due to any degree of scatter along the lines of response defined by the vector maps 41 are projected from the scatter source modeled as the model-based scatter source for the non-parallel-hole collimator. The forward radiation transfer for the scatter sources along lines of response defined by the vector maps 41 accounts for the effects of non-parallel-hole collimators on scatter components. The vector maps 41 are measured for the non-parallel-hole collimator. The measurements are for the type or part number, such as being an average. Alternatively, the measurements are for a specific non-parallel-hole collimator of the SPECT system being used to scan the patient. The model of scatter may account for patient-specific characteristics. The tissues or structures of the patient may affect the scatter. The density, attenuation, and/or absorption may be determining characteristics. A computed tomography (CT) scan of the patient may indicate density, attenuation, and/or absorption. The CT values for a volume of the patient provide patient specific characteristics for modeling scatter. These CT values are used in the model of scatter. As part of the forward radiation transfer, the CT values are used to account for the effects of the patient-specific characteristics on the scatter components. The CT values are used for attenuation correction as a depiction of deviations from water. The density or material properties may be extracted from the CT values. The scatter model may account for the density or other material properties. In act 44, the model of scatter output in act 42 is combined with the data model output in act 34. The combination is an optimization. The reconstruction processor estimates which of the emissions of the data model from act 34 are primary photons (i.e., without scatter) and which of the emissions of the data model are from scatter. The optimization varies voxel values in the image space so that the different between the data and the data model in data space is minimized given an objective function. The combination of results of the forward radiation transfer with a forward projection from a three-dimensional activity distribution for the patient creates the projection data model 45 from the emissions without or with less scatter. The data model from the forward projection is corrected for scatter. The data model includes scatter, as does the measured data. The scatter source is treated as if there is a scatter source of which to track. This corrected data model 45 is used to calculate residuals for back projection. The residuals are then used in the next iteration of altering the image object and forward projecting. Alternatively, the merit criterion or criteria indicate a sufficient match of the projection data model 45 with the detected emissions after correction for scatter. The image object from the sufficient match is used as the reconstruction of the activity distribution for the patient. The counts at each location in the activity distribution may be used or converted to quantification of the activity. In act 46, a SPECT image of the patient is displayed. The SPECT system, such as the reconstruction or other processor, generates an image from the reconstruction of the activity distribution. The image is a function of the projection data model with correction for scatter from the model of scatter, accounting for the non-parallel-hole collimator. The image is displayed on the display of the SPECT system or another display. The image may be of a quantity of activity, such as an image of an alphanumeric value (e.g., level of uptake). The image may be any SPECT image, such as a three-dimensional rendering from the reconstructed object image. An image of the patient or part of the patient is generated from the reconstruction. The results of the reconstruction represent a distribution of emissions or counts of emissions in N-dimensions. For qualitative SPECT, this distribution is used to generate an image. For quantitative SPECT, the activity concentration for each location (e.g., voxel) is determined. The reconstruction provides voxel values representing activity concentration. In one embodiment, data for one or more (e.g., multi-planar reconstruction) planes is extracted (e.g., selected and/or interpolated) from the volume or voxels and used to generate a two-dimensional image or images. The output may be a transmission. The image transmission may be to a memory through a memory interface and/or to a patient medical record, server, or other computer connected through a network interface. While the invention has been described above by reference to various embodiments, it should be understood that many changes and modifications can be made without departing from the scope of the invention. It is therefore intended that the foregoing detailed description be regarded as illustrative rather than limiting, and that it be understood that it is the following claims, including all equivalents, that are intended to define the spirit and scope of this invention. |
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claims | 1. An assembly comprising:a body,one or more collimators supported by the body, each of the one or more collimators having a through hole configured to define a shape or size of a treatment beam having a beam centerline, one or more photon flattening filters supported by the body, and one or more electron scattering foils supported by the body, wherein said one or more collimators and said one or more photon flattening filters are positioned in a circular or arc configuration having a first radius, and said one or more electron scattering foils are positioned in a circular or arc configuration having a second radius different from the first radius; anda first axis operable to move the body, and a second axis operable to move the body, wherein said first axis is a linear axis operable to translate the body and said second axis is a rotational axis operable to rotate the body about a line parallel to the beam centerline, wherein the first and second axes are operable to align one of the one or more collimators, photon flattening filters, or electron scattering foils with the treatment beam. 2. The assembly of claim 1, wherein said first axis is operable to translate the second axis. 3. The assembly of claim 1 wherein said second axis is operable to rotate the first axis. 4. The assembly of claim 1 further comprising a second axis operable to move the body, wherein said first and second axes are linear axes. 5. The assembly of claim 1 wherein said first axis is a linear axis operable to translate the body and said second axis is a rotational axis operable to rotate the body, and said first axis is further operable to translate the second axis. 6. The assembly of claim 1 further comprising a field light assembly comprising a mirror member and one or more light sources, said mirror member being supported by the body. 7. The assembly of claim 1 further comprising a target assembly comprising one or more targets configured to produce radiation upon impingement by electrons and a third axis operable to move the target assembly. 8. The assembly of claim 7 further comprising an ion chamber assembly and a fourth axis operable to move the ion chamber assembly, wherein each of the one or more collimators is configured to allow a portion of the treatment beam to pass through an outside area of the collimator to be projected onto and detected by the ion chamber. 9. The assembly of claim 1 wherein the first axis comprises a servo motor and one or more feedback devices. 10. The assembly of claim 1 wherein each of the one or more collimators has a conically-, cylindrically-, or trapezoidally-shaped hole configured to define a treatment beam for radiosurgery. 11. A radiation apparatus comprising a linear accelerator, a treatment head, and a gantry enclosing the linear accelerator and the treatment head, the radiation apparatus comprising:a radiation source configured to produce a treatment beam;a primary collimator configured to generally define a field of the treatment beam;one or more stereotactic radiosurgery (SRS) collimators supported by a body movable relative to the radiation source to align one of the one or more SRS Collimators with the treatment beam, each of the one or more SRS collimators having a through hole configured to further collimate the treatment beam to provide a beam suitable for radiosurgery, and each of the one or more SRS collimators having a size to allow a portion of the treatment beam to pass around an outside area of the each of the one or more SRS collimators to be projected onto an ion chamber;the ion chamber configured to detect the portion of the treatment beam passing around the outside area of the each of the one or more SRS collimators; andone or more collimation jaws configured to block at least a portion of the treatment beam passing around the outside area of the each of the one or more SRS collimators and through the ion chambers. 12. The radiation apparatus of claim 11 wherein each of the one or more SRS collimators has a conically-, cylindrically-, or trapezoidally-shaped through hole to define the treatment beam. 13. The radiation apparatus of claim 11 wherein the radiation source comprises a target configured to produce radiation upon impingement by electrons, and said target resides in the treatment head. 14. The radiation apparatus of claim 11 wherein said one or more SRS collimators are movable relative to the radiation source. 15. The radiation apparatus of claim 11 further comprisingone or more photon flattening filters and/or one or more electron scattering foils supported by the body; anda first and a second axes operable to move the body. 16. The radiation apparatus of claim 15 wherein said one or more collimators and said one or more photon flattening filters are arranged in a circular or arc configuration having a first radius, said one or more electron scattering foils are arranged in a circular or arc configuration having a second radius different from the first radius. 17. The radiation apparatus of claim 16 wherein said first axis is a linear axis operable to translate the body and said second axis is a rotational axis operable to rotate the body. 18. The radiation apparatus of claim 17, wherein said first axis is operable to translate the second axis. 19. The radiation apparatus of claim 17 wherein said second axis is operable to rotate the first axis. 20. The radiation apparatus of claim 15 wherein the first and second axes are linear axes. 21. A radiation system comprising:a radiation source operable to generate a radiation beam having a beam centerline;an assembly comprising a body, one or more collimators supported by the body each having a through hole configured to define a shape or size of a treatment beam, one or more photon flattening filters supported by the body, and one or more electron scattering foils supported by the body, wherein said one or more collimators and said one or more photon flattening filters are positioned in a circular or arc configuration having a first radius, and said one or more electron scattering foils are positioned in a circular or arc configuration having a second radius different from the first radius, and a first and a second axes operable to move the body relative to the radiation source, each of the first and second axes comprising a servo motor, wherein said first axis is a linear axis operable to translate the body and said second axis is a rotational axis operable to rotate the body about a line parallel to the beam centerline, and the first and second axes are operable to align one of the one or more collimators, photon flattening filters, or electron scattering foils with the treatment beam; anda controller programmed to control the servo motor of the first and second axes in moving the body to a desired position. 22. The radiation system of claim 21 wherein each of the first and a second axes further comprises one or more feedback devices, and the controller is further programmed to control the servo motor of the first and second axes based on at least signals from the one or more feedback devices. 23. The radiation system of claim 21, wherein said first axis is operable to translate the second axis. 24. The radiation system of claim 21 wherein said second axis is operable to rotate the first axis. 25. The radiation system of claim 21 wherein said one or more collimators have a conically-, cylindrically-, or trapezoidally-shaped hole configured to define a treatment beam suitable for radiosurgery. 26. The radiation system of claim 21 further comprising an ion chamber and a third axis operable to move the ion chamber, wherein said third axis comprises a servo motor and said controller is further programmed to control the servo motor of the third axis in driving the ion chamber, wherein each of the one or more collimators is configured to allow a portion of the treatment beam to pass through an outside area of the collimator to be projected onto and detected by the ion chamber. 27. The radiation system of claim 26 further comprising a field light simulation system comprising a light source and a mirror member, said mirror member being supported by the body and movable by the first and second axes, and said light source being movable by the third axis. 28. The radiation system of claim 27 wherein the radiation source comprises a target assembly comprising one or more targets each being configured to produce radiation upon impingement by electrons, and a fourth axis operable to move the target assembly for positioning the one or more targets, wherein said fourth axis comprises a servo motor and the controller is further programmed to control the servo motor of the fourth axis in driving the target assembly to a desired position. |
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052727327 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT Reference will now be made in detail to the preferred embodiment of the invention, an example of which is illustrated in the accompanying drawings. FIG. 1 illustrates the environment in which the preferred embodiment of the present invention is used. As shown in FIG. 1, a reactor vessel 10 having a vessel head 12 is located in a reactor vessel cavity 14 and is suspended therein by the vessel nozzles 16 which are supported within the cavity walls 18. Above the reactor vessel 10 the cavity walls 18 define a refueling canal or pool 20 having a lower boundary or pool floor 23 defining a refueling pool ledge 22 generally opposite a vessel flange 24 of the vessel 10. Refueling of the reactor occurs periodically and consists of filling the refueling pool 20 with water, then removing the vessel head 12 from the vessel 10 so that access may be had to the nuclear fuel (not shown) inside the vessel 10. It is imperative, however, that none of the water in the refueling pool 20 leak into the vessel cavity 14 because refueling water would create contamination problems with the vessel 10 and other equipment in the vessel cavity 14. In addition, the seal must be able to accommodate the thermal expansion and contraction of the reactor vessel due to the cyclic heating and cooling periods to which the vessel is subject. In addition, however, access to the cavity 14 must be maintained during plant operation because of safety and licensing requirements relating to hypothetical accidents which must be accommodated without damage to any components. In particular, if the vessel nozzle 16 breaks, the flashing liquid must have an escape route from the vessel cavity 14 in order to prevent excessive uplift on the vessel 10 which could further complicate an already serious accident. Therefore, the space 28 must be maintained open for ventilation and cooling during normal operations and to permit a steam flow path out of the vessel cavity 14. The space 28 may be sealed only during the reactor refueling operation. In addition access for maintenance during an outage must be provided. FIGS. 2 and 3 illustrate a seal of the prior art. As shown in FIG. 2 a annular seal ring is designated at 30 and includes horizontal plate or deck portions 32 for use as a work platform. The deck sections 32 are joined by splice plates 34. Threaded studs 3 and nuts (not shown) fasten the deck section 32 and plates 34 together. The plate section 32 include openings 42 with removable plugs or covers 44 (not shown). The openings 42 provide for reactor cavity cooling air flow during operation. Radially disposed ribs or members 46 are provided, circumferentially spaced, to span the gap 28. The inner ends 47 of ribs 46 are supported by the vessel flange 24 and the outer ends 48 of the ribs 46 are supported by the ledge 22. As shown in FIG. 3, flexible members 50 and 52 are welded to the reactor vessel flange 2 and the ledge 22 of the pool and the deck member 32 to provide a water tight seal. The seal joint of the present invention is illustrated in FIGS. 4a, 4b, 5, 6 and 8. FIG. 4a shows the reactor vessel 10 with the reactor vessel refueling flange 24 in the reactor vessel cavity 14 and the refueling pool 20 having the pool floor 23 and the refueling pool ledge 22. The gap 28 between the reactor vessel 10 and the pool floor 23 and the cavity walls 18 is sealed by the annual ring seal and refueling deck assembly of the present invention illustrated generally at 60. The seal 60 includes a support 64, a platform 65 with hatches 68 and a seal joint 70. The support 64 and the seal joint 70 are welded to the platform 65 and the pool ledge 22 and the reactor vessel flange 24. The hatches 68 may be opened during operation to provide ventilation. A top view of a portion of the annular seal 60 is illustrated in FIG. 4b. As shown in FIGS. 5 and 6, the seal joint 70 includes and upper cylindrical section 79 and a lower flexible coil or arcuate section 71. The joint 70 is welded at a rim 78 of cylindrical section 79 to the platform 65 and is welded at its other end 72 to the reactor vessel flange 24. The upper section 79 forms a straight walled cylinder. The cylinder may be short or long depending on the design criteria of the plant. The lower section 71 is made in coil shapes resembling a structure composed of narrow rings and short cylinders. The coil or arcuate shaped lower section 71 provides greater flexibility and absorption of forces resulting from movement of the vessel 10. The seal joint 70 is able to withstand greater forces and provide a longer seal than the seals of the prior art because of the greater structural flexibility provided by the combined properties of the arcuate sections and the upper cylindrical section of the joint. As a result, the joint stress induced by the movements of the reactor vessel is kept to a minimum. Thus, the joint 70 is able to accommodate the radial, axial and rotational movements between the reactor vessel 10 and the pool floor 23. The present invention optimizes the fatigue usage life of the joint 70 to meet the usage life requirements of an individual plant, by proper flexibility arrangement between the coil section 71 and the upper cylindrical section 79. The number of coils, the orientation of the ends of the coil section 71, the width of the rim 78 for the cylindrical section 79, and the dimensions of the sections 71, 72 and 79 may vary to suit the design criteria of a specific plant. The seal joint 70, the platform 65 and the support 64 are typically metal, preferably steel, and more preferably Type 304 stainless steel. While the specific dimensions of the joint 70 may vary depending on the design criteria of the plant, the dimensions of the seal joint 70 shown in FIG. 6 may include having the upper cylindrical section 79 between 0.1 and 0.15 inches thick, preferably 0.125 inches thick. The lower coil section may consist of two coils 73 and 74. The coil section 73 may have a radius of 0.625 inches and may be made from Type 304 stainless steel, Sch. 40. While the coil section 74 may also have a radius of 0.625 inches, it may be made from Type 304 stainless steel, Sch. 160. The entire height of the seal joint 70 may typically be 7 to 8 inches, preferable 7.625 inches. The inner radius of the cylindrical portion of the annular seal 60 may be 109 inches. EXAMPLE 1 In order to demonstrate the effectiveness of the seal joint of the present invention, the seal of the prior art and the present invention were analyzed by the Finite Element analysis method and the results were compared. In FIG. 7 a membrane 85 of a seal 80 of the prior art sealed to a reactor flange 84 and a deck or platform 86 is shown. The membrane 85 is of straight cylindrical construction and is 0.125 inches thick. In FIG. 8 a seal joint membrane 95 of a seal 90 of the present invention is shown. The seal joint membrane 95 in FIG. 8 has a straight cylindrical section 91 and a flexible coil section 99. The coil section 99 has two coils or arcuate portions 97 and 98 of three thicknesses. The radius of both coil sections 97 and 98 are 0.95 inches. The thickness of the section of the coil indicated at 92 is 0.145 inches, at 93 is 0.20-inches thick, and at 99 is 0.281 inches thick. the cylindrical section 91 is 0.125 inches thick. The seal joints shown in FIGS. 7 and 8 were assumed to be installed for the same postulated refueling deck and under the same following thermal movements: EQU Reactor Vessel Radial Movement .DELTA.=0.43 inches; and EQU Reactor Vessel Axial Movement.DELTA.=0.70 inches. Both joints were made from the same type 304 Stainless Steel material. The radius of the entire cylindrical sections 91 and 85 is 109 inches. The calculated Tresca Stress of the seal joint 80 is 63.1 ksi at the point indicated at 81 and 134.1 ksi at the point indicated at 82. On the other hand, the calculated Tresca Stress of the seal 90 is 59.5 ksi indicated at 101 and 92 ksi indicated at 102. The calculated Maximum Tresca Stress for joint 80 is 134.1 ksi while it is 92 ksi for joint 90. In accordance with requirements for ASME Boiler and Pressure Vessel Code Section III, the seal 80 can only operate 136 cycles while the seal 90 can operate 702 cycles. In other words, the seal 90 of the present invention lasts five times longer than the seal 80 of the prior art. Typically a plant operates 500 cycles of heating-up and cooling-down. Accordingly, the seal 80 is not a permanent seal since it only lasts for 136 of these cycles. In contrast, the seal 90 lasts for 702 cycles, well beyond the life of the plant. Thus, the seal according to the present invention requires no maintenance or inspection. Further, the seal is not subject to leakage and is not subject to degradation since it is designed from metal. Moreover, the seal of the present invention reduces the seal's joint stress and, as a result, can accommodate greater radial, axial and rotational movements of the vessel and has a useful life as long as the life of the plant. In addition, it should be recognized that the seal arrangement of the present invention can accommodate typical shielding and insulation systems. The foregoing description of a preferred embodiment of the invention has been presented for purposes of illustration and description. It is not intended to be exhaustive or to limit the invention to the precise form disclosed. Many modifications and variations are possible in light of the above teaching. The embodiment was chosen and described in order to best explain the principles of the invention and its practical application to thereby enable others skilled in the art to best utilize the invention and various embodiments and with various modifications as are suited to the particular use contemplated. It is intended that the scope of the invention only be limited by the claims appended hereto. |
abstract | A method and a device for state sensing of a technical system, such as an energy store, in which performance quantities are measured and supplied to a state estimation routine, which determines the state variables characterizing the current system state using a model based on system-dependent model parameters and the measured performance quantities. To improve state estimation, the measured performance quantities may be supplied to a parameter estimation routine, which performs a use-dependent determination of the model parameters. To increase the quality of the estimation and reduce the calculating time and the memory requirements, a selection of state variables and/or parameters determined by estimation are performed depending on the dynamic response of the measured performance quantities. |
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041995392 | claims | 1. A method for monitoring the operation of a dual platen press including a first platen and a second platen subject to displacement with variable velocity toward and away from each other during a cycle of operation, the method comprising the steps of: indicating the displacement of said first platen; indicating the displacement of said second platen; and imposing said displacements on orthogonal axes such that said displacements jointly control the motion of a point which traces a first lissajous figure representative of the displacement and relative velocity of said platens. superimposing a second lissajous figure representing desired platen displacement and over said first lissajous figure to facilitate comparison between said first lissajous figure and said second lissajous figure. indicating the displacement of said first platen; indicating the displacement of said second platen; imposing said indicated displacements on orthogonal axes such that said displacements jointly control the motion of a point which traces a first lissajous figure representative of the displacement and relative velocity of said platens, the completion of said figure coinciding with the completion of the cycle of operation of said platens; detecting the completion of preselected portions of said first lissajous figure; and triggering subsequent operations of the cycle of operation in response to the detection of the completion of said preselected portions of said first lissajous figure for controlling the operation of the press. generating a first electrical signal representative of the displacement of said first platen; generating a second electrical signal representative of the displacement of said second platen; and imposing said signals on orthogonal axes such that said signals jointly control the motion of a point which traces a first lissajous figure representative of the displacement and relative velocity of said platens. superimposing a second lissajous figure representing desired platen displacement and relative velocity on said orthogonal axes over said first lissajous figure to facilitate comparison between said first lissajous figure and said second lissajous figure. generating a first electrical signal representative of the displacement of said first platen; generating a second electrical signal representative of the displacement of said second platen; imposing said signals on orthogonal axes such that said signals jointly control the motion of a point which traces a first lissajous figure representative of the displacement and relative velocity of said platens; detecting the completion of preselected portions of said first lissajous figure; and triggering subsequent operations of the cycle of operation in response to the detection of the completion of said preselected portions of said first lissajous figure for controlling the operation of said press. generating a first electrical signal representative of the displacement of said first platen; generating a second electrical signal representative of the displacement of said second platen; and imposing said signals on orthogonal axes such that said signals jointly control the motion of a point which traces a first lissajous figure representative of the displacement and relative velocity of said platens. 2. The method of claim 1 further including the step of: 3. A method for monitoring and controlling the operation of a dual platen press including a first platen and a second platen subject to displacement with variable velocity toward and away from each other during a cycle of operation, the method comprising the steps of: 4. A method for monitoring the operation of a dual platen press including a first platen and a second platen subject to displacement with variable velocity toward and away from each other during a cycle of operation, the method comprising the steps of: 5. The method of claim 4 further including the step of: 6. A method for monitoring and controlling the operation of a dual platen press including a first platen and a second platen subject to displacement with variable velocity toward and away from each other during a cycle of operation, the method comprising the steps of: 7. A method for monitoring the operation of a dual platen press used in the manufacture of nuclear fuel pellets, said press including a first platen and a second platen subject to displacement with variable velocity toward and away from each other during a cycle of operation, the method comprising the steps of: |
051125668 | description | DETAILED DESCRIPTION OF THE INVENTION Referring initially to FIG. 1, the inventive device is seen in its operable state within pool 10 that is filled with water 12. The device is affixed to pool 10 at curb 14 about its upper end and laterally against wall 16 (see FIG. 4). The predominant components of the inventive device can be seen in FIG. 1 to include instrument package 18 (which will be described in detail later herein), upper strongback section and drive system 20 (see FIG. 2), lower strongback section and rail assembly 22 (see FIG. 3), and carriage assembly 24 (see FIG. 3). The remaining features of the device depicted at FIG. 1 will become apparent from the description that follows. Referring to upper strongback section and drive system 20, reference will be made to FIGS. 1, 2, and 4. Initially, it will be observed that platform 26 is attached to curb 14 of pool 10 by clamp assembly 28 which includes channels 31 and 32 through which threaded swivel pad assemblies 34a-d pass to provide clamping against curb 14. Upper platform 26 additionally contains the drive mechanism for driving carriage assembly 24. Motor 36 is seen to be coupled to gear reducer 38 to which is attached optical encoder 40. Motor 36 desirably is a stepper motor and optical encoder 40 enables the operator to determine the extent of movement of carriage assembly 24 during use of the inventive device. Take-up and idler assembly 42 connect gear reducer 38 to carriage assembly 24 via chain 44 (see FIG. 5). Junction box 46 is the last major element on upper platform 26 and provides for centralized connection to instrument package 18 (as will be described in detail later herein). Connected to upper platform 26 is pipe 48, which in the preferred use of the inventive device in the dimensional characterization of nuclear reactor fuel channels, has an outside diameter of about 12 inches and a wall thickness of one-quarter inch. Pipe 48 is open at both ends so that water fills the interior thereof when the assembly is placed within water 12 of pool 10. About the upper end of pipe 48 can be seen lift assembly 50 which is designed for connection to a crane, for example, for lifting the inventive device into and out of water 12 of pool 10. It should be observed that pipe 48 in the present design is fabricated in two sections which are attached at flanges 52 and 54 (see FIG. 3). Such arrangement enhances the portability of the device by enabling pipe 48 to be disassembled for more compact storage and transport thereof. Referring now to lower strongback section and rail assembly 22, reference in particular will be made to FIGS. 2 and 3. The lower section of pipe 48 assists in aligning pipe 48, and thus the inventive device, in the desirable vertical orientation utilizing threaded swivel pad assembly 56 which is carried by lower platform 58 and which is in contact with wall 16 (see FIG. 5). Lower platform 58 also is fitted with shock absorbers 60a and 60b (60b not visible in the drawings) for carriage assembly 24. Next, the component to be measured and its attachment to the inventive device will be described with particular reference to FIGS. 1-4. Initially, it will be observed that grapple assembly 62 is connected to fuel channel 30 about its upper end. Grapple assembly 62, in turn, is connected to alignment sleeve 84 which is connected to grapple shaft 85 which is fitted with grapple handle 66 about its upper end. Such combination permits the transport of fuel channel 30 from a fuel storage pool, for example, to be placed within the inventive device for dimensional characterization. Grapple shaft 85 is retained at upper platform 26 by retainer bar 68 which is connected by tie channel 70 to upper platform 26. Retaining bar 68 is pivotally connected to tie channel 70 at the end closest to curb 14 for retaining grapple shaft 85. Alignment sleeve 84, just above grapple assembly 62, is secured to the lower section of pipe 48 by grapple clamp 72 which retains alignment sleeve 84 in the V recess contained at the outer projection of grapple rest 74. Grapple rest 74, in turn, is secured to pipe 48. Alignment sleeve 84 has a machined outer surface for precise position alignment with the V recess of grapple rest 74. Grapple clamp 72 is pivotally connected to slide block 76 which, in turn, is connected to spring holder 78 which, in turn, is pivotally connected to clamp arm 80, which is connected to pipe 48 as depicted at FIG. 3. Spring 82 encircles spring holder 78 and is biased against slide block 76 for closing grapple clamp 72 against grapple rest 74. Referring to FIG. 4, grapple clamp 72 is manually actuated to retain alignment sleeve 84 in the V recess of grapple rest 74 by a, for example, 3/4 inch socket tool (not shown in the drawings) with a suitable extension (e.g. 6 feet when the inventive device is used to dimensionally characterize nuclear fuel channels) on hexagonal head actuator 81 (3/4 inch when a 3/4 inch socket tool is used) at the pivot point of clamp arm 80. The operator turns hexagonal head 81 counterclockwise to release alignment sleeve 84 and clockwise to clamp alignment sleeve 84 tight in the V recess of grapple rest 74. Alignment shaft 83 prevents pivot clamp arm 80 attached to spring holder 78 from extending beyond slightly over-center, thus assuring that alignment sleeve 84 is clamped tight and will be held in place when unattended. Referring to FIG. 9, it will be observed that actuator shaft 64 is inserted into grapple shaft 85 and into alignment sleeve 84 of grapple assembly 62. Actuator shaft 64 has flanges 86 and 88 about its lower end. Flange 88 is inserted into grapple body 90 through an aperture therein which rotates dogs 92 and 94 to a closed position indicated by the end of actuator shaft 64 contacting end plate assembly 96. Dogs 92 and 94 pivot about shafts 98 and 100, respectively. Bearings are provided about shafts 98 and 100 in conventional fashion. When dogs 92 and 94 are in place and secure against flange 88, it will be observed that they engage lip 102 about the upper or proximal end of fuel channel 30 which is retained about its upper side by the lower end of grapple body 90. Actuator shaft 64 is moved up and down actuating dogs 92 and 94 by turning threaded knob 67. Actuator shaft 64 will not rotate due to flange 88 fitting into slot 91 in grapple body 90. Actuator shaft 64 moves up and down as threaded knob 67 is turned, as collet 68 prevents such movement by knob 67. Since fuel channel 30 must be precisely located in a predetermined reference position for its dimensional characterization, grapple assembly 62 is fitted with an alignment mechanism which can be seen by referring to FIG. 10. There are two types of alignment mechanisms in grapple assembly 62. There are two fixed alignment mechanisms, identical in construction and two spring-loaded alignment mechanisms identical in construction. For ease of labeling, various parts of the two alignment mechanisms will be used in the following detailed description. The fixed alignment mechanisms consist of bearing 105 which is pressed into sleeve 107 which is held in a fixed position in grapple tool 62 by cap screw 109. Bearing 105 rotates in sleeve 107. The spring-loaded alignment mechanisms operate as follows. Bearing block 104 pivots about shaft 106 and its position is adjusted by set screw 108. Housed within bearing block 104 is roll pin 110 which retains bearing 112 about its free end. Bearing 112 is free to rotate about roll pin 110 and presses against the inner wall of channel 30 by the action of spring 114. Retaining ring 116 secures bearing 112 to roll pin 110. Appropriate spacers are used on either side of bearing 112 in conventional fashion. End plate assembly 96 is secured to grapple body 90 by bolts 118 and 120. It will be observed that the four biased assemblies locate channel 30 in a pre-determined reference position about its upper end. With respect to securing channel 30 at its lower end, reference is made to FIGS. 11-13. Pedestal 122 is secured to lower platform 58 and is surmounted by tapered cap 124 which is adapted to receive the lower or distal end of fuel channel 30. The distal end of fuel channel 30 rests atop sensor standard 126. Within housing 128 is disposed an alignment mechanism similar in construction to that mechanism housed within grapple assembly 62 and described in detail in connection with FIG. 10. It will be observed, however, that four pairs of rollers are provided at two different vertical elevations with each set of four rollers adapted to abutt two oppositely-disposed interior faces of fuel channel 30 each. Referring to FIG. 12, it will be observed that bearing blocks 130a and 130b rotate about shafts 132a and 132b, respectively. Shafts 132a and 132b are conventionally provided with bearing sleeves. Housed within bearing blocks 130a and 130b, respectively, are roll pins 134a and 134b. About the free ends of roll pins 134a and 134b are bearings 136a and 136b which are retained by retaining rings 138a and 138b. It will be observed that bearings 136a and 136b rotate about roll pins 134a and 134b and are pressed against one side of fuel channel 30 by springs 140a and 140b which are housed within plate 142. Oppositely disposed within plate 142 are roll pins 144a and 144b which retain about their free ends bearings 146a and 146b which are secured by retaining rings 148a and 148b. Bearings 146a and 146b are adapted to confront and rotate against the opposite interior surface of fuel channel 30, thus providing a reference position in one direction of fuel channel 30. In order to provide a reference position for the distal end of fuel channel 30 in the opposite direction, reference is made to FIG. 13 where it will be observed that bearing blocks 150a and 150b rotate about shafts 152a and 152b and are provided with sleeve bearings in conventional fashion. Roll pins 154a and 154b are housed within bearing blocks 150a and 150b and about their free ends is located bearings 156a and 156b which are secured in position by retaining rings 158a and 158b. Again, bearings 156a and 156b confront one interior surface of fuel channel 30 while the oppositely-disposed interior surface is confronted by a similar assembly affixed to bearing block 160 which is retained by shaft 162. Disposed within bearing block 162 are roll pin 164a and 164b, which about their free ends retain bearings 166a and 166b which are secured in position by retaining rings 168a and 168b. It will be observed that bearings 166a and 166b are stationary while bearings 156a and 156b confront the interior surface of fuel channel 30 by the action of springs 170a and 170b that are housed within plate 172. It will be observed that bearing blocks 130a and 130b have provision for fine adjustment by set screws 174a and 174b and bearing blocks 150a and 150b can be adjusted by set screws 176a and 176b. Thus, the distal end of fuel channel 30 can be placed in a reference position for its dimensional characterization. It should be observed with reference to FIG. 13 that proximity sensor 178 senses the presence of a fuel channel in position on pedestal 122. With reference to carriage assembly 24, a review of FIG. 3 will reveal that carriage 24 is adapted for vertical movement along the longitudinal extent of pipe 48 in such a manner so as to simultaneously traverse the longitudinal extent of fuel channel 30 from its distal end to its proximal end and in such a manner enabling the dimensional characterization of all sides of fuel channel 30 simultaneously, a feature setting the present device apart from the art. Carriage assembly 24 is mounted to pipe 48 by rail assemblies 180 (FIG. 3) and 182 (see FIG. 6). Rail assembly 180 is composed of mounting bar 184 and rail 186 (FIG. 6). Rail assembly 182 is formed of mounting bar 188 mounted to pipe 48 and rail 190 affixed thereto. Drive from motor 36 is provided to carriage assembly 24 via chain 44 that winds between various idler sprockets, e.g. sprockets 192a, 192b, and 192c, for example (see FIG. 5), thence to return sprocket 194 and thence through chain guide 196 which returns the chain to take-up and idler assembly 42 about upper platform 26. With reference to FIG. 6, it will be observed that chain 44 and chain guide 196 is affixed via plate 198 to cross beam 200 that forms a portion of carriage assembly 24. Bolt 202 secures chain mounting plate 198 to cross-beam 200. In such fashion, drive chain 44 provides movement of carriage assembly 24 along the longitudinal extant of pipe 48 substantially the length of rail assemblies 180 and 182. Referring still to FIG. 6, it will be observed that roller assembly 204 with notched wheels provides secure attachment to rail 186. Roller assembly 206 with its flat wheels rides along rail 190 in following fashion as rail assembly 204 provides the reference location of carriage assembly 24 with respect to pipe 48. Idler sprocket 192 is exemplified in its attachment to pipe 48. Side beams 208 and 210 sandwich cross-beam 200 and provide an attachment location for roller assemblies 204 and 206 which are bolted thereon in conventional fashion. Sensor mounting plate 212 similarly is seen to be bolted to side beams 208 and 210. It will be observed that hole 211 penetrates sensor mounting plate 212 and is of the same configuration as the external circumference of fuel channel 30. It will be appreciated that hole or aperture 211 in sensor mounting plate 212 could be of different configuration mirroring the external circumference or fingerprint of the component to be dimensionally configured by the device of the present invention. While the number of sensors used to dimensionally configure each side of fuel channel 30 is variable, six sensors for each of the four sides of fuel channel 30 are depicted at FIGS. 6 and 7, though such number could be greater or lesser as is necessary, desirable, or convenient. In order not to clutter FIG. 6 to the point of obfuscating the invention, the description of the sensors for the device illustrated in the drawings will be described with particularity by reference to sensors 214a and 214b set forth at FIG. 7, although it will be appreciated that all 24 sensors set forth in the drawings are identical in construction. Sensors 214a and 214b are illustrated in the drawings as linear variable differential transformers (LVDT). Briefly, an LVDT comprises linearly movable spring-loaded plungers 216a and 216b extending from housings 232a and 232b. Movement of plungers 216a and 216b changes the mutual inductance of a pair of coils disposed within housings 218a and 218b. Thus, the mutual inductance of the pair of coils can be measured and interpreted as a function of the linear position of plungers 216a and 216b. A signal representing the mutual inductance of the pair of coils is carried by lines 220a and 220b, as will be more fully described later herein. Nose pieces 222a and 222b are fitted at the outer end of plugers 216a and 216b and are in contact with rollers 224a and 224b. These rollers are secured to bell cranks 226a and 226b, each of which are pivotally mounted to flex pivots 228a and 228b. Rollers 230a and 230b are affixed to each bell crank and are adapted to tactilely confront and ride along the outer surface of fuel channel 30 for its dimensional characterization. Utilizing a roller minimizes the opportunity for any mars or defects in the component being dimensionally characterized from placing an undue strain on the sensor and damaging it. Bell cranks 226a and 226b are biased against fuel channel 30 by springs contained in spring housings 232a and 232b. Sensor clamps 234a and 234b attach the LVDTs to sensor mounting plate 212 which, it will be observed, is fabricated in the form of a case in which the bell cranks are housed. Bail and guard ring assembly 236 similarly is attached to sensor mounting plate 212 and protects the LVDTs from damage. It will be observed that the use of the bell cranks enables the LVDT sensors to be disposed in an upright or vertical state. Since the LVDT sensors have substantial length, mounting them vertically enables a much smaller carriage assembly to be fabricated than, for example, the arrangement depicted in U.S. Pat. No. 4,274,205 which utilizes such LVDT sensors in horizontal disposition with the nosepieces directly riding on the fuel channel. When it is desired to move the bell cranks away from the fuel channel, reset bar 238 is activated by air cylinders 240a, 240b, 240c, and 240d. With reference to FIG. 8, it will be observed that shaft 242 is biasedly connected to reset bar 238 utilizing spring 244. Activating air cylinders 240 causes reset bar 238 to move upwardly, thus causing the bell canks to pivot outwardly and away from the component being dimensionally characterized. In order to ensure that carriage assembly 24 does not exceed its intended path, over-travel switches 246 and 248 (see FIG. 3) are provided. Proximity switch 250 (FIG. 3) is a "home" switch that provides a datum for carriage assembly 24. Proximity switch 250 is tripped by sensor actuator 252 mounted to cross beam 200 (FIG. 6b). In operation, a fuel channel is removed from the fuel bundle which is placed in storage. The grapple/alignment tool composed of grapple assembly 62, actuator shaft 64, and grapple handle 66 are attached to channel 30 making sure that the channel is in the requisite orientation. Fuel channel 30 then is transferred to the dimensional characterization device of the present invention in the requisite orientation and clamped in place with retaining bar 68 and grapple clamp 72. Channel 30 now is in place for measurement. Instrument package 18 (FIG. 1) is seen to include printer 254, computer 256, and signal conditioner 258 to which cable assemblies 260 and 262 run. Cable assembly 262 runs to junction box 46 while cable assembly 262 of data cables runs through data cable strain relief mechanism 264 and thence down to a similar mechanism affixed to carriage assembly 24. Power for the instruments located within wheeled instrument container 266 can be accessed via power cord 268. Again, a modular arrangement for portability and ease of transportability is seen to be provided. Now that the channel is ready for dimensional characterization, the operator enters the needed channel identification information and parameters into computer 256 as at block 272 in FIG. 14. Such parameters include the delay in activation of carriage 24 after initiation of authorization to engage the bell crank assemblies with channel 30, the carriage movement rate (e.g. 2.5 inches per second or other value). Thereafter, the operator initiates the automatic channel dimension measurement system data acquisition program stored within computer 256. The following sequence of events are monitored and/or controlled by the computer program. Initially, the program encounters function 276 wherein the computer looks for a signal from proximity sensor 178 indicative that the channel is in place and properly located within the fixture. If the "channel not in place" message appears on the screen as indicated in block 278, data collection is not allowed and the program is terminated as indicated at block 280. The operator then needs to remove channel 30 and start the sequence again by re-mounting it within the fixture. If the computer properly senses that the channel is in place, the program continues to block 282 wherein the computer next looks to see whether home switch 252 has been activated by proximity switch 250, indicating that the carriage is at the requisite "home" position. If the "home" position is not sensed, then the program ascertains at block 284 whether the computer has information sufficient that it knows where the "home" position is. If the answer is in the negative, then carriage movement is activated as in block 286 by energizing motor 36 with the carriage movement being in the downward direction until it reaches over-limit switch 248 as at block 288 and then movement of the carriage is directed slowly upwardly per the instruction at block 290 until proximity switch 250 is tripped by sensor actuator 252. This home position then is saved in the computer at instruction 292 by correlating optical encoder 40. In this condition or where the computer already knows where home is as at block 294, the program at step 296 then activates automatic channel dimension measurement sequencing. Initially, the program at 298 calls for the carriage to move to a position at sensor standard 126 at which time it stops. Solenoid 270 (FIG. 2) then causes the activation of air cylinders 240 which via reset bar 238 lowers bell cranks 226 into contact with sensor standard 126 according to instruction 300. The operator entered delay at step 272 then is encountered in the program at block 302. This time period of up to 30 seconds ensures that all bell cranks have been sufficiently released and are firmly resting against sensor standard 126. A set of data at 304 then is taken from the LVDTs of the bottom of the three sensor standards. Thereafter, carriage upward movement is commenced at step 306 at the operator specified rate entered at 272. With the carriage moving, data is taken at the second and third levels of sensor standard 126 according to block 308. All of the LVDTs now are calibrated at three different positions by virtue of sensor standard 126. Commencing at one inch from the bottom of fuel channel 30, data is taken at 2.51 mm (0.1 inch) increments substantially the entire longitudinal extent of fuel channel 30 per the instruction at block 310. Data collection is ceased at approximately one inch from the top of channel 30 per the instruction at block 312. Carriage elevation is determined by optical encoder 40 affixed to the drive system, as described above. When data collection terminates, all bell crank assemblies also are retracted from fuel channel 30 via reset bar 238. Per the instruction at block 314, the carriage then returns to the "home" position, e.g. at a rate of 7.5 inches per second. When the carriage reaches the "home" position, the program at step 316 specifies that the data automatically is transferred from RAM disk, where it was recorded during data collection, to a floppy disk housed within computer 256 for providing a permanent record of the dimensional characterization of fuel channel 30. Selected channel bow data also is displayed graphically on the CRT screen for operator verification of the dimensional characterization of fuel channel 30. Channel measurement having been completed, grapple clamp 72 and retaining bar 68 are released so that the channel can be transferred to storage. As to materials of construction, preferably all components are manufactured from stainless steel in order to provide a maximum degree of corrosion protection and for providing dimensional stability of all elements. It will be appreciated that various of the components shown and described herein may be altered or varied in accordance with conventional wisdom in the field and certainly are included within the present invention provided that such variations do not materially vary from the spirit and precepts of the present invention as described herein. |
abstract | A radiation imaging device suitable for SPECT or other nuclear imaging includes a detector (22) which receives radiation. A fan beam-slit collimator (20) is positioned adjacent a radiation receiving face (32) of the detector, intermediate the detector and a radiation source (12, 18). The collimator includes a plurality of slats (30) having a common focus. A body (44) adjacent the slats defines one or more elongate slits (46). The slit is arranged such that radiation passes through the slit and between the slats to the detector face. The body is at least substantially impermeable to the radiation. The fan beam-slit collimator (20) enables higher resolution or efficiency to be achieved from the detector. |
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summary | ||
description | This invention refers to a floor and wall cleaner specially designed to be used in critical areas with difficult accessibility or restricted access, such as pools for housing a reactor vessel at a nuclear power station, in which human presence must be avoided as far as possible and, should this be necessary, this must be for the shortest possible time. According to the invention, the floor cleaner comprises: A casing or housing provided with a suction mouth; Drive belts on each side, driven by respective motors; Inner rollers, provided with mutually independent drive media; A set of outer rollers with permanent opposite rotation; At least one elastic hinge of at least one axle carrying the rollers; Gear motor assemblies for the roller movement A set of sealed connections and a first control body; Lighting systems; At least one camera for taking pictures; A float or buoy of variable volume; A set of turbines for gripping the wall; A set of turbines with lateral movement; and An anchorage for holding the float or buoy to the body of the casing or housing. The pools in which the reactor of a nuclear power station is housed are made up of a cubicle which may be in a regular or irregular shape and have dimensions that can range from one or two dozen meters on the smallest horizontal side to several dozen meters on the larger side, with a height of several meters, able to temporarily house a large number of the components of the reactor in the dismantling stage. The base of the pools tends to be of irregular shape. On one hand there are small-sized recesses which have to be cleaned preferably before emptying the pool, as these could contain radioactive material, and there are also uneven parts of the floor, amongst other reasons due to the bolts for holding the vessel of the reactor. This thus requires a device for cleaning the floors of the pools in which reactors of nuclear power stations are housed which is able to clean narrow spaces, to the maximum width of the apparatus and which is able to get over any small obstacles which it might come up against. As well as the floors, particles are deposited on the walls of these pools. Conventional devices are not nevertheless able to clean the walls, as if they did so it would be the suction force of the absorption system which would have to keep the device attached to the wall. Since these devices have to be made as far as possible of stainless steel or some other material able to be decontaminated, they have a high minimum weight, and the absorption systems conventionally used are not able to maintain their grip. Furthermore, even when the absorption capacity is enough to maintain a grip, any irregularity or space would cause loss of adherence, and the device would fall to the floor and have to be positioned again. Since the positioning task is extremely delicate, this risk in an installation of this sort makes such a system inoperative. There are different types of floor cleaners. First of all there are manual cleaners, which have a rod with which the cleaning head is moved; this head is connected by means of a suction hose to a pump and normally to a filter to be returned to the pool. This type of cleaner cannot be used in the vessel of a nuclear reactor due to several problems: The floor tends to be located more than ten meters below the surface and there is an even greater distance to the accessible upper edge; The pool may not have an upper perimeter strip from which the rod can be handled; The visibility of the floor from the height at which this must be handled is very limited or none at all. This requires a person to be handling the rod, which is not feasible through the height at which this is handled, the lack of visibility and the dose of radiation that the person in question would receive. EP 1472425 describes an independent floor cleaner for pools which comprises a set of support wheels and is provided with filtration and pumping means. It does not have means of controlling the movement at will. EP 1002173 discloses a floor cleaner with movement induced by a suction flow from an exterior pump; like the previous one this does not have any means for controlling its movement. A robot device known on the market as “ZODIAC Sweepy M3”, comprises a pair of lateral drive chains driven by motors and also comprises a motor for pumping water through a filter. The cleaning width is nevertheless interior, between the drive chains, for which reason it ends far from the outer edges. Furthermore, since this is conceived for cleaning swimming pools, it is not designed to get over obstacles. In the nuclear industry, the “WEDA N600” device is also a compact device able to be handled in remote control or in automatic mode, which has, like the previous one, a pair of drive chains, in this case with front and rear brushes of a width roughly equal to that of the body of the device and in which the extraction system installed in the apparatus itself expels the water through filter bags. The “ATOX underwater bottom cleaner” device has a structure similar to the previous ones, in that this is provided with lateral drive chains, with a filtration body operated with an exterior pump. One major disadvantage of this device is its weight, apart from the difficulties of cleaning the side zones, for the reasons given above. Other devices, even whilst complying with some of the characteristics described in the devices mentioned, are machines with a greater size, weight, cost and with the disadvantages also described above, without the manoeuvring capacity which is intended to be solved with this invention. Furthermore, any of these can be held up by a small obstacle, such as a bolt head two or three centimeters high, when said obstacle is not directly confronted by one of the drive chains. There are light swimming pool cleaners made of plastic materials which are able to go up the walls of pleasure swimming pools, but which are not usable in the pool of a nuclear power plant reactor for the reasons stated above, since plastic is not an acceptable material for said use, and neither do they have devices for controlling their movement. None of said devices is able to efficiently clean the walls of the vessel of a nuclear power plant reactor in a controlled manner. It is furthermore desirable for the same apparatus which is able to clean the walls to be able to clean the floor. This has advantages in the cost of the device, since instead of two (one for the walls and one for the floor) one will be enough and the operations can be performed consecutively with no need to perform two decontamination processes; one of these is enough at the end of both operations, for cleaning the floor and the walls. It is furthermore desirable for the same apparatus to be suitable for cleaning sloping surfaces. The invention being proposed consists of a floor cleaner which comprises a structure carrying the other items, which are as follows: A front roller; the front roller is held on a central support, securely held in turn to one of the side elements forming said structure; this roller is elastically hinged to said central support; it is divided into two halves or bodies, each of these being on one side of the central support; A rear roller, essentially identical to the front roller; A front central roller, preferably the front roller and the front central roller should be driven by a single motor, but they could also be driven by means of separate motors; A rear central roller; the rear roller and the rear central roller should preferably be driven by a single motor, but they could also be driven by means of separate motors; A suction bell placed on the casing, with an upper intake (on the side opposite that of the support for the rollers) and a linear suction mouth which is placed between the central rollers; Two sets of drive wheels or belts, one on each side, in which each set of drive wheels or belts is driven by an independent motor; it is preferable for the movement to take place by means of belts, as the possibility of the device being held up on an obstacle, such as a bolt head, is lower if this option is used. The pulling takes place by means of independent motors, with variable speed and rotation direction, meaning that, depending on the rotation direction of the motors, the cleaner can move forward when both belts rotate at the same speed in one direction, move in reverse when they rotate inversely in respect of the above or with displacement when the speeds of the belts are different. For proper cleaning of the floor, there are central interior rollers and front and rear exterior rollers. In particular, according to the preferred embodiment, two interior rollers are used, with the suction bell between them, and two exterior rollers, each of these, the front and rear ones, being placed on a hinged support in a normally central position. The interior rollers have a smaller size than the width of the cleaner, insofar as these are driven from at least one of their sides and between the drive system. The outer rollers are divided into two portions, and driven from the centre, so that the free end of each side reaches the maximum width of the cleaner; in particular the length of the rollers is slightly greater than the width of the cleaner casing. The rollers are made up of a core and a sheath. It has been found that an ideal sheathing for proper cleaning is made up of rubber strips, arranged radially (in a transversal direction to the movement). Hence, at least some of the strips will have to be positioned radially in respect of the roller axis. These transversal strips may be joined to strips arranged on a plane perpendicular to the axle of the roller without impeding their operation. In normal operation, with no obstacles, the exterior rollers and interior rollers turn in a direction so as to move the dirt towards the interior of the suction bell, that is, they drag the dirt along the floor towards the interior of the suction bell. The displacement is caused by the drive belts. The movement of the front belts and of the rear belts in this normal operation will be in mutually opposite rotation directions; however, when they come up against an obstacle, one of the rollers may possibly have a support which exerts significant force, so that the movement inverse to its displacement could block the floor cleaner, without the drive belts having sufficient support. For this reason, since the front rollers and the rear rollers are driven by independent motors, in the event of their coming up against an obstacle, such as a bolt head or a drop or rise in level of some centimeters, all the rollers may be made to run in forward motion, that is, in the same rotation direction as the wheels or drive belts, which helps to get over the obstacle in question. The movement of the rollers is separate from the displacement movement of the cleaner, and is driven by two independent motors, as has already been said. The control device can nevertheless synchronise the motors for optimum operation. For the movement of the rollers and the drive belts, there are respectively motors and mechanical transmission assemblies, each formed of a plurality of pinions engaging each other. As has already been stated, the exterior rollers are driven from the central part; this central drive is made up of an arm or support which houses a mechanism, and sustains the corresponding parts of the lateral roller projecting outward, up to a width slightly over that of the casing. This means that the exterior rollers do not properly clean a central zone in which the support and the drive mechanism for the front and rear rollers are located, which is why this zone has to be cleaned by the interior rollers. The sheath of the interior rollers must thus be continuous on the longitudinal plane on which the mechanism for driving the exterior rollers is located, especially the front rollers. Throughout the cleaning process different obstacles may come up, such as screw heads, bolt covers, etc. These obstacles do not tend to be over 2 or 3 cm in height but no compact conventional system is able to overcome these without getting jammed. If the arm carrying the front or rear rollers were rigidly fixed to with the housing of the cleaner, this would make it jam, since on rising up the obstacle, it also undesirably raises the drive belts, and the device loses traction. For this reason it has been designed for both the front arm and the rear arm to have a hinged support, and be subject to an elastic retaining tension, so that the elevation tension is lower than the cleaner's effective weight in the water and so that when an obstacle is reached said arm rises over the obstacle and the cleaner continues its travel and after the obstacle is reached by the drive belts, these are indeed able to get over this with no further problems, the arm returning to the normal working position when the elastic tension caused on reaching the obstacle has been released. Sometimes small obstacles are nevertheless located in the centre of the cleaner and are not reached by the drive belts. To solve this drawback, at least one of the rollers, and in particular all of these, have been provided with a set of wheels joined to their axle, so that when the cleaner comes up against an obstacle, these wheels continue to pull. The wheels have a smaller diameter than that of the corresponding brush, so that they will not have contact with the floor unless an obstacle with sufficient height is found. This guarantees that the cleaning is correct in routes with no obstacles. Since the rollers are driven by independent motors, two by two (one for the front ones and one for the rear ones) when an obstacle is reached which holds up the floor cleaner, all the rollers will rotate in the same direction, the wheels of said rollers thus pressing on the obstacle and easily getting over this. According to a less preferred option for embodiment, the wheels of one of the rollers can be freely rotating, independently of the roller movement. The alignment of the support wheels of the interior rollers with the position of the arm holding the mechanism for driving the exterior rollers should be avoided, insofar as said exterior rollers do not reach the position of said supporting arm. The suction head is placed held on the cover of the structure, and comprises an upper suction mouth which is connected to a suction pump, either directly or through a conduit; if this is joined to a conduit, a connector is provided, freely rotating at both ends and in a central zone also at 45°, allowing the positioning of the conduit with no restriction both from the upper head and from any lateral position. The structure is made up of lateral elements and means of joining said elements; it also comprises an upper cover holding the suction head, and protectors or covers at the front and rear, essentially symmetrical except for the holes for the corresponding connectors. The structure is closed at the front and rear by the corresponding rollers. According to one option each of the lateral elements is formed of a pair of separate parallel plates which define a chamber housing mechanical transmission and possibly drive assemblies. Even when a turbine has been used for the cleaner to grip the floor in embodiments prior to this invention, this is insufficient. Furthermore, since the suction bell is in a central position, a turbine has to be displaced from said centre, and although this is not critical in cleaning floors, it causes unwanted imbalances when this has to clean walls, which could make the cleaner fall to the floor, requiring further repositioning. The floor and wall cleaner of the invention is thus provided with at least a pair of turbines, which may run simultaneously or independently. The use of turbines for adherence placed symmetrically in respect of the longitudinal and/or transversal central plane has been shown to have a satisfactory result, which cannot be achieved with a single one. Since the device may be used in a dark zone, such as the pool of reactor vessel at a nuclear power plant, the cleaner is designed to have lighting means, at least in the forward motion direction, but possibly also for reverse movement. It is also designed for this to have at least one camera and possible two, one at the front and one at the rear, so that the state of cleaning achieved can be known at all times as well as the directions to be taken. One of the problems for keeping the cleaner on a wall is the weight of the device. As already stated, plastic materials cannot be used in operations in radioactive zones, for which reason the cleaner has a significant weight, of several dozen kilograms. For this reason the casing has been provided on both sides with two supports for joining this to a float. The float has the aim of compensating part of the cleaner's weight. In particular, it has been designed to have a pair of supports on each side, so that when only walls have to be cleaned, the alignment of the float is roughly over the centre of gravity of the cleaner. When this has to clean sloping surfaces the anchorage could nevertheless be hinged, or arranged in any other position. The float comprises a normally prismatic sealed body, with a fixed volume, when the apparatus is operating. This sealed body can also comprise an inflatable interior membrane. It is designed to have inlet/outlet valves for cleaning or ballast, normally with water, when the volume required for the specific application is lower than the total volume of the chamber. This float exerts an upward force of from 40% to 90% of the weight of the cleaner, according to the design specifications, apart from overcoming its own weight. Furthermore, to regulate proper operation of the ascending and descending operations it has also been designed for the body to be provided with a second chamber fitted with an inflatable membrane, with a variable body which totally neutralises the weight of the body or even which makes this float. This second chamber is made with perforated sheet metal, so that when the membrane inflates, any water found inside said second chamber can easily be drained out. The cleaner comprises an electronic control system. The electronic control system determines the actions of speeds and movement directions of each of the motors for driving the displacement or movement of the rollers and the turbine, of the lighting and picture-taking elements, or indicates any fault which might arise in the device. The electronic system comprises a sealed connection plate for connecting electric supply and control cables of the device. The control body is placed outside the device, and joined to this by means of supply cables for the different elements, insofar as it been shown that the radiation received in the pool quickly disables some of the functions. The governing system is normally placed in a remote control unit, which is normally a computer. This could possibly have an intermediate unit, for example a float which minimises the requirements of control cable sections, when the distances are too long, and which also enables control by means of wireless means. The following reference numbers are used in said figures: 1 upper cover 4 gripping turbines 8 turbine or autonomous external suction pump 11 lateral elements 12 front and rear covers 30 suction mouth or nozzle 51 front exterior cleaning roller 52 rear exterior cleaning roller 53 front interior cleaning roller 54 rear interior cleaning roller 55 traction device 56 roller sheathing 100 drive and cleaning body 121 engagement opening 150 suction bell 151 rectangular section of the suction bell 200 floatation body 201 casing or housing of the floatation body 202 coupling arms of the floatation body 203 securing holes of the coupling arms 204 lateral turbines of the floatation body 205 second variable volume chamber 206 connection of the second chamber 210 scraper 211 soft strip 212 sustaining part 521 pivot axis of the exterior arms 525 support of the exterior rollers 526 hinged arm of the support of the exterior rollers 527 spring of the hinged arm 538 core of the interior roller 539 support wheels of the rollers 551 pulleys of the drive device 552 drive belt 558 transmission items 559 drive motor 561 lamellae of the roller housing The invention being proposed consists, as stated in the heading, of a floor and wall cleaner, governed by remote control, suitable for use in cleaning the floors and walls of the pools housing the vessel of nuclear power stations. This is made up of a drive and cleaning body (100) and a floatation body (200). The drive and cleaning body (100) is mainly made up of components of stainless steel and comprises the following elements: Lateral elements (11), joined together to form a structure; according to a preferred embodiment the lateral elements (11) are made up of a double wall on each of their sides, inside which transmission elements (558) are housed; An upper cover (1) provided with a suction mouth (30); Front and rear covers (12); at least one of these will normally be provided with an engagement opening (121), made in the end emerging over the upper cover (1); A traction device (55); the traction device is made up of at least one drive motor (559) with adjustable speed on each side of the drive and cleaning body (100); normally each of the sides will have a gear motor mechanism for distributing the movement to a pair of pulleys (551), one front and one rear, which sustain and move a drive belt (552) or band or chain. The drive motors (559) as well as the transmission mechanisms are independent on each of the sides and are governed by a control system which could determine whether one or both move, the speed of the movement and the rotation direction, so as to enable the following states: The cleaner is at rest, when the motors (559) are idle The cleaner moves in a forward direction, with a variable speed depending on the rotation speed of the motors (559), synchronized by the control body; The cleaner will rotate, by inverting the rotation direction of the motors (559) for a static rotation, or by variation of the speed of one of the motors in respect of the other, when the rotation takes place while moving; Hence, an axle moved by the drive motor (559) transmits the rotation movement to each of the sides, and a mechanical system of gears (transmission elements 558) transmits this to at least one of a pair of pulleys (551) or drive crown wheels set on the corresponding side; the movement is preferably transmitted to the two front and rear pulleys or crown wheels on each of the sides; the belt (552) may have a toothed interior matching the outside of the pulleys (551), so as to guarantee absolute control of the movement with no unwanted sliding; A set of cleaning rollers (51,52,53,54), the rollers are made up of a core (538) and a sheath (56); the sheath is made of an elastic material, such as rubber, formed of or comprising in its outer surface at least one set of tabs or lamellae (561) arranged in a radial position, i.e. transversal in respect of the rotation direction; said lamellae (561) may be complemented by others arranged in planes transversal to the roller axis (51,52,53,54), or in other directions, this cleaning roller assembly comprises: Exterior cleaning rollers (51,52) which are located at the front and rear edges of the casing; Interior cleaning rollers (53,54) which are located inside the casing, between the drive belts or between the lateral elements (11) which sustain these; The front rollers (51,53) are driven by means of a single motor which transmits the movement to the motor axles of both of these by means of the corresponding transmission mechanism, but, within the scope of the invention, they could also be driven by means of independent motors; and the rear cleaning rollers (52,54) are driven by means of a single motor which transmits the movement to the motor axles of both of these by means of the corresponding transmission mechanism, but they could also be driven, within the scope of the invention, by means of independent motors; In the ordinary cleaning operation, on flat surfaces or with fairly low obstacles, the front (exterior and interior) rollers and the rear (interior and exterior) rollers will rotate in opposite directions, dragging the dirt towards the centre of the device; there are nevertheless times at which it is necessary to get over an obstacle of some height; for this purpose the exterior rollers are arranged on a support (525) with an elastically hinged arm (526) which tends to be placed in the lower position, for cleaning, but which is able to rise against the elastic force when an obstacle forces it to do so; also in view of any change in position of the cleaning device, particularly through its forward or backward tilting, it has been seen that it is useful for the rear exterior roller also to be arranged on an elastically rotated arm (526); it has nevertheless been designed for the cleaning of smooth surfaces, especially walls, that the arm (526) can be secured to prevent any movement; a securing pin is enough to do this; The rollers will normally rotate in opposite directions, dragging the dirt towards the suction zone; to assist in getting over obstacles, it is nevertheless designed for the rollers to be able to all turn in the same direction as the drive means; The width of the interior rollers is thus limited by the width of the casing; it is nevertheless a requisite for the cleaning to be carried out at the maximum width of the devices, without a wall or any other similar obstacle being able to limit the lateral cleaning capacity; for this reason the exterior cleaning rollers (51,52) reach the required width on the outside, at the limit of or outside the width of the device; for this purpose they are fitted on respective central arms (526) which support these, and which have the corresponding transmission mechanisms, with the cleaning roller (51,52) formed of two separate portions and sustained only by its central part (by one end of each of the portions); in accordance with one option, the separate portions may be independently sustained, so that the corresponding arm (526) is independent for each side, and in the event of there being any type of hinge of said arm (526) the axis of the two portions could become out of alignment; said option is not nevertheless considered preferential due to its mechanical complexity, even though it is considered within the scope of the invention; as a general rule, the two front and rear arms are elastically hinged so that they can pivot on respective axles (521) located in the body of the casing (1); when an obstacle is reached the elastic retaining of the arm (526) which keeps this in a position aligned with the floor (as for the rest of the rollers) is overcome, so that the arm allows the cleaning roller (51,52) which this sustains to rise, thus preventing the cleaner from becoming jammed on said obstacle; in FIG. 9, one can see a configuration of the hinged arm (526) with elastic retention by means of a spring (527), the power shafts for the movement are represented and in a preferential embodiment the support (525) of the arm (526) normally held to just one of the lateral elements (11) of the structure; the interior rollers (53,54) have a core made up of a single continuous rigid body, and their sheathing divided into portions and the drive mechanism is placed on at least one of their sides; on the other hand, the exterior rollers (51,52) are divided so that these have two external portions, with a central drive mechanism in the arms (526) which constitute the sole support of each of said portions; The housing of the interior rollers (53,54) is made up of several portions, between which there are one or more support wheels, with movement linked to the roller on which these are located, or free in respect of this, in a less preferred embodiment; it is intended for the support wheels (539) of the interior rollers not to be aligned with the wheels and drive mechanisms for the exterior rollers, in which there is no exterior cleaning, and these are not aligned either with the support wheels (539) of the other of the interior rollers; The exterior rollers also preferably have support wheels (539); The casing (1) also comprises at least one pair of gripping turbines (4); the gripping turbines (4) take the water from the outside of the casing and drive this in normal direction (perpendicular) and in the opposite direction to the support surface of the rollers (normally horizontal); the greater the discharge force (flow, speed), the greater the adherence to the surface will be; in particular there are two turbines located on the longitudinal symmetry plane and symmetrically in respect of the transversal symmetry plane; The structure also comprises, according to a preferred option, a suction turbine or pump (8) (represented in FIG. 4), independent and linked with the suction mouth (30); the suction turbine or pump (8) is set on the outside of the suction mouth (30), integrated in the cleaner, allowing independent operation with no need for any external suction source; it can be provided with filtration media or not; The casing (1) comprises a suction mouth (30); in the event of this having to be connected to an external inlet with a suction tube, said suction mouth will be provided with rotating elements and with a rotation body at a 45° angle; The casing (1) also comprises at least a light source (70) and a camera (80) for taking pictures; The casing (1) is provided with a sealed connection plate or sealed connectors; the cleaner comprises a control and governing body; due to the sensitivity of the semi-conductors to radiation, it is intended for the control body to be placed outside the device, preferably outside the intense radiation zone, and joined to this from the outside (in the area around the pool) by means of connection cables; this control and governing body will provide remote control for each of the elements controlled, such as stop-start and speeds and direction of rotation of each of the motors, as well as the light, camera, turbines, etc. The upper cover (1) holds a suction bell (150); in its upper portion it forms the suction mouth (30), and in the lower zone it forms a rectangular section (151) set between the interior rollers at the height of the geometrical plane joining their corresponding axes. In this configuration, the drive and cleaning body (100) has a maximum width of roughly 32 cm and a length of roughly 41 cm, and has a floatation body roughly 90 cm long, which allows great manoeuvring capacity and can reach recesses which would be impossible for other devices due to their dimensions and structure. To give the cleaner the required floatability to be able to move along a wall or sloping surface, normally from the top downwards, it has been designed for said cleaner to comprise a floatation body (200). The floatation body (200) is formed of a casing or housing (201) with at least one first sealed chamber, which can also be provided with an interior inflatable membrane. This first chamber is provided with inlet/outlet connections for filling/emptying and interior cleaning. This will normally be full of air, but in some applications, or to be used with a lighter body, it may be partly full of water, the rest being air, which means that the float force can be regulated. The floatation body (200) is provided with coupling arms (202) on both sides of one of its ends. These coupling arms (202) comprise at least two holes (203) or means of connection to other corresponding ones on the outer walls of the lateral elements (11). These arms will preferably hold the lateral walls by means of securing screws in all their holes. However, especially when used for cleaning sloping surfaces, the arms will be secured with a single screw on each side, allowing the floatation body to tilt (200) in respect of the drive and cleaning body (100), held by the coupling arms (202). The floatation body comprises lateral impulsion turbines (204). In a preferential embodiment, the lateral impulsion turbines (204) are attached to the coupling arms (202), along with the first chamber (201). The activation of these turbines when the cleaner is in a state of weightlessness through the compensation of the weight with the corresponding floating force will enable a lateral displacement to a new cleaning position. The turbines for securing the cleaner body will return the cleaner to the surface of the wall so that the assembly has free travel in all degrees since it is provided with forward and reverse movements, rotation, gripping and withdrawing from the surface to be cleaned, and lateral displacement. According to a preferential embodiment, the floatation body (200) also comprises a second chamber (205) with variable volume, provided with an interior inflatable membrane. Filling/emptying said second chamber (205) with air is done by means of a connection (206) to an external compressor. Depending on whether greater or lesser floatation of the cleaner is needed, the variable volume chamber will be totally empty, thus meaning that the effective weight of the cleaner will be the maximum or will be partly full, or totally full of air, and the effective weight will therefore be the minimum. The second chamber comprises at least one perforated wall, so that when the inner balloon inside this is filled (totally or partly) with air, the water that the volume of air displaces can be drained out. According to a particular embodiment, the first fixed volume chamber and the second variable volume chamber constitute a prismatic body to which the coupling arms (202) are linked; the chamber with variable volume is preferably located in the portion furthest from said prismatic body. Furthermore, insofar as the floatation body (200) will when cleaning walls always be located at the top of the drive and cleaning body (100), it has been designed for the floatation body (200) to be provided in the portion furthest from said drive and cleaning body (100) with a scraper (210). The scraper is formed of a soft strip (211), normally made of rubber, arranged on a holding part (212); in accordance with a preferential embodiment this support is made up of a tube with circular section made of a light material and filled with injected foam, thus minimizing its density and constituting a further floating part. In a preferred embodiment, the holding part is set on one or more supports joined to the floatation body (200) which allow the scraper to take up different angular positions, modifying the distance to the wall in the same way in accordance with operating requirements. |
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abstract | A method of making a nuclear fuel pellet for a nuclear power reactor. The method includes: providing a nuclear fuel material in powder form, pressing the powder such that a green pellet is obtained; providing a liquid that comprises an additive which is to be added to the green pellet; contacting the green pellet with the liquid so the liquid, with the additive, penetrates into the pellet; and sintering the treated green pellet. The additive is such that larger grains in the nuclear fuel material are obtained with the additive. |
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abstract | The present invention provides compositions for the protection against electromagnetic radiation. The compositions include a polymeric material including a polyamide such as nylon 6 or nylon 6, 6, barium sulfate and magnesium sulfate. The polymeric material upon exposure to incident electromagnetic radiation emits subtle electromagnetic oscillations at probiotic frequencies that counter adverse effects of incident electromagnetic radiation. The polymeric material may be formed into a protective housing for electronic devices and may be formed into protective fabrics. |
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061817600 | claims | 1. A sensor electrode for use in measuring an electrochemical corrosion potential comprising: a sensor tip; a conductor electrically connected to the sensor tip; an insulating member which surrounds the conductor; a connecting member which surrounds the conductor; and a sleeve which fits over the sensor tip, the insulating member, and the connecting member, the sleeve having inner threads which engage with corresponding outer threads on at least one of the sensor tip and the connecting member. a sensor tip electrically connected to a conductor; a insulating member comprising sapphire which surrounds the conductor; a connecting member which surrounds the conductor; a sleeve which fits over at least a portion of the sensor tip, the insulating member, the sleeve having inner threads which engage with corresponding outer threads on at least on at least one of the sensor tip and the connecting member and at least a portion of the connecting member, wherein the sleeve is preformed to have a density which is at least 97.5% of theoretical density. 2. The sensor electrode of claim 1, wherein the insulating member comprises sapphire. 3. The sensor electrode of claim 1, wherein the connecting member comprises alloy 42. 4. The sensor electrode of claim 1, further comprising a braze joint between the insulating member and at least one of the connecting member and the sensor tip. 5. The sensor electrode of claim 1, wherein the sensor tip comprises platinum. 6. The sensor electrode of claim 1, wherein the sleeve comprises at least one of magnesia stabilized zirconia, yttria stabilized zirconia, and zircaloy. 7. The sensor electrode of claim 1, wherein the sleeve has a density which is greater than 97.5% of theoretical density. 8. The sensor electrode of claim 1, wherein the sleeve has a density which is greater than 99% of theoretical density. 9. A sensor electrode for use in measuring an electrochemical corrosion potential comprising: 10. The sensor electrode of claim 9, wherein the sleeve comprises zircaloy. 11. The sensor electrode of claim 9, wherein the insulating member is sealed to at least one of the sensor tip and the connecting member, the seal comprising a ceramic powder sprayed onto the insulating member and at least one of the sensor tip and the connecting member. 12. The sensor electrode of claim 11, further comprising a bond coat beneath the ceramic powder. 13. The sensor electrode of claim 12, wherein the bond coat comprises M-Chromium-Alumina-Yttrium alloy, where M=NiCoFe or Ni+Co. 14. The sensor electrode of claim 9, wherein the connecting member comprises alloy 42. 15. The sensor electrode of claim 9, further comprising a braze joint between the insulating member and at least one of the connecting member and the sensor tip. 16. The sensor electrode of claim 9, wherein the sensor tip comprises platinum. 17. The sensor electrode of claim 9, wherein the sleeve comprises at least one of magnesia stabilized zirconia, yttria stabilized zirconia, and zircaloy. 18. The sensor electrode of claim 9, wherein the sleeve has a density which is greater than 99% of theoretical density. 19. The sensor electrode of claim 9, wherein the sleeve is formed by sintering a ceramic powder. 20. The sensor electrode of claim 19, wherein the ceramic powder comprises at least one of magnesia stabilized zirconia and yttria stabilized zirconia. |
054988250 | abstract | A method of removing radioactivity from the interior of a building by transporting radioactive material within a slurry, precipitating out or otherwise filtering out the then contaminated material outside the building, thus removing it in a continuous fluid recirculation system, and storing the precipitated out material while providing shielding of radiation, thereby to provide radiation protection without requiring conventional large mass to block the radioactivity. A toxic waste storage facility includes a building having a portion located below ground level, walls for bounding an interior space in the building, and recirculating fluid for removing thermal energy from the building and for providing radioactive shielding and absorption at least at part of the roof of the building. |
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