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description | Referring to FIGS. 1 and 2, a computed tomograph (CT) imaging system 10 is shown as including a gantry 12 representative of a xe2x80x9cthird generationxe2x80x9d CT scanner. Gantry 12 has an x-ray source 14 that projects a beam of x-rays 16 toward a detector array 18 on the opposite side of gantry 12. Detector array 18 is formed by detector elements 20 which together sense the projected x-rays that pass through an object 22, for example a medical patient. Each detector element 20 produces an electrical signal that represents the intensity of an impinging x-ray beam and hence the attenuation of the beam as it passes through patient 22. During a scan to acquire x-ray projection data, gantry 12 and the components mounted thereon rotate about a center of rotation 24. Detector array 18 may be fabricated in a single slice or multi-slice configuration. In a multi-slice configuration, detector array 18 has a plurality of rows of detector elements 20, only one of which is shown in FIG. 2. Rotation of gantry 12 and the operation of x-ray source 14 are governed by a control mechanism 26 of CT system 10. Control mechanism 26 includes an x-ray controller 28 that provides power and timing signals to x-ray source 14 and a gantry motor controller 30 that controls the rotational speed and position of gantry 12. A data acquisition system (DAS) 32 in control mechanism 26 samples analog data from detector elements 20 and converts the data to digital signals for subsequent processing. An image reconstructor 34 receives sampled and digitized x-ray data from DAS 32 and performs high speed image reconstruction. The reconstructed image is applied as an input to a computer 36 which stores the image in a mass storage device 38. Computer 36 also receives commands and scanning parameters from an operator via console 40 that has a keyboard. An associated cathode ray tube display 42 allows the operator to observe the reconstructed image and other data from computer 36. The operator supplied commands and parameters are used by computer 36 to provide control signals and information to DAS 32, x-ray controller 28 and gantry motor controller 30. In addition, computer 36 operates a table motor controller 44 which controls a motorized table 46 to position patient 22 in gantry 12. Particularly, table 46 moves portions of patient 22 through gantry opening 48. In one embodiment, and referring to FIGS. 3 and 4, detector array 18 comprises a plurality of modules 50. Each module 50 includes a scintillator array 52 and a photodiode array 54. Detector elements 20 include one photodiode of photodiode array 54, and a corresponding scintillator of scintillator array. Each module 50 of detector array 18 comprises a 16xc3x9716 array of detector elements 20, and detector array 18 comprises fifty-seven such modules 50. Dectector array 18 is thus capable of acquiring projection data for up to 16 image slices simultaneously. In one embodiment and referring to FIG. 5, to collimate x-rays 16 after they have passed through an object or patient 22, a post-patient collimator 56 is disposed over detector array 18. Post-patient collimator 56 comprises a top rail 58 and a bottom rail 60 spaced from and parallel to top rail 58. A plurality of collimator plates 62 (e.g., tungsten plates) are arranged radially between each rail 58, 60. (FIG. 5 is a cross-sectional view of post-patient collimator 56 through one collimator plate 62.) To attach collimator plates to rails 58 and 60, collimator plates 62 are each edge-welded at opposite ends to rails 58 and 60 using at least one directed energy beam welder 64. The use of edge welding prevents warping of collimator plates out of the plane of FIG. 5. Distortion inherent in other welding methods, including laser welding not specifically directed at edges of collimator plates 62, is avoided. Suitable types of directed energy beam welders 64 include those utilizing directed energy beams 65 comprising photons (e.g., laser beam welders) and those utilizing particles (e.g., electron beam welders). Directed energy beams 65 are thin beams of energy that concentrate their energy at a single point. (FIG. 5 is intended to show narrow beams 65 directed at different locations, i.e., 66, 68, 70, and 72 rather than two fan beams of energy.) In particular, a top rear corner 66, a top front corner 68 a bottom rear corner 70, and a bottom front corner 72 of collimator plates 62 are edge welded by directed energy beam welding in the plane of FIG. 5. Top rear corner 66 and bottom rear corner 70 are edge welded towards a rear 74 of top rail 58 and towards a rear 76 of bottom rail 60, respectively. Top front corner 68 and bottom front corner 72 are edge welded towards a front 78 of top rail 58 and towards a front 80 of bottom rail 60, respectively. In one embodiment and referring to FIG. 6, a collimator is prepared by assembling a plurality of sections. For each collimator section, a plurality of collimator plates 62 are edge welded, using at least one directed energy beam welder, to curved metal (e.g., steel) top and bottom segments 82 and 84, respectively. Each segment 82 and 84 has a cross sectional area and length smaller than that of rails 58, 60 to form sections 86 of a collimator. Sections 86 are then radially arrayed between and fastened to top and bottom rails 58 and 60. (The radial arrangement of sections 86 is illustrated in FIG. 7, which shows collimator plates 62 that are not actually visible in a top view as hidden lines.) Top segments 82 are affixed to top or upper rail 58 and bottom segments 84 are affixed to bottom or lower rail 60. Wires 92 (such as tungsten wires) are also affixed to collimator plates 62 in a direction transverse to rear edges 88 of the collimator plates 62. A fixture (not shown) is used to hold collimator plates 62 and rails 58, 60 (or segments 82, 84) in position relative to one another. This fixture serves essentially the same purpose as a comb in a conventional post-patient collimator. However, unlike a comb, a fixture is needed only during welding of post-patient collimator 56. The fixture is not, and does not become a part of collimator 56, and can be re-used as needed. It is not necessary to use spacers, such as the molybdenum spacers used in at least one known post-patient collimator. In one embodiment, two directed energy beam welders 64, 90 are used to weld collimator plates 62 to rails 58 and 60. In another embodiment, two welders 64, 90 are used to weld collimator plates 62 to segments 82 and 84. One of the welders produces the rear welds, while the other produces the front welds. For a multislice detector array 18, attenuating wires 92 (e.g., tungsten wires) are strung across collimator 56 in spaced notches 94 on rear edges 88 of collimator plates 62. Wires 92 provide x-ray attenuation between detector rows. In one embodiment of the present invention, a directed energy beam welder 64 is used to weld wires 92 onto collimator plates 62. In another embodiment, the precision of directed energy beam welders allows the use of collimator plates 62 without notches 94. Wires 92 are strung across collimator plates 62 transverse to rear edges 88 and are accurately positioned against the collimator plates, for example, by using a fixture. Wires 94 are then welded to collimator plates 62 using a directed energy beam welder 64. In one embodiment, laser welders are used as welders 64 and 90 and their welds are accurately aimed and operated by computers (not shown) under program control. FIG. 8 is an enlargement of region 96 of FIG. 5, showing how a wire 98 (for example, steel wire) is used in one embodiment to take up collimator plate 62 height and/or rail 58, 60 spacing tolerance in a z-direction. Wire 98 is inserted in chamfered gaps 100 between at least one of top rail 58 or bottom rail 60 and collimator plates 62. (The selection of which one or both of rails 58 and 60 is a design choice.) Wire 98 is welded on one side to the selected rail 58 (or 60) and on the other side to collimator plate 62. The welds of wire 98 to the selected rail 58 (or 60) are at least in chamfered gaps 100. In one embodiment using welded wire 98, a weld at 68 is omitted. Also in a segmented embodiment of the present invention, chamfered gaps 100 are provided between at least one segment 82 or 84 and collimator plates 62 rather than between rail 58 or 60 and plate 62. Chamfers forming chamfered gap 100 can be in either plate 62 or the opposing segment or rail, or both. FIG. 9 is a top view in an x-y plane of the collimator and laser welder configuration shown in FIG. 5 (or FIG. 6) showing a phantom outline of a segment 82 (if used) and the location of one collimator plate 62 welded to rail 58 (or segment 82). (Neither segment 82, if used, nor collimator plate 62 would actually be visible from the top of collimator 56.) FIG. 9 illustrates the curvature of collimator 56, which corresponds to that of detector array 18. The arrangement of collimator plates 62 in collimator 56 is such as to provide collimation between detector elements 20 that are adjacent one another in the same row or slice of detector array 18. In another embodiment and as shown in FIG. 10, laser welding is used in conjunction with a comb 102 affixed to at least one of rail 58 or 60 and optional spacers 104, 106, 108, for example, molybdenum spacers. In the embodiment illustrated in FIG. 10, collimator plates 62 are positioned in slots of combs 102, 110 and directed energy beam welders 64, 90 weld areas 112, 114 and 116. In one embodiment, welder 64 is also used to weld wires 92 into wire notches 94. It is clear that the various embodiments of the invention provide more efficient and less expensive manufacturing methods for producing post-patient collimators. The welded collimators themselves are less expensive and potentially more durable than collimators having adhesive bonds, whether or not a comb is part of the collimator. 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 modification within the spirit and scope of the claims. |
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description | The present application is a continuation application of U.S. patent application Ser. No. 12/527,673, filed on Aug. 18, 2009, the entire contents of which are incorporated herein by reference and priority to which is hereby claimed. The Ser. No. 12/527,673 is a U.S. national stage of application No. PCT/JP2008/052422, filed on 14 Feb. 2008, the entire contents of which are incorporated herein by reference and priority to which is hereby claimed. Priority under 35 U.S.C. §119(a) and 35 U.S.C. §365(b) is claimed from Japanese Application No. 2007-041437, filed 21 Feb. 2007, Japanese Application No. 2007-041440, filed 21 Feb. 2007, and Japanese Application No. 2007-041446, filed 21 Feb. 2007, the disclosures of which are also incorporated herein by reference. The present invention relates to a radiological image capturing apparatus and a radiological image capturing system, and specifically relates to such a radiological image capturing apparatus that employs a Talbot-Lau interferometer method and such a radiological image capturing system that applies various kinds of processing to an image captured by the radiological image capturing apparatus. In recent years, the morbidity rate of the rheumatic disease in Japan has reached to 1% of the national population, and accordingly, the rheumatic disease has been regarded as a kind of a folk disease at present. An abrasion loss at a cartilage portion (destruction of cartilage) and/or subtle changes of a bone shape and a bone trabeculae are observed as its early symptoms, and then, at the time when the symptoms have worsened, considerable changes of shape of the bone sections can be observed significantly. Accordingly, by observing the shape of cartilage portion and subtle changes of the bone shape and the bone trabeculae, it is possible to make a diagnosis with respect to the disease situation of the rheumatic disease at its early stage. Considering the actual condition that only medical treatments for stopping the progress of the symptoms are currently available as the medical treatments for the rheumatic disease, it is important to detect the rheumatic disease at its early stage and to speedily shift the patient into the phase of applying the medical treatments. However, the above-mentioned early symptoms of the rheumatic disease have been hardly detected by observing the X-ray photographic image, which has been widely accepted as a simple and convenient inspection method, and accordingly, it has been difficult for a doctor or the like to determine whether or not the rheumatic disease has actually developed. On the other hand, in recent years, instead of the radiographic images acquired by radiographing the patient, images acquired by employing the MRI (Magnetic Resonance Imaging) has been considered as a tool for making diagnosis, in order to detect the changes of cartilage tissues. Further, recently, in the field of the radiographic image capturing technologies, there has been reported such a technology that extracts a radiant beam, which straightly progress in parallel, so that the above-extracted radiant beam are used for capturing images of the cartilage portion concerned. However, since the patient has been heavily burdened with the MRI photographing operation from the viewpoint of the cost and time required for making a diagnosis, it has been difficult to perform the MRI photographing operation in the framework of the regular physical examination. Therefore, there has been such a problem that it has been difficult to periodically perform the MRI photographing operation so as to observe (inspect) the changes of the joint portions, such as fingers, etc., over time, as a longitudinal diagnosis. Further, since a huge image-capturing installation is necessary for conducting the image-capturing operation that employs the radiant beam, and, sometimes, several tens of minutes are required for completing the image-capturing operation, it has been virtually impossible for a general-purpose medical facility to employ the radiant beam for conducting the image-capturing operation. Due to the present situations as aforementioned, it has been desired to make it possible to simply and easily make a diagnosis on the diseases of cartilage portion at its early stage, such as a subtle change in a shape of joint portion, a subtle change in the bone shape, a swelling, etc. For instance, in order to make a diagnosis on the case of the rheumatic diseases at its early stage, it is indispensable to capture such a radiographic image that has a high sharpness being sufficient for recognizing a subtle change of a symptom in the patient, represented thereon. As the radiological image capturing apparatus that can capture a radiographic image having a sufficiently high sharpness, there has been well-known the technology for capturing a phase contrast image by employing the radiological image capturing apparatus, for instance, set forth in Patent Document 1. According to the technology set forth in Patent Document 1, even for such a subject whose X-ray absorbing rate is specifically lower than other subjects to such an extent that its radiological image having a sufficient contrast cannot be formed by employing the normal X-ray absorbing action, it has been possible to obtain such an radiological image in which contrast of the peripheral portions (edge portions) are specifically emphasized. Further, it has bee possible to apply the above-mentioned technology not only to joint disorders, which are represented by the rheumatic disease, but also to various kinds of sections, such as a breast image capturing operation that should be capable of detecting a micro calcification from a breast, most of which is formed by a soft tissue, an operation for radiographing a child body, almost bones of which are cartilages, etc. Further, as the technology for further emphasizing the contrast of the peripheral portions of the subject, for instance, Patent Document 2 sets forth an X ray radiographing apparatus employing the Talbot interferometer method based on the Talbot effect caused by the diffraction grating. Still further, Non-patent Document 1 sets forth an X ray radiographing method employing the Talbot-Lau interferometer method, which is improved from the Talbot interferometer method. [Patent Document 1] Tokkai 2004-248699 (Japanese Laid-open Non-Examined Patent Publication) [Patent Document 2] WO 2004-058070 (International Publication) [Non-patent Document 1] “RECENT DEVELOPMENT OF X RAY PHASE IMAGING” written by Atsushi Momose, Medical Imaging Technology, Japanese Society of Medical Imaging Technology, November, 2006, vol. 24, No. 5, page 359-366 However, since the radiant beam X ray source, which requires a special facility, is employed in the Talbot interferometer method set forth in Patent Document 2, there has been a problem that it is virtually impossible for general-purpose medical facilities, widely exiting in the society, to employ the Talbot interferometer method. In addition, in such the general-purpose medical facilities, it has been assumed that low energy X rays are to be irradiated onto the subject. This is because, the phase contrast effect, acquired by employing the low energy X rays, is relatively great, and the absorbing contrast effect, acquired by employing the conventional X rays radiological imaging, is relatively strong. However, since the excessively low energy X-rays tend to be absorbed into the human body, and accordingly, since an amount of the X rays arriving at a detector is relatively small, it is necessary to increase the dose of radiation exposure in order to acquire an appropriate S/N (Signal to Noise) ratio at the detector, resulting in an increase of the X-ray exposure. Further, the increase of the X-ray exposure will cause an extension of the image capturing time interval. However, it is difficult to make the movement of the human body, serving as the subject, freeze for a long time during the image capturing time interval. Then, as a result of the movements of the subject during the radiographing, an X-ray image in which the peripheral sections of the subject are blurred would be captured, and accordingly, an advantageous characteristic of the Talbot-Lau interferometer that can emphasize the contrast of peripheral sections of the subject would be deteriorated. On the other hand, if the energy of the X rays to be irradiated onto the human body is excessively high, it has been acquired such a knowledge that an image contrast being sufficient for depicting bone tissues and soft tissues that constitute the human body cannot be obtained. Accordingly, there has been a problem that an X ray image, which is sufficiently usable for making a diagnosis on the human body, serving as the subject, cannot be obtained, unless the contrast in the X ray image can be obtained. As abovementioned, when the radiological image capturing apparatus in conformity with the Talbot interferometer method is employed for the medical purpose, a usable range of the X ray radiation energy (precisely speaking, average energy) is relatively narrow. In addition, in order to generate the Talbot effect so as to realize the Talbot interferometer method, various kinds of strict limitations are applied to a distance between a first diffraction grating and a second diffraction grating, an interval (grating period) between diffraction elements constituting the each of the diffraction gratings, etc., as detailed later. Therefore, in order to apply the Talbot interferometer method to an operation for radiographing the various kinds of sections in the human body, specifically for such the sections that are hardly captured by the X-ray image capturing method, such as the cartilage tissue, etc., the radiological image capturing apparatus should be configured so as to fulfill the extremely strict conditions. Further, according to the Talbot-Lau interferometer method set forth in Non-patent Document 1, a multi-slit element is disposed between the X ray source to be employed in the Talbot interferometer method and the subject. Since the X ray emitting source is converted to multi (plural) radiant sources by employing the multi-slit element, it is possible to effectively utilize the Talbot effect, even if the X ray tube having a large focal diameter is employed in the apparatus. However, the structure and configuration of the concerned apparatus become more complicated than ever, and, further, since various kinds of conditions, such as positional relationships between the multi-slit element and the other elements, etc., are added as new limitations for configuring the apparatus, the structural conditions for the concerned apparatus become still more stricter than ever, and it is required for the concerned apparatus to fulfill such the extremely strict conditions. On the other hand, a dose of X rays, to be irradiated onto the subject by the radiological image capturing apparatus employing the Talbot interferometer method, is relatively small, compared to that to be irradiated by the other radiological image capturing apparatus employing the Talbot-Lau interferometer method. However, since the X rays are irradiated by a single X-ray emitting source, the radiological image capturing apparatus employing the Talbot interferometer method has such the advantage that a very clear X-ray image, sharpness of which is very high, can be obtained. Whereas, since the X ray emitting source is converted to the multi (plural) radiant sources by employing the multi-slit element, the Talbot-Lau interferometer method is inferior to the Talbot interferometer method in sharpness of the reproduced X ray image to some extent. However, since it is possible in the Talbot-Lau interferometer method to irradiate relatively high energy X rays onto the subject, compared to the Talbot interferometer method, the radiological image capturing apparatus employing the Talbot-Lau interferometer method has such the advantage that the X-ray radiographing operation can be completed within a shorter time than ever. Further, if a single radiological image capturing apparatus is so constituted that the abovementioned two methods are provided within the single apparatus so as to make it possible to selectively change them to each other, for instance, by selecting one of the methods corresponding to the current purpose of capturing the X-ray image, it becomes possible to obtain the X-ray image to which the advantage of the selected method is fully applied, resulting in a very convenient apparatus. Still further, if the apparatus concerned is constituted as abovementioned, it becomes possible to appropriately make a diagnosis by adaptively selecting either the Talbot interferometer method or the Talbot-Lau interferometer method. As aforementioned, it has been desired that the Talbot interferometer method or the Talbot-Lau interferometer method is employed for operations not only for capturing X-ray images of the joint disorders, which are represented by the rheumatic disease, but also for capturing X-ray images of various kinds of sections in human body, such as the breast image capturing operation that should be capable of detecting the micro calcification from the breast, most of which is formed by the soft tissue, an operation for radiographing the child body, almost bones of which are cartilages, etc. However, in order to achieve the abovementioned goals, it is indispensable to configure the apparatus so as to fulfill such the extremely strict conditions as aforementioned. To overcome the abovementioned drawbacks in conventional radiological image capturing apparatuses and systems, it is one of objects of the present invention to provide a radiological image capturing apparatuses, which makes it possible to obtain a good X ray image in which contrast of the peripheral portions (edge portions), such as a cartilage tissue of human body, etc., are emphasized by employing the Talbot interferometer method and the Talbot-Lau interferometer method, and to provide a radiological image capturing system in which the above-captured X ray image is processed. Accordingly, at least one of the objects of the present invention can be attained by any one of the radiological image capturing apparatuses and the radiological image capturing systems described as follows. (1) According to a radiological image capturing apparatus reflecting an aspect of the present invention, the radiological image capturing apparatus, comprises: an X ray tube to emit X rays having an average energy in a range of 15-60 keV; a subject placing plate to place a subject thereon; a multi-slit element that is disposed at a position located on an optical path of the X rays, emitted by the X ray tube, and that has plural slits formed therein; a first diffraction grating to diffract the X rays penetrated through the subject, so as to yield a Talbot effect; a second diffraction grating to diffract the X rays diffracted by the first diffraction grating; and an X-ray detector to detect the X rays diffracted by the second diffraction grating; wherein the multi-slit element and the second diffraction grating are in contact with each other; and wherein a first distance between the multi-slit element and the first diffraction grating, a second distance between the first diffraction grating and the second diffraction grating, and a slit interval distance of the multi-slit element are set at a value equal to or greater than 0.5 m, a value equal to or greater than 0.05 m and a value equal to or greater than 2 μm, respectively.(2) According to another aspect of the present invention, the radiological image capturing apparatus, recited in item 1, further comprises: a control device that compares a Moiré stripe image, captured before an actual operation of the radiological image capturing apparatus is commenced, with another Moiré stripe image captured after the actual operation of the radiological image capturing apparatus is commenced, to determine whether or not a distortion is generated in a diffraction member of the first diffraction grating or the second diffraction grating.(3) According to still another aspect of the present invention, in the radiological image capturing apparatus recited in item 1, the control device issues a warning notification corresponding to a result of determining whether or not the distortion is generated.(4) According to still another aspect of the present invention, the radiological image capturing apparatus, recited in item 1, further comprises: a first temperature sensor to measure a first temperature of the first diffraction grating; a second temperature sensor to measure a second temperature of the second diffraction grating; and a control device to determine whether or not at least one of the first temperature and the second temperature, measured through the first temperature sensor and/or the second temperature sensor, is equal to or greater than a reference temperature established in advance.(5) According to still another aspect of the present invention, in the radiological image capturing apparatus recited in item 4, the control device issues a warning notification corresponding to a result of determining whether or not at least one of the first temperature and the second temperature is equal to or greater than the reference temperature.(6) According to still another aspect of the present invention, in the radiological image capturing apparatus recited in item 1, the first diffraction grating, the second diffraction grating and the X-ray detector are made to rotate around a peripheral space of the subject, so as to continuously capture X ray images of the subject from various directions.(7) According to still another aspect of the present invention, in the radiological image capturing apparatus recited in item 1, the multi-slit element is capable of entering into and withdrawing from the optical path of the X rays emitted by the X ray tube; and the radiological image capturing apparatus further comprising: a control device to control the multi-slit element to enter into and withdraw from the optical path.(8) According to still another aspect of the present invention, in the radiological image capturing apparatus recited in item 7, when the multi-slit element is disposed at the position located on the optical path, the control device sets the first distance between the multi-slit element and the first diffraction grating at a value equal to or greater than 0.5 m, while, when the multi-slit element withdraws from the optical path, the control device sets a third distance between the X ray tube and the first diffraction grating at a value equal to or greater than 0.5 m and sets a focal point diameter of the X ray tube at a value equal to or greater than 1 μm.(9) According to still another aspect of the present invention, in the radiological image capturing apparatus recited in item 7, the control device is configured to detect an abnormal shadow candidate from the X ray image captured, so as to change a Talbot-Lau interferometer method to a Talbot interferometer method when detecting the abnormal shadow candidate, as a method to be currently employed in the radiological image capturing apparatus.(10) According to a radiological image capturing apparatus reflecting still another aspect of the present invention, the radiological image capturing apparatus, comprises: an X ray tube to emit X rays having an average energy in a range of 15-60 keV; a subject placing plate to place a subject thereon; a multi-slit element that is disposed at a position located on an optical path of the X rays, emitted by the X ray tube, and that has plural slits formed therein; a first diffraction grating to diffract the X rays, so as to yield a Talbot effect; a second diffraction grating to diffract the X rays diffracted by the first diffraction grating and penetrated through the subject; and an X-ray detector to detect the X rays diffracted by the second diffraction grating; wherein the multi-slit element and the second diffraction grating are in contact with each other; and wherein a first distance between the multi-slit element and the first diffraction grating, a second distance between the first diffraction grating and the second diffraction grating, and a slit interval distance of the multi-slit element are set at a value equal to or greater than 0.5 m, a value equal to or greater than 0.05 m and a value equal to or greater than 2 μm, respectively.(11) According to a radiological image capturing system reflecting still another aspect of the present invention, the radiological image capturing system, comprises: a radiological image capturing apparatus that is provided with: an X ray tube to emit X rays having an average energy in a range of 15-60 keV; a subject placing plate to place a subject thereon; a multi-slit element that is disposed at a position located on an optical path of the X rays, emitted by the X ray tube, and that has plural slits formed therein; a first diffraction grating to diffract the X rays penetrated through the subject, so as to yield a Talbot effect; a second diffraction grating to diffract the X rays diffracted by the first diffraction grating; and an X-ray detector to detect the X rays diffracted by the second diffraction grating; wherein the multi-slit element and the second diffraction grating are in contact with each other; and wherein a first distance between the multi-slit element and the first diffraction grating, a second distance between the first diffraction grating and the second diffraction grating, and a slit interval distance of the multi-slit element are set at a value equal to or greater than 0.5 m, a value equal to or greater than 0.05 m and a value equal to or greater than 2 μm, respectively; an image processing apparatus to apply various kinds of image processing to image data representing an image captured by the radiological image capturing apparatus; and an image outputting apparatus to output the image based on the image data processed by the image processing apparatus.(12) According to a radiological image capturing system reflecting still another aspect of the present invention, the radiological image capturing system, comprises: a radiological image capturing apparatus that is provided with: an X ray tube to emit X rays having an average energy in a range of 15-60 keV; a subject placing plate to place a subject thereon; a multi-slit element that is disposed at a position located on an optical path of the X rays, emitted by the X ray tube, and that has plural slits formed therein; a first diffraction grating to diffract the X rays, so as to yield a Talbot effect; a second diffraction grating to diffract the X rays diffracted by the first diffraction grating and penetrated through the subject; and an X-ray detector to detect the X rays diffracted by the second diffraction grating; wherein the multi-slit element and the second diffraction grating are in contact with each other; and wherein a first distance between the multi-slit element and the first diffraction grating, a second distance between the first diffraction grating and the second diffraction grating, and a slit interval distance of the multi-slit element are set at a value equal to or greater than 0.5 m, a value equal to or greater than 0.05 m and a value equal to or greater than 2 μm, respectively; an image processing apparatus to apply various kinds of image processing to image data representing an image captured by the radiological image capturing apparatus; and an image outputting apparatus to output the image based on the image data processed by the image processing apparatus.(13) According to a radiological image capturing system reflecting still another aspect of the present invention, the radiological image capturing system, comprises: a radiological image capturing apparatus that is provided with: an X ray tube to emit X rays having an average energy in a range of 15-60 keV; a subject placing plate to place a subject thereon; a multi-slit element that is disposed at a position located on an optical path of the X rays, emitted by the X ray tube, and that has plural slits formed therein; a first diffraction grating to diffract the X rays penetrated through the subject, so as to yield a Talbot effect; a second diffraction grating to diffract the X rays diffracted by the first diffraction grating; and an X-ray detector to detect the X rays diffracted by the second diffraction grating; wherein the multi-slit element and the second diffraction grating are in contact with each other; and wherein a first distance between the multi-slit element and the first diffraction grating, a second distance between the first diffraction grating and the second diffraction grating, and a slit interval distance of the multi-slit element are set at a value equal to or greater than 0.5 m, a value equal to or greater than 0.05 m and a value equal to or greater than 2 μm, respectively; and a diagnosis assistance apparatus to detect an abnormal shadow candidate from an X ray image captured by the radiological image capturing apparatus; wherein the multi-slit element is capable of entering into and withdrawing from the optical path of the X rays emitted by the X ray tube, and the radiological image capturing apparatus is further provided with a control device to control the multi-slit element to enter into and withdraw from the optical path; and wherein the control device is configured to detect an abnormal shadow candidate from the X ray image captured, so as to change a Talbot-Lau interferometer method to a Talbot interferometer method when the diagnosis assistance apparatus detects the abnormal shadow candidate, as a method to be currently employed in the radiological image capturing apparatus.(14) According to a radiological image capturing system reflecting yet another aspect of the present invention, the radiological image capturing system, comprises: a radiological image capturing apparatus that is provided with: an X ray tube to emit X rays having an average energy in a range of 15-60 keV; a subject placing plate to place a subject thereon; a multi-slit element that is disposed at a position located on an optical path of the X rays, emitted by the X ray tube, and that has plural slits formed therein; a first diffraction grating to diffract the X rays, so as to yield a Talbot effect; a second diffraction grating to diffract the X rays diffracted by the first diffraction grating and penetrated through the subject; and an X-ray detector to detect the X rays diffracted by the second diffraction grating; wherein the multi-slit element and the second diffraction grating are in contact with each other; and wherein a first distance between the multi-slit element and the first diffraction grating, a second distance between the first diffraction grating and the second diffraction grating, and a slit interval distance of the multi-slit element are set at a value equal to or greater than 0.5 m, a value equal to or greater than 0.05 m and a value equal to or greater than 2 μm, respectively; and a diagnosis assistance apparatus to detect an abnormal shadow candidate from an X ray image captured by the radiological image capturing apparatus; wherein the multi-slit element is capable of entering into and withdrawing from the optical path of the X rays emitted by the X ray tube, and the radiological image capturing apparatus is further provided with a control device to control the multi-slit element to enter into and withdraw from the optical path; and wherein the control device is configured to detect an abnormal shadow candidate from the X ray image captured, so as to change a Talbot-Lau interferometer method to a Talbot interferometer method when the diagnosis assistance apparatus detects the abnormal shadow candidate, as a method to be currently employed in the radiological image capturing apparatus. Referring to the drawings, the radiological image capturing apparatus and the radiological image capturing system, both embodied in the present invention, will be detailed in the following. However, the scope of the present invention is not limited to the examples indicated in the drawings. In the present embodiment, a radiological image capturing system 100 is constituted by: a radiological image capturing apparatus 1 that irradiates X rays, serving as radial rays, onto a subject so as to generate radiological image data of the subject; an image processing apparatus 30 that applies various kinds of image processing to the radiological image data generated by the radiological image capturing apparatus 1; and an image outputting apparatus 50 that outputs a radiological image, etc., onto a display screen or a film, based on processed image data generated by applying the various kinds of image processing to the radiological image data in the image processing apparatus 30. Each of the radiological image capturing apparatus 1, the image processing apparatus 30 and the image outputting apparatus 50 is coupled to a communication network N (hereinafter, referred to as a network N, for simplicity), such as a LAN (Local Area Network), etc., for instance, through a switching hub, etc., (not shown in the drawings). In this connection, the scope of the configuration of the radiological image capturing system 100 is not limited to the system exemplified in FIG. 1. For instance, it is also applicable that the radiological image capturing system is so constituted that the image processing apparatus 30 and the image outputting apparatus 50 are integrated into a single apparatus, so that the integrated single apparatus conducts both the image processing operations and the radiological image outputting operation (onto the display screen or the film) based on the processed image data. As shown in FIG. 2 and FIG. 3, the radiological image capturing apparatus 1 is provided with a base frame 2 that is fixed on the floor surface with bolts or the like and a supporting base member 3 that is movable in both up and down directions relative to the base frame 2. Further, an image capturing main section 4 is supported by the supporting base member 3 through a supporting shaft 5. The supporting shaft 5 is constituted by an outer supporting cylinder 5a shaped in a cylinder and an inner supporting shaft 5b disposed inside the outer supporting cylinder 5a, so as to make the outer supporting cylinder 5a rotatable in either a clockwise direction or a counterclockwise direction around the inner supporting shaft 5b, serving as the rotating axis. The supporting base member 3 is provided with a driving device 6 for driving its up-and-down movements and the rotational motion of the supporting shaft 5, and the driving device 6 is provided with a conventional driving motor (not shown in the drawings). The image capturing main section 4 is fixed to the outer supporting cylinder 5a so as to elevate or descend synchronized with the up-and-down movements of the supporting base member 3 through the supporting shaft 5. Further, the image capturing main section 4 is rotated around the inner supporting shaft 5b, serving as the rotating axis, by making the outer supporting cylinder 5a rotate in either the clockwise direction or the counterclockwise direction. A supporting bar member 7, shaped in substantially a bar, is fixed in the image capturing main section 4 in such a manner that the supporting bar member 7 can expand and contract in both up and down directions. An X ray tube 8 that irradiates X rays onto a subject H is disposed at the upper section of the supporting bar member 7 in such a manner that the X ray tube 8 can freely elevate and descend. The X ray tube 8 is driven to elevate or descend by a position adjustment device 9, which is provided with a conventional driving motor, etc. (not shown in the drawings), so as to adjust the position of the X ray tube 8. Further, a power source 10 to supply electric power is coupled to the X ray tube 8 through the supporting base member 3, the supporting shaft 5 and image capturing main section 4. Still further, an aperture 8a to adjust the X-ray irradiation field is disposed at an X-ray irradiation opening of the X ray tube 8 in such a manner that the aperture 8a can be freely opened and closed, and the aperture 8a elevates and descend with the X ray tube 8. A X ray tube that can irradiate X rays, having an average energy in a range of 15-60 keV, is employed as the X ray tube 8 abovementioned. This is because, when the average energy of the X rays to be irradiated, is smaller than 15 keV, since almost of all part of the irradiated X rays are absorbed into the subject, a dose of X ray exposure becomes extremely great, and accordingly, such the setting is not suitable for clinical use. On the other hand, when the average energy of the X rays to be irradiated, is greater than 60 keV, it has been impossible to acquire such the X-ray radiation image that has sufficient contrasts so as to clearly represent bones, soft tissue sections, etc., which constitute the human body, and therefore, there is a possibility that the acquired X-ray radiation image cannot be used for a medical diagnosis or the like. It is preferable that, for instance, the Coolidge X-ray tube or the rotation anode X-ray tube, which has been widely used in the actual medical field, is employed as the X ray tube 8. On that occasion, when a molybdenum (Mo) material is employed for the target (anode) of the X-ray tube, as widely employed in the breast image radiographing operation (mammography, in this case, a molybdenum filter, having a thickness of 30 μm, is normally added), generally speaking, the X rays having the average energy of 15 keV are emitted from the X-ray tube at the time when a set voltage of 22 kVp is applied to the X-ray tube concerned, while the X rays having the average energy of 21 keV are emitted from the X-ray tube at the time when a set voltage of 39 kVp is applied to the X-ray tube concerned. Further, when a tungsten (W) material is employed for the target (anode) of the X-ray tube, as widely employed in the normal radiographing operation, generally speaking, the X rays having the average energies of 22, 32, 47 and 60 keV are emitted from the X-ray tube at the time when set voltages of 30, 50, 100 and 150 kVp is applied to the X-ray tube concerned, respectively. In the case of the radiological image capturing apparatus 1, embodied in the present invention, in which not only operations for radiographing joint disorders, which are represented by the rheumatic disease, but also various kinds of other radiographing operations, such as a breast image radiographing operation that should be capable of detecting a micro calcification from a breast, most of which is formed by a soft tissue, an operation for radiographing a child body, almost bones of which are cartilages, etc., are objects to be conducted, since the sharpness of the captured image can be improved due to the phase contrast effect by irradiating the X rays, specifically having a low X-ray energy (voltage to be applied to the X-ray tube concerned is set at a low voltage), among the X rays having various levels of the average energies, it is preferable that the average energy of the X rays to be irradiated is in a range of 15-32 keV. Further, considering the does of radiation exposure, it is more preferable that the average energy of the X rays to be irradiated is in a range of 20-27 keV. This can be achieved by employing the tungsten (W) material for the target of the X-ray tube concerned. The radiological image capturing apparatus 1 is so constituted that the Talbot interferometer method and the Talbot-Lau interferometer method, both detailed later, can be selectively changed to each other. When the radiological image capturing apparatus 1 is used as the Talbot interferometer, the focal diameter of the X ray tube 8 is set at such a value that is equal to or greater than 1 μm, so as to irradiate the X rays having the average energy in the abovementioned range and to acquire a practical output intensity. In this connection, in order to acquire a sufficient X-ray intensity, it is preferable that the focal diameter of the X ray tube 8 is set at a value being equal to or greater than 7 μm. Further, the X rays to be incident onto the first diffraction grating, detailed later, should have a coherence property. From the point that the X rays to be employed has the average energy in a range of 15-60 keV, and from the other point that the upper limit of the length of the radiographing apparatus is around 2 meters at the longest as detailed later, it is preferable that the focal diameter of the X ray tube 8 is set at a value being equal to or smaller than 50 μm, in order to posses the coherence property. Further, in order to improve the coherence property and to acquire a clear image by effectively using the Talbot effect detailed later, it is more preferable that the focal diameter of the X ray tube 8 is set at a value being equal to or smaller than 30 μm. Further, when the radiological image capturing apparatus 1 is used as the Talbot-Lau interferometer, the X rays to be incident onto the first diffraction grating, detailed later, should have a coherence property, as well as the above, and it is preferable that the focal diameter of the X ray tube 8 is set at a smaller value, in order to posses the coherence property. However, according to the present invention, since the X rays emitted by the X ray tube 8 are converted to multi (plural) radiant sources by employing a multi-slit element 11 detailed later, and in addition, the high power outputting capability is required for the X ray tube 8, it is not necessary to make the focal diameter of the X ray tube 8 so small. Accordingly, in the Talbot-Lau interferometer method embodied in the present invention, the focal diameter of the X ray tube 8 is set at a value being equal to or greater than 10 μm. Concretely speaking, it is preferable that the focal diameter of the X ray tube 8 is in a range of 10-500 μm, and more preferable that the focal diameter of the X ray tube 8 is set at a value being equal to or greater than 50 μm. Practically, the focal diameter is preferably set at a value being in a range of 100-300 μm. In this connection, it is possible to measure the focal diameter of the X ray tube 8 by employing the method established in “JIS Z4704-1994, 7.4.1 FOCAL POINT TEST, (2.2) Slit Camera”. Further, an X-ray irradiation time (exposure time) for completing every radiographing operation can be set at around several parts of one second, or two-three seconds at the longest. In the present embodiment, a control device, detailed later, conducts an operation for changing the focal diameter of the X ray tube 8 from one to another by changing the angle of the target of the X ray tube. Various kinds of methods, such as a method for inclining the target, a method for providing the target having two angles in advance and changing the position of the target onto which the electron beam is irradiated, etc., can be employed for changing the angle of the target of the X ray tube. Other than the above, it is also possible that the system is so constituted that, for instance, the focal diameter of the X ray tube 8 is changed from one to another by changing the area of the electron beam to be irradiated onto the target, or plural X ray tubes, focal diameters of which are different from each other, are provided in the system, so as to change the X ray tube 8 itself from one to another at the time of the changeover operation between the Talbot interferometer method and Talbot-Lau interferometer method. In this connection, it is preferable that the X ray tube 8 fulfills such a condition that the half-value width of the wavelength distribution of the X rays to be irradiated is equal to or smaller than 0.1 times of the peak wavelength of the X rays concerned. As far as the X ray tube 8 fulfills such the condition as abovementioned, the scope of the applicable X ray tube is not limited to the Coolidge X-ray tube or the rotation anode X-ray tube aforementioned, and the micro-focus X ray source or the like may be applicable as the X ray tube 8. As shown in FIG. 3, the multi-slit element 11 is disposed at a lower side of the X ray tube 8. When the radiological image capturing apparatus 1 is used as the Talbot-Lau interferometer method, the multi-slit element 11 is inserted into the optical path of the X rays emitted from the X ray tube 8, while, when the radiological image capturing apparatus 1 is used as the Talbot interferometer method, the multi-slit element 11 is made to withdraw from the optical path concerned. As shown in FIG. 4, the multi-slit element 11 is constituted by plural thin plates, which are arranged so as to make a plurality of slits 111 line up in parallel to each other. Each of the plural thin plates is made of such a material that can shield the X rays (X ray absorbing capability is great), for instance, a lead, a tungsten, etc. Further, an aperture width of each of the slits 111 (namely, a slit width, so to speak) is set at a value in a range of 1-50 μm. In order to effectively utilize the Talbot effect and to acquire a sufficient amount of X rays, it is preferable that the slit width is formed at a value in a range of around 7-30 μm. Accordingly, the X rays that are incident onto the first diffraction grating, detailed later, are converted into the multi (plural) radiant sources while having the coherency property. In this connection, a space distance d0 between the slits 111 of the multi-slit element 11 will be detailed later on. Further, the slits 111 of the multi-slit element 11 are formed only within an area of the X-ray irradiation field of the X rays emitted from the X ray tube 8. As shown in FIG. 3, the multi-slit element 11 is supported by the supporting bar member 7 through a supporting member 112 in such a manner that the multi-slit element 11 can freely elevate and descend, and a position adjusting device 9 moves it upward or downward along the supporting bar member 7 so as to adjust the position of the multi-slit element 11. In the present embodiment, the multi-slit element 11 is mounted onto the supporting bar member 7 in such a manner that the multi-slit element 11 is made to rotate around the axis of the supporting bar member 7 by the driving action of the position adjusting device 9. Accordingly, when the radiological image capturing apparatus 1 is used as the Talbot interferometer, the multi-slit element 11 is made to rotate around the axis of the supporting bar member 7 so as to withdraw from the optical path aforementioned. On the other hand, when the radiological image capturing apparatus 1 is used as the Talbot-Lau interferometer, the multi-slit element 11 is made to rotate around the axis of the supporting bar member 7 so as to insert it into the optical path aforementioned. In this connection, it is also applicable that another driving device is employed for conducting the abovementioned rotating action or the radiological image capturing apparatus 1 is so constituted that the abovementioned rotating action can be manually achieved. For instance, other than the above, it is also applicable that the coupling portion between the multi-slit element 11 and the supporting bar member 7 is configured as being capable of freely expanded and contracted, so that the multi-slit element 11 is inserted into or is made to withdraw from the optical path of the X rays by moving it toward the supporting bar member 7 or in a direction apart from the supporting bar member 7. The multi-slit element 11 is disposed in such a manner that the extended directions of the slits 111 are parallel to those of diffraction members 152 of a first diffraction grating 15 detailed later. Further, as shown in FIG. 2, since the farther the X rays irradiated from the X ray tube 8 depart from the X ray tube 8, the wider the irradiation area of the X rays becomes, if the multi-slit element 11 is disposed at a position being far apart from the X ray tube 8, the area of the multi-slit element 11 should be widened, and possibly causes a physical interference with the subject H. In order to avoid the above inconveniences, it is preferable that the multi-slit element 11 is disposed at such a position that is apart from the focal point of the X ray tube 8 with a distance in a range of around 1-10 cm. In this connection, hereinafter, precisely speaking in the present invention, the distance between the X ray tube 8 and another member represents the distance between the focal point of the X ray tube 8 and another member. A subject placing plate 12, on which the subject H is to be placed, is disposed at a position located below the X ray tube 8, in such a manner that the subject placing plate 12 is extended from the inner supporting shaft 5b of the supporting shaft 5 substantially in parallel to the floor surface. The subject placing plate 12 and the inner supporting shaft 5b are fixed neither to the image capturing main section 4 nor to the supporting bar member 7. Therefore, even if the image capturing main section 4 is driven to rotate clockwise or counterclockwise by the rotating action of the outer supporting cylinder 5a of the supporting shaft 5, the subject placing plate 12 does not rotate in conjunction with the rotating action of the outer supporting cylinder 5a. The subject placing plate 12 can also rotate around the inner supporting shaft 5b, etc., as needed, and further, a pressing plate 13 presses the subject H onto the subject placing plate 12 so as to fix the subject H thereon, as needed. The pressing plate 13 is supported by the subject placing plate 12 through a supporting member (not shown in the drawings). It is applicable that the pressing plate 13 is made to move either automatically or manually. As mentioned in the above, the subject placing plate 12 elevates and descends in conjunction with the up-and-down movements of the supporting base member 3 through the supporting shaft 5, for instance, so that the position of the subject placing plate 12 is adjusted at such a position that a patient can take an easy stance (natural posture) while putting his arm, serving as the subject H, on the subject placing plate 12. Further, a protector 14, which is extended in substantially a vertical direction, is mounted to the lower surface of the subject placing plate 12, so that the patient can take his position at the radiographing position without hitting his leg to a structure equipped under the subject placing plate 12 and without receiving X-ray exposure. In this connection, the pressing plate 13 and the protector 14 are not necessary indispensable structural elements, but it is applicable that the system can be configured without employing them. The first diffraction grating 15 is disposed at a central section of the supporting bar member 7, located below the subject placing plate 12, in such a manner that the first diffraction grating 15 is made to freely elevate and descend, and a second diffraction grating 16 is disposed at a lower section of the supporting bar member 7, in such a manner that the second diffraction grating 16 is made to freely elevate and descend. The first diffraction grating 15 and the second diffraction grating 16 are held so as to arrange them in parallel to each other. The structures of the first diffraction grating 15 and the second diffraction grating 16, and positional relationships between an X-ray detector 17, detailed later, and them will be detailed later on. As aforementioned, since the X rays irradiated from the X ray tube 8 are converted to the multi radiant sources by the multi-slit element 11, it is possible to regard the multi-slit element 11 as a radiant source. Further, it is necessary to appropriately adjust the distance between the first diffraction grating 15 and the radiant source. Accordingly, when the radiological image capturing apparatus 1 is used as the Talbot interferometer in which the first diffraction grating 15 is made to withdraw from the optical path of the X rays, the position adjusting device 9 makes the first diffraction grating 15 elevate and descend with respect to the supporting bar member 7 so as to adjust a distance L between the X ray tube 8 and the first diffraction grating 15. While, when the radiological image capturing apparatus 1 is used as the Talbot-Lau interferometer in which the first diffraction grating 15 is inserted into the optical path of the X rays, since the X rays irradiated from the X ray tube 8 are converted to the multi radiant sources by the multi-slit element 11, it is possible to regard the multi-slit element 11 as the radiant source, as aforementioned. Accordingly, the position adjusting device 9 makes the first diffraction grating 15 elevate and descend with respect to the supporting bar member 7, so as to adjust a distance L between the multi-slit element 11, serving as the radiant source, and the first diffraction grating 15. Further, the position adjusting device 9 makes the second diffraction grating 16 elevate and descend with respect to the supporting bar member 7, so as to adjust a distance Z1 between the first diffraction grating 15 and the second diffraction grating 16. In this connection, in the present embodiment, the position adjusting device 9 makes each of the first diffraction grating 15 and the second diffraction grating 16 elevate and descend independently from each other. Further, for instance as shown in FIG. 5, a first temperature sensor 15a and a second temperature sensor 16a are mounted on the first diffraction grating 15 and the second diffraction grating 16 and disposed at such positions that cannot be captured as X-ray images, respectively. In this connection, for instance, it is applicable that sheets, which are made of material having a good thermal conductivity and does not impede the X-ray radiographing operation, are adhered onto the first diffraction grating 15 and the second diffraction grating 16, respectively, so as to keep the temperature uniform within the surface of each of them. Further, for instance, it is also applicable that Peltier elements, which are capable of conducting heating and cooling operations by controlling the direction and amplitude of electric currents flowing through them, are installed into the first diffraction grating 15 and the second diffraction grating 16, respectively, so as to make it possible to conduct the heating and cooling operations of them. As shown in FIG. 2 and FIG. 3, a detector supporting plate 18 for supporting an X-ray detector 17 is supported at a lower section of the second diffraction grating 16, in such a manner that the X-ray detector 17 is made to freely elevate and descend with respect to the supporting bar member 7. Further, the position adjusting device 9 makes the detector supporting plate 18 elevate and descend independently from the first diffraction grating 15, etc., so as to adjust the position thereof. The X-ray detector 17 is supported on the detector supporting plate 18 so as to oppose to the X ray tube 8. Although the X-ray detector 17 and the second diffraction grating 16 are depicted in the schematic diagrams shown in FIG. 2, FIG. 3, etc., in such a manner that some distance Z2 exists between them, in order to indicate that the X-ray detector 17 and the second diffraction grating 16 are separate elements, in reality, it is preferable that the X-ray detector 17 and the second diffraction grating 16 are disposed in such a state that both of them are in contact with each other. This is because, the farther the X-ray detector 17 departs from the second diffraction grating 16, the more the Moiré fringes become blurred. In other words, both of them are arranged so as to make the distance Z2 equal to substantially zero. In this connection, it is also applicable that the X-ray detector 17 and the second diffraction grating 16 are integrally structured as a single element. Further, in order to prevent a part of the human body, residing below the X-ray detector 17, from the X-ray exposure caused by the X rays emitted by the X ray tube 8, various kinds of radiation shielding members (not shown in the drawings) are disposed at the lower side of X-ray detector 17 and installed into the detector supporting plate 18, etc. The X-ray detector 17 is constituted by a panel, a detector controlling section, etc. (not shown in the drawings), which are coupled to each other through a bus. Further, the X-ray detector 17 detects an amount of X rays penetrated through the subject H after emitted from the X ray tube 8, so as to output X-ray image data, representing the detected amount of X rays, to the image processing apparatus 30 through a network N. It is preferable that a detector that employs any one of a FPD (Flat Panel Detector), a CR (Computed Radiography) and CCD (Charge Coupled Device), each of which output the amount of X rays as the digital information for every pixel, is used as the X-ray detector 17. Among them, the FPD, which is superior to the others as the two dimensional image sensor, is specifically preferable for this purpose. The overall size of the panel is selected as needed. A distance Ltotal between the X-ray detector 17 and the X ray tube 8 or the multi-slit element 11, serving as the radiant source, is set at such a value that is equal to or greater than 0.5 m. Further, considering the fact that the radiological image capturing apparatus 1 is used in a room environment and accuracy, strength, etc. of the radiological image capturing apparatus 1, the upper limit of the distance Ltotal is set at around 2 m. A control device 20, indicated in the schematic diagram shown in FIG. 6, conducts various kinds of setting operations and controlling operations for controlling actions to be conducted in the radiological image capturing apparatus 1. The control device 20 is provided with a computer constituted by a CPU (Central Processing Unit), a ROM (Read Only Memory), a RAM (Random Access Memory), etc., which are coupled to each other through a bus. Although it is possible to install the control device 20 into the same room in which the radiological image capturing apparatus 1 is already installed, in the present embodiment, the control device 20 is constituted by employing a computer that is provided in the image processing apparatus 30 coupled to the radiological image capturing apparatus 1 through the network N. In other words, the control device 20 and the image processing apparatus 30 are constituted by employing the same computer. In this connection, it is also applicable that the control device 20 is constituted by employing another computer, which is equipped separately from the image processing apparatus 30 and which is coupled to the control device 20 through the network N. As shown in FIG. 6, the control device 20 is coupled to the X ray tube 8, the power source 10, the driving device 6, the position adjusting device 9, the first temperature sensor 15a and the second diffraction grating 16, which are described in the foregoing. Other than the above, the control device 20 is also coupled to a radiation amount detecting device 21 to detect an amount of irradiated X rays, an operating device 22 that is provided with an input device 22a and a display device 22b, etc. A controlling program for controlling various kinds of sections included in the radiological image capturing apparatus 1 and various kinds of processing programs are stored in a storage section of the control device 20, which includes the ROM, etc. Based on information inputted by the operator from the input device 22a, such as a keyboard, a mouse, a controller, etc., the control device 20 reads out the controlling program and the various kinds of processing programs from the storage section, so as to totally control operations to be conducted in the various kinds of sections included in the radiological image capturing apparatus 1, while making the display device 22b, such as a CRT display, a LCD (Liquid Crystal Display), etc., display contents of current controlling actions thereon. For instance, when the operator inputs information for selecting any one of the Talbot interferometer method or Talbot-Lau interferometer method as a method to be currently employed in the radiological image capturing apparatus 1, and other information for setting the tube voltage to be currently applied for the X ray tube 8 as aforementioned, from the input device 22a, the average energy of the X rays to be irradiated from the X ray tube 8 is determined, and further, the distance L between the X ray tube 8 and the first diffraction grating 15 or the other distance L between the multi-slit element 11 serving as the radiant source and the first diffraction grating 15, and the distance Z1 between the first diffraction grating 15 and the second diffraction grating 16 are also determined. Further, when the second diffraction grating 16 and the X-ray detector 17 are made to tightly contact with each other as aforementioned, assuming that the distance between the X ray tube 8 and the subject placing plate 12 is established as R1, and the other distance between the subject placing plate 12 and the X-ray detector 17 is established as R2, with respect to the schematic diagram shown in FIG. 2, the magnification factor of the subject H is determined by the following Equation, depending on the position of the subject placing plate 12.(magnification factor)=(R1+R2)/R1 Accordingly, in the present embodiment, when the operator inputs the selected method to be currently employed, the tube voltage of the X ray tube 8, the distance L, the distance Z1, etc., through the input device 22a, the control device 20 makes the position adjusting device 9 drive various kinds of sections, based on the inputted information, so as to conduct the operations for adjusting the positions of the X ray tube 8, the multi-slit element 11, the first diffraction grating 15, the second diffraction grating 16 and the X-ray detector 17. With respect to the multi-slit element 11, other than the position adjusting operation abovementioned, corresponding to the selected method currently established in the apparatus, the multi-slit element 11 is rotated around the axis of the supporting bar member 7 by the driving action of the position adjusting device 9 so as to insert it into the optical path of the X rays (in the case of the Talbot-Lau interferometer method), or to make it withdraw from the optical path (in the case of the Talbot interferometer method). Successively, while maintaining the positional relationships of them, the positional adjusting operations are conducted by making the subject placing plate 12 elevate and descend in conjunction with the up and down movements of the supporting base member 3, so that the subject person can take a posture of being hardly fatigued. In this connection, since the positional relationships should be adjusted so that the subject placing plate 12 and the first diffraction grating 15 are not in contact with each other, certain limitations are applied to the distances R1 and R2 aforementioned, and accordingly, a settable range of the magnification factor (=(R1+R2)/R1) is also limited. Accordingly, it is applicable that the radiological image capturing apparatus 1 is so constituted that the settable range of the magnification factor is displayed on the display device 22b at the time when the selected method to be currently employed, the tube voltage of the X ray tube 8, the distance L, the distance Z1 and the magnification factor are inputted. Further, it is also applicable that the radiological image capturing apparatus 1 is so constituted that an LUT (Look Up Table), in which the distances L and Z1 being appropriate for the apparatus method and the tube voltage of the X ray tube 8 to be employed are stored, is provided in advance, so as to automatically set the distance L and the distance Z1 at the time when the selected method to be currently employed and the tube voltage of the X ray tube 8 are inputted. In this case, when the selected method to be currently employed and the tube voltage of the X ray tube 8 are inputted, the operations for adjusting the positions of the X ray tube 8, the first diffraction grating 15, the second diffraction grating 16 and the X-ray detector 17 are automatically implemented, and further, when the magnification factor is inputted, the operations for adjusting the positional relationships between the subject placing plate 12 and them is implemented, corresponding to the above-inputted magnification factor. Still further, although the radiological image capturing apparatus 1 embodied in the present invention is so constituted that, when the Talbot-Lau interferometer method is employed, the distance between the multi-slit element 11 disposed below the X ray tube 8 and the X ray tube 8 is set at a predetermined distance value, it is also applicable that the abovementioned distance is set at another value by inputting it at the same time when inputting the apparatus method and the tube voltage of the X ray tube 8 to be currently established in the apparatus, and/or an LUT for establishing a distance value, being optimum for the apparatus method and the tube voltage of the X ray tube 8 to be currently employed, is provided in advance. Still further, as aforementioned, when any one of the Talbot interferometer method or the Talbot-Lau interferometer method is inputted as the selected apparatus method and the tube voltage of the X ray tube 8 is established, the control device 20 conducts the operation for changing the diameter of focal point of the X ray tube 8 corresponding to the selected apparatus method. As abovementioned, the control device 20 activates the driving device 6 to rotate the supporting shaft 5 clockwise or counterclockwise, as shown in FIG. 3, so that the image capturing main section 4 is rotated around the subject H so as to adjust the radial-ray irradiation angle. Further, when the radiological image capturing apparatus 1 is activated, the control device 20 irradiates the X rays emitted from the X ray tube 8 onto the subject H according to the electric power supplied from the power source 10 (in the case of the Talbot-Lau interferometer method, the X rays emitted from the multi radiant sources generated by the multi-slit element 11 are irradiated onto the subject H). Then, at the time when an amount of X rays detected by the radiation amount detecting device 21 reaches the predetermined amount of X rays, established in advance, the control device 20 stops the electric power currently supplied to the X ray tube 8 from the power source 10 so as to deactivates the X ray irradiating action. In this connection, the conditions for irradiating the X rays are established as needed by considering factors other than the amount of X rays detected by the radiation amount detecting device 21, namely, by taking a kind of the X-ray detector 17, etc. into account. According to the present embodiment, since the control device 20 activates the driving device 6 to rotate the supporting shaft 5, so that the image capturing main section 4 is rotated around the subject H, so as to rotate the X ray tube 8, the first diffraction grating 15, the second diffraction grating 16 and the X-ray detector 17 (in the case of the Talbot-Lau interferometer method, the multi-slit element 11 is further added) around the subject H, it is possible to continuously capture X ray images by irradiating the X rays onto the subject H from plural directions. In this connection, the rotation amount of the image capturing main section 4 and an image capturing timing (corresponding to a rotated angle, at every which the X ray images are captured) are established by inputting them from the input device 22a. Further, the control device 20 determines whether or not temperatures of the first diffraction grating 15 and the second diffraction grating 16, which are measured by the first temperature sensor 15a and the second temperature sensor 16a, respectively, are equal to or greater than the predetermined temperature established in advance. In the present embodiment, when at least one of the temperatures of the first diffraction grating 15 and the second diffraction grating 16 becomes equal to or greater than the predetermined temperature established in advance, a warning operation is conducted. The warning operation is achieved in such a manner that the control device 20 controls the display device 22b to display a kind of visual warning message thereon or to issue a kind of audible warning notification therefrom. In this connection, in the case that the Peltier elements, which are capable of conducting heating and cooling operations by controlling the directions and amplitudes of electric currents flowing through them, are installed into the first diffraction grating 15 and the second diffraction grating 16, respectively, as aforementioned, when the temperatures of the first diffraction grating 15 and the second diffraction grating 16, which are measured by the first temperature sensor 15a and the second temperature sensor 16a, are increase or decrease, it is possible to activate the Peltier elements so as to control the temperatures of the first diffraction grating 15 and the second diffraction grating 16 to be kept within a predetermined temperature range. Further, as detailed later, the radiological image capturing apparatus 1 embodied in the present invention is so constituted that the control device 20 can determine whether or not a distortion due to the temperature change or another distortion due to the change over time has generated on the first diffraction grating 15 or the second diffraction grating 16, based on an image of the Moiré fringes detected in such a state that the subject H is not placed on the subject placing plate 12 (refer to Moiré fringes M indicated in the schematic diagram shown in FIG. 7, detailed later). Concretely speaking, at the time stage when the radiological image capturing apparatus 1 is installed into a certain room after shipped from a factory, or when the first diffraction grating 15 and/or the second diffraction grating 16 are/is replaced with a new one, the control device 20 stores an image of the Moiré fringes, which is captured in the state that the subject H is not placed on the subject placing plate 12 before the apparatus is put in its actual operation, into the storage section including a RAM, etc. After that, at the next time stage when a predetermined condition established in advance, such as a condition that the operating time of the radiological image capturing apparatus 1, established in advance, has elapsed, a condition that a number of X ray irradiation times has reached to a predetermined number of times, etc., is fulfilled, the control device 20 newly conducts the operation for capturing an image of the Moiré fringes in the state that the subject H is not placed on the subject placing plate 12. Otherwise, it is also applicable that the apparatus is so constituted that the control device 20 periodically conducts the operation for capturing an image of the Moiré fringes. Successively, the control device 20 reads out the image of the Moiré fringes, captured and stored before the apparatus is put in its actual operation, from the storage section, to compare it with the other image of the Moiré fringes currently captured at this time. As a result of the abovementioned comparison, when the currently-captured image of the Moiré fringes fulfills at least one of the conditions that: an interval of the Moiré fringes in the currently-captured image is expanded or reduced at a value equal to or greater than a predetermined value, compared to that in the image of the Moiré fringes captured and stored before the apparatus is put in its actual operation; a part of or all of the Moiré fringes are curved; a difference between a maximum part and a minimum part of detected amount of the irradiated X rays among the Moiré fringes is expanded or reduced at a value equal to or greater than a predetermined value; etc., the control device 20 determines that a distortion (deformation) has generated in the diffraction member(s) (grating) of the first diffraction grating 15 and/or the second diffraction grating 16. In this connection, it is applicable that the abovementioned operation is conducted in any one of the Talbot interferometer method and the Talbot-Lau interferometer method, or in both methods. When determining that a distortion (deformation) has generated in the diffraction member(s) of the first diffraction grating 15 and/or the second diffraction grating 16 as abovementioned, the control device 20 controls the display device 22b to display a kind of visual warning message thereon or to issue a kind of audible warning notification therefrom. Further, in the present embodiment, the apparatus is so constituted that the control device 20 detects an abnormal shadow candidate from the captured X-ray image. Further, the apparatus is so constituted that, when detecting the abnormal shadow candidate, the control device 20 changes the apparatus method from the Talbot-Lau interferometer method to the Talbot interferometer so as to capture the abnormal shadow candidate being sharper than ever. The operation for detecting the abnormal shadow candidate from the captured X-ray image can be conducted by employing, for instance, the technology of the medical image diagnosis assisting system, set forth in Tokkai 2005-102936, which has been previously submitted by the applicant of the present invention. In this system, an image analysis processing is applied to the medical image, such as an X-ray image, etc., so as to extract its featuring amount, and then, based on the extracted featuring amount, an abnormal shadow candidate is detected from the image concerned. In this connection, it is applicable that the system is so constituted that the abovementioned apparatus that detects the abnormal shadow candidate from the X-ray image, captured by the radiological image capturing apparatus 1, is installed as the diagnosis assisting apparatus (not shown in the drawings) separately from the radiological image capturing apparatus 1, and is coupled to the radiological image capturing apparatus 1, etc. through the network N, so as to provide it within the radiological image capturing system 100. For instance, it is also applicable that the system is so constituted that, when the diagnosis assisting apparatus detects an abnormal shadow candidate and transmits information in regard to the detected abnormal shadow candidate, the control device 20 of the radiological image capturing apparatus 1 changes the method to be employed in the radiological image capturing apparatus 1 from the Talbot-Lau interferometer method to the Talbot interferometer method, based on the information sent from the diagnosis assisting apparatus. As shown in FIG. 1, the image processing apparatus 30 and the image outputting apparatus 50 are coupled to the radiological image capturing apparatus 1 through the network N. The image outputting apparatus 50 includes: a display device, such as a CRT display, a LCD (Liquid Crystal Display), etc.; a developing device for outputting an image onto a film; etc. Receiving X-ray image data for every captured image sent from the X-ray detector 17 of the radiological image capturing apparatus 1 through the network N, the image processing apparatus 30 temporarily stores the received X-ray image data into a storage section (not shown in the drawings). In this connection, at least one of an HDD (Hard Disc Drive) serving as a high-speed accessible mass memory, an HDD Array, such as a RAID (Redundant Array of Independent Disks), etc., a silicone disc, etc. can be employed as the storage section. Further, the image processing apparatus 30 makes the radiological image capturing apparatus 1 capture an image of the Moiré fringes, and then, transmit X-ray image data of the image, so as to store the X-ray image data into storage section. This X-ray image data is temporarily established as the reference X-ray image data. After that, when the radiological image capturing apparatus 1 commences the operation for capturing an X-ray image of the subject H, the image processing apparatus 30 corrects the transmitted X-ray image data representing the image that is captured by the radiological image capturing apparatus 1 in a state that the subject H is present, based on the reference X-ray image data. The correcting operation is conducted with respect to, for instance, a positional deviation on the image, sensitivity unevenness (namely, non-uniformity in the signal values detected by the detector), etc. Concretely speaking, when it is recognized in advance from the reference X-ray image data that a positional deviation is generated in a fixed pixel area on the image, it is possible to correct the positional deviation by conducting such the correcting operation that turns the fixed pixel area, existing in the image represented by the transmitted X-ray image data, back to the original position by an amount of the positional deviation. Further, by dividing the X-ray image data, the positional deviation of which has been corrected, by the reference X-ray image data, for every pixel, it is possible to acquire an X-ray image having no sensitivity non-uniformity to be caused by the existence of the diffraction grating. The image processing apparatus 30 also stores the above-processed X-ray image data into the storage section. Still further, it is possible for the image processing apparatus 30 not only to convert the X-ray image (the image of the Moiré fringes) detected by the X-ray detector 17 to a distribution image of angles at which the X rays are curved by the refraction effect caused by the subject H (phase shift differential image), but also to obtain such an image that represents the phase sift itself, acquired by integrating the phase shift differential image. The well-known methods, such as the method set forth in the International Publication 2004/058070, etc., are employed for the conversion processing and the image acquisition processing, both abovementioned. Yet further, in the present embodiment, when receiving plural X-ray image data sets, which represent a plurality of X-ray images continuously captured by changing the direction for radiographing the subject H and are sent from the radiological image capturing apparatus 1, the image processing apparatus 30 creates a three dimensional image of the subject H, based on the above-received plural X-ray image data sets. The image outputting apparatus 50 displays the created three-dimensional image on the LCD, etc., or outputs it onto a film, etc., so as to output the three dimensional image created in the above. In this connection, a certain well-known method can be employed as the method for creating the three dimensional image from the plurality of two-dimensional images acquired by capturing the subject H from the various directions. In this connection, it is also possible to further apply another kind of processing to the plurality of two-dimensional images and/or the three dimensional image, which are/is acquired through the abovementioned processes. For instance, it becomes possible to acquire such the image, etc., that is displayed on the screen or outputted on the film, in such a manner that the brightness of the image, in which the concerned cartilage section is represented with the deep color in contrast to the pale color of the background, is reversed, or in such a manner that the considerably changed portion, compared to the standard model of the cartilage section, is colored, etc. Further, when a manifestation of the rheumatic disease emerges on the finger, it becomes possible to observe the diseased part as if it were a moving image, by acquiring a plurality of three-dimensional images in which the bending angles of the finger joints are varied in various kinds of directions. In this connection, it is applicable that the abovementioned operations are implemented in any one of the Talbot interferometer method and the Talbot-Lau interferometer method, or in both of them. Next, the Talbot-Lau interferometer to be configured in the radiological image capturing apparatus 1, embodied in the present invention, will be detailed in the following. Further, the functions of the radiological image capturing apparatus 1 will be detailed in the following, in conjunction with the explanations of the configurations of the multi-slit element 11, the first diffraction grating 15 and the second diffraction grating 16, and the explanations of the positional relationships between the X-ray detector 17 and them. As shown in FIG. 7 and FIG. 8, in the present embodiment, the X rays irradiated from the X ray tube 8 penetrate through the multi-slit element 11 in the case of the schematic diagram shown in FIG. 8, and successively penetrate through the subject H, and then, penetrate through the first diffraction grating 15 and the second diffraction grating 16, and finally, are incident onto the X-ray detector 17. As shown in FIG. 7, the Talbot interferometer is constituted by the X ray tube 8, the first diffraction grating 15 and the second diffraction grating 16, while, as shown in FIG. 8, the Talbot-Lau interferometer is constituted by the X ray tube 8, the multi-slit element 11, the first diffraction grating 15 and the second diffraction grating 16. FIG. 9 shows a cross sectional schematic diagram at the I-I cross section indicated in the schematic diagrams shown in FIG. 7 and FIG. 8. As shown in FIG. 7 through FIG. 9, the first diffraction grating 15 is provided with a substrate 151 and a plurality of the diffraction members 152 that are arranged on the substrate 151, so as to yield a Talbot effect, detailed later, by diffracting the X rays that penetrate through the subject placing plate 12 and the subject H, held by the subject placing plate 12, and are irradiated thereon. The substrate 151 is made of, for instance, a glass material or the like. In this connection, a surface of the substrate 151, on which the diffraction members 152 are arranged, is referred to as a diffraction grating surface 153. Each of the diffraction members 152 is such a linear member that is extended in a direction orthogonal to the irradiation direction of the X rays irradiated from the X ray tube 8, namely, for instance, that is extended in a up-down direction of the schematic diagrams shown in FIG. 7 and FIG. 8. The thicknesses of the diffraction members 152 are substantially the same, for instance, each of them is formed in a range of 10-50 μm. Further, as shown in FIG. 9, an interval distance d1, being one of relative distances between the plural diffraction members 152, is set at a fixed value, and the relative distances between the diffraction members 152 are substantially the same. The interval distance d1 is formed at a value in a range of around 3-10 μm. The interval distance d1 is also referred to as a grating period or a grating interval. In this connection, both the range of the interval distance d1 in the plural diffraction members 152 and the other range of the width of each of the diffraction members 152 are not limited specifically. It is applicable that the diffraction members 152 are formed in such a manner that the interval distance between the diffraction members and the width of each of the diffraction members are either the same as each other or different from each other. It is preferable that a material to be employed for structuring the diffraction members 152 is superior in the X rays absorbing property, and, for instance, a metallic material, such as a gold, a silver, a platinum, etc., can be employed for this purpose. The diffraction members 152 are formed on the substrate 151, for instance, by plating or vapor-depositing the abovementioned metal thereon. The diffraction members 152 is such a member that changes the phase velocity of the X rays irradiated onto the diffraction members 152, and it is preferable that the diffraction members 152 is such a member that structures, so called, the phase-type diffraction grating, which yields a phase modulation at an angle in a range of about 80°-100°, preferably at 90°. The X rays are not necessary a single color, but it is applicable that the X rays have such an energy width (namely, a wavelength spectral width) that fulfils the abovementioned conditions. FIG. 10 shows a cross sectional schematic diagram at the II-II cross section indicated in the schematic diagrams shown in FIG. 7 and FIG. 8. As shown in FIG. 7, FIG. 8 and FIG. 10, the second diffraction grating 16 is provided with a substrate 161 and a plurality of diffraction members 162 in the same manner as those of the first diffraction grating 15. In this connection, a surface of the substrate 161, on which the diffraction members 162 are arranged, is referred to as a diffraction grating surface 163. Wherein, an interval distance d2, being one of relative distances between the plural diffraction members 162, is set at such a value that the ratio of the distance (L+Z1) from the X ray tube 8 to the second diffraction grating 16 and the interval distance d2 is substantially equal to the other ratio of the distance L from the X ray tube 8 to the first diffraction grating 15 and the interval distance d1. In this connection, for instance, it is also possible to set the interval distance d2, being one of relative distances between the plural diffraction members 162 of the second diffraction grating 16, at such a value that is substantially the same as the interval distance d1, being one of relative distances between the plural diffraction members 152 of the first diffraction grating 15. Further, the width of each of the diffraction members 162 of the second diffraction grating 16 is substantially the same as the width of each of the diffraction members 152 of the first diffraction grating 15. As detailed later, the second diffraction grating 16 is disposed in such a state that the extended direction of the diffraction members 162 is rotated relative to the other extended direction of diffraction members 152 of the first diffraction grating 15 by a minute angle θ, so as to form an image contrast by diffracting the X rays previously diffracted by the first diffraction grating 15. Although it is desirable that the second diffraction grating 16 is an amplitude-type diffraction grating in which the diffraction members 162 are made to be thicker than ever, it is also possible to employ such a structure that is similar to that of the first diffraction grating 15. Next, the structure of the multi-slit element 11 will be detailed in the following. As shown in FIG. 11, when the radiological image capturing apparatus 1 is used as the Talbot-Lau interferometer method, the multi-slit element 11 and the first diffraction grating 15 are apart from each other by the distance L. Further, for instance as detailed later, an X ray passing through a slit 111a, serving as one of the multi-slits of the multi-slit element 11, forms self-images of a diffraction member 152a and a diffraction member 152b of the first diffraction grating 15 on the second diffraction grating 16, which is disposed at the position being apart from the first diffraction grating 15 by the distance Z1, (namely, on the X-ray detector 17 that is closely adjacent to the second diffraction grating 16). Further, another X ray passing through a slit 111b that is located at position adjacent to the slit 111a also forms self-images of a diffraction member 152a and a diffraction member 152b of the first diffraction grating 15 on the second diffraction grating 16, respectively. In other words, each of the X rays passing through each of the slits 111 of the multi-slit element 11 forms each of self-images of the diffraction members 152 on the second diffraction grating 16, resulting in a striped pattern of the self-images. On that occasion, unless a slit interval distance d0 between the slits 111 of the multi-slit element 11 is appropriate, the self-images in the striped pattern, which are formed by the X rays passing through the slit 111a and the slit 111b of the multi-slit element 11, counteract with each other. However, if the slit interval distance d0 is adjusted, so as to make the self-image of the diffraction member 152a, formed by the X ray passing through the slit 111a, and the other self-image of the diffraction member 152b, formed by the X ray passing through the slit 111b, overlap with each other at a position Y on the second diffraction grating 16, the self-image and the other self-image in the striped pattern can be superimposed with each other, resulting in achievement of an in-focus state. In the above case, the slit interval distance d0 between the slits 111 of the multi-slit element 11, the interval distance (grating period) d1, being one of relative distances between the plural diffraction members 152 of the first diffraction grating 15, the distance L between the multi-slit element 11 and the first diffraction grating 15 and the distance Z1 from the first diffraction grating 15 to the second diffraction grating 16, fulfill the Equation indicated as follow.d0:d1=(L+Z1):Z1 (1) Deriving from the Equation (1), the slit interval distance d0 can be represented by the Equation (2) indicated as follow. d 0 = L + Z 1 Z 1 d 1 ( 2 ) Further, referring to FIG. 11, although there has been considered such the case that the X rays, passing through the slit 111a and the slit 111b, further pass through the portions of the diffraction member 152a and the diffraction member 152b, which are adjacent to each other on the first diffraction grating 15, for instance, even if the X rays pass through the portions of the diffraction member 152a and the diffraction member 152c or the diffraction member 152d, each of which resides at a position being apart from the diffraction member 152a by an integral multiple of the grating period d1, namely, even when the Equation (3), which is indicated as follow and in which “d1” in the Equation (2) is substituted by “pd1” acquired by multiplying “d1” by “p”, is fulfilled, the self-images in the striped pattern formed on the first diffraction grating 15 are just superimposed with each other, resulting in achievement of the in-focus state. d 0 = L + Z 1 Z 1 pd 1 ( 3 ) Still further, since the Equation of L+Z1+Z2=Ltotal has been established as aforementioned, and the distance Z2 between the second diffraction grating 16 and the X-ray detector 17 are approximately zero, the Equation (3) can be also expressed by the Equation (4) indicated as follow. d 0 = Ltotal Z 1 pd 1 ( 4 ) In other words, if the slits 111 of the multi-slit element 11 are appropriately formed in such a manner that the slit interval distance d0 of the slits 111 fulfills the Equation (3) and the Equation (4), the X rays respectively passing through the slits 111 of the multi-slit element 11 effectively form the self-images of the first diffraction grating 15 on the second diffraction grating 16 so as to make the self-images overlap each other, and as a result, it becomes possible to acquire the self-images being in-focus. Next, when the radiological image capturing apparatus 1 is used in either the Talbot interferometer method or Talbot-Lau interferometer method, the conditions that the X ray tube 8, the multi-slit element 11, the first diffraction grating 15 and the second diffraction grating 16 constitute the interferometer, will be detailed in the following. Initially, when the radiological image capturing apparatus 1 is used in the Talbot interferometer method, the conditions that the X ray tube 8, the first diffraction grating 15 and the second diffraction grating 16 constitute the interferometer will be detailed in the following. On the premise that the first diffraction grating 15 is the phase-type diffraction grating, the distance Z1 between the first diffraction grating 15 and the second diffraction grating 16 should fulfill the condition indicated as follow. In this connection, “m” represents an integer number and “d1” represents the interval distance, being one of relative distances between the plural diffraction members 152 of the first diffraction grating 15, as aforementioned. Z 1 = ( m + 1 2 ) d 1 2 λ ( 5 ) Explaining the Talbot effect while referring to the schematic diagram shown in FIG. 12, in the case that the first diffraction grating 15 is the phase-type diffraction grating, when the plane wave of the X ray passes through the first diffraction grating 15, the Talbot effect is to form the self-image of the diffraction grating at the distance given by the Equation (5). In the state that the subject H is absence, the self-image of the first diffraction grating 15, namely, the image of diffraction members 152 in which the grating period for every interval distance d1 is slightly expanded, emerges at such a position that is apart from the first diffraction grating 15 by the distance Z1 given by the Equation (5). In this connection, at a position other that the position of the distance Z1 given by the Equation (5), the self-image cannot be observed or may be out of focus. However, in the vicinity of the position of the distance Z1 given by the Equation (5), the relatively in-focus state of the self-image is maintained. Accordingly, the distance Z1 defined by Equation (5) includes allowable distances in the vicinity of the distance Z1. Further, when setting the actual distance Z1, some allowance for the distance Z1 given by the Equation (5) can be introduced into a distance to be actually set. Then, when the second diffraction grating 16 is positioned at a position of the distance Z1 in such a state that the extended direction of the diffraction members 162 is rotated relative to the other extended direction of diffraction members 152 of the first diffraction grating 15 by a minute angle θ, Moiré fringes emerge, and the X-ray detector 17 detects a Moiré stripe image M, which is formed by projecting the Moiré fringes, as shown in FIG. 7. In this case, an interval distance between the Moiré stripes of the Moiré stripe image M, generated in the above, is given by d1/θ, from the interval distance d1 of the diffraction members 152 and the minute angle θ. On the other hand, when the subject H exists between the X ray tube 8 and the first diffraction grating 15, the phase of the X rays emitted from the X ray tube 8 would shift in mid course of passing through the subject H. This phase sift causes a distortion of the wave front of the X rays being incident into the first diffraction grating 15. Accordingly, the self-image of the first diffraction grating 15 is deformed, depending on the distortion of the wave front. Successively, when the X rays diffracted by the first diffraction grating 15 passes through the second diffraction grating 16, the Moiré stripe image M is distorted according to the distortion of the wave front of the X rays, namely, according to the shape of the subject H. On that occasion, since the X rays penetrate through the inside section of the subject H, the X rays would be distorted by the shape of the inside section, and accordingly, those distortions will be projected into the Moiré stripe image M. On that occasion, actually, the self-image of the first diffraction grating 15 is also reflected by the distortion caused by the subject H, and accordingly, at the position of the distance Z1 given by the Equation (5), it emerges such a state that the shape of the subject H and the shape of its inner section are reflected into the diffraction stripe of the diffraction members 152 in which the grating period for every interval distance d1 is slightly expanded. However, it has been virtually impossible for the normal-type X-ray detector 17 to detect the diffraction stripe above-mentioned with its resolution capability. Accordingly, since it is also impossible to detect the distortion caused by the subject H, it has been difficult to obtain the X ray image of the subject H as it is. However, if the apparatus is so constituted that the second diffraction grating 16 is rotated relative to first diffraction grating 15 by a minute angle θ so as to form such a Moiré stripe image in which the interval distance between the stripes is far greater than the grating period, it becomes possible for the normal-type X-ray detector 17 to detect the diffraction stripe abovementioned even with its resolution capability. Further, by employing the normal-type X-ray detector 17 for detecting the Moiré stripe image M distorted according to the shape of the subject H and the shape of its inner section, it becomes possible to obtain the X ray image of the subject H, into which the shape of the subject H and the shape of its inner section are projected. In the radiological image capturing apparatus 1 employing the Talbot interferometer, embodied in the present invention, as described in the foregoing, in order to heighten the coherence property of X rays emitted from the X ray tube 8 and having an average energy in a range of 15-60 keV, as aforementioned, at the time when the concerned X rays are incident into the first diffraction grating 15, it is necessary to set the distance L between the X ray tube 8 and the first diffraction grating 15 at a value equal to or greater that a certain fixed distance. As aforementioned, when the radiological image capturing apparatus 1 is used as the Talbot interferometer method, a focal point diameter “a” is set at a value equal to or greater that 1 μm. When the focal point diameter “a” is set at 1 μm as its minimum value and the average energy of the X rays is set at 60 keV as its maximum value, it is necessary to set the distance L between the X ray tube 8 and the first diffraction grating 15 at a value equal to or greater that 0.5 m. However, since the coherency (coherency distance) is in proportion to the distance L while in inverse proportion to the average energy of the X rays and the focal point diameter, in the case that the coherency is acquired at 60 keV of the X ray average energy, for instance, the distance L between the X ray tube 8 and the first diffraction grating 15 can be set at a value equal to or greater than 0.125 m (12.5 cm) as far as the average energy of the X rays is 15 keV, or, even if the focal point diameter “a” of the X ray tube 8 is widened up to 4 μm, the equivalent degree of the coherency can be obtained. Further, although the distance Z1 between the first diffraction grating 15 and the second diffraction grating 16 is given by the Equation (5) aforementioned, as being recognizable from the fact that the Equation (5) includes the wavelength λ, the distance Z1 depends on the average energy of the X rays. Accordingly, as aforementioned, when the interval distance d1, being one of relative distances between the plural diffraction members 152 of first diffraction grating 15, is set at around 3 μm, which is technically formable value, and the average energy of the X rays to be irradiated is set at a value in a range of 15-60 keV, it is necessary to set the distance Z1 at a value equal to or greater than 0.05 m. In this connection, the lower limit of the range of a value, which is settable as the distance Ltotal from X ray tube 8 to the X-ray detector 17, is specified by the limitations for the distance L and the distance Z1 (the distance Z2 from second diffraction grating 16 to the X-ray detector 17 is zero). Further, although its upper limit is not limited to a specific value, considering the fact that the radiological image capturing apparatus 1, embodied in the present invention, would be virtually used in the room environment, the upper limit may be set at around 2 meters. Next, when the radiological image capturing apparatus 1 is used in the Talbot-Lau interferometer method, the conditions that the X ray tube 8, the multi-slit element 11, the first diffraction grating 15 and the second diffraction grating 16 constitute the interferometer will be detailed in the following. Even in this case, the conditions are principally similar to those of the Talbot interferometer method abovementioned, the distance Z1 between the first diffraction grating 15 and the second diffraction grating 16 is set at such a value that fulfills the Equation (5) aforementioned. Further, as shown in FIG. 13, each of the X rays passing through each of the slits 111 of the multi-slit element 11 yields the Talbot effect aforementioned. Then, each of the X rays passing through each of the slits 111 forms the self-image of the first diffraction grating 15 at the position being apart from the first diffraction grating 15 by the distance Z1. On that occasion, if the slit interval distance d0 is structured so as to fulfill the Equation (3) and the Equation (4), the self-images just overlap with each other at the position being apart from the first diffraction grating 15 by the distance Z1, resulting in a just in-focus state. Accordingly, when the second diffraction grating 16 is positioned at a position of the distance Z1 in such a state that the extended direction of the diffraction members 162 is rotated relative to the other extended direction of diffraction members 152 of the first diffraction grating 15 by a minute angle θ, Moiré fringes emerge, and the X-ray detector 17 detects the Moiré stripe image M, which is formed by projecting the Moiré fringes, as shown in FIG. 8. On the other hand, when the subject H exists between the multi-slit element 11 and the first diffraction grating 15, the phase of each of the X rays, emitted from the X ray tube 8 and passing through the multi-slit element 11, would shift in mid course of passing through the subject H. This phase sift causes a distortion of the wave front of each of the X rays being incident into the first diffraction grating 15. Accordingly, the self-image of the first diffraction grating 15 is deformed, depending on the distortion of the wave front. Successively, when each of the X rays diffracted by the first diffraction grating 15 passes through the second diffraction grating 16, the Moiré stripe image M is distorted according to the distortion of the wave front of each of the X rays, namely, according to the shape of the subject H. On that occasion, since each of the X rays penetrates through the inside section of the subject H, each of the X rays would be distorted by the shape of the inside section, and accordingly, those distortions will be projected into the Moiré stripe image M. As described in the above, by employing the normal-type X-ray detector 17 for detecting the Moiré stripe image M distorted according to the shape of the subject H and the shape of its inner section, it becomes possible to obtain the X ray image of the subject H, into which the shape of the subject H and the shape of its inner section are projected. In the case that the Talbot-Lau interferometer method as abovementioned is employed, when a width of aperture of each of the slits 111 of the multi-slit element 11, corresponding to the focal point diameter of the X ray tube 8, is set at 1 μm as its minimum value and the average energy of the X rays is set at 60 keV as its maximum value, it is necessary to set the distance L between the multi-slit element 11 and the first diffraction grating 15 at a value equal to or greater that 0.5 m. However, since the coherency (coherency distance) is in proportion to the distance L while in inverse proportion to the average energy of the X rays and the width of aperture of each of the slits 111, in the case that the coherency is acquired at 60 keV of the X ray average energy, for instance, the distance L between the multi-slit element 11 and the first diffraction grating 15 can be set at a value equal to or greater than 0.125 m (12.5 cm) as far as the average energy of the X rays is 15 keV, or, even if the width of aperture of each of the slits 111 is widened up to 4 μm, the equivalent degree of the coherency can be obtained. As well as in the case of the Talbot interferometer method, it is necessary in the Talbot-Lau interferometer method to set the distance Z1 between first diffraction grating 15 and the second diffraction grating 16 at a value equal to or greater than 0.05 m. Further, the slit interval distance d0 of the multi-slit element 11 has the relationship, represented by the aforementioned Equation (2), with respect to the interval distance (grating period) d1, being one of relative distances between the plural diffraction members 152 of the first diffraction grating 15, the distance L between the multi-slit element 11 and the first diffraction grating 15, and the distance Z1 between first diffraction grating 15 and the second diffraction grating 16. Accordingly, when the abovementioned limitations are applied to the interval distance (grating period) d1, the distance L, the distance Z1 and further, the slit width, the slit interval distance d0 is set at a value equal to or greater than 2 μm as its setting range. As described in the foregoing, various kinds of setting conditions are different from each other between in the case that the radiological image capturing apparatus 1 is used as the Talbot interferometer method and in the other case as the Talbot-Lau interferometer method. Accordingly, for instance as aforementioned, when detecting an abnormal shadow candidate from the captured X ray image, or when receiving the information in regard to an abnormal shadow candidate detected and transmitted by the diagnosis assistance apparatus of the radiological image capturing system 100, the control device 20 changes the Talbot-Lau interferometer method, shown in FIG. 2 and FIG. 3, to the Talbot interferometer method, in order to capture the abnormal shadow candidate as a farther sharper and clearer image. In the above case, the control device 20 makes the multi-slit element 11 rotate around the axis of the supporting bar member 7 so as to make it withdraw from the optical path of the X rays, and at the same time, changes the angle of the target of the X ray tube so as to change the focal point diameter of the X ray tube 8. Further, in the Talbot-Lau interferometer method, since the objects to be adjusted are somewhat changed, for instance, such that the distance L, which has been adjusted as the distance between the multi-slit element 11 and the first diffraction grating 15 in the Talbot-Lau interferometer method, is adjusted as the distance between the X ray tube 8 and the first diffraction grating 15 after making the multi-slit element 11 withdraw from the optical path of the X rays in the Talbot interferometer method, etc., the positional adjustments of the X ray tube 8, the first diffraction grating 15, the second diffraction grating 16 and the X-ray detector 17 are arbitrarily conducted as needed. In the case that the Talbot interferometer method is changed to the Talbot-Lau interferometer method, the control device 20 conducts operations being reverse to the above. As aforementioned, when the radiological image capturing apparatus 1, embodied in the present invention, is employed for the medical use, the apparatus can merely irradiate the X rays having an average energy in a relatively narrow range of 15-60 keV. However, even in such the case, by arranging the second diffraction grating 16 and the X-ray detector 17 so that both of them contact each other, and by specifying the distance L between multi-slit element 11 and the first diffraction grating 15, the distance Z1 between first diffraction grating 15 and the second diffraction grating 16, and the slit interval distance d0 of the multi-slit element 11 as aforementioned, it becomes possible to make the apparatus sufficiently bring out the Talbot effect so as to accurately detect the shapes of the subject H and its inner section in the Moiré stripe image. Further, when the average energy of the X rays to be irradiated, is smaller than 15 keV, since almost of all part of the irradiated X rays are absorbed into the subject, a dose of X ray exposure for the subject becomes extremely great, and accordingly, such the setting is not suitable for clinical use. However, by setting the average energy of the X rays at a value equal to or greater than 15 keV, it becomes possible not only to avoid such the problem as mentioned in the above, but also to obtain such the X ray image that has no blur caused by the movement of the human body, serving as the subject H, since the operation for irradiating the X rays can be completed within several per second or 2-3 seconds at longest for every time of the single X-ray radiographing operation. Further, by setting the average energy of the X rays to be irradiated at a value equal to or smaller than 60 keV, it becomes possible to acquire such the X-ray radiation image that has sufficient contrasts so as to clearly represent bones, soft tissue sections, etc., which constitute the human body. As a result, even for the tissue sections, such as the cartilage tissue of the human body, etc., from which the normal-type X-ray radiographing apparatus hardly captures a clear X ray image, it becomes possible to acquire the good X ray image in which the contrast of peripheral sections of the subject is emphasized by employing the Talbot-Lau interferometer method, and therefore, it becomes possible to effectively use the clearly contrasted X ray image acquired in above for the diagnosis purpose or the like. Further, on that occasion, by making the multi-slit element having a plurality of slits insert into or withdraw from the optical path of the X rays irradiated from the X ray tube, it becomes possible to make the Talbot interferometer method and the Talbot-Lau interferometer method switchable between them. In addition, by appropriately setting the distance between the X ray source or the multi-slit element and the X-ray detector, the other distance between the X ray source or the multi-slit element and the first diffraction grating, the focal point diameter of the X ray tube and the slit interval distance of the multi-slit element, corresponding to each of the abovementioned methods, it becomes possible to obtain a sufficiently clear X ray image within a short X rays irradiation time, while taking advantage of good points possessed by the corresponding one of methods above-mentioned. For this purpose, for instance, by employing the Talbot-Lau interferometer method to widely radiograph the subject at first, and then, employing the Talbot interferometer method, switched from the Talbot-Lau interferometer method, to radiograph a specific diseased part, etc., so as to acquire its clearer X ray image, it becomes possible to acquire a good X ray image in which the contrast of peripheral sections of the subject is emphasized, even for the operations for radiographing tissue sections from which the normal-type X-ray radiographing apparatus hardly captures a clear X ray image, including not only radiographing the joint disorders, which are represented by the rheumatic disease, but also radiographing various kinds of sections, such as a breast image capturing operation that should be capable of detecting a micro calcification from a breast, most of which is formed by a soft tissue, an operation for radiographing a child body, almost bones of which are cartilages, etc. Further, since the image processing apparatus 30 of the radiological image capturing system 100 appropriately conducts the various kinds of image processing, it becomes possible to obtain not only the X ray image being clearer than ever, but also the three-dimensional image of the subject H and such the image in which a concerned lesion area is emphasized. In this connection, instead of the configuration in which the subject placing plate 12, on which the subject H is to be placed, is disposed between the multi-slit element 11 and the first diffraction grating 15 (when the apparatus is used as the Talbot interferometer method, between the X ray tube 8 and the first diffraction grating 15) as structured in the radiological image capturing apparatus 1 embodied in the present invention and shown in FIGS. 2 and 3, for instance, as the radiological image capturing apparatus shown in FIG. 14, it is also applicable that the apparatus is so constituted that the subject placing plate 12, on which the subject H is placed, is disposed between the first diffraction grating 15 and the second diffraction grating 16. On that occasion, compared to the radiological image capturing apparatus 1 embodied in the present invention, the first diffraction grating 15 closely approaches the multi-slit element 11 and the X ray tube 8. It is necessary that the X rays to be incident into the first diffraction grating should have the coherency property, and for this purpose, an aperture width of each of the slits 111 (namely, a slit width, so to speak) of the multi-slit element 11, which is to be employed when the apparatus is used as the Talbot-Lau interferometer method as shown in FIG. 14, is set at a value in a range of around 1-50 μm, and it is preferable that the slit width is formed at a value in a range of around 7-30 μm. According to the above, the X rays to be incident into the first diffraction grating 15 posses the coherency property and the X rays irradiated from the X ray tube 8 are converted to the multi (plural) radiant sources. Further, when the apparatus is used as the Talbot interferometer method by rotating the multi-slit element 11, it becomes necessary to set the focal point diameter “a” of the X ray tube 8 at a smaller value. For this reason, the focal point diameter of the X ray tube 8 is set at a value equal to or greater than 0.1 μm, so as to make it possible to irradiate the X rays having an average energy within the abovementioned range and to acquire the output intensity being practically available. Further, when the focal point diameter “a” is set at 1 μm as its minimum value and the average energy of the X rays is set at 60 keV as its maximum value, it is necessary to set the distance L between the X ray tube 8 and the first diffraction grating 15 at a value equal to or greater that 0.5 m. However, since the coherency (coherency distance) is in proportion to the distance L while in inverse proportion to the average energy of the X rays and the focal point diameter, in the case that the coherency is acquired at 60 keV of the X ray average energy, for instance, the distance L between the X ray tube 8 and the first diffraction grating 15 can be set at a value equal to or greater than 0.125 m (12.5 cm) as far as the average energy of the X rays is 15 keV, or, even if the focal point diameter “a” of the X ray tube 8 is widened up to 4 μm, the equivalent degree of the coherency can be obtained. As indicated in the schematic diagram shown in FIG. 14, since the apparatus is so constituted that the first diffraction grating 15 is disposed at the space located between the X ray tube 8 and the subject H, it becomes possible not only to employ such the first diffraction grating 15 that is formed within a small area, resulting in an easiness of the creation working, but also to reduce the influence of slurs, etc., generated in the X ray image and caused by the manufacturing variation of the diffraction grating, etc., and therefore, it becomes possible to acquire the high-resolution X ray image being shaper than ever. According to the present invention, the following effects can be attained. (1) It becomes possible to make the apparatus sufficiently exhibit the Talbot effect so as to accurately detect a shape of the subject in the Moiré stripe image. On that occasion, by the converting X rays irradiated from the X ray tube to the multi (plural) radiant sources by employing the multi-slit element, the apparatus is made to be in such a state as if micro focus X ray tubes exist in the apparatus. By employing the multi-slit element having a sufficiently small slit width so as to acquire the Talbot effect, and by employing such an X ray tube that has a large focal diameter so as to acquire high power X rays, it becomes possible to acquire an X ray image without blur caused by the movements of human body, serving as the subject, only by irradiating the X rays in a short time equal to or smaller than several parts of one second. As a result, it becomes possible not only reduce an amount of X ray exposure to be irradiated onto the human body, but also to obtain an image having higher contrast to such an extent that the image can be used for a diagnosis purpose.(2) By setting a distance between the multi-slit element serving as multi radiant source and the X-ray detector 17, another distance between the multi-slit element and the first diffraction grating, and a slit interval of the multi-slit element, at appropriate values, respectively, it becomes possible to obtain a sufficiently clear X ray image even if the X ray irradiation time is short. Accordingly, it becomes possible to acquire a good X ray image in which the contrast of peripheral sections of the subject is emphasized, by employing the Talbot-Lau interferometer method for the operations for radiographing tissue sections from which the normal-type X-ray radiographing apparatus hardly captures a clear X ray image, including not only radiographing the joint disorders, which are represented by the rheumatic disease, but also radiographing various kinds of sections, such as a breast image capturing operation that should be capable of detecting a micro calcification from a breast, most of which is formed by a soft tissue, an operation for radiographing a child body, almost bones of which are cartilages, etc. |
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claims | 1. A micro-nuclear battery, comprising:a base frame comprising a bottom, a top and a side wall;a cantilever structure having a free end and a fixed end fixed to the side wall of the base frame and provided with a piezoelectric component thereon; anda radiation unit comprising an upper radioactive source and a lower radioactive source configured to emit electrons to the free end and respectively arranged at positions in inner surfaces on the top and bottom of the base frame corresponding to that are over opposite sides of the free end of the cantilever structure,wherein a width of the free end is greater than a width of the fixed end;wherein a distance between the cantilever structure and one of the upper radiation source and the lower radiation source is smaller than a distance between the cantilever structure and the other of the upper radiation source and the lower radiation source. 2. The micro-nuclear battery according to claim 1, wherein a ratio of the width of the free end to the width of the fixed end is between 2.5 and 6. 3. The micro-nuclear battery according to claim 1, further comprising: a radiation source control module electrically connected to the radiation unit and configured to control the radiation unit to emit electrons periodically. 4. The micro-nuclear battery according to claim 3, wherein the radiation source control module controls the upper radiation source and the lower radiation source to emit electrons alternately. 5. The micro-nuclear battery according to claim 1, wherein the piezoelectric component is formed of a piezoelectric material and is configured to generate a current pulse based on mechanical oscillation of the cantilever structure. 6. The micro-nuclear battery according to claim 1, further comprising:a direct current conversion module electrically connected to the piezoelectric component of the cantilever structure and configured to convert a current pulse generated by the piezoelectric component into a direct current. 7. The micro-nuclear battery according to claim 6, wherein the direct current conversion module further comprises a functional module configured to implement at least one of the following functions: transformation, rectification, filtering, and voltage regulation. 8. The micro-nuclear battery according to claim 1, wherein the radiation unit comprises a β radiation source. 9. The micro-nuclear battery according to claim 1, wherein the free end and the fixed end are connected via a connection portion therebetween. 10. The micro-nuclear battery according to claim 1, wherein the cantilever structure is formed of an elastic material. 11. An energy conversion method using the micro-nuclear battery according to claim 1, comprising:controlling, by the radiation source control module, the radiation unit to periodically emit electrons to the free end of the cantilever structure;generating, by the piezoelectric component, a current pulse based on mechanical oscillation of the cantilever structure; andconverting, by the direct current conversion module, the current pulse generated by the piezoelectric component into a direct current. 12. The energy conversion method of claim 11, wherein controlling the radiation unit to periodically emit electrons to the free end of the cantilever structure comprises: controlling the upper radiation source and the lower radiation source to emit electrons alternately. |
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052280705 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT A mechanical gantry portion 10 includes a stationary gantry portion 12 and a rotating gantry portion 14. The stationary gantry portion includes a stationary cylinder 16 in which a subject to be imaged is received. The rotatable gantry portion 14 is mounted on the stationary cylinder 16 by suitable bearings (not shown) to allow free rotational movement therearound. A motor 18 selectively rotates the rotatable gantry portion 14 around the cylinder 16. The motor may be a separate motor as shown in FIG. 1 connected by a chain or other suitable drive. Alternately, the cylinder 16 may be an integral portion of the "rotor" with the "stator" windings mounted to the rotatable gantry 14. An x-ray tube 20 is mounted to the rotatable gantry portion 14 to rotate therewith. Appropriate slip ring electrical connections (not shown) are mounted between the cylinder 16 and the rotatable gantry portion 14 to provide electrical operating power to the x-ray tube 20. More specifically, a tube voltage (kV) is provided to bias the anode and cathode and a filament current is provided to adjust the tube current (mA) between the cathode and the anode. At a higher tube current, the x-ray beam generated by the interaction of the tube current and the anode has a higher x-ray energy fluence. The x-ray tube has an outlet port 22 through which radiation is directed toward the cylinder 16. A collimator and shutter arrangement 24 shapes the emitted radiation into a thin, fan-shaped beam which spans a scan circle, i.e. a circular imaging region 26 within the cylinder 16. A shutter selectively gates the radiation beam on and off. Radiation from the fan beam which has traversed the scan circle 26 impinges upon an array of radiation detectors 28. In the preferred embodiment, the radiation detectors 28 are arranged in a complete circle on the stationary gantry portion. Alternately, an arc of radiation detectors can be mounted to the rotating gantry portion to rotate with the radiation source. An angular position monitoring means or resolver 30 monitors the angular position of the x-ray source 20 relative to a subject disposed in the scan circle 26. The angular position resolver 30 produces an indication of the current angular position of the x-ray source relative to the cylinder 16, hence the subject. Each of the detectors 28 is connected With a sampling means 40 which samples a group of the detectors 28 which are receiving incident radiation. Each time the x-ray source rotates a preselected angular increment relative to the subject, the sampled group of detectors is incremented around the circle. In this manner, electronic data is collected which represents radiation attenuation along a preselected multiplicity of paths through the subject. A reconstruction means 42 reconstructs the radiation attenuation data using a filtered backprojection or other conventional algorithm into n image representation which is stored in an image memory 44 for display on a video monitor 46. A digital motor speed controller 50 controls the motor 18 in order to control the angular velocity of the rotating gantry portion 14 relative to the stationary cylinder 16. A velocity versus angular position means 52, such as a look-up table, provides rotational speed information to the digital motor controller. More specifically to the preferred embodiment, the angular position resolver 30 addresses the angular position versus motor speed look-up table 52 with the current angular position to retrieve a corresponding motor speed designation which is communicated to the digital motor control 50. Each time the angular position resolver 30 senses a preselected increment f angular rotation, such as the angle spanned by one of detectors 28, the angular position resolver 30 indexes the address to the angular position versus speed look-up table 52 to provide the digital motor control 50 with an updated angular velocity. Of course, other means may be provided for converting angular position indications into speed control signals. The angular position versus velocity look-up table may be programmed various ways. For example, the height and width dimensions of the region of interest of the patient can be measured and compared with a plurality of preselected height to width dimensions. A preprogrammed look-up table corresponding to the most similar height and width dimensions is loaded into the look-up table memory 52 to control rotational speed during the upcoming scan. This table can be derived through manual calculations, trial and error, or the like. In the preferred embodiment, the angular position versus angular velocity look-up table 52 is derived empirically for each patient. An x-ray energy fluence measuring means 60 determines the average x-ray energy fluence which is being received by the arc of radiation detectors irradiated by the fan beam at each angular position of the x-ray tube 20 during a prediagnostic imaging scan. An x-ray energy fluence rate versus angular position table 62 correlates the measured x-ray energy fluence with angular position around the subject. An x-ray energy fluence rate reference means 64, such as a computer memory, stores one or more preselected reference x-ray energy or fluence rate at which the detectors 28 operate optimally. Preferably, the look up table includes a first reference x-ray energy fluence for diagnostic scanning and a second, lower reference x-ray energy fluence for screening. Optionally, additional higher and lower reference levels may be provided for other scanning procedures. A difference means 66 subtracts the x-ray energy fluence rate measured at each angle from the reference x-ray energy fluence rate to determine an x-ray energy fluence error or deviation for each angular position of the x-ray tube. An x-ray energy fluence error versus angular position memory means 68 correlates the x-ray energy fluence error or deviation corresponding to each angular position of the x-ray tube. An x-ray energy fluence correcting means 70 calculates a speed or speed change which is projected to eliminate the x-ray energy fluence deviation. For example, the x-ray energy fluence correction means 70 may determine the percentage by which the x-ray energy fluence deviation or error differs from the standard reference x-ray energy fluence and increase or decrease the selected rotation speed by the same percentage. Optionally, the x-ray energy fluence correction means may take other variables into account. Optionally, a maximum x-ray energy fluence detecting means 72 detects the maximum x-ray energy fluence detected by any detector at each angular position and stores it in a maximum x-ray energy fluence versus angular position look-up table 74. The correction means 70 then determines the effect which the determined speed correction should have on each maximum x-ray energy fluence. For example, if the x-ray energy fluence deviation is 5% of the reference x-ray energy fluence, slowing the rotational speed by 5% can be expected to increase both the maximum and the average x-ray energy fluence by 5%. The maximum speed change allowed may be limited by the percentage difference between the maximum x-ray energy fluence for each angular position and the reference. As another alternative, the technique discussed above may be used on the fly during a diagnostic scan to adjust speed, monitor for malfunctions, and the like. The average x-ray energy fluence means 60 can determine the currently measured x-ray energy fluence and use it to adjust the motor speed. This procedure is always one angular increment behind. When the motor speed corrections are calculated on the fly, a complete look-up table is not necessary. Rather, the speed indicating means 52 may be a counter, or the like, which is indexed up or down by the appropriate percentage to indicate the now desired motor speed to the digital motor control means 50. A rate of angular velocity change limiting means, such as an integrating or averaging circuit in the speed indicating means, may be used to assure that the gantry changes angular velocity smoothly without jerky movements. Of course, the x-ray energy fluence impinging on the subject at each angular position of the x-ray tube can be adjusted in ways other than changing the speed at which the x-ray tube rotates. For example, the x-ray energy fluence produced by the x-ray tube may be varied in accordance with the angular position. To this end, a tube current means 80 identifies a selected tube current for each angular position. This may again be a look-up table analogous to table 52 for which the tube currents or current variations are calculated as described above. An x-ray tube control circuit 82 controls the operating parameters of the x-ray tube 20. More specifically, the x-ray tube operating circuit includes a control circuit for controlling the tube current (mA). The x-ray tube control circuit 82 also controls the tube voltage applied between a filament or cathode 90 and an anode or target 92. Electrons boiled off from the heated filament are propelled by the tube voltage to travel from the filament to the anode to create the tube current, commonly measured in milliamps (mA). Preferably, the tube current is controlled by the bias on a grid disposed between the cathode and the anode. Optionally, the tube current can be adjusted by adjusting the current supplied to an x-ray tube filament to control its heating. The invention has been described with reference to the preferred embodiment. Obviously, modifications and alterations will occur to others upon reading and understanding the preceding detailed description. It is intended that the invention be construed as including all such modifications and alterations insofar as they come within the scope of the appended claims or the equivalents thereof. |
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046817263 | summary | BACKGROUND OF THE INVENTION (Field of the Invention) The present invention relates to a fast breeder reactor and, more particularly, to a technique most suitable for use in a tank-type fast breeder reactor in which a reactor core is supported at an inner lower portion of the reactor container of a nuclear reactor. A typical conventional tank-type fast breeder reactor (referred to simply as a "tank-type FBR", hereinafter) is arranged such that a reactor core, intermediate heat exchangers and circulation pumps are installed inside a nuclear reactor container known as a main container which contains sodium as a liquid metal. A roof slab having a rotational plug is attached to the upper end portion of the nuclear reactor container such as to cover the upper side thereof. The intermediate heat exchangers and the circulation pumps are mounted on the roof slab. The reactor core in which a multiplicity of fuel assemblies are disposed is supported by a core-supporting structural member which is mounted on the bottom portion of the nuclear reactor container. Examples of such core-supporting structure are respectively shown in FIG. 1 of the specification of Japanese Patent Laid-Open No. 15,895/1974, FIG. 1 of the specification of Japanese Patent Laid-Open No. 126,887/1979 and FIG. 1 of the specification of Japanese Patent Laid-Open No. 133,379/1982. The core-supporting structural member in these examples is adapted to support the reactor core and to apply the load of the reactor core to the pressure vessel of a nuclear reactor. The core-supporting structural member has a tapered shape, such as a truncated cone shape, in cross-section in order to support the reactor core stably. Even if the core-supporting structural member is formed into a specific cross-sectional configuration as described above, it is still preferable to take a measure to suppress the generation of vibration due to an exciting force produced by, for example, an earthquake, since the reactor core has a large weight and a center of gravity which is located higher than the core-supporting structural member. SUMMARY OF THE INVENTION (Object of the Invention) Accordingly, a primary object of the invention is to suppress the vibration of the reactor core of a fast breeder reactor. (Feature of the Invention) According to a first aspect of the invention, there is provided a fast breeder reactor having a nuclear reactor container filled with a liquid metal, a reactor core disposed within the nuclear reactor container, and a first supporting structural member mounted to the nuclear reactor container such as to support the reactor core, characterized by comprising: a cylindrical structural member which surrounds the periphery of the reactor core such as to define an annular gap between the cylindrical structural member and the reactor core for allowing the liquid metal to exist therein, the cylindrical structural member being mounted to the nuclear reactor container by means of a second supporting structural member. Further, according to a second aspect of the invention, there is provided in a tank-type fast breeder reactor having a nuclear reactor container filled with a liquid metal, a reactor core disposed within the reactor core container, a roof slab mounted on the upper portion of the nuclear reactor container, a core upper mechanism which is mounted on the roof slab and inserted into the nuclear reactor container, heat exchangers and circulation pumps, a fast breeder reactor characterized by comprising: a first supporting structural member for supporting the reactor core, the first supporting structural member being mounted on the bottom portion of the nuclear reactor container; a cylindrical structural member which surrounds the periphery of the reactor core such as to define an annular gap between the cylindrical structural member and the reactor core for allowing the liquid metal to exist therein, the cylindrical structural member being mounted to the nuclear reactor container by means of a second supporting structural member; and a partition wall which separates the inside of the nuclear reactor container into a low-temperature plenum and a high-temperature plenum, the partition wall being disposed above the cylindrical structural member and between an outer shroud of the reactor core and the nuclear reactor container. |
claims | 1. A method, comprising: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; andsupplying 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. 2. The method of claim 1, wherein the 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 comprises: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. 3. The method of claim 2, wherein the 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 comprises: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. 4. The method of claim 3, wherein the 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 comprises: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. 5. The method of claim 4, wherein the 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 comprises: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. 6. The method of claim 5, wherein the 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 comprises: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. 7. The method of claim 4, wherein the 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 comprises: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. 8. The method of claim 2, wherein the 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 comprises: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. 9. The method of claim 8, wherein the 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 comprises: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. 10. The method of claim 9, wherein the 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 comprises: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. 11. The method of claim 1, wherein the 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 comprises: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. 12. The method of claim 1, wherein the 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 comprises: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. 13. The method of claim 1, further comprising:supplementing the at least one auxiliary thermal reservoir with an additional portion of thermal energy from at least one additional energy source. 14. The method of claim 13, wherein the supplementing the at least one auxiliary thermal reservoir with an additional portion of thermal energy from at least one additional energy source comprises: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. 15. The method of claim 1, wherein the 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 comprises: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. 16. The method of claim 15, wherein the, 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 comprises: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. 17. The method of claim 15, wherein the, 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 comprises: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. 18. The method of claim 15, wherein the, 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 comprises: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. 19. The method of claim 15, wherein the, 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 comprises: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. 20. The method of claim 19, wherein the, 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 comprises: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. 21. The method of claim 15, wherein the, 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 comprises: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. 22. The method of claim 21, wherein the, 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 the plurality of nuclear reactor systems to the at least one auxiliary thermal reservoir comprises: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. 23. The method of claim 15, wherein the, 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 comprises: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. 24. The method of claim 15, wherein the, 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 comprises:upon a pre-selected 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. 25. The method of claim 15, wherein the, 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 comprises: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. 26. The method of claim 25, wherein the, 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 comprises: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. 27. The method of claim 15, wherein the, 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 comprises: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. 28. The method of claim 15, wherein the, 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 comprises: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. 29. The method of claim 1, wherein the 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 comprises: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. 30. The method of claim 1, wherein the 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 comprises: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. 31. The method of claim 1, wherein the 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 comprises: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. 32. The method of claim 31, wherein the, 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 comprises: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. 33. The method of claim 32, wherein the, 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 comprises: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. 34. The method of claim 31, wherein the, 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 comprises: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. 35. The method of claim 31, wherein the, 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 comprises: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. 36. The method of claim 31, wherein the, 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 comprises: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. 37. The method of claim 31, wherein the, 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 comprises: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. 38. The method of claim 37, wherein the, 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 comprises: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. 39. The method of claim 31, wherein the, 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 comprises: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. 40. The method of claim 31, wherein the, 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 comprises: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. 41. The method of claim 31, wherein the, 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 comprises: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. 42. The method of claim 1, wherein the 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 comprises: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. 43. The method of claim 1, further comprising:monitoring at least one condition of the at least one auxiliary thermal reservoir. 44. The method of claim 43, wherein the monitoring at least one condition of the at least one auxiliary thermal reservoir comprises:monitoring at least one condition of the at least one auxiliary thermal reservoir using at least one reservoir monitoring system. 45. The method of claim 43, wherein the monitoring at least one condition of the at least one auxiliary thermal reservoir comprises:monitoring the temperature of the at least one auxiliary thermal reservoir. 46. The method of claim 43, wherein the monitoring at least one condition of the at least one auxiliary thermal reservoir comprises:monitoring the pressure of the at least one auxiliary thermal reservoir. 47. The method of claim 43, wherein the monitoring at least one condition of the at least one auxiliary thermal reservoir comprises:determining the amount of energy stored in the at least one auxiliary thermal reservoir. 48. The method of claim 43, wherein the monitoring at least one condition of the at least one auxiliary thermal reservoir comprises:determining the amount of available energy storage capacity in the at least one auxiliary thermal reservoir. 49. The method of claim 1, wherein the 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 comprises: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. 50. The method of claim 49, wherein the 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 comprises: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. 51. The method of claim 49, wherein the 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 comprises: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. 52. The method of claim 49, wherein the 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 comprises: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. 53. The method of claim 49, wherein the 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 comprises: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. 54. The method of claim 49, wherein the 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 comprises: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. 55. The method of claim 49, wherein the 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 comprises: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. 56. The method of claim 55, wherein the 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 comprises: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. 57. The method of claim 55, wherein the 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 comprises: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. 58. The method of claim 1, further comprising:maintaining the temperature of at least one heat storage material of at least one auxiliary thermal reservoir above a selected temperature. 59. The method of claim 58, wherein the maintaining the temperature of at least one heat storage material of at least one auxiliary thermal reservoir above a selected temperature comprises: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. 60. The method of claim 1, wherein the 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, comprises: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. 61. The method of claim 1, wherein the 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, comprises: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. 62. The method of claim 1, wherein the 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, comprises: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. |
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056407017 | abstract | Soil comprising small soil particles, clay and silt particles, humus, fine vegetation, and contaminated with soluble or insoluble radioactive species is treated by first introducing an aqueous extracting solution comprising a mixture of sodium and potassium carbonate (or bicarbonate), or ammonium carbonate (or bicarbonate) into the soil to solubilize and disperse the radioactive species into solution. The extracting solution has a pH greater than or equal to about 7.5. Contaminated fine vegetation then is separated from the soil and extracting solution. Next, an acid like hydrochloric acid is introduced into the soil. The acid is added in an amount sufficient to lower the pH of the extracting solution at which point desirable organic material will substantially precipitate or coagulate from the extracting solution. The cleansed soil particles, including organic matter, is separated from the contaminated extracting solution. Radioactive species are then removed from the extracting solution, which then may be reused. |
043280706 | claims | 1. A method for the achievement of fusion energy release by inertial confinement comprising: (a) positioning a target within a cavity filled with a tenuous gas, the atomic weight A of the gas being given by ##EQU6## where R is the gas constant, Z is the degree of ionization given by Z=6.6.times.10.sup.-3 T.sup.0.41, T varies from 5.times.10.sup.6 .degree.K. to 10.sup.7 .degree.K., V varies from 20 km/sec to 200 km/sec; (b) using at least one power source to implode said cavity, the implosion being selected from the group consisting of ablatively driven implosion or direct high velocity impact implosion, said tenuous gas, having an initial atomic number density of between 10.sup.18 cm.sup.-3 and 10.sup.21 cm.sup.-3, becoming a source of black body radiation upon the high implosion velocity (V); (c) confining, compressing and thereby amplifying the black body radiation; (d) ablatively imploding the target positioned inside said cavity by means of black body radiation absorbed on the surface of said target. 2. The application of the method according to claim 1 for the generation of useful power by letting a sequence of fusion microexplosions take place inside a reactor chamber. 3. The application of the method according to claim 1 for the propulsion of spacecraft by letting a sequence of fusion microexplosions take place behind a structure being selected from the group consisting of either a pusher plate or concave reflector. 4. The application of the method according to claim 1 where said target includes a small amount of fissile material is compressed with the goal of controlled release of fission energy in small bursts. 5. The application of the method according to claim 1 where the target is composed of both fissile and fusionable material and together being ablatively compressed by the black body radiation for enhanced fission and fusion yield of the ensuing microexplosions. 6. The application of the method according to claim 1 where the source to implode the cavity is selected from the group consisting of either intense laser or charged particle beams. 7. The application of the method according to claim 1 where the initial energy source to implode the cavity is drawn from a macroscopic projectile accelerated to high velocities. 8. The application of the method according to claim 1 where the energy source to implode the cavity is obtained from focused chemical explosion waves. |
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claims | 1. An apparatus, comprising:at least one energy transfer system arranged to divert a selected portion of energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir; andat least one heat supply system configured to supply at least a portion of the diverted selected portion of energy to at least one energy conversion system of the nuclear reactor system in response to a shutdown event. 2. The apparatus of claim 1, wherein the at least one energy transfer system arranged to divert a selected portion of energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir comprises:at least one energy transfer system arranged to divert at least a portion of excess energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir. 3. The apparatus of claim 2, wherein the at least one energy transfer system arranged to divert at least a portion of excess energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir comprises:at least one energy transfer system arranged to divert at least a portion of energy exceeding operational demand of at least one energy conversion system from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir. 4. The apparatus of claim 1, wherein the at least one energy transfer system arranged to divert a selected portion of energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir comprises:at least one energy transfer system arranged to divert a specified percentage of the energy output of a portion of at least one nuclear reactor system from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir. 5. The apparatus of claim 1, wherein the at least one energy transfer system arranged to divert a selected portion of energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir using at least one energy transfer system comprises:at least one energy transfer system arranged to divert a selected portion of thermal energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir. 6. The apparatus of claim 5, wherein the at least one energy transfer system arranged to divert a selected portion of thermal energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir comprises:at least one heat transfer system arranged to divert a selected portion of thermal energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir. 7. The apparatus of claim 6, wherein the at least one heat transfer system arranged to divert a selected portion of thermal energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir comprises:at least one heat transfer system arranged to divert a selected portion of thermal energy from a portion of at least one nuclear reactor system in thermal communication with at least one heat source of the least one nuclear reactor system to at least one auxiliary thermal reservoir. 8. The apparatus of claim 7, wherein the at least one heat transfer system arranged to divert a selected portion of thermal energy from a portion of at least one nuclear reactor system in thermal communication with at least one heat source of the least one nuclear reactor system to at least one auxiliary thermal reservoir comprises:at least one heat transfer system arranged to divert a selected portion of thermal energy from a portion of at least one nuclear reactor system in thermal communication with at least one nuclear reactor core of the at least one nuclear reactor system to at least one auxiliary thermal reservoir. 9. The apparatus of claim 8, wherein the at least one heat transfer system arranged to divert a selected portion of thermal energy from a portion of at least one nuclear reactor system in thermal communication with at least one nuclear reactor core of the at least one nuclear reactor system to at least one auxiliary thermal reservoir comprises:at least one heat transfer system arranged to divert a selected portion of thermal energy from a portion of at least one primary coolant system of at least one nuclear reactor system to at least one auxiliary thermal reservoir. 10. The apparatus of claim 9, wherein the at least one heat transfer system arranged to divert a selected portion of thermal energy from a portion of at least one primary coolant system of at least one nuclear reactor system to at least one auxiliary thermal reservoir comprises:at least one heat transfer system arranged to divert a selected portion of thermal energy from a portion of at least one primary coolant loop of at least one nuclear reactor system to at least one auxiliary thermal reservoir. 11. The apparatus of claim 9, wherein the at least one heat transfer system arranged to divert a selected portion of thermal energy from a portion of at least one primary coolant system of at least one nuclear reactor system to at least one auxiliary thermal reservoir comprises:at least one heat transfer system arranged to divert a selected portion of thermal energy from at least one coolant pool of at least one nuclear reactor system to at least one auxiliary thermal reservoir. 12. The apparatus of claim 9, wherein the at least one heat transfer system arranged to divert a selected portion of thermal energy from a portion of at least one primary coolant system of at least one nuclear reactor system to at least one auxiliary thermal reservoir comprises:at least one heat transfer system arranged to divert a selected portion of thermal energy from a portion of at least one primary coolant system of at least one nuclear reactor system 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 at least one nuclear reactor system and at least one secondary coolant system of the nuclear reactor system. 13. The apparatus of claim 6, wherein the at least one heat transfer system arranged to divert a selected portion of thermal energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir comprises:at least one direct fluid exchange heat transfer system arranged to divert a selected portion of thermal energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir. 14. The apparatus of claim 13, wherein the at least one direct fluid exchange heat transfer system arranged to divert a selected portion of thermal energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir comprises:at least one direct fluid exchange heat transfer system arranged to intermix at least one reservoir fluid of at least one auxiliary thermal reservoir with at least one coolant of at least one nuclear reactor system. 15. The apparatus of claim 6, wherein the at least one heat transfer system arranged to divert a selected portion of thermal energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir comprises:at least one heat exchanger arranged to divert a selected portion of thermal energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir. 16. The apparatus of claim 4, wherein the energy transfer system arranged to divert a selected portion of energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir comprises:at least one energy transfer system arranged to divert at least a portion of electrical energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir. 17. The apparatus of claim 16, wherein the at least one energy transfer system arranged to divert at least a portion of electrical energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir comprises:at least one electrical-to-thermal conversion system arranged to divert at least a portion of electrical energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir. 18. The apparatus of claim 17, wherein the at least one electrical-to-thermal conversion system arranged to divert at least a portion of electrical energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir comprises:at least one electrical-to-thermal conversion system arranged to divert at least a portion of electrical energy from at least one energy conversion system of at least one nuclear reactor system to at least one auxiliary thermal reservoir. 19. The apparatus of claim 17, wherein the at least one electrical-to-thermal conversion system arranged to divert at least a portion of electrical energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir comprises:at least one resistive heating device arranged to divert at least a portion of electrical energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir. 20. The apparatus of claim 1, wherein the at least one heat supply system configured to supply at least a portion of the diverted selected portion of energy to at least one energy conversion system of the nuclear reactor system in response to a shutdown event comprises:at least one heat exchange loop. 21. The apparatus of claim 1, wherein the at least one heat supply system configured to supply at least a portion of the diverted selected portion of energy to at least one energy conversion system of the nuclear reactor system in response to a shutdown event comprises:at least one heat pipe. 22. The apparatus of claim 1, wherein the at least one heat supply system configured to supply at least a portion of the diverted selected portion of energy to at least one energy conversion system of the nuclear reactor system in response to a shutdown event comprises:at least one heat exchanger. 23. The apparatus of claim 1, wherein the at least one heat supply system configured to supply at least a portion of the diverted selected portion of energy to at least one energy conversion system of the nuclear reactor system in response to a shutdown event comprises:at least one thermoelectric device. 24. The apparatus of claim 1, wherein the at least one energy conversion system of the nuclear reactor system comprises:at least one primary energy conversion system of the nuclear reactor system. 25. The apparatus of claim 1, wherein the at least one energy conversion system of the nuclear reactor system comprises:at least one auxiliary energy conversion system of the nuclear reactor system. 26. The apparatus of claim 1, wherein the at least one energy conversion system of the nuclear reactor system comprises:at least one emergency energy conversion system of the nuclear reactor system. 27. The apparatus of claim 1, wherein the at least one heat supply system configured to supply at least a portion of the diverted selected portion of energy to at least one energy conversion system of the nuclear reactor system in response to a shutdown event comprises:at least one heat supply system configured to supply at least a portion of the diverted selected portion of energy to at least one boiling loop of the nuclear reactor system in response to a shutdown event. 28. The apparatus of claim 1, wherein the at least one heat supply system configured to supply at least a portion of the diverted selected portion of energy to at least one energy conversion system of the nuclear reactor system in response to a shutdown event comprises:at least one heat supply system configured to supply at least a portion of the diverted selected portion of energy to at least one turbine of the nuclear reactor system in response to a shutdown event. 29. The apparatus of claim 28, wherein the at least one heat supply system configured to supply at least a portion of the diverted selected portion of energy to at least one turbine of the nuclear reactor system in response to a shutdown event comprises:at least one heat supply system configured to supply at least a portion of the diverted selected portion of energy to at least one working fluid of at least one turbine of the nuclear reactor system in response to a shutdown event. 30. The apparatus of claim 1, wherein the at least one energy conversion system of the nuclear reactor system comprises:at least one topping cycle of the nuclear reactor system. 31. The apparatus of claim 1, wherein the at least one energy conversion system of the nuclear reactor system comprises:at least one bottoming cycle of the nuclear reactor system. 32. The apparatus of claim 1, wherein the at least one energy conversion system of the nuclear reactor system comprises:at least one low grade heat dump. 33. The apparatus of claim 1, further comprising:at least one additional energy source configured to supplement the at least one auxiliary thermal reservoir with an additional portion of energy. 34. The apparatus of claim 33, wherein the at least one additional energy source configured to supplement the at least one auxiliary thermal reservoir with an additional portion of energy comprises:at least one energy conversion system of the nuclear reactor system configured to supplement the at least one auxiliary thermal reservoir with an additional portion of energy. 35. The apparatus of claim 33, wherein the at least one additional energy source configured to supplement the at least one auxiliary thermal reservoir with an additional portion of energy comprises:at least one additional nuclear reactor system configured to supplement the at least one auxiliary thermal reservoir with an additional portion of energy. 36. The apparatus of claim 1, wherein the at least one energy transfer system arranged to divert a selected portion of energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir comprises:at least one energy transfer system arranged to divert a selected portion of energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir in response to at least one condition. 37. The apparatus of claim 36, wherein the at least one energy transfer system arranged to divert a selected portion of energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir in response to at least one condition comprises:at least one energy transfer system arranged to divert a selected portion of energy from a portion of the at least one nuclear reactor system to at least one auxiliary thermal reservoir in response to at least one operation system of at least one nuclear reactor system. 38. The apparatus of claim 37, wherein the at least one energy transfer system, arranged to divert a selected portion of energy from a portion of the at least one nuclear reactor system to at least one auxiliary thermal reservoir in response to at least one operation system of at least one nuclear reactor system comprises:at least one energy transfer system arranged to divert a selected portion of energy from a portion of the at least one nuclear reactor system to at least one auxiliary thermal reservoir in response to at least one signal from at least one operation system of at least one nuclear reactor system. 39. The apparatus of claim 36, wherein the at least one energy transfer system arranged to divert a selected portion of energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir in response to at least one condition comprises:at least one energy transfer system arranged to divert a selected portion of energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir in response to at least one reservoir operation system of at least one auxiliary thermal reservoir. 40. The apparatus of claim 39, wherein the at least one energy transfer system arranged to divert a selected portion of energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir in response to at least one reservoir operation system of at least one auxiliary thermal reservoir comprises:at least one energy transfer system arranged to divert a selected portion of energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir in response to at least one signal from at least one reservoir operation system of at least one auxiliary thermal reservoir. 41. The apparatus of claim 36, wherein the at least one energy transfer system arranged to divert a selected portion of energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir in response to at least one condition comprises:at least one energy transfer system arranged to divert a selected portion of energy from a portion of the at least one nuclear reactor system to at least one auxiliary thermal reservoir in response to at least one signal from at least one operator of at least one nuclear reactor system. 42. The apparatus of claim 36, wherein the at least one energy transfer system arranged to divert a selected portion of energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir in response to at least one condition comprises:at least one energy transfer system arranged to divert a selected portion of energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir upon a pre-selected diversion start time. 43. The apparatus of claim 36, wherein the at least one energy transfer system arranged to divert a selected portion of energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir in response to at least one condition comprises:at least one energy transfer system arranged to divert a selected portion of energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir in response to a shutdown event. 44. The apparatus of claim 36, wherein the at least one energy transfer system arranged to divert a selected portion of energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir in response to at least one condition comprises:at least one energy transfer system arranged to divert a selected portion of energy from a portion of the at least one nuclear reactor system to the at least one auxiliary thermal reservoir in response to determination of the amount of energy stored in at least one auxiliary reservoir. 45. The apparatus of claim 1, wherein the at least one heat supply system configured to supply at least a portion of the diverted selected portion of energy to at least one energy conversion system of the nuclear reactor system in response to a shutdown event comprises:at least one heat supply system configured to supply at least a portion of the diverted selected portion of energy to at least one energy conversion system of the nuclear reactor system in response to a scheduled shutdown event. 46. The apparatus of claim 1, wherein the at least one heat supply system configured to supply at least a portion of the diverted selected portion of energy to at least one energy conversion system of the nuclear reactor system in response to a shutdown event comprises:at least one heat supply system configured to supply at least a portion of the diverted selected portion of energy to at least one energy conversion system of the nuclear reactor system in response to an emergency shutdown event. 47. The apparatus of claim 1, wherein the at least one heat supply system configured to supply at least a portion of the diverted selected portion of energy to at least one energy conversion system of the nuclear reactor system in response to a shutdown event comprises:at least one heat supply system configured to supply at least a portion of the diverted selected portion of energy to at least one energy conversion system of the nuclear reactor system in response to at least one condition indicative of a shutdown event. 48. The apparatus of claim 1, wherein the at least one heat supply system configured to supply at least a portion of the diverted selected portion of energy to at least one energy conversion system of the nuclear reactor system in response to a shutdown event comprises:at least one heat supply system configured to supply at least a portion of the diverted selected portion of energy to at least one energy conversion system of the nuclear reactor system in response to a shutdown event established by at least one operation system of the nuclear reactor system. 49. The apparatus of claim 48, wherein the at least one heat supply system configured to supply at least a portion of the diverted selected portion of energy to at least one energy conversion system of the nuclear reactor system in response to a shutdown event established by at least one operation system of the nuclear reactor system comprises:at least one heat supply system configured to supply at least a portion of the diverted selected portion of energy to at least one energy conversion system of the nuclear reactor system in response to a shutdown event established by at least one reactor control system of the nuclear reactor system. 50. The apparatus of claim 49, wherein the at least one reactor control system of the nuclear reactor system comprises:at least one reactor control system of the nuclear reactor system configured to respond to at least one signal from at least one safety system. 51. The apparatus of claim 50, wherein the at least one safety system comprises:at least one safety system configured to respond to at least one sensed condition of the nuclear reactor system. 52. The apparatus of claim 1, wherein the at least one heat supply system configured to supply at least a portion of the diverted selected portion of energy to at least one energy conversion system of the nuclear reactor system in response to a shutdown event comprises:at least one heat supply system configured to supply at least a portion of the diverted selected portion of energy to at least one energy conversion system of the nuclear reactor system upon determination of a selected amount of stored energy in the at least one auxiliary thermal reservoir in response to a shutdown event. 53. The apparatus of claim 1, wherein the at least one heat supply system configured to supply at least a portion of the diverted selected portion of energy to at least one energy conversion system of the nuclear reactor system comprises:at least one heat supply system configured to supply at least a portion of the diverted selected portion of energy to at least one energy conversion system of the nuclear reactor system upon determination of a selected amount of available energy storage capacity in the at least one auxiliary thermal reservoir in response to a shutdown event. 54. The apparatus of claim 1, further comprising:at least one reservoir monitoring system configured to monitor at least one condition of the at least one auxiliary reservoir. 55. The apparatus of claim 54, wherein the reservoir monitoring system comprises:at least one reservoir temperature monitoring system. 56. The apparatus of claim 54, wherein the reservoir monitoring system comprises:at least one reservoir pressure monitoring system. 57. The apparatus of claim 54, wherein the reservoir monitoring system comprises:at least one monitoring system configured to determine the amount of energy stored in the at least one auxiliary reservoir. 58. The apparatus of claim 1, wherein the at least one energy transfer system arranged to divert a selected portion of energy from a portion of at least one nuclear reactor system to at least one auxiliary thermal reservoir comprises:at least one energy transfer system arranged to divert a selected portion of energy from a portion of at least one nuclear reactor system to a mass of at least one heat storage material of at least one auxiliary thermal reservoir. 59. The apparatus of claim 58, wherein the at least one heat storage material comprises:at least one solid heat storage material reservoir. 60. The apparatus of claim 58, wherein the at least one heat storage material comprises:at least one liquid heat storage material. 61. The apparatus of claim 58, wherein the at least one heat storage material comprises:at least one pressurized gaseous heat storage material. 62. The apparatus of claim 58, wherein the at least one heat storage material comprises:at least one mixed phase heat storage material. 63. The apparatus of claim 58, wherein the at least one heat storage material comprises:at least one material having a phase transition within the operating temperature of the at least one auxiliary thermal reservoir. 64. The apparatus of claim 58, wherein the at least one heat storage material comprises:at least one heat storage material contained in a reservoir containment system. 65. The apparatus of claim 64, wherein the reservoir containment system comprises:at least one external vessel. 66. The apparatus of claim 65, wherein the at least one external vessel comprises:at least one external high pressure gas vessel. 67. The apparatus of claim 65, wherein the at least one external vessel comprises:at least one external liquid vessel. 68. The apparatus of claim 64, wherein the reservoir containment system comprises:at least one external liquid pool. 69. The apparatus of claim 1, wherein the at least one auxiliary thermal reservoir comprises:at least one auxiliary thermal reservoir configured to store the selected portion of energy in the form of a temperature change in at least one heat storage material of the auxiliary thermal reservoir. 70. The apparatus of claim 1, wherein the at least one auxiliary thermal reservoir comprises:at least one auxiliary thermal reservoir configured to store the selected portion of energy in the form of a phase change in at least one heat storage material of the auxiliary thermal reservoir. 71. The apparatus of claim 1, further comprising:at least one reservoir temperature control system configured to maintain the temperature of at least one heat storage material of at least one auxiliary thermal reservoir above a selected temperature. 72. The apparatus of claim 71, wherein the at least one reservoir temperature control system configured to maintain the temperature of at least one heat storage material of at least one auxiliary thermal reservoir above a selected temperature comprises:at least one reservoir temperature control system configured to maintain 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. 73. The apparatus of claim 1, wherein the at least one nuclear reactor system, comprises:at least one thermal spectrum nuclear reactor system. 74. The apparatus of claim 1, wherein the at least one nuclear reactor system, comprises:at least one fast spectrum nuclear reactor system. 75. The apparatus of claim 1, wherein the at least one nuclear reactor system, comprises:at least one multi-spectrum nuclear reactor system. 76. The apparatus of claim 1, wherein the at least one nuclear reactor system, comprises:at least one breeder nuclear reactor system. 77. The apparatus of claim 1, wherein the at least one nuclear reactor system, comprises:at least one traveling wave nuclear reactor system. |
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claims | 1. A method of patterning a substrate, comprising:providing a first set of patterned resist features on the substrate;exposing the first set of patterned resist features to a first exposure of ions extracted from a plasma sheath modifier operable to provide ions incident on the substrate over an angular range relative to the substrate; andperforming a lithographic patterning process on the substrate to form a second set of patterned resist features, wherein the plasma sheath modifier comprises a first insulator portion and a second insulator portion that define a gap therebetween, such that a shape of a boundary of the plasma proximate the gap is a convex shape relative to a plane of the substrate. 2. The method of claim 1, wherein the first and second set of patterned resist features are formed using a double patterning lithographic process. 3. The method of claim 1, wherein the first exposure is operable to harden the first set of patterned resist features which remain intact during the lithographic patterning process used to form the second set of patterned resist features. 4. The method of claim 1, wherein the first exposure comprises an exposure to inert gas ions. 5. The method of claim 1, wherein ion energy of the ions of the first exposure is less than about 20 keV. 6. The method of claim 1 wherein a linewidth roughness of the first set of patterned resist features is substantially reduced after the first exposure. 7. The method of claim 1, further comprising exposing the first set and second set of patterned resist features to a second exposure of ions extracted from the plasma sheath modifier operable to provide ions from the second exposure incident on the substrate over a wide angular range, wherein a linewidth roughness of the second set of patterned resist features is reduced after the second exposure. 8. The method of claim 1, wherein the angular range is between positive 60° and negative 60° centered about 0°. 9. The method of claim 1, wherein the first exposure is sufficient to produce a substantial decrease in low frequency linewidth roughness. 10. The method of claim 1, wherein the first and second sets of patterned resist features are silicon-based features. |
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041347919 | summary | This invention relates to a nuclear reactor fuel assembly comprising a stack of parallel fuel plates disposed vertically in uniformly spaced relation and surrounded by cladding. Said fuel plates are joined together and maintained in position with respect to each other by connecting means which extend at right angles to the plane of the fuel plates and are disposed at intervals in the vertical direction on the lateral edges of said plates. As disclosed in particular in American Patent Application Ser. No. 484,743 of July 1, 1974 fuel assemblies of the type mentioned above are already known. Each clad fuel plate of a fuel assembly is made up of small parallelepipedal plates of fuel material which are formed especially of uranium oxide and each covered with thin metal foil. Said small plates are suitably disposed in spaced relation by means of metallic strips and distributed over the entire surface of the fuel plate to be formed. The complete set of small plates which are each covered with foil is enclosed along the lateral faces of each fuel plate between two metallic cladding sheets which are separated by the thickness of said small plates. By way of alternative, consideration has also been given to the possibility of covering each small plate with a thin metallic strip, in which case the cladding sheets are directly in contact with the uncovered faces of the small plates. In another patent Application which was filed in France on Oct. 2nd, 1975 under No. 75 30247 in the name of Commissariat a 1'Energie Atomique, a different type of fuel assembly made up of a stack of clad plates was also disclosed. The means employed for interconnecting these plates consisted of lateral combs, cross-members or wires which served to connect the parallel fuel plates together so as to constitute a single-unit structure. In this design, provision is made for at least two of these cross-members or the like at the end of the stack in order to be welded or made integral with an end component of parallelepipedal shape which is provided with means for supporting the fuel assembly thus formed. This latter may or may not be associated with a laterally closed outer fuel wrapper, the clad fuel plates being cooled by circulation of a fluid which usually consists of water and flows under pressure in contact with said fuel plates within the wrapper. In other design solutions which are also known in the technique and generally applicable to clad fuel plates of all types, the plates can be maintained within the fuel wrapper by forming parallel longitudinal grooves in the lateral internal faces of the fuel wrapper so as to permit engagement of said plates in said grooves. In contradistinction to clad fuel plates, other forms of nuclear reactor fuel assemblies are also known in which the fuel assembly consists of a cluster of fuel pencils of appreciable length and of generally cylindrical shape. These fuel pencils are maintained in parallel relation and disposed on a uniform lattice by means of spacer grids traversed by the fuel pencils and disposed at intervals along the height of these latter. The fuel-pencil cluster rests on a bottom support component and this latter is in turn connected to a top component in parallel relation thereto by means of tubular connecting-members or tie-rods which are located at intervals in the stack at certain nodes of the lattice. Said tubular tie-rods are advantageously designed to serve as guides for the rods of neutron-absorbing material which are displaced in sliding motion for controlling neutron flux and making reactivity changes during reactor operation. As has already been disclosed in American Pat. No. 3,954,560 of Dec. 11, 1972 consideration has already been given to a particular solution of this type in which the spacer grids are freely mounted in a floating assembly both with respect to the coupling tie-rods and with respect to the clad fuel pencils. Displacement of said spacer grids is limited by spacing sleeves which are so designed as to provide a suitable clearance space between said grids and also between these latter and the end components of the fuel assembly. The aim of the present invention is to make an improvement in plate-type fuel assemblies for nuclear reactors. The primary objective of this improvement is to give these fuel assemblies a general shape which is similar to that of a fuel-pencil assembly of the type recalled above, especially in regard to the external contour of said assemblies. Within a reactor core formed by the side-by-side arrangement of fuel assemblies of a first type such as fuel-pencil assemblies, plate-type assemblies can accordingly be substituted for one or a number or even all these latter if necessary without having to modify the environment and structures of the reactor core. In particular, the invention is intended to permit the substitution mentioned above while permitting adaptation of the plate-type assemblies employed to the mechanical means adopted for handling fuel-pencil assemblies and while also permitting accommodation of the reactivity control systems which are designed and employed for these latter. A further aim of the invention is to ensure substantially uniform cooling of the reactor core assemblies while making it possible, especially in the event that a plate-type assembly is placed next to a pencil-type assembly, to prevent any unbalance in the flow of coolant as a result of the different structures of these two types of fuel assembly. The plate-type fuel assembly under consideration is accordingly distinguished by the fact that at least a number of plates of the fuel stack are provided with hollow sleeves which are rigidly fixed to the fuel plates and extend vertically in the plane of said plates, said sleeves being uniformly spaced in the transverse direction of said plates. In a first embodiment of the invention, the sleeves which are rigidly fixed to the fuel plates have a longitudinal dimension which exceeds that of the fuel plates and are secured at the ends thereof to two support components of the fuel assembly. In another alternative embodiment, tubular tie-rods are capable of passing freely through the sleeves which are rigidly fixed to the fuel plates and are greater in length than said plates, said tie-rods being secured to the support components. In either of the two particular embodiments of the invention for producing in one case a single-unit structure in which the fuel plates are rigidly fixed to the sleeves and for producing in the other case a floating assembly in which the fuel plates are permitted to slide with respect to the tubular tie-rods, at least a certain number of the sleeves or tubular tie-rods serve to guide the reactor control rods of neutron-absorbing material as said control rods are displaced in sliding motion within the fuel assembly. In accordance with a particular feature, the means for interconnecting the fuel plates of the stack are constituted by combs having teeth which are flush-mounted in the cladding of said fuel plates or by transverse rods which are welded to the edges of said fuel plates. In accordance with another distinctive feature, the parallel fuel plates of the stack are braced with respect to each other by means of transverse spacers located at right angles to the plane of the fuel plates and distributed over the surface of said plates. As an advantageous feature, the transverse spacers are constituted by flat lugs having a width equal to the spacing between two fuel plates, said flat lugs being joined together in pairs by means of cylindrical connecting portions having a length equal to the thickness of the fuel plates. Preferably, the transverse spacers tranverse the fuel plates through elongated slots formed in said plates in a zone which does not contain fuel and are positioned by means of a movement of rotation through an angle of 90.degree. in order to bring the plane of the spacing lugs into position at right angles to the direction of the elongated slots. In accordance with another alternative embodiment, the hollow sleeves extend in a single piece to the full height of the fuel plates or else are constituted by spaced tubular elements located in the line of extension of each other. As an advantageous feature, the hollow sleeves project from the contour of the fuel plates at that end of these latter which is located opposite to the bottom support component so as to form a given spacing between said component and said fuel plates. In the particular case in which the plate-type fuel assembly is intended to be placed within the reactor in the vicinity of at least one assembly of parallel fuel pencils, the means for interconnecting the fuel plates are provided with extensions in the form of lateral sheet-metal strips of small thickness applied against the edges of the fuel plates in order to limit the flow of coolant which penetrates between said fuel plates through the sides of the fuel assembly. Moreover, the stack of fuel plates is provided with means for inducing turbulence in the flow between the fuel plates, said means being such as to comprise thin metallic cross-strips which are parallel to the fuel plates or which have a wavy shape. In accordance with a distinctive feature, the thin metallic cross-strips have at least one cut-out edge, the portions of strips thus formed being folded-back in one direction and in the other or in one direction alone. |
claims | 1. A passive containment cooling and filtered venting system of a nuclear power plant, the plant including:a core,a reactor pressure vessel that accommodates the core,a containment vessel including:a dry well that contains the reactor pressure vessel,a wet well that contains in its lower portion a suppression pool connected to the dry well via a LOCA vent pipe and includes in its upper portion a wet well gas phase,a vacuum breaker that circulates gas in the wet well gas phase to the dry well, anda pedestal that supports the reactor pressure vessel in the containment vessel via an RPV skirt and an RPV support and forms a pedestal cavity inside,the passive containment cooling and filtered venting system comprising:an outer well that is provided outside the dry well and the wet well, surrounds at least a part of the dry wall and the wet wall, adjoins the dry well via a dry well common part wall, adjoins the wet well via a wet well common part wall, and has pressure resistance and gastightness equivalent to pressure resistance and gastightness of the dry well and the wet well;a scrubbing pool that is arranged in the outer well and stores water inside;a cooling water pool that is installed above the dry well and the outer well and reserves cooling water;a heat exchanger that includes an inlet plenum, an outlet plenum, and a heat exchanger tube, and is submerged at least in part in the cooling water;a gas supply pipe that is connected to the inlet plenum of the heat exchanger at a first end of the gas supply pipe and connected to a gas phase of the containment vessel at a second end of the gas supply pipe to lead gas in the containment vessel to the heat exchanger;a condensate return pipe that is connected to the outlet plenum of the heat exchanger at a first end of the condensate return pipe, passes through the outer well, and is connected to inside the containment vessel at a second end of the condensate return pipe to lead condensate in the heat exchanger into the containment vessel; anda gas vent pipe that is connected to the outlet plenum of the heat exchanger at a first end of the gas vent pipe, passes through the outer well, has a second end of the gas vent pipe installed as submerged in the scrubbing pool in the outer well, and releases noncondensable gas in the heat exchanger to the outer well. 2. The passive containment cooling and filtered venting system according to claim 1, further comprising:a lid that covers a top of the scrubbing pool to form a space above a surface of the water of the scrubbing pool; anda first outlet pipe that is connected to the lid at a first end of the first outlet pipe and that opens to a space of the outer well at a second end of the first outlet pipe. 3. The passive containment cooling and filtered venting system according to claim 2, further comprising:a filter that is connected to the first outlet pipe at the second end of the first outlet pipe; anda second outlet pipe that is connected to the filter at a first end of the second outlet pipe and opens to the outer well at a second end of the second outlet pipe. 4. The passive containment cooling and filtered venting system according to claim 1, further comprising:a flooder pipe that opens to the suppression pool at a first end of the flooder pipe and opens in the pedestal cavity at a second end of the flooder pipe;a flooder valve that is arranged on a portion of the flooder pipe inside the pedestal cavity; anda flooder check valve that is arranged on a portion of the flooder pipe inside the suppression pool. 5. The passive containment cooling and filtered venting system according to claim 1, whereinthe condensate return pipe includes a U-shaped water seal, and the second end of the condensate return pipe is connected to the dry well through the dry well common part wall, the U-shaped water seal being a U-bent portion storing sealing water inside. 6. The passive containment cooling and filtered venting system according to claim 5, further comprising a spray sparger at the second end of the condensate return pipe in the dry well. 7. The passive containment cooling and filtered venting system according to claim 1, further comprising:a PCCS drain tank that stores water inside, has a gas phase in its upper portion, and is arranged inside the outer well; andan overflow pipe that connects the gas phase of the PCCS drain tank at a first end of the overflow pipe with the dry well at a second end of the overflow pipe, whereinthe second end of the condensate return pipe is submerged in the water in the PCCS drain tank. 8. The passive containment cooling and filtered venting system according to claim 7, further comprising a spray sparger at the second end of the overflow pipe in the dry well. 9. The passive containment cooling and filtered venting system according to claim 7, further comprising:a water injection pipe that is connected to below a water surface of the water in the PCCS drain tank at a first end of the water injection pipe and led into the pedestal cavity at a second end of the water injection pipe;a water injection valve that is arranged on the water injection pipe; anda drain pit that is located in the PCCS drain tank and stores water inside the drain pit, whereinthe second end of the condensate return pipe is submerged in the drain pit. 10. The passive containment cooling and filtered venting system according to claim 1, further comprising a condensate check valve arranged on the condensate return pipe. 11. The passive containment cooling and filtered venting system according to claim 1, whereinthe gas supply pipe is connected to the inlet plenum of the heat exchanger at the first end of the gas supply pipe, passes through the outer well, and is connected to the dry well through the dry well common part wall at the second end of the gas supply pipe to lead the gas in the dry well to the heat exchanger. 12. The passive containment cooling and filtered venting system according to claim 11, further comprising:a cyclone separator that is arranged in the outer well; andan inlet pipe that connects the dry well with the cyclone separator and leads gas in the dry well to the cyclone separator, whereinthe gas supply pipe is connected to an outlet of the cyclone separator at the first end to lead gas discharged from the cyclone separator to the heat exchanger. 13. The passive containment cooling and filtered venting system according to claim 11, further comprising:a gas supply isolation valve that is installed on the gas supply pipe; anda wet well gas supply pipe that is connected at a first end of the wet well gas supply pipe to the inlet plenum of the heat exchanger or a portion of the gas supply pipe between the inlet plenum and the gas supply isolation valve, passes through the outer well, and is connected at a second end of the wet well gas supply pipe to inside the wet well gas phase through the wet well common part wall to lead the gas in the wet well to the heat exchanger. 14. The passive containment cooling and filtered venting system according to claim 13, wherein:the scrubbing pool is a filtered venting tank storing decontamination water inside;the second end of the gas vent pipe is connected to an inlet pipe of the filtered venting tank; andthe filtered venting tank opens to an interior of the outer well via an outlet pipe. 15. The passive containment cooling and filtered venting system according to claim 1, further comprising a screen located on a portion of the gas supply pipe that extends inside the dry well. 16. A nuclear power plant comprising:a containment vessel that contains a reactor pressure vessel, the containment vessel including a dry well and a wet well;an outer well that is provided outside the dry well and the wet well, surrounds at least a part of the dry well and the wet well, and has pressure resistance and gastightness;a scrubbing pool that is arranged in the outer well and stores water inside;a cooling water pool that is installed above the dry well and the outer well and reserves cooling water;a heat exchanger that includes an inlet plenum, an outlet plenum, and a heat exchanger tube, and is submerged at least in part in the cooling water;a gas supply pipe that is connected to the inlet plenum of the heat exchanger at a first end of the gas supply pipe and connected to a gas phase of the containment vessel at a second end of the gas supply pipe to lead gas in the containment vessel to the heat exchanger;a condensate return pipe that is connected to the outlet plenum of the heat exchanger at a first end of the condensate return pipe, passes through the outer wall, and is connected to inside the containment vessel at a second end of the condensate return pipe to lead condensate in the heat exchanger into the containment vessel; anda gas vent pipe that is connected to the outlet plenum of the heat exchanger at a first end of the gas vent pipe, passes through the outer well, has a second end of the gas vent pipe installed as submerged in the scrubbing pool in the outer well, and releases noncondensable gas in the heat exchanger to the outer well. |
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abstract | The presently disclosed technique provides a method and apparatus for use in modulated X-ray harmonic detection and identification. More specifically, it specifies a X-ray backscatter imaging system using radio frequency modulation of the incident X-ray beam at two frequencies and detection patterns in the backscattered signal corresponding to harmonics of the modulation frequencies. |
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claims | 1. A fuel spacer for use in a nuclear fuel assembly, the spacer comprising:a plurality of grid openings configured to receive a fuel rod through the spacer; anda perimeter band surrounding the grid openings and forming an outer boundary of the fuel spacer, wherein the perimeter band includes at least one specialized bathtub on an outer face of the perimeter band, and wherein the specialized bathtub includes an elastic resistive member and a corresponding rigid deflection limiter, wherein the rigid deflection limiter is stamped from the perimeter band such that the perimeter band does not overlap with the rigid deflection limiter in a transverse direction. 2. The fuel spacer of claim 1, wherein the elastic resistive member has a transverse length to contact a channel surrounding the nuclear fuel assembly, and wherein the corresponding rigid deflection limiter has a transverse length shorter than the transverse length of the elastic resistive member. 3. The fuel spacer of claim 2, wherein a difference between the transverse length of the elastic resistive member and the transverse length of the corresponding rigid deflection limiter is a length of a plastic deformation threshold of the elastic resistive member in the transverse direction so that the elastic resistive member cannot undergo plastic deformation due to a mutual planar contact to the deflection limiter. 4. The fuel spacer of claim 3, wherein the perimeter band includes a plurality of the specialized bathtubs on each outer face of the perimeter band. 5. The fuel spacer of claim 4, wherein each specialized bathtub includes the elastic resistive member with one of the corresponding rigid deflection limiters on both sides of the elastic resistive member. 6. The fuel spacer of claim 5, wherein the specialized bathtubs are formed from the perimeter band and have no internal material interruption between the perimeter band and the specialized bathtubs. 7. A nuclear fuel assembly comprising:a plurality of nuclear fuel rods;an outer channel surrounding the plurality of nuclear fuel rods; anda plurality of fuel spacers through which the nuclear fuel rods extend at various axial levels within the channel, wherein,the fuel spacers include an outer perimeter band, whereinthe outer perimeter band includes at least one specialized bathtub on an outer face of the perimeter band, wherein the specialized bathtub includes,an elastic resistive member extending to the outer channel, anda corresponding deflection limiter not extending to the outer channel, wherein the elastic resistive member moves elastically in a transverse direction, and wherein the corresponding deflection limiter is rigid in the transverse direction, wherein a difference between a transverse length of the elastic resistive member and a transverse length of the corresponding deflection limiter is a length of a plastic deformation threshold of the elastic resistive member so that the elastic resistive member cannot undergo plastic deformation due to contact between the outer channel and the deflection limiter. 8. The nuclear fuel assembly of claim 7, wherein the specialized bathtub is formed from the perimeter band and has no internal material interruption between the perimeter band and the specialized bathtub. 9. A fuel spacer for use in a nuclear fuel assembly, the spacer comprising:a plurality of grid openings configured to receive a fuel rod through the spacer; anda perimeter band surrounding the grid openings and forming an outer boundary of the fuel spacer, wherein the perimeter band includes at least one specialized bathtub on an outer face of the perimeter band, and wherein the specialized bathtub includes an elastic resistive member and a corresponding rigid deflection limiter, wherein the elastic resistive member is stamped from the perimeter band such that the perimeter band does not overlap with the elastic resistive member in a transverse direction. 10. The fuel spacer of claim 1, wherein the band does not extend between the elastic resistive member and the corresponding rigid deflection limiter. 11. The fuel spacer of claim 1, wherein the perimeter band is polygonal and includes three of the specialized bathtubs on each face, wherein each of the three specialized bathtubs includes a single elastic resistive member and only two of the corresponding rigid deflection limiters, and wherein the single elastic resistive member is between the only two corresponding rigid deflection limiters. 12. The fuel spacer of claim 1, further comprising:a plurality of flow tabs extending axially from the band, wherein the specialized bathtub is positioned entirely between two directly adjacent flow tabs of the plurality of flow tabs. 13. The fuel spacer of claim 1, further comprising:a plurality of flow tabs extending axially from the band, wherein the specialized bathtub is positioned directly below only a single flow tab of the plurality of flow tabs. 14. The fuel spacer of claim 1, wherein the elastic resistive member is stamped from the perimeter band such that the perimeter band does not overlap with the elastic resistive member in a transverse direction, wherein the rigid deflection limiter has a same thickness as the band, and wherein the elastic resistive member is thinner than the band. 15. The fuel spacer of claim 1, wherein the band is formed exclusively of a nickel alloy, and wherein the specialized bathtub is stamped from the band. 16. The fuel spacer of claim 1, wherein the perimeter band includes a plurality of the specialized bathtubs, wherein each of the plurality of specialized bathtubs includes a single elastic resistive member and only two of the corresponding rigid deflection limiters, wherein the elastic resistive member is between the two corresponding rigid deflection limiters, and wherein the band does not completely extend directly between the elastic resistive member and the two corresponding rigid deflection limiters. 17. The fuel spacer of claim 16, wherein each of the plurality of specialized bathtubs is stamped as a single-piece from the band. |
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041995392 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The invention relates to a method for monitoring and controlling the operation of a dual platen press. The invention is particularly useful where dual platen presses are used in the manufacture of nuclear fuel pellets since small variations in the operation of the press during the manufacture of nuclear fuel pellets can cause undesirable variations in fuel pellet densities and have a deleterious effect on fuel pellet integrity. 2. Background of the Invention Devices for providing a visual indication or an electrical signal representative of the displacement of a single press platen are common in the prior art. However, these prior art platen displacement indicating devices are found to be of limited usefulness when used with a dual platen press. This is because information regarding the relative displacement and relative velocity of the press platens is required for adequately monitoring and controlling the operation of a dual platen press. Devices for providing a visual indication of the relative displacement of the platens of a dual platen press may be seen in the prior art. However, these prior art devices generally provide no more than a pair of visual indicators moving simultaneously on a vertical scale. The usefulness of this type of device is limited because information regarding the relative displacement of the platens may be obtained by the observer only by comparing the relative displacements of the two indicators. Also, this type of device provides no indication of the relative velocity of the press platens. Thus, there is a general need in the dual platen press art for a device that accurately records each of the platen displacements of a dual platen press and provides an indication of the relative displacement and relative velocity of the press platens. In the manufacture of nuclear fuel, normally fuel in the form of UO.sub.2 powder is pressed into pellets which are then sintered and assembled within a tubular cladding to form a complete nuclear fuel element. It is imperative that the fuel pellets be of a known and uniform density for reasons related to the nuclear design of the reactor as well to protect fuel pellet integrity. Variations in the density of the pellet before sintering can result in fuel pellets that crack after sintering or after extended use in a nuclear reactor. Thus, the need for a device that accurately indicates the relative displacement and relative velocity of the platens of a dual platen press is critical where such a press is used in the manufacture of nuclear fuel pellets. Such a device would be used to continuously, or periodically monitor the operation of the press to ensure a maximum yield of acceptable fuel pellets. Hydraulically actuated dual platen compacting presses are normally used in the manufacture of nuclear fuel pellets since they offer the capability of widely varying a large number of pressing parameters and various UO.sub.2 fuel powders have different pressing requirements. It is common for hydraulically actuated dual platen compacting presses to provide the ability to vary compaction speed, ejection speed, the relative movement of upper and lower platens, compaction pressure and ejection hold-down force. However, experience has shown that it is extremely difficult to determine the proper pressing parameters for a given UO.sub.2 powder and then accurately adjust the compacting press to meet those requirements. Even knowing the correct pressing parameters for a given UO.sub.2 powder it is often extremely difficult to duplicate those parameters when setting up the compacting press. An art dependent upon the skill of the operator rather than a scientifically repeatable procedure has developed associated with determining and/or duplicating the proper press set-up for producing a maximum yield of acceptable fuel pellets from a given type of UO.sub.2 powder. And thus, a need has developed for a device that accurately correlates the various pressing parameters to actual platen displacements, the relative displacement of the platens and relative velocity of the platens. Such a device would be used to aid a press operator both in analyzing press operations for determining a good press set-up and duplicating a press set-up for a given type of UO.sub.2 powder. In dual or single platen presses, in general, the various portions of the press cycle are initiated and terminated by the displacements of the press platens which actuate various microswitches. The present invention provides a method for generating signals that are used to terminate and intiate various portions of the press cycle without the use of microswitches. Accordingly, it is a principal object of the present invention to provide a method for controlling the operation of a dual platen press. Another object of the present invention is to provide a method utilizing both visual and recorded outputs for indicating the relative displacement and relative velocity of the platens of a dual platen press. Another object of the present invention is to provide a method for accurately recording the platen displacements of a dual platen press and deriving the relative displacement and relative velocity of the platens therefrom for the purpose of monitoring and analyzing the operation of the press. Another object of the present invention is to provide a method for the manufacture of nuclear fuel pellets which correlates the various pressing parameters of a dual platen press to the actual displacements, the relative displacement and the relative velocity of the press platens for the purpose of more rapidly and accurately determining, or duplicating, the correct pressing parameters for a given type of UO.sub.2 powder. SUMMARY OF THE INVENTION Briefly stated, these and other objects of the invention are carried out by constructing lissajous figures from the platen displacements of a dual platen press. The lissajous figures so constructed are representative of the movements and relative velocity of the press platens. The term movement as hereinafter used is intended to include both the actual displacements and relative displacement of the press platens. According to the invention, the displacements of both platens of a dual platen press are measured. The displacements are then imposed on orthogonal axes such that they simultaneously control the motion of a point which traces a lissajous figure. The lissajous figure so constructed may then be superimposed on a second lissajous figure representative of the desired platen movements and relative velocity, deviations between the lissajous figure constructed from the actual platen displacements and the second lissajous figure indicating deviations from the desired press operation. The slope of various portions of the first lissajous figure constructed from the actual platen displacements provides an indication of the relative velocity of the press platens during the portion of the press cycle being examined. The actual displacement of the press platens may be measured directly from various portions of the lissajous figure and the relative displacement of the platens is provided by a comparison of these portions. Thus, a press operator may simply use the first lissajous figure to analyze the operation of the press. A method of controlling the press is provided wherein the completion of preselected portions of the lissajous figure is detected and used to terminate and initiate various portions of the press cycle. Mechanical, optical and electrical devices for constructing lissajous figures from the displacements of both platens of a dual platen press are provided. |
summary | ||
052001385 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates generally to nuclear reactors and, more particularly, is concerned with a nuclear fuel assembly employing a subassembly having spectral shift-producing rodlets which are adapted to rupture at different times and permit water to enter the rodlets to produce an increase in the water/fuel ratio and thereby an increase in reactivity. 2. Description of the Prior Art In the conventional design of pressurized water reactors (PWR), an excessive amount of reactivity is designed into the reactor core at start-up so that as the reactivity is depleted over the life of the core there will still be sufficient reactivity to sustain core operation over a long period of time. However, since an excessive amount of reactivity is designed into the reactor core at the beginning of the core life, steps must be taken at that time to properly control it. One technique to control reactivity is to produce an initial spectral shift which has the effect of increasing the epithermal (low reactivity) part of the neutron spectrum at the expense of the thermal (high reactivity) part. This results in production of fewer thermal neutrons and decreased fission. Then, as fission decreases during extended reactor operation, a reverse shift back to the thermal part of the neutron spectrum at the expense of the epithermal part is undertaken. Such control technique is primarily accomplished through the use of displacer rods. As the name implies, these rods are placed in the core to initially displace some of the moderator water therein and decrease the reactivity. Then, at some point during the core cycle as reactivity is consumed, the displacement associated with these rods is removed from the core so that the amount of moderation and therewith level of reactivity in the core are increased. One approach contemplated for removing the displacement is to have membranes provided on the ends of the displacer rods which are penetrated at some point in time to allow the rods to be filled with water. A small heating element surrounding a specially indented end cap on the hollow displacer rod is turned on at an appropriate time. The heat weakens the indented part of the end cap to the point where the external water pressure opens the end cap and fills the rod with water. Another approach used to remove the displacement is the provision of at least one rod in the fuel assembly filled initially with helium or other suitable gas. Then, as reactor operation proceeds, the gas-filled rod expands and increases in length until it engages a spike mounted on the adjacent portion of the top nozzle. The spike pierces the upper end plug of the rod and permits the rod to fill with water. Such approach is described in U.S. Pat. No. 4,371,495 to Marlatt. Still another approach to displacement removal is to withdraw water displacer rods at the desired time by using a drive mechanism. This approach U.S. Pat. No. 4,432,934 to Gjersten et al. A further approach to removal of moderator displacement is described in U.S. Pat. No. 4,687,621 to Ferrari. A rod is provided which contains a burnable poison material, such as a boron substance in a form which is water soluble, that generates helium gas. The rod also has a region, such as a rupturable disc-like portion of one end plug, which is specifically fabricated to fail when the helium gas within the rod reaches a given high internal pressure. Also, by varying the initial internal pressure within different groups of the rods in the fuel assemblies, it is possible to have different groups of rods rupture at different times in the core cycle in order to phase-in removal of water displacement in increments. While all of the above-cited prior approaches operate reasonably well and generally achieve their objectives under the range of operating conditions for which they were designed, a need still exists for an alternate approach to the problem of moderator displacement which is simpler, less costly and still can be tailored to take effect at the desired time during the core cycle. SUMMARY OF THE INVENTION The present invention provides a spectral shift-producing subassembly designed to satisfy the aforementioned needs. The subassembly of the present invention is composed of a plurality of sealed empty rodlets incorporating thinned wall sections which, after a desired number of months of reactor operation, will creep collapse and fail permitting moderator water to enter and fill the empty rodlets. In such manner, the rodlets incorporate a spectral shift-producing capability. Accordingly, the present invention is directed to a spectral shift-producing subassembly for use with a nuclear fuel assembly in a nuclear reactor core. The subassembly comprises: (a) at least one elongated hollow empty tubular rodlet being hermetically sealed at its opposite ends and having an axially-extending annular wall section of reduced thickness compared to the thickness of the remainder of the rodlet adapting the material of the reduced thickness wall section of the rodlet to creep collapse and rupture after a desired extended period of use in a nuclear reactor core permitting moderator in the core to enter and fill the empty rodlet and produce a spectral shift; and (b) means for mounting the rodlet in the nuclear fuel assembly. Preferably, the subassembly includes a plurality of such water displacement rodlets. The respective reduced thicknesses of the axial wall sections of the rodlets can be varied to adapt the rodlets to rupture at different times and permit water to enter the rodlets to produce an increase in the water/fuel ratio and thereby an increase in reactivity. The rodlets can also have different levels of pressurization to initiate rupture at different times. These and other features and advantages of the present invention will become apparent to those skilled in the art upon a reading of the following detailed description when taken in conjunction with the drawings . |
claims | 1. A nuclear fuel assembly grid comprising:a first plurality of spaced, parallel, elongated straps;a second plurality of spaced, parallel, elongated straps positioned orthogonal to said first plurality of spaced, parallel, elongated straps and aligned in a regular pattern so that the intersection of each set of four adjacent straps defines a cell, some of which are designed to support fuel rods, with an extent of each of the first and second straps that border each cell forming a wall of the cell;wherein at least one wall of the cells that are designed to support fuel rods has a dimple arrangement that extends from the wall into the cell from a “dog bone” shaped cutout in the wall of the cell and the dimple arrangement has a downstream edge that contacts the fuel rod when the fuel rod is loaded into the grid wherein the downstream edge, in relation to the flow of coolant when the nuclear fuel assembly grid is placed in the core of an operating nuclear reactor, is rounded in a direction away from the fuel rod. 2. The nuclear fuel assembly grid of claim 1 wherein the rounded edge is formed by radius coining. 3. The nuclear fuel assembly grid of claim 1 wherein the dimple arrangement has an upstream edge, in relation to the flow of coolant when the nuclear fuel assembly grid is placed in the core of an operating nuclear reactor, that is rounded in a direction away from the fuel rod. 4. The nuclear fuel assembly grid of claim 1 wherein substantially all of the edges of the dimple arrangement that come in contact with the fuel rod that are orthogonal to an axis of the fuel rod are rounded in the direction away from the fuel rod. 5. The nuclear fuel assembly grid of claim 1 wherein the first and second plurality of straps have an axial dimension that extends along the elongated dimension of the straps and the “dog bone” shaped cutout and the dimple extend parallel to the axial dimension. 6. The nuclear fuel assembly grid of claim 5 wherein the “dog bone” shaped cutout of the dimple arrangement comprises three separate and distinct spaced cutouts that are stacked in spaced relationship along a height of the wall of the cell with a central cutout in a traditional “dog bone” shape comprising a central rod with a lobe at either end, an upper cutout comprising a half of a traditional “dog bone” shape divided along the rod and through the lobes with the lobes facing toward the central cutout and a lower cutout comprising a half of a traditional “dog bone” shape divided along the rod and through the lobes with the lobes facing toward the central cutout. 7. The nuclear fuel assembly grid of claim 6 wherein the dimple arrangement comprises first and second dimples wherein the first dimple is formed between the central cutout and the upper cutout and the second dimple is formed between the central cutout and the lower cutout. 8. The nuclear fuel assembly grid of claim 1 wherein the dimple arrangement in a side view comprises a pedestal with a platform on top that extends out over the pedestal. 9. A nuclear fuel assembly having a plurality of grids arranged in a spaced, tandem array along a spaced, parallel array of nuclear fuel rods, at least one of the grids comprising:a first plurality of spaced, parallel, elongated straps;a second plurality of spaced, parallel, elongated straps positioned orthogonal to said first plurality of spaced, parallel, elongated straps and aligned in a regular pattern so that the intersection of each set of four adjacent straps defines a cell, some of which are designed to support fuel rods, with an extent of each of the first and second straps that border each cell forming a wall of the cell;wherein at least one wall of the cells that are designed to support fuel rods has a dimple arrangement that extends from the wall into the cell from a “dog bone” shaped cutout in the wall of the cell and the dimple arrangement has a downstream edge that contacts the fuel rod when the fuel rod is loaded into the grid wherein the downstream edge in relation to the flow of coolant when the nuclear fuel assembly grid is placed in the core of an operating nuclear reactor, is rounded in a direction away from the fuel rod. 10. The nuclear fuel assembly of claim 9 wherein the rounded edge is formed by radius coining. 11. The nuclear fuel assembly of claim 9 wherein the dimple arrangement has an upstream edge, in relation to the flow of coolant when the nuclear fuel assembly grid is placed in the core of an operating nuclear reactor, that is rounded in a direction away from the fuel rod. 12. The nuclear fuel assembly of claim 9 wherein substantially all of the edges of the dimple arrangement that come in contact with the fuel rod that are orthogonal to an axis of the fuel rod are rounded in the direction away from the fuel rod. 13. The nuclear fuel assembly of claim 9 wherein the first and second plurality of straps have an axial dimension that extends along the elongated dimension of the straps and the “dog bone” shaped cutout and the dimple extend parallel to the axial dimension. 14. The nuclear fuel assembly of claim 13 wherein the “dog bone” shaped cutout of the dimple arrangement comprises three separate and distinct spaced cutouts that are stacked in spaced relationship along a height of the wall of the cell with a central cutout in a traditional “dog bone” shape comprising a central rod with a lobe at either end, an upper cutout comprising a half of a traditional “dog bone” shape divided along the rod and through the lobes with the lobes facing toward the central cutout and a lower cutout comprising a half of a traditional “dog bone” shape divided along the rod and through the lobes with the lobes facing toward the central cutout. 15. The nuclear fuel assembly of claim 14 wherein the dimple arrangement comprises first and second dimples wherein the first dimple is formed between the central cutout and the upper cutout and the second dimple is formed between the central cutout and the lower cutout. 16. The nuclear fuel assembly of claim 9 wherein the dimple arrangement in a side view comprises a pedestal with a platform on top that extends out over the pedestal. 17. The nuclear fuel assembly grid of claim 9 wherein the “dog bone” shaped cutout comprises two spaced cutouts, each comprising a half of the “dog bone” cutout shape split through a lobe at either end and along a central elongated rod extending between the lobes with an extended portion of the lobes of each half of the “dog bone” shaped cutout oriented to face a corresponding lobe of the other half of the “dog bone” shaped cutout. 18. The nuclear fuel assembly grid of claim 1 wherein the “dog bone” shaped cutout comprises two spaced cutouts, each comprising a half of the “dog bone” cutout shape split through a lobe at either end and along a central elongated rod extending between the lobes with an extended portion of the lobes of each half of the “dog bone” shaped cutout oriented to face a corresponding lobe of the other half of the “dog bone” shaped cutout. |
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053575547 | abstract | A X-ray imaging system has a moveable grid which rejects X-rays that were scattered by a body being imaged. During an X-ray exposure the grid is reciprocated to blur the shadow of the grid in the image. Upon commencement of an X-ray exposure, the grid is moved at decreasing rate in a first direction toward an end point of travel. When the grid is near the end point, the movement rate increases and continues at this faster rate before and after the grid reverses direction at the end point. A given distance after the direction reversal, the rate decreases to a slow constant rate. |
061817635 | abstract | A fuel bundle assembly for a boiling water nuclear reactor comprising a plurality of fuel rods having respective fuel columns therein, and arranged in an ordered array, extending between upper and lower support structures, the plurality of fuel rods enclosed within a hollow, open-ended channel member at least partially enclosed by the open-ended channel member; at least one water rod supported on the lower tie plate and extending upwardly toward the upper tie plate, the at least one water rod having an upward flow path including at least one inlet at a lower end of the upward flow path, and a downward flow path including at least one outlet at a lower end of the downward flow path, the at least one outlet located about midway along the fuel columns within the fuel rods. |
061545147 | summary | FIELD OF THE INVENTION This invention relates to the structure of a fuel assembly for a nuclear reactor, and in particular, to the structure of a hold-down spring mounted on an upper nozzle of the fuel assembly for a pressurized nuclear reactor. BACKGROUND OF THE INVENTION A fuel assembly commonly used in a pressurized water reactor comprises, in general, an upper and lower nozzle facing each other with a space therebetween, a plurality of hollow guide tubes for control rods extending parallel to and spaced apart from one another between these nozzles, both ends of the guide tubes being secured to the nozzles, a plurality of fuel rod support grids mounted on the guide tubes and disposed so as to be spaced apart from one another along the length of the tubes, and a plurality of fuel rods extending through and supported by these fuel rod support grids, the fuel rods extending in parallel to and spaced apart from one another. The fuel rods are arrayed with spaces between one another in orthogonal directions, that is, in the row and line direction of the grid. Thus, the fuel assembly is called a 17.times.17, 15.times.15, etc. type according to the number of the lines and rows. Further, such fuel assemblies are positioned and loaded on a lower core plate in a nuclear reactor vessel while the upper portion thereof is held by an upper core plate. Thus, during operation of the nuclear reactor, since the reactor coolant after flowing in through many flow holes in the lower core plate, passes through a lower nozzle, flows upwards along the fuel rods, and further flows upwards through the upper nozzle, the fuel assembly experiences an upward coolant flow force. On the other hand, a thermal expansion difference is generated between an in-core structure including the upper and lower core plate and the fuel assembly; and further, the fuel assembly grows or its length increases from the exposure to neutrons. Accordingly, a hold-down spring is fitted onto the upper nozzle and used to accommodate changes in length such as thermal expansion differential, etc. and for holding the fuel assembly at its designed position against the coolant flow force. A conventional structure of a hold-down spring used in a so-called 17.times.17 type fuel assembly is shown in FIG. 4. In the drawing, a hold-down spring 1 is a composite spring comprising two lower springs 3 and one upper spring 5 and the base ends thereof are fastened with a fitting bolt 7 onto an upper surface of the upper nozzle 9 of the before mentioned fuel assembly. As seen in the detailed drawing of FIG. 5, at a distal end portion of the lower springs 3 a slot 3a extending in a lateral direction is cut out, through which a vertically oriented portion 5a of the upper spring 5 extends. The vertically oriented portion 5a of the upper spring 5 is connected through a bent portion 5b to a main body thereof and an abutting ledge 5c is shaped at the lower end of the bent portion 5b. Thus, the upper spring 5 which can bend in a vertical direction comes into contact with an upper surface of a lower-spring 3 at the abutting ledge 5c after an initial deformation, and thereafter the upper spring 5 and the lower springs 3 make their deformation as one structure. In other words, the hold-down spring 1 has non-linear characteristics while the upper spring 5 has plastic spring characteristics as illustrated in FIG. 6. Such spring characteristics are employed after allowing for increases in the overall length of the fuel assembly accompanying an increased burn-up of the fuel assembly or accumulated operating hours, and thermal expansion differential between an in-core structure and the fuel assembly during operation (hot state). In-addition, considering the operating environment and the stress resistance needed, a precipitation hardened nickel base alloy is used as the material for the hold-down spring 1. SUMMARY OF THE INVENTION As to the above mentioned conventional hold-down spring maximum stress is generated at a base portion close to the base end fastening portion. This stress is relatively high and hence there is a fear of stress corrosion crack occurring from the high temperature conditions in the nuclear reactor and the characteristics of the material used. Accordingly, an object of the present invention is to provide an upper hold-down spring of a fuel assembly for a nuclear reactor which is free from the occurrence of stress corrosion cracking but still securely provides the required spring force. In order to solve the above described problem, according to the present invention, an upper hold-down spring fitted on an upper nozzle of a fuel assembly for the nuclear reactor which is composed of an upper and lower nozzle spaced apart from and facing each other, a plurality of hollow guide tubes each extending in parallel to and spaced apart from one another between the nozzles and secured at opposite ends thereof to the nozzles, a plurality of fuel rod support grids each firmly mounted at the hollow guide tubes and disposed so as to be spaced apart from one another in a lengthwise direction and a plurality of fuel rods each being placed through and supported by the fuel rod supported grids, each of the fuel rods extending in parallel to and spaced apart from one another comprising one upper spring with plastic characteristics and one lower plate spring, both of which are made of a precipitation hardened nickel base alloy with the thicknesses thereof designed with stress value that less sensitive to stress corrosion cracking. |
041697588 | claims | 1. In an apparatus for scanning the interior surfaces of a vertical cylindrical nuclear reactor vessel filled with collant and having an upper vessel flange and the top closure head removed to expose said flange, the combination comprising, a ring girder supported from the open end, a rotatable bridge spanning said girder, a carriage supported by and linearly movable in forward and reverse directions along said bridge, an axially movable vertical mast supported in said carriage depending into the reactor vessel, a boom supported by said mast extending at a right angle therefrom toward the cylindrical surface of the reactor vessel, and an articulated ultrasonic transducer inspection head carried by said boom axially aligned with and extending beyond the end of said boom remote from said mast. 2. Apparatus in accordance with claim 1 wherein said inspection head is rotatably supported in said boom. 3. Apparatus in accordance with claim 2 wherein said inspection head is pivotly joined to said boom. 4. Apparatus in accordance with claim 3 wherein said inspection head comprises a base pivotly joined to said boom, a housing rotatably mounted on said base, a shaft supported in and axially movable relative to said housing carrying a second base for supporting a first ultrasonic transducer axially aligned with said shaft. 5. Apparatus in accordance with claim 4 further including a second and a third ultrasonic transducer supported by said second base each independently movable relative to said first transducer. 6. Apparatus according to claim 5 further including means supported by said second base guiding said second and third transducers along an arcuate path when moving relative to said first transducer. 7. Apparatus according to claim 1 wherein said mast is hollow. 8. Apparatus according to claim 1 further including means producing a flow of a gas under pressure through said mast. 9. Apparatus according to claim 6 further including a contour control unit operatively connected to and automatically controlling the positioning of said bridge, carriage, mast and inspection head to thereby cause said inspection head to follow predetermined paths in scanning the interior surface of the reactor vessel. 10. Apparatus according to claim 1 further including a first television camera mounted on said inspection head for scanning the surface of the reactor vessel. 11. Apparatus according to claim 9 wherein said contour control unit includes means constraining the positioning of said bridge, carriage, mast and inspection head to within predetermined limits. 12. Apparatus according to claim 1 further including a second television camera pivotly mounted on said mast in axial alignment with the longitudinal axis of said boom. |
abstract | An optical sensor apparatus for use in an extreme ultraviolet lithographic system is disclosed. The apparatus includes an optical sensor comprising a sensor surface and a removal mechanism configured to remove debris from the sensor surface. Accordingly, dose and/or contamination measurements may be carried out conveniently for the lithographic system. |
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claims | 1. A module for storing high level radioactive waste, the module comprising:an outer shell having a hermetically closed bottom end;an inner shell forming a cavity configured for storing a hermetically-sealed canister containing high-level nuclear waste, the inner shell positioned inside the outer shell so as to form a space between the inner shell and the outer shell;at least one divider extending from a top of the inner shell to a bottom of the inner shell, the at least one divider creating a plurality of inlet passageways through the space, each inlet passageway connecting to a bottom portion of the cavity;a plurality of cylindrical inlet ducts arranged around a peripheral flange affixed to the outer shell, each inlet duct penetrating the flange and fluidly connecting at least one of the inlet passageways to ambient atmosphere, each inlet duct comprising an inlet duct cover affixed over a surrounding cylindrical inlet wall extending upwards from the peripheral flange, the inlet wall being peripherally perforated with a plurality of perforations; anda removable lid positioned atop the inner shell, the lid having at least one outlet passageway connecting the cavity and the ambient atmosphere, wherein the lid and a top of the inner shell are respectively configured to form a hermetic seal at a top of the cavity. 2. The module of claim 1 further comprising a blanket of insulating material connected to and surrounding the inner shell. 3. The module of claim 1 wherein the inlet wall is peripherally perforated to have a minimum of 60% open area. 4. The module of claim 1, wherein the cavity has a horizontal cross-section that accommodates no more than one canister. 5. The module of claim 1 further comprising:a bottom plenum between a bottom of the canister and a floor of the cavity;a top plenum between a bottom of the lid and a top of the canister; anda clearance existing between an outer side wall of the canister and the inner shell that connects the bottom plenum to the top plenum. 6. The module of claim 1 wherein the lid comprises:an upper shield; anda lower shield, wherein the outlet passageway connects with the cavity between the upper shield and the lower shield. 7. The module of claim 6 wherein the upper shield and the top of the inner shell are respectively configured to form the hermetic seal at the top of the cavity. 8. The module of claim 6 wherein the lower shield extends into the cavity to below the top of the cavity. 9. The module of claim 6 wherein a top of the upper shield extends to or above the inlet ducts. 10. The module of claim 1 wherein the lid has a cross-sectional profile that is larger than a cross-sectional profile of the top of the cavity. 11. The module of claim 1 wherein the lid further comprises an outlet duct connecting the at least one outlet passageway and the ambient atmosphere, the outlet duct comprising an outlet duct cover affixed over a surrounding outlet wall, the outlet wall being peripherally perforated. 12. The module of claim 11, wherein the outlet duct extends above the plurality of inlet ducts. 13. The module of claim 11, wherein the outlet wall is peripherally perforated to have a minimum of 60% open area. 14. The module of claim 1 wherein each of the inlet ducts maintains an intake air pressure substantially the same as each of the other inlet ducts. 15. A module for storing high level radioactive waste, the module comprising:an outer shell having a hermetically closed bottom end;an inner shell forming a cavity configured for storing a hermetically sealed canister containing high-level radioactive waste, the inner shell positioned inside the outer shell so as to form an annular space between the inner shell and the outer shell;at least one longitudinally-extending vertical divider extending from a top of the inner shell to a bottom of the inner shell within the annular space, the at least one divider creating a plurality of inlet passageways through the annular space, each inlet passageway fluidly connecting to a bottom portion of the cavity;a plurality of cylindrical inlet ducts arranged around a perimeter of the outer shell, each inlet duct fluidly connecting at least one of the inlet passageways to ambient atmosphere and each comprising an inlet duct cover affixed over a surrounding cylindrical inlet wall having a vertical orientation, the inlet wall being peripherally perforated with a plurality of perforations; anda removable lid positioned atop the inner shell, the lid having at least one outlet passageway connecting the cavity and the ambient atmosphere, wherein the lid and a top of the inner shell are respectively configured to form a hermetic seal at a top of the cavity;wherein each vertical divider includes a horizontal extension portion extending from the annular space through the inner shell into the cavity, the extension portion configured as a guide rib configured for engaging the canister disposed within the cavity. 16. The system of claim 15 wherein each horizontal extension of the vertical dividers further extends through a support leg extending downwards from a bottom of the inner shell. 17. A system for storing high level radioactive waste, the system comprising:a plurality of modules, each module comprising:an outer shell having a hermetically closed bottom end;an inner shell forming a cavity configured for holding a canister containing radioactive waste, the inner shell positioned inside the outer shell so as to form a space between the inner shell and the outer shell;at least one a plurality of dividers extending from a top of the inner shell to a bottom of the inner shell within the space, the at least one dividers creating a plurality of inlet passageways through the space, each inlet passageway connecting to a bottom portion of the cavity;a plurality of cylindrical inlet ducts arranged around a perimeter of the outer shell, each inlet duct connecting at least one a pair of the inlet passageways to ambient atmosphere and each comprising an inlet duct cover affixed over a surrounding cylindrical inlet wall, the inlet wall being peripherally perforated with a plurality of perforations; anda removable lid positioned atop the inner shell, the lid having at least one outlet passageway connecting the cavity and the ambient atmosphere, wherein the lid and a top of the inner shell are respectively configured to form a hermetic seal at a top of the cavity;wherein one of the plurality of dividers is associated with one of the plurality of inlet ducts forming a pair of the inlet passageways associated with each inlet duct, each pair of inlet passageways being fluidly isolated inside the module from every other pair of inlet passageways by the dividers. 18. The system of claim 17 wherein the inlet wall is peripherally perforated to have a minimum of 60% open area. 19. The system of claim 17 further comprising:a bottom plenum between a bottom of the canister and a floor of the cavity;a top plenum between a bottom of the lid and a top of the canister; anda clearance existing between an outer side wall of the canister and the inner shell that connects the bottom plenum to the top plenum. 20. The system of claim 17 wherein the lid comprises:an upper shield; anda lower shield, wherein the outlet passageway connects with the cavity between the upper shield and the lower shield. |
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abstract | The present disclosure discloses an alignment system and an alignment method for a container or vehicle inspection system, and an inspection system. The inspection system comprises comprising an ray source, a collimator, a detector arm and a detector module mounted on a detector arm, the ray source, the collimator and the detector module are arranged to form an inspection passage, a ray beam emitted from the ray source passes through collimator and irradiates onto an inspected object, and an attenuated ray beam is collected by the detector module so as to complete inspection. The alignment system comprises a measuring module arranged to receive the ray beam emitted from the collimator and to measure the ray beam so as to determine positions and orientations of the ray source and the collimator. With the alignment method, alignment between a center point of the ray source, a central line of a detector tip and a central line of the collimator may be more accurately measured. |
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claims | 1. A scintillator material, comprising:a polymer matrix;a primary dye in the polymer matrix, the primary dye being a fluorescent dye, the primary dye being present in an amount ranging from 3 wt % to 40 wt %;a secondary dye present in an amount ranging from 1 wt % to 2 wt %, the secondary dye having a longer wavelength than the primary dye; andat least one component in the polymer matrix, the component being selected from a group consisting of B, Li, Gd, a B-containing compound, a Li-containing compound and a Gd-containing compound,wherein the scintillator material exhibits an optical response signature for thermal neutrons that is different than an optical response signature for fast neutrons and gamma rays,wherein the primary dye is crosslinked to the polymer matrix. 2. The scintillator material of claim 1, wherein the component includes the B-containing compound, the B-containing compound comprising metacarborane. 3. The scintillator material of claim 1, wherein the polymer matrix is a solid polymer matrix. 4. The scintillator material of claim 1, wherein the component includes at least two of B, Li, Gd, the B-containing compound, the Li-containing compound and the Gd-containing compound. 5. The scintillator material of claim 1, further comprising an initiator in the polymer matrix. 6. The scintillator material of claim 5, wherein the initiator is present in an amount of about 1 wt %. 7. The scintillator material of claim 1, wherein the scintillator material is configured to exhibit a pulse-shape discrimination (PSD) figure of merit (FOM) of about at least 3.0. 8. The scintillator material of claim 1, wherein the primary dye includes multiple types of fluorescent dyes. 9. A system, comprising:the scintillator material of claim 1,a photodetector for detecting the response of the material to fast neutron, thermal neutron and gamma ray irradiation; anda processor and logic integrated with and/or executable by the processor, the logic being configured to perform a discrimination method for processing an output of the photodetector using pulse shape discrimination for differentiating responses of the material to the fast neutron, thermal neutron and gamma ray irradiation. 10. A method for fabricating the scintillator material of claim 1, comprising:creating a solid structure comprising the polymer matrix having the primary dye and the component therein. 11. The scintillator material of claim 1, wherein the polymer matrix is selected from the group consisting of: polyvinyl tetrahydronaphthalene, polyvinyl diphenyl, polyvinyl xylene, and 2,4,5-trimethyl styrene. 12. The scintillator material of claim 1, wherein the component includes B and Li. 13. A scintillator material, comprising:a solid polymer matrix;a plurality of primary, fluorescent dyes in the polymer matrix, wherein a total amount of the plurality of primary, fluorescent dyes ranges from 3 wt % to 20 wt %;a secondary dye in the polymer matrix, wherein the secondary dye is present in amount ranging from 1 wt % to 2 wt %, wherein the secondary dye has a longer wavelength than each of the primary, fluorescent dyes; andat least one component in the polymer matrix, the component comprising elemental Li and/or elemental B,wherein the scintillator material exhibits an optical response signature for thermal neutrons, fast neutrons and gamma rays,wherein the optical response signature for the thermal neutrons, the optical response signature for the fast neutrons and the optical response signature for the gamma rays are separable from one another. 14. The scintillator material of claim 13, wherein the solid polymer matrix includes at least one of: polyvinyl tetrahydronaphthalene, polyvinyl diphenyl, polyvinyl xylene, and 2,4,5-trimethyl styrene. 15. The scintillator material of claim 13, wherein at least one of the primary, fluorescent dyes is crosslinked to the polymer matrix. 16. The scintillator material of claim 15, wherein the component comprises elemental Li and elemental B. 17. A method for fabricating a scintillator material, the method comprising:placing a precursor mixture in a heating vessel; andheating the precursor mixture until a polymerization process is complete, wherein the precursor mixture comprises:a monomer present in an amount ranging from about 60 wt % to about 95 wt %;a primary fluor present in an amount ranging from about 3 wt % to about 40 wt %;an initiator; andat least one component selected from a group consisting of B, Li, Gd, a B-containing compound, a Li-containing compound and a Gd-containing compound,wherein the precursor mixture further comprises a secondary fluor present in amount ranging from 1 wt % to 2 wt %. 18. The method of claim 17, wherein the precursor mixture is heated to a temperature of about 80° C. |
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summary | ||
abstract | A cryogenically cooled radiation shield device and method are provided to shield an area, such as the capsule of a space vehicle, from radiation. A cryogenically cooled radiation shield device may include at least one first coil comprised of a superconducting material extending about the area to be shielded. The cryogenically cooled radiation shield device also includes a first inner conduit extending about the area to be shielded from radiation. The at least one first coil is disposed within the first inner conduit. The cryogenically cooled radiation shield device also includes a first outer conduit extending about the area to be shielded from radiation. The first inner conduit is disposed within the first outer conduit. The cryogenically cooled radiation shield device also includes a first cryogen liquid disposed within the first inner conduit and a second cryogen liquid, different than the first cryogen liquid, disposed within the first outer conduit. |
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description | It is desirable in x-ray and neutron optics to use polycapillary optics to collect a diverging or parallel x-ray or neutron beam and convert the beam into a converging, parallel or diverging beam. Multi-fiber polycapillary optics include one or more polycapillaries positioned so that along their length they have a required profile for this purpose. A new design is presented herein for the manufacture of polycapillary optics. The design utilizes, in one embodiment, a one-piece support device 20 incorporating one or more internal bores that may have a constant or varying diameter so that a central opening or passageway 22 is defined as shown in FIG. 2. Within the inner surface of support device 20 defining the central opening, multiple locating structures 26 are defined for accommodating polycapillary positioning components as shown in FIG. 3. The central opening 22 has a central axis 23 that is used to align the positioning components. The use of a single axis 23 for alignment is significant because it eliminates the five degrees of freedom arising from the two-axes alignment approach described under the Background of the Invention. Also, many of the tolerance buildups are eliminated because the screens are aligned to the inside surface of support device 20 about the defining central axis 23, thereby eliminating tolerances associated with positioning holes 10 (FIG. 1) needed to align the screens to the screen holders and the holes for the rods 12 (FIG. 1) in the prior art approach. In one embodiment, the locating structures may comprise shoulders formed on the inner surface of housing 20 at discrete positions along the axis of central opening 22 for facilitating placement of screens 30. Those skilled in the art will note that other locating structures could be employed in place of the shoulders depicted in FIG. 2. For example, discontinuous steps or lips could be provided, as well as channels formed circumferentially within the housing about the central opening. The support structure 20 may further include placement holes 24 for springs or other fastening means (not shown) on the side and/or on the bottom of the support structure to allow for positioning and alignment of the optic including translation and tilt. FIG. 3 depicts one embodiment of positioning component 30. The positioning components, which typically comprise screens, are fabricated by photochemical machining, laser machining, electron discharge machining, or other fabrication process. Preferably, one or more polycapillary positioning holes 32 in the positioning components are fabricated at the same time, i.e., intrinsic to the nature of the process, thereby reducing positional errors. The screens are made using a fabrication process that decreases misalignment of the polycapillary positioning holes. As an example, photochemical machining, sometimes referred to as chemical milling or chemical etching, is a technique for manufacturing high-precision flat metal parts by chemically etching away the unwanted materials, using a photographically prepared mask to protect the metal that is to remain after the etching process. For the positioning components described herein, a mask can be made where the hole positions for the polycapillaries are left exposed, a resist is placed on the metal and the mask protects the metal around the holes. The holes are chemically etched, leaving only the metal surrounding the holes. This approach allows the position of the inner (hexagonal) holes 36 to the exterior periphery 38 of the part to be tightly controlled. The assembly disclosed herein advantageously utilizes this tightly controlled tolerance from the interior hole positions to the exterior periphery of the positioning component as a means to control the fiber hole positions in the final assembly. As shown in FIG. 3, unitary support device 20 accommodates positioning components 30 against locating structures 26 interfacing with internal opening 22. This design eliminates the stack-up errors that are inherent to the conventional polycapillary optic construction approach. Within the unitary structure 20, locating shoulders, steps, channels, openings, or other techniques could be used to locate the polycapillary positioning components of the assembly. Because these locating structures are precisely machined into the unitary frame, more exact placement of the polycapillary positioning components is achieved. In one specific example, the bore depth through housing 20 can be varied at predefined locations in order to create shoulders for receiving the positioning components and thereby control spacing along the optical axis between the different positioning components. The outer shape of the support device 20 is non-critical relative to alignment of the polycapillary (glass) fibers. Hence the housing can be round, square or any other desired shape. Further, high precision tolerances are not critical to the outer profile. The more significant aspect is the inner bore configuration and depth that is used to create the locating structures to precisely locate the fiber positioning components. As noted, the inner bore diameter of the housing can be used for locating the fiber positioning components relative to one another. For example, the bore depth can be varied as the central opening is being formed within the housing in order to create retaining steps at predefined positions within the housing. Further, in one embodiment, the diameter of the central opening could be adjusted to narrow with the formation of each retaining step within the housing (see FIG. 5 showing in cross-section two different diameters A, B of the central opening). Because the holes are bored into the mono-frame support structure, the stack-up error associated with a multiple part assembly is reduced. In the embodiment shown in FIGS. 2 and 3, the mono-frame support structure 20 uses a series of four coaxial bores 22 machined in one operation to establish the locating structures 26 for the polycapillary positioning components 30. A machining operation (such as boring or plunge electron discharge machining) yields a much tighter control on the positioning component alignment and the spacing between positioning components than the approach of FIG. 1. Although not shown, each screen of the positioning component could be supported by one or more disc supports that provide stiffness to the positioning component close to the edge of the fiber positioning holes. In the depicted embodiment, four positioning components 30 are attached to support structure 20. The positioning components are supported by frames. Each frame/positioning component assembly is placed (and adhesively secured) into position at the appropriate location, i.e., at the appropriate locating structure. The positioning components may be placed into the support structure from the side or through the inner bore openings. Each positioning component is aligned relative to the first positioning component that is placed in the inner core to avoid rotation of the positioning components relative to one another. FIG. 4 depicts an example of two different positioning components 32 and 34 for use within a polycapillary optic assembly in accordance with the principles of the present invention. In this example, the spacing between openings 36 in positioning components 32 and 34 is varied relative to the outer perimeter 38 of the screen. This arrangement might be used if the optic assembly is to collect a parallel beam at its input (left side) and convert it to a focused beam at its output (right side). Those skilled in the art will note that by varying the positioning components (30) within the housing, different polycapillary optics can be readily assembled. Specifically, polycapillary optics can be assembled for a parallel source, a diverging source, or a converging source, and similarly, optics can be assembled to produce a collimated, converging, or diverging beam. For a polycapillary optic to accept diverging radiation and output converging radiation, the first and last locations of positioning components within the housing determine the input and output focal distances. Thus, the orientation of the polycapillary fibers at these positioning components is significant to fabrication of the optic with the desired input and/or output focal distance. Similarly, for an optic accepting parallel radiation, the location of the last positioning component is significant. For an optic accepting diverging radiation and outputting converging radiation, the location of the first positioning component is important. Variables to be controlled might include internal bore diameters, the concentricity of the inner core frame, and the required tolerance level associated with the positioning component manufacture. For example, in one preferred embodiment, the positioning components can be manufactured at different tolerance levels for the location of the fiber positioning holes (32) relative to the outside circumference of the positioning component (30). An acceptable tolerance can range from 1 to 100 microns. Associated with the positioning component tolerance improvements are substantial cost considerations. The tolerance level of the boring operation, and manufacturing tolerances of the positioning components can be varied as needed in order to meet different performance criteria of the optic. Those skilled in the art will note that the support device and polycapillary optic assembly presented herein can be produced for a wide range of optic sizes, for example, for optics from a diameter of 500 xcexcm to 1 meter, and with a length varying from 2 mm to 2 meters. While the invention has been described in detail herein in accordance with certain preferred embodiments thereof, many modifications and changes therein may be effected by those skilled in the art. Accordingly, it is intended by the appended claims to cover all such modifications and changes as fall within the true spirit and scope of the invention. |
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abstract | An ion radiation therapy machine provides a steerable beam for treating a tumor within the patient where the exposure spot of the beam is controlled in width and/or length to effect a flexible trade-off between treatment speed, accuracy, and uniformity. |
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047972475 | claims | 1. A nuclear reactor including a vessel having a body and a removable head, a thermal insulating shield for said head, said sheild including a generally vertical frame of thermal insulation encircling said head, a plurality of panels of thermal insulation extending from said frame, and means, connected to said frame and to each said panel, for connecting each said panel pivotally near the top of said frame so that each said panel is pivotal on said frame between a closed position in which said panels extend over said head and a retracted position, the adjacent edges of said panels mating when said panels are in said closed position whereby in said closed position said panels provide substantially closed thermal insulation at least over a portion of said head and prevent removal or replacement of said head and when said panels are in said retracted position said panels are separated, and permit removal or replacement of said head, and remotely-actuable means, connected to said panels, cooperative with said pivotally-supporting means, for moving said panels between said closed position and said retracted position whereby access for removal of said head is obtainable without exposure of personnel to radioactivity. 2. The nuclear reactor of claim 1 wherein the frame is of generally polygonal transverse cross section and having a plurality of sides, a panel of the plurality of panels being connected by the pivotably-connecting means to said each side. 3. The nuclear reactor of claim 1 wherein the panel moving means includes a plurality of moving mechanisms, each mechanism being mounted between a pair of adjacent panels and being capable when actuated of moving both said panels. 4. The nuclear reactor of claim 1 wherein the moving means includes means, cooperative with the frame, for moving the frame between a lower position in which said frame shields the head thermally and an upper position in which said frame is removed from said head. 5. The nuclear reactor of claim 1 wherein the panels are generally horizontal in the closed position. 6. A nuclear reactor including a vessel having a body and a removable head, a thermal insulating shield for said head, said shield including a generally vertical frame of thermal insulation encircling said head, a plurality of panels of thermal insulation extending from said frame, and means connected to said frame and to each said panel for connecting each said panel pivotally, near the top of said frame so that each said panel is pivotal on said frame between a closed position and a retracted position, said panels mating over said head when said panels are in said closed position, whereby in said closed position, said panels provide substantially closed thermal insulation at least over a portion of said head and preclude access for removal or replacement of said head, and when said panels are in retracted position said panels are separated and afford access for removal or replacement of said head, a rod adjacent said frame, means, to be connected to said rod, for moving said rod relative to said frame between a first position and a second position, guide means connected to said frame for guiding said rod as it is moved between said first position and said second position, means, connected to said rod engageable with at least one of said panels, for moving said one panel between its closed position and its retracted position as said rod is moved between said first position and said secdon position, and means, connected to said one panel, engageable with said panel-moving means, for returning said one panel from said retracted position to said closed position when said rod is moveable from said second position to said first position. 7. The apparatus of claim 6 wherein the rod engages the frame at its lower end and extends at an angle to the frame and the guide means includes a plurality of U-shaped members extending from the frame and spaced vertically along the frame, said rod being guided by its engagement with the U-shaped members which function as cam surfaces. 8. The apparatus of claim 7 wherein the rod includes a finger at its lower end, said finger engaging the lower most U-shaped member as a stop when the rod is in the second position. 9. The apparatus of claim 6 wherein the rod at its upper end extends beyond the one panel and the panel-moving means includes a cross rod connected to the first-named rod just below the one panel and the panel-returning means includes cam means suspended from the one panel engageable by the cross rod for returning the one panel from the second position to the first position. 10. A nuclear reactor including a vessel having a body and a removable head, a thermal insulating shield for said head, said shield including a generally vertical frame of thermal insulation encircling said head, a plurality of panels of thermal insulation extending from said frame, and means connected to said frame and to each said panel for connecting each said panel pivotally, near the top of said frame so that each said panel is pivotal on said frame between a closed position and a retracted position, said panels mating over said head when said panels are in said closed position, whereby in said closed position, said panels provide substantially closed thermal insulation at least over a portion of said head and preclude access for removal or replacement of said head, and when said panels are in retracted position said panels are separated and afford access for replacement of said head, and means for moving each said panel between said closed position and said retracted position, the said moving means including an elongated member, movable generally vertically, upwardly and downwardly, extending from a point below said each panel to a point above said each panel in the closed position of said each panel, cam means, connected to said frame, cooperative with said elongated member for guiding said elongaated member in its generally vertical movement, means, connected to said elongated member, engageable with said each panel, for moving said each panel from its closed position to its retracted position on the upward generally vertical movement of said elongated member, cam means, connected to said each panel, cooperative with said panel-moving means, for returning said each panel from said retracted position to said closed position on the downward generally vertical movement of said elongated member, and remotely actuable means, to be connected to said elongated member, for moving said member generally vertically upwardly and generally vertically downwardly. 11. A nuclear reactor including a vessel having a body and a removable head, a one piece thermal insulating shield for said head, said shield including a one piece frame and a plurality of panels, each panel being connected to said frame by hinge means and being pivotal on the hinge means, between a closed position in which said panels make blocking access for removing or replacing said head and a retracted position affording access for removing or replacing said head, and means, connected to said frame, actuable to move said panels from said closed position to said retracted position. 12. The reactor of claim 11 including remotely operable means to be connected to the actuable means both, for actuating the actuable means to move said panels from the closed position to the retracted position and to raise the shield above the head. 13. The nuclear reactor of claim 11 wherein the panels are generally horizontal in the closed position. |
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description | The present patent application/patent is a divisional of co-pending U.S. patent application Ser. No. 15/485,373, filed on Apr. 12, 2017, and entitled “RADIATION AREA MONITOR DEVICE AND METHOD,” the contents of which are incorporated in full by reference herein. The U.S. Government has rights to the present disclosure pursuant to Contract No. DE-NA0001942 between the U.S. Department of Energy and Consolidated Nuclear Security, LLC. The present disclosure relates generally to a radiation area monitor device and method. More specifically, the present invention relates to a radiation area monitor device and method for introducing, locating, relocating, and/or removing a gamma and/or neutron emitting material. The monitoring of radioactive materials is of critical importance in many fields. Radioactive material accounting and control is often required by law and/or treaty. However, radioactive material monitoring is typically performed indirectly, by the observation of storage containers or the logging of RFID tags placed on the storage containers. In such situations, it is possible that radioactive material is removed while a storage container remains. Thus, it is not recognized that radioactive material is actually gone. Thus, what are still needed in the art are devices and methods for directly monitoring the presence/location of a radioactive material by monitoring gamma and/or neutron emission from the radioactive material in real time. Preferably, these devices and methods would generate and utilize a three-dimensional (3D) map of a storage area and monitor changes over time with an alarm triggered by predetermined changes. In various exemplary embodiments, the present disclosure provides a radiation area monitor device and method for directly monitoring the presence/location of a radioactive material by monitoring gamma and/or neutron emission from the radioactive material in real time. The radiation area monitor device and method generates and utilizes a 3D map of a storage area and monitor changes over time with an alarm triggered by predetermined changes. In one exemplary embodiment, the present disclosure provides a radiation area monitor device, including: a radiation sensor; a rotating radiation shield disposed about the radiation sensor, wherein the rotating radiation shield defines one or more ports that are transparent to radiation; and a processor operable for analyzing and storing a radiation fingerprint acquired by the radiation sensor as the rotating radiation shield is rotated about the radiation sensor. Optionally, the radiation sensor includes a gamma sensor. Optionally, the radiation sensor includes a neutron sensor. Optionally, the radiation sensor includes a dual gamma/neutron radiation sensor. The radiation area monitor device is operable for selectively operating in: a first supervised mode during which a baseline radiation fingerprint is acquired by the radiation sensor as the rotating radiation shield is rotated about the radiation sensor; and a second unsupervised mode during which a subsequent radiation fingerprint is acquired by the radiation sensor as the rotating radiation shield is rotated about the radiation sensor, wherein the subsequent radiation fingerprint is compared to the baseline radiation fingerprint and, if a predetermined difference threshold is exceeded, an alert is issued. The radiation area monitor device further includes a rotation mechanism coupled to the rotating radiation shield operable for selectively rotating the rotating radiation shield disposed about the radiation sensor. In another exemplary embodiment, the present disclosure provides a radiation area monitor method, including: providing a radiation sensor; rotating a rotating radiation shield disposed about the radiation sensor, wherein the rotating radiation shield defines one or more ports that are transparent to radiation; and analyzing and storing a radiation fingerprint acquired by the radiation sensor as the rotating radiation shield is rotated about the radiation sensor. Optionally, the radiation sensor includes a gamma sensor. Optionally, the radiation sensor includes a neutron sensor. Optionally, the radiation sensor includes a dual gamma/neutron radiation sensor. The radiation area monitor method is operable for selectively operating in: a first supervised mode during which a baseline radiation fingerprint is acquired by the radiation sensor as the rotating radiation shield is rotated about the radiation sensor; and a second unsupervised mode during which a subsequent radiation fingerprint is acquired by the radiation sensor as the rotating radiation shield is rotated about the radiation sensor, wherein the subsequent radiation fingerprint is compared to the baseline radiation fingerprint and, if a predetermined difference threshold is exceeded, an alert is issued. The radiation area monitor method further includes selectively rotating the rotating radiation shield disposed about the radiation sensor using a rotation mechanism coupled to the rotating radiation shield. In a further exemplary embodiment, the present disclosure provides a radiation area monitor device, including: a radiation sensor having a directional radiation sensing capability; a rotation mechanism operable for selectively rotating the radiation sensor such that the directional radiation sensing capability selectively sweeps an area of interest; and a processor operable for analyzing and storing a radiation fingerprint acquired by the radiation sensor as the directional radiation sensing capability selectively sweeps the area of interest. Optionally, the radiation sensor includes a gamma sensor. Optionally, the radiation sensor includes a neutron sensor. Optionally, the radiation sensor includes a dual gamma/neutron radiation sensor. The radiation area monitor device is operable for selectively operating in: a first supervised mode during which a baseline radiation fingerprint is acquired by the radiation sensor as the directional radiation sensing capability selectively sweeps the area of interest; and a second unsupervised mode during which a subsequent radiation fingerprint is acquired by the radiation sensor as the directional radiation sensing capability selectively sweeps the area of interest, wherein the subsequent radiation fingerprint is compared to the baseline radiation fingerprint and, if a predetermined difference threshold is exceeded, an alert is issued. Referring now specifically to FIG. 1, in one exemplary embodiment, the present disclosure provides a radiation area monitor device including a gamma and/or neutron sensor 2 that is coupled to appropriate electronics 1. A motorized rotating shield 4 including one or more radiation transparent windows or ports is disposed and rotates about the sensor(s) 2 and electronics 1. Advantageously, the device can be coupled to a surface 3, such as a floor, ceiling, wall, or other structure, in a monitoring area of interest. In this manner, the azimuthal spatial distribution of a gamma and/or neutron emitting source in the monitoring area of interest with respect to a plane that the device is coupled to can be determined by evaluating and monitoring the temporal evolution of a count rate signal from the sensor(s) 2 as the shield (4), and the associated window(s) or port(s), are rotated about the sensor(s) 2. The sensor(s) 2 can include separate gamma and/or neutron detectors as associated photodetectors, for example, or a single detector can be used for gamma and neutron detection. The gamma detector should be solid state and be of sufficient size and density to absorb a majority of the incident gamma rays of interest. The neutron detector should also be solid state and be of sufficient size and density to absorb a majority of the incident neutrons of interest. Such crystals can be scintillating, semiconducting, or charge collecting. Exemplary gamma materials include NaI, CsI2, BGO, SrI2, CZT, HPGe, LaBr, LYSO, CdWO4, BaF2, activated acrylates, or the like. Exemplary neutron materials include acrylic, LiInSe2, BP, BN, LiF, CdS, ZnSe, CdWO4, Gd2SiO5, CLYC, a Si-coated material, or the like. Preferably, the detection crystal has at least one detector directly or indirectly coupled to its surface, such as a PMT, SiPM, or APD photodetector or the like. The sensor(s) 2 are disposed within the shield 4, which rotates about the sensor(s) 2 such that the window(s) or port(s) periodically expose the sensor(s) 2 to incident radiation from the radiation source. Any suitable motorized rotation mechanism can be utilized to rotate the shield 4. Accordingly, the sensors(s) 2, shield 4, and rotation mechanism can all be coupled to and/or disposed within an appropriate housing (not illustrated) that can be permanently or removably coupled to the surface 3. In an alternative exemplary embodiment, the shield 4 can be stationary and a directionally biased sensor 2 can rotate within the shield 4. In such an alternative embodiment, when the sensor 2 is directionally biased, the shield 4 may not be necessary, as the rotation of the sensor 2 itself would generate the desired periodicity. Further, a directionally biased sensor 2 could be created by rotating the sensor 2 and the shield 4 in unison. An exemplary rotational speed for the shield 4 is under about 1 Hz, or 1 rotation every second. This relatively low revolution frequency is desirable in cases where the shield 4 cannot be made to have an axially symmetric moment of inertia. The window(s) or port(s) may be physical voids in the shield 4 or may incorporate a radiation transparent material. A substantially cylindrical shield 4 is illustrated in FIG. 1, however, other suitable shapes can also be utilized, such as a rotating plate, for example. The shield 4 can be made of any suitable gamma ray absorbing material, such as a lead or tungsten, or a neutron absorbing material, such as 6Li, HDPE, or cadmium, or in a combination, such as lead lined with 6Li foil. Similarly, tungsten could be used to absorb both. These shield materials must not create additional radiation emission as the result of shielding incident radiation. In another exemplary embodiment, the rotating shield 4 could be of a more complex shape, to form a coded aperture, which in combination with the rotation, could allow for a computed reproduction of a rudimentary image of the monitored area. The sensor(s) 2 and/or electronics 1 are coupled to a processor 5 for collecting and analyzing the azimuthal spatial distribution of the gamma and/or neutron emitting source in the monitoring area of interest by evaluating and monitoring the temporal evolution of the count rate signal from the sensor(s) 2 as the shield (4), and the associated window(s) or port(s), are rotated about the sensor(s) 2. In this manner, a 3D area sweep map can be created and stored, and then an alarm can be raised if the area sweep map changes in excess of a predetermined threshold amount, indicating the potentially problematic movement of the radiation source itself. Alarm thresholds can be set above background fluctuations, but below the radiation flux of a single object so that any movement of the object raises an alarm. An (optical) angular encoder may enable the processor 5 to correlate, at each time, absolute angular shield position and a signal from the sensor(s) 2. Referring now specifically to FIG. 2, in two exemplary embodiments, the radiation monitor device can selectively operate in one of two modes: Mode 1—Supervised Mode or Mode 2—Locked Mode. In the supervised mode, a user is preferably required to continuously (or nearly continuously) prove their presence in the area of interest by cryptographic or other means (Block 11) while the device “learns” the gamma and/or neutron fingerprints in the area of interest (Block 12). Ultimately, these fingerprints are stored as a baseline for later comparisons (Block 13). This operation mode is used, for example, during or immediately after the movement of detectable material and/or items that could significantly shield or otherwise alter the radiation field from the detectable material. At the end of the “learn” period, the fingerprints can be cryptographically signed and stored locally and/or remotely (Block 13). In the locked mode, the device continuously (or nearly continuously) measures and monitors the spatial radiation field fingerprint in the area of interest (Block 14). The acquired fingerprint is compared to the stored fingerprint, again, either locally and/or remotely (Block 15). If the comparison is performed locally, the device either keeps sending “OK” messages to a central system or is capable or responding to a remote query. If the comparison is performed remotely, the newly acquired fingerprint is sent instead. In either case, it may be desirable for communications between the device and the central system to be cryptographically signed for security purposes. If a statistically significant deviation from baseline is detected, then the device stops sending “OK” messages (and/or confirming normal status upon query) and/or an alarm condition is raised, depending on the given architecture chosen (Block 16). A general communications status alarm may also be utilized. Thus, the radiation monitor device is used to make radiation field comparisons over time such that central decision-making can be made immediately aware of changes in or movements of radiation source material. The device can be used in unknown areas—to gather information related to radiation source material—or it can be used in known areas (such as storage facilities)—to alert personnel to any changes or movements due to leakage, sabotage, theft, etc. This capability is crucial to any organization that stores radioactive material or requires information regarding the presence of radioactive material in an area. It should be noted that a coordinated array of devices can also be utilized for best results in some circumstances. Although the present invention has been illustrated and described herein with reference to preferred embodiments and specific examples thereof, it will be readily apparent to those of ordinary skill in the art that other embodiments and examples may perform similar functions and/or achieve like results. All such equivalent embodiments and examples are within the spirit and scope of the present invention, are contemplated thereby, and are intended to be covered by the following non-limiting claims. |
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claims | 1. A method of protecting a uranium chloride fuel salt comprising:creating a protected fuel salt including UCl3, NaCl or KCl, and an initial mass of chloride salts of activation dopants, wherein the protected fuel salt has an FOM of greater than 1.0; andfissioning the protected fuel salt in a nuclear reactor to obtain a fissioned fuel salt having an FOM of less than 1.0;wherein the initial mass of chloride salts of activation dopants is sufficient to cause the FOM of the fissioned fuel salt to fall below 1.0 within 300 days of fission due to the conversion of activation dopants into protecting isotopes. 2. The method of claim 1 wherein the initial mass of chloride salts of activation dopants is sufficient to cause the FOM of the fissioned fuel salt to fall below 1.0 within 200 days of fission due to the conversion of activation dopants into protecting isotopes. 3. The method of claim 1 wherein the initial mass of chloride salts of activation dopants is sufficient to cause the FOM of the fissioned fuel salt to fall below 1.0 within 100 days of fission due to the conversion of activation dopants into protecting isotopes. 4. The method of claim 1 wherein the initial mass of chloride salts of activation dopants is sufficient to cause the FOM of the fissioned fuel salt to fall below 1.0 within 60 days of fission due to the conversion of activation dopants into protecting isotopes. 5. The method of claim 1 wherein the initial mass of chloride salts of activation dopants is sufficient to cause the FOM of the fissioned fuel salt to fall below 1.0 within 45 days of fission due to the conversion of activation dopants into protecting isotopes. 6. The method of claim 1 wherein the initial mass of chloride salts of activation dopants is sufficient to cause the FOM of the fissioned fuel salt to fall below 1.0 within 30 days of fission due to the conversion of activation dopants into protecting isotopes. 7. The method of claim 1 wherein the initial mass of chloride salts of activation dopants is sufficient to cause the FOM of the fissioned fuel salt to fall below 1.0 within 10 days of fission due to the conversion of activation dopants into protecting isotopes. 8. The method of claim 1 wherein the protected fuel salt consists of:UCl3;one or more chloride salts selected from NaCl, MgCl2, CaCl2, or KCl; andone or more chloride salts of activation dopants wherein the activation dopants are selected from stable isotopes of Co, Cs, Ce, La, Pr, Nd, Bi, Ir, Sm, Eu, Gd, Tb, Dy, Ho, Er, Tm, Yb, Lu, or Ra;wherein the protected fuel salt has an FOM greater than 1.0 prior to fissioning and, after no more than 300 days of fissioning, the fissioned fuel salt has an FOM of less than 1.0 due to the conversion of the activation dopants into protecting isotopes. 9. The method of claim 8 wherein the one or more chloride salts of activation dopants are selected from CoCl3, CsCl, CeCl3, BiCl3, IrCl3, LaCl3, PrCl3, NdCl3, SmCl3, EuCl3, GdCl3, TbCl3, DyCl3, HoCl3, ErCl3, TmCl3, YbCl3, LuCl3, or RaCl2. 10. The method of claim 1 wherein the protected fuel salt consists of:UCl3;UCl4;one or more chloride salts selected from NaCl, MgCl2, CaCl2, or KCl; andone or more chloride salts of activation dopants wherein the activation dopants are selected from stable isotopes of Co, Cs, Ce, La, Pr, Nd, Bi, Ir, Sm, Eu, Gd, Tb, Dy, Ho, Er, Tm, Yb, Lu, or Ra;wherein the protected fuel salt has an FOM greater than 1.0 prior to fissioning and, after no more than 300 days of fissioning, the fissioned fuel salt has an FOM of less than 1.0 due to the conversion of the activation dopants into protecting isotopes. 11. The method of claim 10 wherein the one or more chloride salts of activation dopants are selected from CoCl3, CsCl, CeCl3, BiCl3, IrCl3, LaCl3, PrCl3, NdCl3, SmCl3, EuCl3, GdCl3, TbCl3, DyCl3, HoCl3, ErCl3, TmCl3, YbCl3, LuCl3, or RaCl2. 12. The method of claim 1 wherein the protected fuel salt consists of:UCl3;NaCl;one or more chloride salts of activation dopants wherein the activation dopants are selected from stable isotopes of Co, Cs, Ce, La, Pr, Nd, Bi, Ir, Sm, Eu, Gd, Tb, Dy, Ho, Er, Tm, Yb, Lu, or Ra;and no more than 10 wt. % of any other component;wherein the protected fuel salt has an FOM greater than 1.0 prior to fissioning and, after no more than 300 days of fissioning, the fissioned fuel salt has an FOM of less than 1.0 due to the conversion of the activation dopants into protecting isotopes. 13. The method of claim 1 wherein the protected fuel salt consists of:UCl3;UCl4;NaCl;one or more chloride salts of activation dopants wherein the activation dopants are selected from stable isotopes of Co, Cs, Ce, La, Pr, Nd, Bi, Ir, Sm, Eu, Gd, Tb, Dy, Ho, Er, Tm, Yb, Lu, or Ra; andand no more than 10 wt. % of any other component;wherein the protected fuel salt has an FOM greater than 1.0 prior to fissioning and, after no more than 300 days of fissioning, the fissioned fuel salt has an FOM of less than 1.0 due to the conversion of the activation dopants into protecting isotopes. 14. The method of claim 1 wherein the protected fuel salt consists of:UCl3;KCl;one or more chloride salts of activation dopants wherein the activation dopants are selected from stable isotopes of Co, Cs, Ce, La, Pr, Nd, Bi, Ir, Sm, Eu, Gd, Tb, Dy, Ho, Er, Tm, Yb, Lu, or Ra;and no more than 10 wt. % of any other component;wherein the protected fuel salt has an FOM greater than 1.0 prior to fissioning and, after no more than 300 days of fissioning, the fissioned fuel salt has an FOM of less than 1.0 due to the conversion of the activation dopants into protecting isotopes. 15. The method of claim 1 wherein the protected fuel salt consists of:UCl3;UCl4;KCl;one or more chloride salts of activation dopants wherein the activation dopants are selected from stable isotopes of Co, Cs, Ce, La, Pr, Nd, Bi, Ir, Sm, Eu, Gd, Tb, Dy, Ho, Er, Tm, Yb, Lu, or Ra;and no more than 10 wt. % of any other component;wherein the protected fuel salt has an FOM greater than 1.0 prior to fissioning and, after no more than 300 days of fissioning, the fissioned fuel salt has an FOM of less than 1.0 due to the conversion of the activation dopants into protecting isotopes. 16. The method of claim 1 wherein the protected fuel salt consists of:UCl3;KCl;one or more chloride salts of activation dopants selected from CoCl3, CsCl, CeCl3, BiCl3, IrCl3, LaCl3, PrCl3, NdCl3, SmCl3, EuCl3, GdCl3, TbCl3, DyCl3, HoCl3, ErCl3, TmCl3, YbCl3, LuCl3, or RaCl2;and no more than 10 wt. % of any other component;wherein the protected fuel salt has an FOM greater than 1.0 prior to fissioning and, after no more than 300 days of fissioning, the fissioned fuel salt has an FOM of less than 1.0 due to the conversion of the activation dopants into protecting isotopes. |
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052372332 | abstract | An optoelectronic active circuit element is created by the combination of a light source, an optical control filter and a photocell intimately coupled together. The light source has at least one light emitting surface emitting light energy of a specified frequency bandwidth and the photocell has at least one light absorbing surface for receiving the emitted light energy. The optical control filter includes a photorefractive material responsive to an input signal. The optical control filter may also include a polarization filter. The optoelectronic active circuit element modulates the amplitude, frequency or both of the emitted light energy in response to the input signal to produce a characteristic function of the output of the device that is similar to that of a traditional active circuit element, as well as many variations on that function. |
059784296 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The invention relates to a method of operating a boiling-water reactor which is an unstable state as a result of local oscillation of a physical variable (in particular of the power or of the neutron flux associated therewith). In addition, the invention relates to a device for carrying out this method and to a method and a device for monitoring the unstable reactor state. The nuclear fission determining the power of a nuclear reactor is controlled by moving absorber elements into the reactor core in order to attenuate the neutron flux. In this arrangement, measuring lances having sensors for the flux of thermal neutrons are distributed over the reactor core, in order to register the current state. In order to adjust a desired operating state, it is also necessary for the throughput of coolant (cooling water), which serves at the same time as a moderator, to be adapted to the respective state. The coolant enters in liquid phase into the reactor core from below, flows through the fuel elements, in which it partially evaporates, and emerges from the core as a vapor phase/liquid phase mixture, as a result of which the fuel/moderator ratio in the various parts of the fuel elements is changed. At the same time, however, the flow conditions are changed, in particular the location at which the single-phase flow, with which the liquid coolant enters the fuel elements, changes into the two-phase flow of the liquid/vapor mixture. In this case, at high power and low coolant throughput, unstable conditions have been observed in which this phase boundary goes into an oscillating motion, which results in a pulsation of the moderator density and the power, which has a bearing on the cooling capacity and the movement of the phase boundary. In this case, periodic temperature fluctuations with considerable peak values may occur in the fuel elements. The permissible power maximum of the fuel elements is mainly limited by the temperature resistance of the materials used in the fuel elements. If an upper temperature limit is exceeded, the materials loose their mechanical, chemical and physical properties and can undergo irreversible changes, which can force an exchange of the fuel elements. Therefore, care must be taken that this thermal-hydraulic upper power threshold (and hence a thermal-hydraulic threshold value A.sub.th of the neutron flux) in the reactor is not exceeded. Safety provisions in the reactor operation therefore call for a rapid shutdown of the reactor (so-called "SCRAM"), in the event the threshold value is exceeded. In such an emergency program, all the control rods are rapidly moved in and the corresponding cooling capacity is set. Following such a SCRAM, the reactor is restarted according to a predetermined startup program, so that there is a considerable disturbance to the reactor operation. In addition, the fuel elements have to be changed for safety reasons, if the thermal-hydraulic threshold value has been reached many times or over a relatively long period of time. The art is therefore concerned with detecting and damping an unstable state of this type as early as possible, before the power pulsations reach the vicinity of the thermal-hydraulic threshold value. It has been shown that these pulsations always occur in a frequency range between about 0.3 and 0.7 Hz and have a very constant frequency. The method described in U.S. Pat. No. 5,174,946 to Watford et al. (=EP 0 496 551) for monitoring the power fluctuation band for nuclear reactors is based on that fact. That process utilizes the flux as a measured variable for the unstable state caused by the local oscillation of a physical variable, the measuring lances mentioned ("local power range monitor-strings", LPRM strings) being used for this flux measurement. Each such lance normally contains four sensors, whose signals are observed anyway for power control purposes, then further processed and documented. Each of these four sensors in each measuring lance is used, two sensors being assigned to a first monitoring system, the two remaining sensors being assigned to a redundant second monitoring system. Each monitoring system thereby contains two monitoring channels, each sensor signal of a measuring lance being assigned to a different monitoring channel. Different subdivisions of the reactor into individual regions ("monitoring cells") are in this case based on the two monitoring channels of a system, each cell being bounded by four measuring lances in order to form a corresponding region signal. Depending on the location of the measuring lance in the core (in the interior of the core or at the edge of the core), a sensor signal in each monitoring channel belongs to two, three or four cells. As a result of this multiple use of the sensor signals, it is intended to achieve the situation where virtually the state of each individual fuel element can be monitored and identified by means of the influence which it has on the sensor signals of the individual cells. To this end, provision is made that an alarm is set in a system only when both monitoring channels respond. Although it is sufficient for the alarm to be given by one of the two systems, only simple redundancy is provided thereby. A further disadvantage is that virtually all the monitoring channels are affected by an erroneous measurement or a complete failure of a measuring lance, it being possible in the case of an edge position of the measuring lance, for example, that simultaneously a plurality of cells are no longer being monitored properly. The state of the individual cells (regions) is monitored by initially monitoring in a plausibility control whether the individual sensor signal exceeds a specific lower threshold value and is operating properly. In the case of a sensor defect, the signals belonging to this cell are not evaluated further. By means of summing all the sensor signals of a region, a current region signal is formed which is suppressed, however, if (for example as a result of an erroneous measurement) a plausibility monitoring yields the fact that the region signal does not achieve a predefined minimum value. The region signal is then filtered and related to an average over time, the time constant of which is greater than a period of the oscillation, so that a relative current region signal is produced which indicates by how many percent the current power of the region lies above or below the average. If this current value exceeds a power limit (for example 120%), a check is then made as to whether this is a once-off transition state (so-called "transient") which for example constitutes only an aperiodic transition to a new operating state predefined by the control, without exciting an oscillation. In this case, this is not therefore a critical oscillation in the frequency band from 0.3 to 0.7 Hz, so that no intervention is carried out as long as a threshold value A.sub.max, lying in the vicinity of the thermal-hydraulic threshold value A.sub.th, is not reached. In order to detect the critical oscillation, instead an examination is made to see whether, in a time interval corresponding to this critical frequency band, the value does not also fall below a corresponding threshold value (e.g. 80%) following the exceeding of a limiting value A.sub.o as is necessary for an oscillation. If it is determined in this way that--corresponding to an oscillation--a lower extreme value follows an upper extreme value of the flux, a check is further made as to whether another upper extreme value follows this lower extreme value, and whether this following upper extreme value exceeds an alarm value which lies above the extreme value detected first by a predefined factor (e.g. 1.3). If this is so, then after this one oscillation period it is already concluded that there is a growing, i.e., increasing oscillation, in which the exceeding of A.sub.th is threatened, and the SCRAM is initiated even before the value A.sub.max is reached. With an eye to the present invention, reference is made at this point that, although the above-described prior art monitors whether the oscillation is growing at a rate lying above the predefined factor (here 1.3), the growth (rate of increase) of the extreme values is not itself measured. This factor (1.3) is also relative in as much as it is related to the extreme value detected first, but it independent of the rate of increase. In addition, reference is made to the fact that although it is checked whether the time interval between the detected extreme values corresponds to the critical frequency band of 0.3 and 0.7 Hz, no check is made as to whether the next extreme value A.sub.n+1 follows in practice at the same interval DT.sub.n, which is given by the previously detected upper extreme value (denoted A.sub.n-1, point in time T.sub.n-1) and the presently detected lower extreme value (A.sub.n, point in time T.sub.n), after this point in time T.sub.n. Those skilled in the art of reactor control and monitoring will appreciate that the usual techniques for the monitoring and documentation of the sensor signals apply and they will therefore readily be able not only to register the extreme values A.sub.n-1, A.sub.n, A.sub.n+1 . . . but also the points in time T.sub.n-1, T.sub.n, T.sub.n+1 . . . at which these extreme values occur. The person in charge of monitoring could therefore readily suppress the corresponding region signal if the time interval DT.sub.n =T.sub.n -T.sub.n-1 deviates significantly (for example 0.1 seconds) from the time interval DT.sub.n+1 =T.sub.n+1 -T.sub.n. However, U.S. Pat. No. 5,174,946 contains no advice on this point. In the state of that prior art, therefore, no attention is initially paid to an oscillation whose (unmeasured) rate of increase lies below the set factor (1.3); rather, intervention is considered in the reactor operation only when its extreme values exceed the threshold value A.sub.max. Only rapidly increasing oscillations cause this extremely critical state to be recognized in good time and to the initiation of suitable countermeasures. Apparently, it is assumed that slowly increasing oscillations inherently decay by themselves and normally do not require a SCRAM. To be specific, that prior art provides as counter-measure only to damp the oscillation by means of rapidly moving in virtually all the control rods (total SCRAM). That is to say, apart from the SCRAM, this strategy provides no further measure for damping the oscillation and does not reduce the probability of the SCRAM either, which constitutes a considerable intervention in the reactor operation. Instead, in the event that there is a rapidly increasing oscillation, damping only takes place earlier (i.e., below A.sub.max). As a result, only the thermal loading of the fuel elements is reduced. SUMMARY OF THE INVENTION It is accordingly an object of the invention to provide a method and device for operating a reactor in an unstable state, which overcomes the above-mentioned disadvantages of the heretofore-known devices and methods of this general type and which improves the oscillation detection and damping so as to entirely obviate any SCRAM, i.e., to manage without an intervention in the reactor operation, or with an intervention which is the least disturbing. It is a further object to allow the monitoring of the critical state with a system and method which is least susceptible to interference. With the foregoing and other objects in view there is provided, in accordance with the invention, in a reactor operated in accordance with operationally dependent input parameters, a method of operating the reactor which is unstable as a result of an oscillation of an internal physical variable, which comprises: measuring the physical variable during at least two oscillations and calculating at least one measured value for a rate of increase of the oscillation; and deciding, in dependence on the measured value, whether a stabilization strategy is to be initiated with changed input parameters for damping the instability or a reactor operation is to be continued with unchanged input parameters. By measuring the physical variable (i.e., the neutron flux, in the case of the thermally-hydraulically induced oscillations) the invention provides for the formation of local measured values in a plurality of regions of the reactor core, the measured values being assigned to the respective regions. The monitoring of the measured values leads to the formation of a current alarm stage from among a hierarchy of alarm stages with associated monitoring criteria, and the selection of the highest alarm stage, whose monitoring criterion is satisfied by the measured values in a predefined minimum number of the regions. (The monitoring criterion can in this case be composed of a plurality of individual conditions, for example the exceeding of separate threshold values for the amplitude and for the rate of increase of the extreme values.) Depending on the current alarm stage, a stabilization strategy is then initiated. As a stabilization strategy which belongs to a low-ranking alarm stage, provision is made to intervene in the operational control and regulation of the reactor only so as to block a removal of the control rods, as is envisaged in the case of an operational increasing of the reactor power: the power of the reactor cannot then be raised by the operating personnel of the reactor; instead only such control commands which correspond to the control of the reactor to a constant or decreasing power become effective in the reactor control system. In at least one higher-ranking alarm stage, provision is made as stabilization strategy for a plurality of control rods to be introduced into the core in the sense of a reduction in the reactor power (alarm stage I). Advantageously, at least two higher-ranking alarm stages (alarm stage II and alarm stage III) are provided, in alarm stage II only a plurality of control rods, corresponding to a fraction of the total number, being moved into the core slowly and in such a way as corresponds to an operational reduction in the power (that is to say the reactor control system performs an operational reduction in the power, even if, for example, a higher power consumption would intrinsically require a higher reactor power and the operating personnel wish to increase the reactor power). In the second higher-ranking alarm stage (alarm stage III)--in a manner similar to the case of a total rapid shutdown of the reactor (total SCRAM)--control rods are moved in rapidly, however likewise not all thereof but only some of the control rods being involved ("partial SCRAM"). A total SCRAM is then no longer necessary, but an option for an alarm stage IV which triggers the SCRAM can be retained. In particular, during the monitoring of the measured values, at least two periods of the oscillation are evaluated, so that the reactor is therefore initially further operated in an unchanged manner, although an oscillation is already indicated. Furthermore, a method of operating a reactor which is unstable as a result of oscillation of a physical variable occurring in the core makes provision, by measuring the physical variable, for forming a measured value which registers the rate of increase of the oscillation (if appropriate, also further measured values). Depending on this measured value, a decision is made as to whether a stabilization strategy should be initiated in order to damp the instability or the reactor is initially further operated in accordance with measured values entered as a function of operation. In particular, in this case the reactor can continue to be operated for at least two more oscillations during the measurement of the rate of increase, without an intervention being made in the reactor control system--provided that no measured value reaches a threshold value which calls for the initiation of a total SCRAM. Thus, for example, it is possible that when a threshold value A.sub.max for the oscillation amplitudes is exceeded, the SCRAM--corresponding to the highest alarm stage IV--is initiated only at high rates of increase, but at low rates of increase the reactor is still operated with relatively high amplitudes, since in the case of amplitudes which are growing so weakly, a SCRAM which is initiated only later (in the event that the oscillation then does not intrinsically decay) still has sufficient time to become effective before A.sub.th is reached. A threshold value, dependent on the rate of increase, is preferably predefined for the extreme values of the oscillating physical variable, and the stabilization strategy is triggered if the extreme values exceed this threshold value. However, a threshold value for the rate of increase can also be predefined, the stabilization strategy then being initiated when the rate of increase exceeds this threshold value. In a similar embodiment of the invention, a number of oscillations can be predefined, the number depending on the rate of increase, and the stabilization strategy can then be triggered only when the oscillation of the physical variable persists over the duration of these oscillation periods. In accordance with an added feature of the invention, a plurality of stabilization strategies are provided, from which the stabilization strategy to be triggered is selected as a function of the rate of increase. The (unstable) state of the reactor core is monitored with a plurality of sensors which are strategically distributed about the core. The sensor locations are divided into a plurality of regions of the reactor core and the sensors measure the behavior of the physical variable in those regions. The output signals of the sensors are combined into a number Mp of region channels and each region channel is assigned a region and sensors arranged therein for generating a region signal. The region signals are then combined into a number P of system channels, with a plurality of region channels being assigned to a system channel, in that they generate a system signal. The system signals are finally assigned to an output channel and they generate an output signal. By means of monitoring stages and selection stages, in this case an alarm output signal is set in the output signal as soon as, at least in a predefined number N.sub.p of the system channels, particularly in a minimum number N.sub.mp of region channels of the system, a monitoring criterion is satisfied over a plurality of oscillation periods. In this case, the output signal of each sensor influences a maximum of one single region signal and each region signal influences a maximum of one system signal. The region signals of a system channel are in each case formed from the output signals of sensors which are located in regions which are distributed over the cross section of the reactor core in such a way that the regions which are adjacent to such a region contain sensors whose output signals are assigned to region channels of other system channels. The invention thus effectively dispenses with multiple evaluations and region overlaps. Although each individual fuel element is no longer as precisely monitored as in the Watford et al. patent, experience and model calculations with unstable states have shown that it is always relatively large parts of the reactor, but not isolated fuel elements, which begin oscillating. In other words, fine resolution of the measured value registration is not necessary. In addition, the redundancy and interference immunity of the registration is increased. With the above and other objects in view, there is further provided, in accordance with the invention, a device for monitoring a reactor core of a boiling-water reactor with regard to local oscillations of a physical variable causing an unstable state of the reactor. The device comprises: a system selection stage, a plurality of region selection stages connected to the system selection stage, a given number of region monitoring stages connected to each the region selection stage, and a sensor stage connected to each the region monitoring stage with a plurality of sensors strategically disposed in regions of a reactor core of a boiling water reactor, wherein a) measured signals supplied by the sensors to a respective the region monitoring stage are combined into a region signal for the physical variable; each the region signal is monitored in the respective the region monitoring stage in accordance with a monitoring criterion, and a region signal containing a region monitoring signal is output by each region monitoring stage; b) each region signal is connected to at least one of the region selection stages, and the region selection stages forming respective system monitoring signals from a predefined minimum number of region monitoring signals; and c) the region selections stages each outputting a respective system monitoring signal to the system selection stage, and the system selection stage outputting an output monitoring signal according to a predefined minimum number of systems. There is further provided, in accordance with the invention, a device for monitoring a reactor core of a boiling-water reactor with regard to a state which is unstable as a result of local oscillation of a physical variable in the reactor core, comprising: a) a plurality of sensors disposed in a plurality of regions of a reactor core of a boiling-water reactor, the sensors measuring a physical variable of the reactor core and outputting output signals, the output signals of a plurality of the sensors of a given region being combined into an associated region signal; b) a plurality of evaluation stages each receiving a respective region signal, the evaluation stages identifying in the region signal an occurrence of extreme values of the physical variable and, given an oscillation of constant frequency, determining a rate of increase of the extreme values in the respective region; and c) at least one monitoring stage receiving output signals from the evaluation stages, the monitoring stage setting an alarm signal when the extreme values of a predefined number of regions satisfy a local monitoring criterion which depends on the rate of increase of the extreme values In other words, there is provided a system selection stage, a number P of region selection stages, for each region selection stage a number Mp of region monitoring stages and for each region monitoring stage a sensor stage having a plurality of sensors which are arranged inside a region of the core and are assigned to this region monitoring stage. The device is constructed in such a way that the sensors which are respectively assigned to a region monitoring stage supply measured signals for the physical variable which are combined into a region signal, and each region signal is monitored in accordance with a monitoring criterion in the region monitoring stage assigned to the sensors. Each region monitoring stage supplies a region signal which contains a region monitoring signal. Each region monitoring signal is connected to at least one region selection stage which forms a system monitoring signal from a predefined minimum number of region monitoring signals. Each system monitoring signal is then fed to the system selection stage; the latter supplies an output monitoring signal by means of a predefined minimum number of system monitoring stages. The system sensors which are strategically distributed about a plurality of regions of the reactor core for measuring the physical variable. The output signals of a plurality of sensors of a region are combined into an associated region signal. Each region signal is assigned an evaluation stage, which identifies in the region signal the occurrence of extreme values of the physical variable (in particular over a plurality of oscillation periods) and, given an oscillation of constant frequency and appropriate duration, determines the rate of increase of the extreme values in this region. The evaluation stages are assigned at least one monitoring stage which sets an alarm signal as soon as the extreme values at least in a predefined number of regions satisfy a local monitoring criterion which depends on the determined rate of increase. With a view to the proposed stabilization criteria, a device for monitoring the local oscillations can contain sensors for measuring the physical variable, which sensors are arranged in a plurality of regions of the reactor core, and the output signals of a plurality of sensors of a region being combined into an associated region signal. Each region signal is then assigned an evaluation stage which identifies the occurrence of an oscillation of constant frequency in the region signal. The evaluation stages are assigned an output monitoring stage which selects an alarm stage from a hierarchy of alarm stages in accordance with predefined monitoring criteria for the oscillations identified in at least a predefined number of region signals. In this case, the output monitoring stage, corresponding to the selected alarm stage, defines a point in time (or at least the criteria for the point in time) at which an emergency instruction is output to initiate a stabilization strategy corresponding to the alarm stage. This point in time can be predefined, for example, by means of a number of oscillation periods which are allowed to elapse before the initiation of a stabilization measure. However, by this means it can also be defined that, depending on the instantaneous current values (for example current values of the rate of increase) a threshold value (for example a threshold value for the amplitude) is defined, which leads to the triggering of the stabilization measure at a later point in time, at which a monitored current value (e.g., the amplitude) then exceeds this predetermined threshold value. 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 method and device for operating a reactor in an unstable state, 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. |
045270660 | summary | FIELD OF THE INVENTION The invention relates to a concrete shielding housing for receiving and storing a fuel element container filled with spent nuclear reactor fuel elements. The container is suitable for transport and storage. The outer dimensions of the container are somewhat smaller than the clear interior dimensions of the concrete shielding housing. The concrete shielding housing has a pallet-like base, a concrete shielding wall placeable on the base, and a cover which can be placed atop the upper end of the concrete shielding wall. At the lower region of the concrete shielding housing, at least one air inlet opening is provided and, in the upper region of the concrete shielding housing, at least one air outlet opening is provided. BACKGROUND OF THE INVENTION In efforts to provide a temporary storage for fuel element containers in the open, it has been suggested to accommodate the containers in silo-like housings made of concrete or steel-reinforced concrete. The silo-like housings can be of different configurations and each is suitable for accommodating one fuel element container. In one configuration of a shielding housing for receiving fuel element containers, the shielding housing is provided with lateral air inlet passages at the lower end of the concrete shielding wall and lateral air outlet passages in the region of the upper end of the shielding wall beneath the cover. With this arrangement of the air inlet and air outlet openings, a natural ventilation within the housing is obtained for directing away heat produced by the radioactive decay of materials stored in the container. The base of the concrete shielding housing is configured as a separate pallet which can be moved about from one location to another with the aid, for example, of a fork-lift truck. The fuel element container and the concrete shielding wall of the concrete shielding housing can be set down upon this base. The pallet-like base makes it possible to move the entire concrete shielding housing and container to a location on the storage field after the fuel element container is delivered and the concrete shielding housing is put together. Again, a suitable vehicle for moving this entity from one location to another could be, for example, a fork-lift truck. The storage field is preferably in the open air and is therefore subjected to the weather elements. SUMMARY OF THE INVENTION It is an object of the invention to provide a concrete shielding housing of the type referred to above wherein the surface water which collects on the surface of the concrete shielding housing during a rainfall can be run off. The concrete shielding housing according to the invention includes a pallet-like base and a concrete shielding wall placeable upon the base. A removable cover is placeable atop the shielding wall. The outer dimensions of the container are somewhat smaller than the clear interior dimensions of the housing. According to a feature of the invention, the base has a plan profile smaller than the plan profile of the concrete shielding wall whereby the shielding wall overlaps the base when placed thereon. By constructing the pallet-like base pursuant to the invention as described above, the water collecting on the outer surface of the concrete shielding housing can run off and drip from the overlapping concrete shielding wall to the ground. This advantageous runoff of water minimizes the disadvantageous effects of the weather elements. A further significant advantage of the invention is that this configuration enables the transporting corridor in the container field to be dimensioned narrower. In another advantageous embodiment of the invention, it is a feature to provide air inlet means formed at the lower end of the concrete shielding wall in the region of the latter overlapping the pallet-like base. The air inlet means are in the form of segment-like inner recesses arranged at the lower end of the concrete shielding wall. With this configuration, the conventional radial air inlet openings can be dispensed with. A further minimization of the radiation emanating from the shielding housing is an additional advantageous consequence of this arrangement. The simplification in the production of the air inlet openings is likewise a significant advantage. The invention achieves advantageous runoff of the surface water from the concrete shielding housing and permits the transport corridors in the container storage field to be made narrower. |
description | This application is based upon and claims the benefit of priority from Japanese Patent Application No. 2014-119267, filed on Jun. 10, 2014; the entire content of which is incorporated herein by reference. Embodiments of the present invention relate to a nuclear power plant and a reactor building gas treatment system. With reference to FIGS. 8 and 9, a reactor building gas treatment system of a conventional boiling water-type nuclear power plant will be outlined. FIG. 8 shows an example of a conventional plant called ABWR. In FIG. 8, a core 1 is housed inside a reactor pressure vessel 2. The reactor pressure vessel 2 is housed inside a containment vessel 3. The inner portion of the containment vessel 3 is divided into a dry well 4, which houses the reactor pressure vessel 2, and a wet well 5. The wet well 5 contains a suppression pool 6. Above the suppression pool 6, a wet well gas phase portion 7 is formed. The atmosphere in the containment vessel 3 is replaced with nitrogen in order to keep oxygen concentration low, in the case of a boiling water light water reactor. Based on the material thereof, containment vessels 3 are generally categorized into a steel containment vessel, a reinforced concrete containment vessel (RCCV), a steel concrete composite structure (SC structure) containment vessel (SCCV), and the like. In the case of RCCV, a steel liner is put on the inner surface. FIG. 8 shows an example of RCCV, which is used in ABWR. On top of the containment vessel 3, a containment vessel head 8 made of steel is provided. The containment vessel head 8 is joined to the containment vessel 3 via a containment vessel head flange 9. The containment vessel head 8 can be detached at the time of refueling. On an outer peripheral portion of the containment vessel head 8, there is a space known as a reactor well 10. The reactor well 10 is a space formed by a reactor well sidewall 11, which surrounds the periphery of the containment vessel head 8 and extends upward, a reactor well bottom portion 12, which is connected to a lower end of the reactor well sidewall 11 in such a way as to support the reactor well sidewall 11, the containment vessel head 8, and a shield plug 13. The reactor well bottom portion 12 is part of the containment vessel 3 in the case of RCCV. However, the reactor well bottom portion 12 is part of shield concrete that surrounds the periphery of a steel containment vessel in the case of the steel containment vessel. In general, the reactor well 10 is circular in horizontal cross section. However, the reactor well 10 may be elliptical or polygonal. Above the reactor well 10, the shield plug 13 is placed. The shield plug 13 is mainly made of concrete, and is divided into several blocks 13a. The reason is to lighten the weight of one block 13a. The function of the shield plug 13 is to block radiation generated during the operation of a reactor. The joint areas of blocks 13a are therefore formed into a stepwise pattern, thereby blocking radiation from leaking via a gap 13b between the blocks toward an upper area. The gap 13b between the joint areas of blocks 13a may be about 1 cm, for example. Therefore, when the reactor starts to operate, the air inside the reactor well 10 is heated and expands. Part of the air passes through the gap 13b into the upper area. Outside the reactor well 10, an operation floor 14 is provided in such a way as to connect to an upper end of the reactor well sidewall 11. An upper portion of the operation floor 14 is covered with an operation floor area wall 14c, which is part of a reactor building 15, in such a way as to form an operation floor area 14a, which is part of the space inside the reactor building 15. To the reactor pressure vessel 2, major penetrating pipes, such as main steam lines and feed water lines, are connected. These lines or pipes penetrate the containment vessel 3 and then the reactor building 15, before being connected to a turbine and a main condenser inside a turbine building (not shown). Those major pipes, such as the main steam lines and the feed water lines, are collectively referred to as penetrating pipe 16 in the figures. On the penetrating pipe 16, a first isolation valve (penetrating pipe isolation valve) 17 and a second isolation valve (penetrating pipe isolation valve) 18 are placed near the wall surfaces of the containment vessel 3. The figure shows an example in which the first isolation valve 17 is placed near the inner wall surface of the containment vessel 3, and the second isolation valve 18 is placed near the outer wall surface of the containment vessel 3. However, both of the two valves may be placed outside the containment vessel 3 in some cases. If radioactive materials are released into the inner portion of the containment vessel 3 in the event of a design basis accident such as a loss-of-coolant accident, the isolation valves are automatically closed in order to prevent the leak of radioactive materials via the penetrating pipe 16 to the outside as much as possible. However, there is a design leakage rate set for the isolation valves, meaning that very small amount of radioactive materials would leak to the outside. Moreover, there is a design leakage rate (e.g., 0.4%/d in the example of ABWR) set for the containment vessel 3, meaning that very small amount of radioactive materials would leak from inside the containment vessel 3 into the inside of the reactor building 15. In the reactor building 15, a standby gas treatment system (SGTS) 19 is provided. The standby gas treatment system 19 is designed to take in radioactive materials that are leaked into the reactor building 15 along with the atmosphere inside the reactor building 15, and removes the radioactive materials through a filter, and then releases the mainly decontaminated air into the environment from a high position. The standby gas treatment system 19 includes many branched suction pipes 20, an exhaust fan 21, a filter (filter train) 22, a standby gas treatment system exhaust pipe 23, and a heater 60. The heater 60 is disposed on the upstream side of the filter train 22. The standby gas treatment system 19 also includes isolation valves, which are not shown in the figure. Inside the filter train 22, a charcoal filter filled with activated carbon is housed. The charcoal filter is capable of removing 99% or more of radioactive materials such as cesium iodide (CsI), for example. However, if the charcoal filter is wet, its performance becomes deteriorated. Accordingly, the atmosphere needs to be heated by the heater 60 in advance in order to limit the moisture. The standby gas treatment system exhaust pipe 23 is led into a stack 24 so that gas is released from an upper end thereof. Inside the stack 24, the standby gas treatment system exhaust pipe 23 extends upward, thereby forming a double-tube structure, which is made up of the standby gas treatment system exhaust pipe 23 and the stack 24. The exhaust fan 21, the heater 60, and the isolation valves of the standby gas treatment system 19 require an electric power source to operate. In the event of a design basis accident, electric power is supplied from an emergency DG (diesel generator) 25. However, in the accident at the Fukushima Daiichi nuclear power plant, the offsite power was lost due to the earthquake and tsunami. Moreover, all emergency DGs 25 failed, and the system could not receive any supply of power from AC power sources, which is known as station blackout (SBO). The standby gas treatment system 19 therefore could not operate. Moreover, the core 1 could not be sufficiently cooled, resulting in a core melt accident. The cladding tube of the melted core fuel reacted with high temperature water, and the metal-water reaction generated large amounts of hydrogen, and the inside of the containment vessel 3 was over-pressurized. In such a severe accident, the cooling of the containment vessel 3 could be insufficient, and the atmosphere inside the containment vessel 3 could become high in temperature, probably causing damage to the containment vessel head flange 9. As a result, hydrogen could leak into the reactor well 10 via the containment vessel head flange 9, and then into the operation floor area 14a via the gap 13b of the shield plug 13. Moreover, a penetrating portion of the penetrating pipes 16 or hatch (not shown) portion could deteriorate at high temperatures, causing hydrogen to leak into the reactor building 15. Then, the hydrogen could rise up due to buoyancy, and be accumulated inside the operation floor area 14a. Because part of the operation floor 14 has an opening, such as staircase (not shown) the hydrogen can get into the operation floor area 14a via the opening. Then, the detonation of the hydrogen inside the operation floor area 14a caused damage to the reactor building 15. In order to prevent such an event, an external water injection pipe 26 is provided so that water can be poured into the reactor well 10 from the outside. In the event of a severe accident, water can be poured from a fire truck 27 or the like in order to cool the containment vessel head flange 9. In this manner, new measures have been taken. Moreover, a new hydrogen vent system 28 is provided in the ceiling of the reactor building 15 so that the hydrogen accumulated in the operation floor area 14a can be released to the external environment. Although the above description is for the containment vessel 3 and the reactor building 15 of an ABWR, those basic features are identical to those of conventional BWR/2, BWR/3, BWR/4, and BWR/5, which have been available prior to ABWR. With reference to FIG. 9, an example of a conventional passive safety BWR, which uses a passive safety system, will be described. The conventional passive safety BWR includes passive cooling system pools 30a and 30b that keep cooling water above a containment vessel 3. In many cases, the passive cooling system pools 30a and 30b are connected together via a communicating pipe (not shown) so that the cooling water can move therebetween. Inside the passive cooling system pools 30a and 30b, a passive containment cooling system heat exchanger (PCCS Hx) 31a and a reactor isolation cooling system heat exchanger (IC Hx) 31b are provided. The PCCS Hx 31a cools the steam that is released into the containment vessel 3 in the event of an accident, and sends condensed water back into the containment vessel 3. The IC Hx 31b cools the steam inside the reactor pressure vessel 2 in the event of reactor isolation or an accident, and sends condensed water back into the reactor pressure vessel 2. The heat that is generated at a time when the steam is cooled by the PCCS Hx 31a or the IC Hx 31b is transferred to the cooling water inside the passive cooling system pools 30a and 30b. After a certain period of time, the cooling water becomes so high enough in temperature that the cooling water starts boiling. The steam generated by the boiling of the cooling water is released to the external environment via exhaust ports 32a and 32b, which are provided in upper portions of the passive cooling system pools 30a and 30b. In many cases, the tips of the exhaust ports 32a and 32b are equipped with insect screens (not shown) in order to prevent insects and the like from getting into from the outside. The upper portions of the passive cooling system pools 30a and 30b are covered with an operation floor 14. In a reactor well 10, shielding water 33 is always stored during normal operation. Radiation shielding effect of the shielding water 33 is almost equal to that of the shield plug 13 (FIG. 8). Therefore, no shield plug is placed. Above the operation floor 14 is an operation floor area 14a. The portion of the reactor building 15 (operation floor area wall) that covers an upper portion of the operation floor area 14a may be dome-shaped, as shown in FIG. 9. In such a case, the operation floor area wall is referred to as an operation floor dome 14b. In many cases, the reactor building 15 is built outside the operation floor dome 14b and the containment vessel 3, in such a way as to encircle the sidewalls of the containment vessel 3. In this case, as shown in FIG. 9, the operation floor area 14a makes up the space independent of a portion of the reactor building 15 that surrounds the sidewalls of the containment vessel 3. Suction pipes 20 of a standby gas treatment system 19 are a large number of ramified pipes, which can take in the atmosphere from the operation floor area 14a inside the operation floor dome 14b as well as from other parts in the reactor building 15. In another example of the passive safety BWR, the containment vessel 3, the passive cooling system pools 30a and 30b, and the operation floor area 14a may be housed in a reactor building (not shown) whose structure is the same as the reactor building 15 of ABWR (See FIG. 8). Even in this case, the outlets of the exhaust ports 32a and 32b of the passive cooling system pools lead to the environment outside the reactor building 15. However, there is a case where a standby gas treatment system is not provided like ESBWR (Economic Simplified Boiling Water Reactor) whose safety systems consists only of a passive safety system. As an example of the reactor building gas treatment system for reactor accident, for example, the technology disclosed in Patent Document 1 (Japanese Patent Application Laid-Open Publication No. 2005-43131; the entire content of which is incorporated herein by reference) is known. In the conventional BWR, there is a possibility in the event of a severe accident that the operation floor area 14a is filled with hydrogen as a result of leakage via the containment vessel head flange 9, the first isolation valve 17 and the second isolation valve 18 provided on the penetrating pipes 16, and the like. In order to prevent the detonation of hydrogen in a severe accident, the hydrogen vent system 28, which is installed near the ceiling of the reactor building 15, can be opened. However, radioactive materials that leak from the containment vessel 3 are released to the environment along with hydrogen. Therefore, in terms of reducing radiation exposure, the hydrogen vent system 28 should not be opened as much as possible. Moreover, there is a possibility that radioactive materials that leak through the penetrating pipes 16 and the isolation valves 17 and 18 could be released directly to the environment outside the reactor building 15. If the standby gas treatment system 19 can operate even in a severe accident, the filter (filter train) 22 can remove 99% or more of radioactive materials such as cesium iodide (CsI). The residual radioactive materials, such as radioactive noble gas, and hydrogen can be released from a high position via the main stack 24. In this manner, it is desirable that, in the event of a severe accident, hydrogen that has leaked into the operation floor area 14a be released through the standby gas treatment system 19 which at the same time removes radioactive materials. It is also desirable that radioactive materials leaking to the environment outside the reactor building 15 through the isolation valves 17 and 18 of the penetrating pipes 16 be removed by the standby gas treatment system 19. Accordingly, electric power needs to be supplied from a alternate power source even at the time of a severe accident so that the standby gas treatment system 19 can operate. However, if water is poured by a fire truck 27 or the like into the reactor well 10 from the outside in order to cool the containment vessel head flange 9, the poured water starts boiling due to the heat from the containment vessel head 8. As a result, the steam leaks through the gap 13b of the shield plug 13, and fills the operation floor area 14a. The standby gas treatment system 19 takes in the steam via the suction pipes 20 to the point where the amount of the steam exceeds the processing capacity of the heater 60. As a result, the steam gets into the filter train 22, and the filter train 22 loses its function to remove radioactive materials. In the case of the conventional passive safety BWR, if the shielding water 33 gets boiling by heat from the containment vessel head 8 at the time of a severe accident, the steam fills the operation floor area 14a. As a result, the standby gas treatment system 19 loses its function to remove radioactive materials in the same way. The object of embodiments of the present invention is to prevent the operation floor area from being filled with steam generated by boiling of water in the reactor well in the event of a severe accident, to allow the standby gas treatment system to operate, to treat radioactive materials inside the reactor building and release gas inside the reactor building to the environment, and to enable the standby gas treatment system to treat radioactive materials leaking to the environment outside the reactor building through the penetrating pipes. According to an embodiment, there is provided a nuclear power plant comprising: a core; a reactor pressure vessel that houses the core; a containment vessel that houses the reactor pressure vessel; a containment vessel head of the containment vessel; a reactor building that surrounds at least part of the containment vessel; a sidewall that surrounds a periphery of the containment vessel head and extends upward; a bottom portion that is connected to a lower end of the sidewall and is connected to the containment vessel; a reactor well that is formed by the containment vessel head, the sidewall, and the bottom portion; a reactor well upper lid that is provided in an upper portion of the reactor well; an operation floor that is joined to an upper end of the sidewall and is provided on a periphery of the sidewall; an operation floor area wall that surrounds a periphery and upper portion of the operation floor to form an operation floor area and is part of the reactor building; a standby gas treatment system including: a suction pipe to take in gas inside the reactor building; an exhaust fan to drive the gas that the suction pipe takes in; a standby gas treatment system exhaust pipe to release the gas that the suction pipe takes in to an environment outside the reactor building via the exhaust fan; a heater that is disposed between the suction pipe and the standby gas treatment system exhaust pipe and uses electric power to heat the gas that the suction pipe takes in; and a filter to filter the gas heated by the heater to send the gas to the standby gas treatment system exhaust pipe; and a reactor well exhaust section to release the gas inside the reactor well to the environment without releasing the gas into the operation floor area in an event of a severe accident. According to another embodiment, there is provided a reactor building gas treatment system of a nuclear power plant that includes a containment vessel housing a reactor pressure vessel, and a reactor building surrounding at least part of the containment vessel, wherein: the nuclear power plant includes: a sidewall that surrounds a periphery of a containment vessel head of the containment vessel and extends upward, a bottom portion that is connected to a lower end of the sidewall and is connected to the containment vessel, a reactor well that is formed by the sidewall and the bottom portion surrounding the containment vessel head, a reactor well upper lid that is provided in an upper portion of the reactor well, an operation floor that is joined to an upper end of the sidewall and is provided on a periphery of the sidewall, and an operation floor area wall that surrounds a periphery and upper portion of the operation floor to form an operation floor area and is part of the reactor building; and the reactor building gas treatment system comprises: a standby gas treatment system including: a suction pipe to take in gas inside the reactor building; an exhaust fan to drive the gas that the suction pipe takes in; a standby gas treatment system exhaust pipe to release the gas that the suction pipe takes in to an environment outside the reactor building via the exhaust fan; a heater that is disposed between the suction pipe and the standby gas treatment system exhaust pipe and uses electric power to heat the gas that the suction pipe takes in; and a filter to filter the gas heated by the heater to send the gas to the standby gas treatment system exhaust pipe; an alternate power source that supplies power to the standby gas treatment system in an event of a severe accident, and a reactor well exhaust section to release the gas inside the reactor well to the environment without releasing the gas into the operation floor area in an event of a severe accident. According to yet another embodiment, there is provided a reactor building gas treatment system of a nuclear power plant, the nuclear power plant including: a reactor building, a containment vessel that is provided inside the reactor building and whose upper portion is equipped with a containment vessel head; a reactor pressure vessel that is housed in the containment vessel and below the containment vessel head; a reactor well that is provided above the containment vessel head of the reactor building; and an operation floor area that is provided above the reactor well of the reactor building, wherein the reactor building gas treatment system comprises: a suction pipe to take in gas inside the reactor building; a filter to filter the gas that the suction pipe takes in; a standby gas treatment system exhaust pipe to release the gas filtered by the filter to outside the reactor building; and a reactor well exhaust section to release the gas inside the reactor well to outside the reactor building without releasing the gas into the operation floor area. Embodiments of the present invention will be described with reference to FIGS. 1 to 7. In FIGS. 1 to 7, the components that are the same as, or similar to, those in FIGS. 8 and 9 are represented by the same reference symbols, and will not be described again. Only essential parts will be described. With reference FIG. 1, a first embodiment of a nuclear power plant that includes a reactor building gas treatment system of the present invention will be described. According to the first embodiment of the present invention, a containment vessel and a reactor building of ABWR are used. However, the type of the containment vessel and reactor building is not limited to them. The present invention can be generally applied to the containment vessels and the reactor buildings of BWR/2, BWR/3, BWR/4, and BWR/5, which share common characteristics with the containment vessel and the reactor building of ABWR. According to the first embodiment, a reactor well upper lid 40 is disposed on the shield plug 13. The reactor well upper lid 40 is made of a material that has sealing properties to block steam and can withstand high-temperature steam. For example, the reactor well upper lid 40 may be made of iron, aluminum, heat-resistant rubber, or heat-resistant resin. A reactor well exhaust section is formed in such a way as to connect the reactor well 10 to the environment outside the reactor building 15. The reactor well exhaust section includes a reactor well exhaust pipe 41 and a reactor well exhaust pipe isolation valve 42. One end (first end) 411 of the reactor well exhaust pipe 41 is opened inside the reactor well 10. The reactor well exhaust pipe 41 passes through the reactor well sidewall 11, and the other end (second end) 412 of the reactor well exhaust pipe 41 is open to the environment outside the reactor building 15. The reactor well exhaust pipe isolation valve 42 is installed on the reactor well exhaust pipe 41 in the reactor building 15. As the reactor well exhaust pipe isolation valve 42, any type of valve that can be opened and closed can be used. In FIG. 1, what is used is a remote-operated motor-driven valve which comes with a handle to enable manual operation on site as well. An alternate power source 43 is provided to supply electric power to the reactor well exhaust pipe isolation valve 42. As the alternate power source 43, an air-cooled diesel generator (DG), a gas turbine generator (GTG), or the like may be used. The alternate power source 43 is installed on the reactor building 15, in the case of FIG. 1. However, the installation location is not limited to this. For example, the alternate power source 43 may be placed on a hill or inside a building that is protected against a natural disaster such as tsunami or earthquake. The alternate power source 43 is not required to be permanent equipment. The alternate power source 43 may be portable equipment, which can be stored in a warehouse built on high ground or the like. In this case, the equipment will be carried and connected via a plug in the event of an accident in order to supply electric power. Electric power is also supplied from the alternate power source 43 to the standby gas treatment system 19. On an outer side of the second isolation valve 18 of penetrating pipes 16, a third isolation valve (penetrating pipes isolation valve) 44 is installed in the reactor building 15. The third isolation valve 44 is an motor-driven valve. The third isolation valve 44 is supplied with electric power from the alternate power source 43; or another DC power source may be provided for it. One end (first end) 451 of a leakage control pipe 45 branches from an inter-isolation-valve pipe portion 44a, which is located between the second isolation valve 18 and the third isolation valve 44 of the penetrating pipes 16, while the other end (second end) 452 is connected to the suction pipes 20 of the standby gas treatment system 19. In FIG. 1, the leakage control pipe 45 may appear to traverse the containment vessel 3. However, the leakage control pipe 45 is placed in the outer peripheral region of the cylindrical containment vessel 3 in the reactor building 15. A leakage control fan 46 is placed in the middle of the leakage control pipe 45. The leakage control fan 46 is intended to improve the ability to take in the gas. However, the leakage control fan 46 may not be required if an exhaust fan 21 of the standby gas treatment system 19 has a large capacity. According to the above configuration of the first embodiment, even if a core melt accident occurs as a result of long-term SBO triggered by an earthquake or tsunami, the reactor well exhaust pipe isolation valve 42 can be opened as electric power is supplied from the alternate power source 43. Therefore, the steam that is generated from the water poured into the reactor well 10 and heated by heat from the containment vessel head 8 can be released to the environment outside the reactor building 15 via the reactor well exhaust pipe 41. The reactor well upper lid 40 prevents the steam inside the reactor well 10 from flowing into an operation floor area 14a via the gap 13b of the shield plug 13. In this manner, even if the standby gas treatment system 19 is operated by electric power supplied from the alternate power source 43, it is possible to prevent excessive amounts of steam getting into the filter (filter train) 22, thereby ensuring that the standby gas treatment system 19 continues operating safely. Accordingly, even if radioactive materials and hydrogen leak into the reactor building 15 from the containment vessel 3 at around a design leakage rate, the standby gas treatment system 19 can treat radioactive materials such as CsI and then release hydrogen, radioactive noble gas, and the like from a high position via the stack 24. In this manner, it is possible to prevent detonation of hydrogen inside the reactor building 15, and to sufficiently reduce an amount of radioactive materials to be released to the surrounding area. Moreover, the radioactive materials that could have been directly released to the environment outside the reactor building 15 after leaking and passing through the first isolation valve 17 and the second isolation valve 18 of the penetrating pipes 16 are accumulated in the inter-isolation-valve pipe portion 44a as the third isolation valve 44 is closed. Then, the radioactive materials are introduced by the leakage suppression pipe 45 and the leakage suppression fan 46 into the suction pipe 20 of the standby gas treatment system 19, where the radioactive materials can be treated. As described above, according to this embodiment, even if a huge earthquake and a tsunami, like those that hit the Fukushima Daiichi nuclear power plant, trigger a long-term station blackout at a nuclear plant and a core melt accident occurs as a result, the standby gas treatment system is able to operate and treat radioactive materials leaking from the containment vessel, as well as to release hydrogen inside the reactor building to the environment safely. Moreover, it is possible to prevent radioactive materials from leaking through the isolation valves to the environment outside the reactor building from the penetrating pipes of the containment vessel. Therefore, even in the event of a core melt accident, the advantage is to be able to safely treat radioactive materials such as CsI and hydrogen that leak from the containment vessel. The advantage is that, since the radioactive contamination of the surrounding areas by CsI and the like is limited, residents in the surrounding areas can return home immediately after the accident is settled even if temporary evacuations have been required. With reference FIG. 2, a second embodiment of a nuclear power plant that includes a reactor building gas treatment system of the present invention will be described. According to the present embodiment, the reactor well exhaust pipe 41 is led to outside the reactor building 15 after passing through the reactor well upper lid 40. At the time of refueling, the reactor well upper lid 40 needs to be detached. Accordingly, in the middle of the reactor well exhaust pipe 41, two flanges 47 are provided: the flanges 47 face each other and are joined together. This configuration allows part of the reactor well exhaust pipe 41 to be detached. According to the present embodiment, even in cases where the reactor well exhaust pipe 41 cannot be placed in such a way as to pass through the reactor well sidewall 11, the reactor well exhaust pipe 41 can be installed. With reference FIG. 3, a third embodiment of a nuclear power plant that includes a reactor building gas treatment system of the present invention will be described. The present embodiment is a variant of the second embodiment. One end (first end) 411 of the reactor well exhaust pipe 41 is opened inside the reactor well 10, and the other end (second end) 412 is joined to the standby gas treatment system exhaust pipe 23 of the standby gas treatment system 19. The rest of the configuration is the same as that of the second embodiment. According to the third embodiment, the exhaust gas can be released through the stack 24. Even if the exhaust gas contains radioactive noble gas or the like that leaks from the containment vessel head 8, the gas is released from a high position. Therefore, it is possible to lower radioactive concentration through atmospheric dispersion. With reference FIG. 4, a fourth embodiment of a nuclear power plant that includes a reactor building gas treatment system of the present invention will be described. The present embodiment is a variant of the third embodiment. The second end 412 of the reactor well exhaust pipe 41 is not connected to the standby gas treatment system exhaust pipe 23 but is directly opened inside the stack 24. The rest of the configuration is the same as that of the third embodiment. According to the fourth embodiment, as in the case of the third embodiment, the exhaust gas can be released through the stack 24, and the same advantageous effects as in the third embodiment can be achieved. With reference FIG. 5, a fifth embodiment of a nuclear power plant that includes a reactor building gas treatment system of the present invention will be described. According to the present embodiment, reactor well exhaust pipes 41a and 41b are provided in such a way as to pass through the reactor well sidewall 11 and to be opened in gas phase portions of passive cooling system pools 30a and 30b. That is, one end (first end) 411a of the reactor well exhaust pipe 41a is opened inside the reactor well 10, and the other end (second end) 412a is opened in the gas phase portion of the passive cooling system pool 30a. Similarly, one end (first end) 411b of the reactor well exhaust pipe 41b is opened inside the reactor well 10, and the other end (second end) 412a is opened in the gas phase portion of the passive cooling system pool 30b. The suction pipes 20 of the standby gas treatment system 19 include an operation floor area upper opening 201 and an operation floor area lower opening 202. The operation floor area upper opening 201 is located near the top of the operation floor dome 14b. The operation floor area lower opening 202 is located near the operation floor 14. According to the present embodiment, the reactor well exhaust pipes 41a and 41b are not required to extend over a long distance. The steam inside the reactor well 10 is once led to the passive cooling system pools 30a and 30b via the reactor well exhaust pipes 41a and 41b, and is then released to the outside environment via exhaust ports 32a and 32b. Moreover, the operation floor dome 14b has a dome-shaped ceiling, and hydrogen is likely to be accumulated in the top portion thereof. However, the suction pipes 20 that are placed in the top portion helps the standby gas treatment system 19 take in hydrogen efficiently, and the gas can be safely released to the outside environment via the stack 24. With reference FIG. 6, a sixth embodiment of a nuclear power plant that includes a reactor building gas treatment system of the present invention will be described. The present embodiment is a variant of the fifth embodiment. On reactor well exhaust pipes 41a and 41b, reactor well exhaust pipe isolation valves 48a and 48b are provided. The reactor well exhaust pipe isolation valves 48a and 48b are motor-driven valves, for example, and use the alternate power source 43 as a power source. The rest of the configuration is the same as that of the fifth embodiment. The reactor well exhaust pipe isolation valves 48a and 48b are closed, and, at the time of refueling, the water level inside the reactor well 10 can rise to near the height of the operation floor 14. As the power source of the motor-driven valves, besides the alternate power source 43, another DC power source (not shown) may be provided. With reference FIG. 7, a seventh embodiment of a nuclear power plant that includes a reactor building gas treatment system of the present invention will be described. The present embodiment is a variant of the sixth embodiment. Reactor well exhaust pipes 41a and 41b pass through the reactor well upper lid 40 and then the operation floor 14 before being led to gas phase portions of the passive cooling system pools 30a and 30b. At the time of refueling, the reactor well upper lid 40 is detached. Accordingly, in the middle of the reactor well exhaust pipes 41a and 41b, flanges 49a and 49b and 50a and 50b are provided so that the reactor well exhaust pipes 41a and 41b can be detached. The rest of the configuration is the same as that of the sixth embodiment. According to the present embodiment, in cases where the reactor well exhaust pipes 41a and 41b cannot be installed in such a way as to pass through the reactor well sidewall 11, the reactor well exhaust pipes 41a and 41b can be placed through the space above the operation floor 14. While certain embodiments have been described, these embodiments have been presented by way of example only, and are not intended to limit the scope of the inventions. Indeed, the novel embodiments described herein may be embodied in a variety of other forms; furthermore, various omissions, substitutions and changes in the form of the embodiments described herein may be made without departing from the spirit of the inventions. The accompanying claims and their equivalents are intended to cover such forms or modifications as would fall within the scope and spirit of the inventions. |
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abstract | The drive assembly includes annular drive magnets extending around a top end of a drive shaft and annular drive coils extending around the drive magnets, separated by a pressure boundary. A latch assembly is coupled to the drive magnets and engages with the drive shaft in response to actuation of the drive assembly. The drive coils also rotate the drive magnets and the engaged latch assembly to axially displace the drive shaft. Deactivating the drive coils disengages the latch assembly from the drive shaft, dropping a connected control rod assembly via gravity into a nuclear fuel assembly. |
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052767213 | claims | 1. A nuclear reactor fuel assembly, comprising: two end plates having inner and outer surfaces; elongated, nuclear fuel-filled fuel rods being disposed between said inner surfaces and having longitudinal axes perpendicular to said inner surfaces; a leaf spring being disposed on said outer surface of one of said end plates and being bent at an acute angle defining first and second legs, a leaf spring being disposed on the outer surface of one of the end plates and being bent at an acute angle defining first and second legs, 2. In a nuclear reactor fuel assembly having two end plates with inner and outer surfaces, a resilient seat for the end plates comprising: |
description | 1. Field of the Invention The present invention concerns a diaphragm device for an x-ray apparatus that is provided for scanning a subject, and a method for scanning a subject with such a diaphragm device. 2. Description of the Prior Art A diaphragm device is known from DE 102 42 920 A1 with which the radiation beam can be adjusted in a very precise manner on the measurement field of the detector to avoid an unnecessary radiation exposure given a scanning of a subject with an x-ray apparatus, for example in the form of a computed tomography apparatus. To adjust the radiation beam emanating from the x-ray radiator, the diaphragm device has a diaphragm with two beam-proximate absorber elements. The diaphragm is fashioned such that both absorber elements can be positioned (set) with a high adjustment precision before the beginning of an examination. Given a helical scan of a subject in which the acquisition system rotates around a system axis of the computed tomography apparatus and with the subject being simultaneously displaced relative to the acquisition system in the direction of the system axis, for reconstruction of an image within a usable volume it is necessary to irradiate a scan volume that is larger in the direction of the system axis than the usable volume. The larger (in comparison to the usable volume) scan volume significantly depends on the algorithm used for reconstruction of the image. the total irradiated volume results from usable volume itself (defined by the reconstruction algorithm that will be used) plus a number of additionally-required rotations, or the additionally required fraction of a rotation that must be implemented during a leading or advance movement and a trailing movement of the x-ray radiator. Only a fraction of the information acquired during the leading movement and trailing movement, however is used later for reconstruction, such that the subject is exposed to an unnecessary radiation exposure during these segments of the scan. With increasing volume coverage of the detectors in the direction of the system axis, the number of the rotations of the acquisition system for scanning the usable volume can in fact be reduced, but the number of the rotations of a spiral scan during the leading and trailing movements that are necessary for complete reconstruction of the usable volume remain unaffected by the volume coverage of the detector. As a consequence, the proportion of the rotations due to the leading and the trailing movement thus increases in comparison to the rotations that are required for the total scanning. The dose efficiency, thus the actual proportion of the radiation used for reconstruction, is thus simultaneously reduced. For example, in a computed tomography apparatus having a detector with Z=16 lines and a line width of B=0.75 mm and an x-ray radiator that radiates the 12 mm-wide measurement field in the direction of the system axis of the computed tomography apparatus with a radiation beam adjusted (gated) by a diaphragm device, the proportion of the scan due to the leading -and the trailing movement in comparison to the total scan of the sample volume amounts to 12%, given a usable volume to be scanned of L=200 mm, a set pitch of P=1 and given one full rotation of the acquisition system required for reconstruction during each of the leading movement and the trailing movement of the scan. Only approximately half of the radiation applied during the leading and trailing movement contributes to the reconstruction of an image, such that the radiation exposure of the patient is approximately 6% of the radiation dose applied in total. Given the use of a detector with Z=128 lines and a line width of B=0.6 mm, the proportion of the leading movement and the trailing movement in the entire scan would increase to approximately 77%, such that the additional radiation exposure of the patient increases to approximately 39% of the radiation dose applied overall. An object of the present invention is to provide a diaphragm device for an x-ray apparatus for scanning a subject, and a method for scanning a subject with such a diaphragm device, that allow scanning of a volume of the subject with a reduced radiation exposure composed to the conventional situation described above. The invention proceeds from the insight that the radiation exposure of a subject can be reduced when both a very precise adjustment of the radiation beam for exposure of the measurement field of a detector and a dynamic masking of an unneeded portion of the x-ray radiation are implemented with the diaphragm device. Two very different requirements, however, must be satisfied for the adjustment of a radiation beam on the measurement field of the detector and for dynamic masking of an unneeded portion of the radiation beam. The adjustment of the radiation beam on the measurement field of the detector must ensue very precisely with an adjustment precision of the diaphragm of a few micrometers. This high requirement results from the fact that, due to the high image scale (magnification), a slight position error of the diaphragm leads to the situation that the radiation beam being significantly displaced on the detector. In competition with this need is the necessity of the dynamic masking of the portion of the radiation beam being implemented with a high speed. A precise and, at the same time, fast adjustment of the radiation beam can be achieved only insufficiently with known diaphragm devices, due to these very different requirements. An optimization of one of these requirements is possible only at the cost of the other requirement. To achieve a faster masking possibility, the diaphragm device would exhibit tolerances that would be too high for the precise adjustment of the radiation beam on the measurement field of the detector. Conversely, for achieving precise adjustment, the diaphragm device would exhibit an inertia that would be too high for dynamically masking the radiation beam. Conventional diaphragm devices therefore are normally operated such that, when in doubt, a larger region of the subject is exposed than would be necessary for reconstruction so that an artifact-free reconstruction of an image is ensured. In this case, however, the subject is exposed to an increased radiation exposure. The invention furthermore is based on the insight that an acquisition of the projections necessary for reconstruction of an artifact-free image with optimally-low radiation exposure of the subject is possible when the adjustment of the radiation beam on the measurement field of the detector, and the dynamic masking of the unneeded portion of the radiation beam, are effected separately from one another. According to the invention, the diaphragm device therefore has at least two diaphragms, with the radiation beam that has been adjusted with the first diaphragm being at least partially, dynamically masked by the second diaphragm for at least one segment of the scan of the subject. The adjustment and the dynamic masking of the radiation beam thus ensues with two different diaphragms that are separate from one another, so that the different requirements for scanning a subject with a low radiation exposure can be simultaneously fulfilled. The first diaphragm is used for precise adjustment of the radiation beam on the measurement field of the detector, while the second diaphragm enables the dynamic masking of the radiation beam. By this separation, each diaphragm thus can be adapted in a safer and simpler manner to the function associated with it. The diaphragms are designed such that the adjustment of the radiation beam by the first diaphragm can be implemented with a high adjustment precision and the dynamic masking by the second diaphragm can be implemented with a high adjustment speed. As a result of the dynamic masking of the radiation beam, the remaining radiation beam exposes essentially only a region of the subject that contributes to the reconstruction of an image, so that an unnecessary radiation exposure of the subject is avoided. The segment of the scan in which the dynamic masking of the radiation beam ensues advantageously corresponds to a leading movement of the scan (for example in the form of a spiral scan) of the subject. Likewise, it is naturally also possible for the segment of the scan to correspond to a trailing movement of the scan of the subject. As already mentioned, the radiation exposure of the subject can be reduced to a significant degree by a dynamic masking in the leading movement and in the trailing movement during, for example, a spiral scan. In an embodiment of the invention, the dynamic masking ensues dependent on a scan position in the direction of a system axis of the x-ray apparatus. It is also possible to control the masking dependent on the rotation angle of the acquisition system or dependent on a sample time. Relative to the first diaphragm, the second diaphragm preferably is closer to the focus of the radiation beam. Due to the fan geometry of the radiation beam, positioning the second diaphragm closer to the focus of the radiation is advantageous because the transmission ratio between an adjustment of the second diaphragm and the change of the radiation beam cause thereby is increased. Nevertheless, reverse order of the diaphragm positions is also possible. In an embodiment of the invention, both diaphragms of the diaphragm device are designed such that they can be adjusted in parallel with one another, such that a situation-dependent adaptation of the diaphragm device to the ray geometry of the acquisition system of the x-ray apparatus can be effected. An adaptation to beam geometry can be necessary, for example, when the focus of the radiator shifts due to thermal loads. An x-ray apparatus (here a computed tomography apparatus) is shown in FIG. 1. The acquisition system thereof includes a radiator 15 (for example in the form of an x-ray tube) with a source-proximate diaphragm device 1 and a mediation detector 13 fashioned as a laminar array. The array has a number of detector elements 14 arranged in rows and columns, with only one thereof being provided with a reference character. The radiator 15 and the detector 13 are mounted opposite one another in a rotary frame (not explicitly drawn, known as a gantry) such that a fan-shaped radiation beam 10 emanating from a focus 12 of the radiator 15 in the operation of the computed tomography apparatus and masked by the diaphragm device 1, strikes the detector 13. The detector elements 14 generate respective attenuation value dependent on the attenuation of the radiation passing through the measurement region, each attenuation value being designated as a measurement value in the following. The translation of the radiation into measurement values ensues, for example, by means of a photodiode optically coupled with a scintillator, or by means of a directly-converting semiconductor. A set of measurement values of the detector 13 is known as a projection. By means of a drive device (not shown) controlled by a control unit 18, the rotary frame can be set in rotation around a system axis 11 in the shown ω-direction. In this manner, a number of projections of a subject 2 located in the measurement region of the acquisition system can be produced from different projection directions. By rotation of the gantry with simultaneous, continuous feed of the subject 2 in the direction of the system axis 11, an examination volume of the subject 2 can be scanned that is larger than that of the measurement region formed by the acquisition system. The measurement values of the projections are read out from a data acquisition unit 16 and supplied to a computer 17 for calculation of a reconstructed image. The reconstructed image can be visually displayed to operating personnel on a display unit 19. For reconstruction of an image of an examination volume (designated in the following as a usable volume 8), it is necessary to expose a sample volume that is larger than the usable volume 8 in the direction of the system axis 11. The larger (in comparison with the usable volume 8) sample volume essentially depends on the algorithm used for reconstruction of the image. It results from the number of the additionally-required rotations that must be implemented during a leading movement 7 (shown in FIG. 2) and trailing movement 9 of the helical scan 5 during reconstruction. Without dynamic masking of a portion of the radiation beam 10 emanating from the radiator 15, during the leading movement 7 and the trailing movement 9, regions of the subject 2 are irradiated that do not contribute to the reconstruction of the image, causing the subject 2 (for example a patient) to be exposed to an unnecessary radiation exposure due to x-ray radiation during these segments of the helical scan 5. To reduce the radiation exposure of the subject 2 in a scan procedure, the diaphragm device 1 has two different diaphragms 3 and 4. The first diaphragm 3 serves for precise adjustment of the radiation beam 10 on the measurement field of the detector 13. The first diaphragm 3 is designed such that the adjustment of the radiation beam 10 is possible with a very high adjustment precision. By contrast, the second diaphragm 4 serves for dynamic masking of the portion of the radiation beam 10 that is not needed for reconstruction. For example, the dynamic masking of the radiation beam 10 ensues during the leading movement 7 and the trailing movement 9 of the scan. The second diaphragm 4 is designed such that a particularly high adjustment speed can ensue, for example at multiple cm per second. The adjustment and the masking of a portion of the radiation beam 10 are thus implemented separate from one another, such that the respective diaphragms 3 and 4 can be adapted to the appropriate requirements with regard to the adjustment precision and the adjustment speed. Relative to the first diaphragm 3, the second diaphragm 4 is located closer to the focus 14, such that large alterations of the fan geometry can be made with slight adjustments of the second diaphragm 4. In principle, diaphragm devices can naturally also be realized with a reverse arrangement of the diaphragms. As shown in FIG. 3 in a lateral detail view of the diaphragm device 1, the diaphragm 3 has two diaphragm elements 3.1, 3.2 and the diaphragm 4 has two diaphragm elements 4.1, 4.2. In each of the diaphragms 3 and 4, the diaphragm elements thereof can be adjusted independently of one another so as to gate the radiation beam 10. The movement of the diaphragm elements 3.1, 3.2 and 4.1, 4.2 can also ensue synchronously, particularly when slit diaphragms with a fixed opening are used. Also, only one of the two diaphragms 3 or 4 can be a slit diaphragm, and the other diaphragm can be designed with two diaphragm elements that can be adjusted independent of one another. The diaphragm elements 3.1, 3.2 of the first diaphragm 3 are, as just described, executed such that they can be adjusted, such that a very precise adjustment of the radiation beam 10 to the measurement field of the detector 13 can ensue. For example, each of these diaphragm elements 3.1, 3.2 can interact with an adjustment motor provided for this purpose, the adjustment motor exhibiting an adjustment precision of a few micrometers. Conventionally, the high adjustment precision is achieved at the cost of a fast dynamic adjustment of the diaphragm elements. For this reason, the first diaphragm 3 is not suited for dynamic masking of a portion of the radiation beam 10 in the region of the leading movement 7 and the trailing movement 9 of the helical scan 5. In these regions, predominantly fast adjustment speeds of multiple centimeters per second (depending on the operating mode of the computed tomography apparatus) are required to prevent the radiation exposure of the subject 2. For fast dynamic masking of a corresponding portion of the radiation beam 10, the additional second diaphragm 4 (which can be operated independently of the first diaphragm 3) is used. The high adjustment speed of multiple centimeters per second of the second diaphragm 4 is achievable, for example, by the use of corresponding adjustment motors that interact with the diaphragm elements 4.1, 4.2. The high adjustment speed of the second diaphragm 4 can also lead to larger tolerances of the adjustable precision of the adjustment position of the diaphragm elements 4.1, 4.2. For this reason, the second diaphragm 4 is operated such that the masking of the radiation beam 10 to reduce the radiation exposure ensues by taking the possible tolerances into account, such that at each point in time the sub-region of the detector 13 necessary for reconstruction is exposed during the leading movement 7 and the trailing movement 9. FIG. 4 illustrates the interaction of both diaphragms 3 and 4 during the helical scan 5 of the subject 2, wherein the adjustment positions 22, 23, 24, 25 of the respective diaphragm elements in the direction of the system axis 11 of the computed tomography apparatus during the leading movement 7, the scan of the usable volume 8, and the trailing movement 9 are shown relative to the gantry in the form of a diagram. The adjustment positions of the first diaphragm element 3.1 of the first diaphragm 3 are provided with the reference character 24; those of the second diaphragm element 3.2 are provided with the reference character 25. The adjustment positions of the first diaphragm element 4.1 of the second diaphragm 4 are provided with the reference character 22; those of the associated second diaphragm element 4.2 are provided with the reference character 23. Moreover, the aperture region 21 of the diaphragm device, which is required for an artifact-free reconstruction of an image, is shown hatched. The first diaphragm 3 is adjusted before the beginning of the examination in a precise manner such that the entire measurement field of the detector 13 can be exposed. Typically no adjustment of the first diaphragm 3 ensues during the examination, under the condition that the geometry of the acquisition system, in particular the geometry between focus 12 and detector 13, does not change. The adjustment positions 24, 25 of the two diaphragm elements 3.1, 3.2 of the first diaphragm 3 are thus constant during the course of the examination, as shown in FIG. 4. As can be seen from the hatched aperture region 21, the complete region of the diaphragm aperture of the first diaphragm 3 is not used for reconstruction of the image during the leading movement 7; but only a sub-region of this is used. In this example, at the beginning of the scan the sub-region amounts to approximately half of the entire diaphragm aperture of the first diaphragm 3. The sub-region used for reconstruction increases in line with the subject feed 20 and, given scanning of the usable volume 8, achieves the entire size of the diaphragm aperture of the first diaphragm 3, which is adjusted such that the entire measurement field of the detector 13 is exposed in a very precise manner. Reversed, in this example the used diaphragm aperture shrinks again with the subject feed 20 to approximately half of the diaphragm aperture of the first diaphragm 3. The second diaphragm 4 is dynamically adjusted to reduce the radiation exposure during the leading movement 7, such that the unneeded part of the radiation beam 10 is essentially masked. The masking is effected by means of the first diaphragm element 4.1 of the second diaphragm 4. In the exemplary embodiment described here, at the beginning of the examination this diaphragm element 4.1 is extended in the direction of the system axis 11 until approximately half of the diaphragm aperture of the first diaphragm 3 is covered and the unneeded part of the radiation beam 10 is masked. During the leading movement 7 of the scan, the first diaphragm element 4.1 is continuously backed up inline with the subject feed 20, whereby during the leading movement 7 the first diaphragm element 4.1 of the second diaphragm 4 is stationary relative to the subject 2 and only the sub-region of the measurement field that is necessary for reconstruction is irradiated. In order to allow for the possible poorer adjustment precision of the second diaphragm 4, the adjustment positions of this diaphragm element 4.1 are selected such that in each case the sub-region of the measurement field of the detector 13 that is necessary for reconstruction is exposed even given greater occurring tolerances. The second diaphragm 4 is thus set back by a small amount further than this would normally be necessary. The adjustment position of the second diaphragm element 4.2 of the second diaphragm 4 is selected during the leading movement 7 such that the radiation beam 10 is delimited by the precisely set second diaphragm element 3.2 of the first diaphragm 3. In the region (subsequent to the leading movement 7) of the scan for acquisition of the usable volume 8, both diaphragm elements 4.1, 4.2 of the second diaphragm 4 are moved back so that the radiation beam 10 is merely faded in by the first diaphragm 3 in a precise manner. Only upon the trailing movement 9 of the scan is the second diaphragm element 4.2 of the second diaphragm 4 moved (inline with the subject feed 20) into the diaphragm aperture formed by the first diaphragm 3, such that the portion of the radiation beam 10 that is not needed for reconstruction is dynamically masked. During the trailing movement 9, the second diaphragm element 4.2 of the second diaphragm 4 is thus aligned stationary relative to the subject. Due to the lesser adjustment precision of the second diaphragm 4, adjustment positions, as in the leading movement 7 adjustment positions are taken in which in each case the sub-region of the measurement field necessary for reconstruction is exposed even given greater tolerances in the respectively-adopted position. In this exemplary embodiment, the adjustment positions of the diaphragm elements 4.1, 4.2 that are adopted at the beginning of and during the scan in the leading movement 7 and in the trailing movement 9 to reduce the radiation exposure are only examples and significantly depend on which algorithm is used for reconstruction. Deviating from the above example, it is also possible for at least one of the two diaphragms 3 or 4 to be a slit diaphragm 30 with a fixed set diaphragm aperture 31. FIG. 5 exemplarily shows the interaction of the two diaphragms during a scan for the case that the second diaphragm is executed as a slit diaphragm 30. In this case, the dynamic masking of the unneeded portion of the radiation beam 10 ensues via an adjustment of the slit diaphragm 30 as a whole. The diaphragm aperture 31 of the slit diaphragm 30 is dimensioned such that, in the region of the scan of the usable volume 8, the delimitation of the radiation beam 10 is possible solely via the first diaphragm 3. If the diaphragm aperture 31 of the slit diaphragm 30 is greater than that of the first diaphragm 3, as shown in the example, the slit diaphragm 30 is adjusted during the scan of the usable volume 8 such that a dynamic delimitation of the radiation beam 10 during the trailing movement 9 is possible with the opposite (with regard to the leading movement 7) part of the slit diaphragm 30. During the operation of the computed tomography apparatus, the thermal load of the radiator 15 can lead to the focus 12 being shifted from its original position. For this reason it can be necessary to correct the position of both diaphragms 3 and 4. For this purpose, both diaphragms 3, 4 as depicted in FIG. 6 can be adjusted in parallel with one another, with the displacement being implemented corresponding to the shift of the focus 12 of the radiation. The displacement is possible, for example, by mounting both diaphragms 3 and 4 on a rail system. 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. |
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summary | ||
description | This invention was made with government support under Contract Number DE-AC07-05-ID14517 awarded by the United States Department of Energy. The government has certain rights in the invention. Embodiments of the disclosure relate generally to heat transfer systems for nuclear reactors and to related methods of removing heat from a nuclear reactor core. More particularly, embodiments of the disclosure relate to heat removal devices (e.g., enhanced radiative heat removal devices, heat transfer valves, and transitioning heat insulation to heat transmission devices) that, depending on circumstances, impede then allow heat transfer to heat sinks and to ultimate heat sinks for removal of heat from a nuclear reactor core, and to related systems for selectively removing heat from a nuclear reactor core. Generation of power by nuclear fuel includes the generation of heat, such as by fission of nuclear fuel materials within a nuclear reactor core and decay of fission and neutron activation products. Heat generated in the nuclear reactor core is removed from the nuclear reactor core by circulation of a heat transfer fluid (e.g., working fluid, also referred to as a coolant) to create a heated fluid that, in turn, is used to generate power. For example, the heat transfer fluid may comprise water (e.g., liquid water) that cools the reactor core to generate steam. The steam may pass through a turbine coupled to an electric generator to generate electricity. After passing through the turbine, the steam may at least partially condense and/or pass through a condenser to condense the steam to liquid water. The liquid water is reheated by the nuclear reactor core to repeat the cycle of heat removal from the nuclear reactor core to generate steam, and power generation from the steam. During normal operation, the fission and decay heat that is generated within the nuclear reactor core is cooled by circulation of the heat transfer fluid. However, during emergency situations, such as instances when the circulation of the heat transfer fluid is insufficient to remove all of the heat generated by the nuclear reactor core, loss of circulation of the heat transfer fluid, or during or after shut down procedures, the decay heat of the nuclear reactor core must be removed from the nuclear reactor to prevent a meltdown or damage to components of the nuclear reactor (e.g., the nuclear fuel, the reactor core internal components, and the reactor vessel). Reactor shut down may be achieved by inserting control rods into the nuclear fuel to deprive the nuclear fuel of neutrons required for fission reactions. Even with the use of reactor poisons or the insertion of control rods to reduce or substantially prevent further fission reactions, decay of the fission products in the nuclear reactor core may continue to generate significant heat that must be dissipated from the nuclear reactor core to prevent damage to the nuclear reactor core. Accordingly, nuclear reactors for power generation require several safety systems in the event that an emergency shutdown becomes necessary and/or in circumstances when the heat generation of the nuclear reactor is out of balance with the heat removal from the nuclear reactor. Auxiliary cooling systems, which may also be referred to as paths to an “ultimate heat sink,” or more simply, by extension, as a “heat sink,” are commonly utilized to safely remove heat from the nuclear reactor during shut down. Such auxiliary cooling systems reduce the temperature of the nuclear reactor with natural circulation of air or water. However, if the circulation of the air or water is hindered, the auxiliary cooling system may not sufficiently remove heat from the nuclear reactor. Embodiments disclosed herein include heat transfer systems for selectively removing heat from a nuclear reactor, and related systems. For example, in accordance with one embodiment, a system for transferring heat from a nuclear reactor comprises a nuclear reactor comprising a nuclear fuel and a heat transfer system surrounding the nuclear reactor. The heat transfer system comprises an inner wall surrounding the nuclear reactor (i.e., nuclear reactor vessel with the nuclear reactor core therein), first fins coupled to an outer surface of the inner wall, an outer wall between the inner wall and a surrounding environment, and second fins coupled to an inner surface of the outer wall and extending in a volume between the outer surface of the inner wall and the inner surface of the outer wall, the outer surface of the inner wall and the first fins configured to transfer heat from the nuclear reactor to the second fins and the inner surface of the outer wall by thermal radiation. Additional embodiments are directed to a heat transfer system for a nuclear reactor, comprising an inner wall disposed around at least a portion of a nuclear reactor, first fins coupled to an outer surface of the inner wall, an outer wall disposed around the inner wall, and second fins coupled to an inner surface of the outer wall and alternating with the first fins in a volume between the outer wall and the inner wall. In accordance with additional embodiments of the disclosure, a system comprises a nuclear reactor, a heat exchanger in operable communication with the nuclear reactor and configured to circulate a fluid through the nuclear reactor to cool the nuclear reactor and generate electricity, and a heat transfer device disposed around at least a portion of the nuclear reactor. The heat transfer device comprises an inner wall surrounding at least a portion of the nuclear reactor, an outer wall disposed around the inner wall and defining a volume between the inner wall and the outer wall, first fins coupled to an outer surface of the inner wall and extending into the volume, and second fins coupled to an inner surface of the outer wall and extending into the volume, the second fins alternating with the first fins. Illustrations presented herein are not meant to be actual views of any particular material, component, or system, but are merely idealized representations that are employed to describe embodiments of the disclosure. The following description provides specific details, such as material types, dimensions, and processing conditions in order to provide a thorough description of embodiments of the disclosure. However, a person of ordinary skill in the art will understand that the embodiments of the disclosure may be practiced without employing these specific details. Indeed, the embodiments of the disclosure may be practiced in conjunction with conventional fabrication techniques employed in the industry. In addition, the description provided below does not form a complete process flow, system, or method for forming an enhanced heat transfer system (e.g., an enhanced path to an ultimate heat sink (UHS)) for a nuclear reactor, or for removing heat from a nuclear reactor. Only those process acts and structures necessary to understand the embodiments of the disclosure are described in detail below. Additional acts to form a heat transfer system to remove heat from a nuclear reactor may be performed by conventional techniques. Further, any drawings accompanying the present application are for illustrative purposes only and, thus, are not drawn to scale. Additionally, elements common between figures may retain the same numerical designation. According to embodiments described herein, a heat transfer system for selectively removing heat from a nuclear reactor utilizes a passive heat transfer mechanism for removing heat from the nuclear reactor when other heat removal paths and heat removal mechanisms are no longer sufficient to cool the nuclear reactor (e.g., when the rate of heat generation of the nuclear reactor is greater than the rate of heat removal from the nuclear reactor by normal convection to the energy conversion system), during reactor shut down status during which normal heat removal systems are not operating, or during equipment failure (e.g., loss of coolant or loss of coolant circulation). The heat transfer system may be sized, shaped, and configured to exhibit thermally insulative properties during normal operation of the nuclear reactor and to exhibit thermally conductive properties at temperatures exceeding normal operating temperatures of the nuclear reactor. As such, the heat transfer system may be referred to herein as a “heat transfer valve.” The heat transfer system may include an inner wall around at least a portion of the nuclear reactor at the level of the nuclear reactor core, and an outer wall surrounding the inner wall. In some embodiments, a volume between the inner wall and the outer wall may include fins that facilitate radiative heat transfer from the inner wall to the outer wall. The outer wall may be in contact with an external environment to conductively transfer heat from the outer wall to the external environment (an external heat sink, such as one or more of the ground of the Earth, the atmosphere, and a body of water), may be configured to convectively transfer heat therefrom to the external environment, or both. In some embodiments, heat is transferred from the nuclear reactor to the inner wall via convective and/or conductive heat transfer and from the inner wall to the outer wall by (e.g., substantially completely by, completely by) thermal radiation. In some embodiments, heat may not be substantially transferred from the inner wall to the outer wall by other means, such as by convective heat transfer or conductive heat transfer. The heat transfer system may be structured to facilitate removal of heat from the nuclear reactor at temperatures greater than a predetermined temperature based on the desired operating parameters of the nuclear reactor core, while not substantially removing heat from the nuclear reactor during normal use and operation. Accordingly, the heat transfer system may not undesirably remove heat from the nuclear reactor and reduce the efficiency thereof during normal use and operation. The heat transfer system does not include moving parts or moving fluids (e.g., moving liquids) and may, therefore, not be prone to failure, as conventional heat transfer systems that require the circulation of air, water, or other fluids (and the associated blowers, pumps, etc.) for active cooling, such as convective cooling. The heat transfer system may be sized, shaped, and configured to facilitate sufficient removal of heat from the nuclear reactor core during emergency situations and may facilitate increasing the safe power generation capacity (e.g., the power density) of the nuclear reactor core. FIG. 1A is a simplified schematic illustrating a nuclear power plant system 100 for generation of power, in accordance with embodiments of the disclosure. The nuclear power plant system 100 includes a nuclear reactor 101 including a nuclear reactor core (which may also be referred to herein simply as a “reactor core”) 102 configured to generate heat for the production of electricity. The reactor core 102 may comprise, for example, a pebble bed core. In some such embodiments, the reactor core 102 includes fuel pebbles 104 of a nuclear fuel formulated to undergo fission reactions to generate heat. The reactor core 102 may be surrounded by a vessel 105, which may contain, for example, the components of the reactor core 102 (e.g., the fuel pebbles 104, a neutron reflector, a heat shield, etc.). The vessel 105 (which may also be referred to herein as a “reactor vessel”) may comprise, for example, a metal (e.g., steel, stainless steel, etc.). In some embodiments, the vessel 105 comprises steel. As will be described herein with reference to FIG. 1H, in some embodiments, the vessel 105 is encased in an outer reflector that is thermally conductive. In some embodiments, a fuel pebble feed 106 is in operable communication with the reactor core 102 and may provide fresh fuel pebbles 104 to the reactor core 102 from a fuel pebble storage tank 108. During use and operation of the reactor core 102, the fuel pebbles 104 are transferred from an upper portion of the reactor core 102 downward toward an outlet 110 where fuel pebbles 104 that have been transported through the reactor core 102 exit the reactor core 102 and enter a storage tank 112. In some embodiments, the fuel pebbles 104 may be transferred through the reactor core 102 more than once before they are spent. In other words, in some such embodiments, the fuel pebbles 104 from the storage tank 112 may be recirculated to the fuel pebble storage tank 108. In some embodiments, the fuel pebbles 104 comprise tri-structural-isotropic (TRISO) fuel particles. In some such embodiments, the fuel pebbles 104 may comprise, for example, a core comprising a nuclear fuel (e.g., TRISO particles embedded in a carbon matrix) and an outer carbon layer. In some such embodiments, the TRISO particles may comprise a core containing one or more of UCO, UO2, ThO2, a combination of UCO, UO2 and ThO2, a transuranic carbide material, a buffer layer around the core, an inner carbon layer around the buffer layer, a ceramic layer around the inner carbon layer, and an outer carbon layer around the ceramic layer. The buffer layer may be formulated and configured to accommodate expansion of the core and of released fission gases from the core. The inner carbon layer and the outer carbon layer may attenuate migration of radionuclides and may comprise, for example, pyrolytic carbon. The ceramic layer may comprise one or more of a silicon carbide material, a zirconium carbide material, or another material. A reactor reflector 115 (also referred to as a “neutron reflector”) may be disposed around the reactor core 102 and located between the reactor core 102 and the vessel 105. The reactor reflector 115 may include one or more of graphite, beryllium, beryllium oxide, and depleted uranium. One or more control rods 107 formulated and configured to control a fission rate of the nuclear fuel pebbles 104 may be located within the reactor reflector 115. The control rods 107 may include one or more materials formulated and configured to absorb neutrons. By way of non-limiting example, the control rods 107 may include one or more of boron, cadmium, silver, lithium, and indium. In some embodiments, the control rods 107 comprise boron. Although not illustrated in FIG. 1A, in some embodiments, at least a portion of the nuclear reactor core 102 is surrounded by a thermal insulation material. In some such embodiments, the thermal insulation material may be located between the nuclear reactor core 102 and the vessel 105. With continued references to FIG. 1A, a coolant fluid 114 may flow through the bed of fuel pebbles 104 of the reactor core 102 to form a heated coolant fluid 116. In some embodiments, the heated coolant fluid 116 exiting the reactor core 102 has a temperature greater than about 600° C., such as greater than about 700° C., greater than about 800° C., or greater than about 900° C. The coolant fluid 114 may include a material that is inert with respect to the composition of the fuel pebbles 104. In some embodiments, the coolant fluid 114 is substantially free of oxygen. In some embodiments, the coolant fluid 114 comprises helium. The heated coolant fluid 116 may exit the reactor core 102 and enter a heat exchanger 118 where heat from the heated coolant fluid 116 is transferred to a working fluid 120 to cool the heated coolant fluid 116 to the temperature of the coolant fluid 114. The cooled coolant fluid 114 is recirculated to the reactor core 102 to flow through the bed of fuel pebbles 104 to generate the heated coolant fluid 116 and the cycle is repeated. The working fluid 120 may be heated by the heated coolant fluid 116 in the heat exchanger 118 to form a heated working fluid 122. In some embodiments, the working fluid 120 comprises water (e.g., liquid water) and the heated working fluid 122 may comprise steam. In some such embodiments, the working fluid 120 undergoes a phase change within the heat exchanger 118. The heated working fluid 122 may pass through a turbine 124 that is, in turn, coupled to an electric generator 126 for generating electricity. In some embodiments, the heated working fluid 122 is condensed after the turbine 124 to form the working fluid 120. In some embodiments, the working fluid 120 is circulated back to the heat exchanger 118 to be heated by the heated coolant fluid 116. As will be understood, in some embodiments, the working fluid 120 may be condensed to form a liquid working fluid 120 that is circulated to the heat exchanger 118. During normal operation (e.g., steady state operation) of the nuclear reactor core 102 and the balance of the nuclear power plant system 100, the heat removed from the nuclear reactor 101 by the coolant fluid 114 may be substantially equal to the heat generated within the nuclear reactor core 102 by the fuel pebbles 104. However, in situations where the nuclear reactor 101 generates more heat than is removed by the coolant fluid 114, the nuclear power plant system 100 may include a heat transfer system for preventing damage to the fuel pebbles 104 and the nuclear reactor 101 that would otherwise be caused by being exposed to excessive temperatures. With continued reference to FIG. 1A, a heat transfer system 125 may be disposed around the vessel 105. For simplicity of illustration, details of the heat transfer system 125 are not illustrated in FIG. 1A and are illustrated in FIG. 1B, which is a simplified cross-sectional view of the nuclear reactor 101 and the heat transfer system 125 surrounding the vessel 105 taken through section line B-B of FIG. 1A, in accordance with embodiments of the disclosure. In order to show details, the scale and proportions of the elements represented in FIG. 1B are not the same as in FIG. 1A. The heat transfer system 125 may be located between the vessel 105 surrounding the nuclear reactor core 102 and a surrounding environment 150. The heat transfer system 125 may comprise an inner wall 103 in contact with an outer surface of the vessel 105 of the nuclear reactor core 102. The inner wall 103 may include an inner surface 128 and an outer surface 130 opposite the inner surface 128. The heat transfer system 125 may further include an outer wall 135 defined by an inner surface 136 and an outer surface 140 (shown as an arrow to distinguish from heat transfer structures 142). A space between the inner surface 136 of the outer wall 135 and the outer surface 130 of the inner wall 103 may define a volume 134. In some embodiments, the volume 134 is filled with air (e.g., a composition including about 78 atomic percent nitrogen, about 21 atomic percent oxygen, and about 1 atomic percent argon). In other embodiments, the volume 134 may include one or more materials (e.g., gases) having a thermal conductivity less than a thermal conductivity of air, such as one or more of argon or carbon dioxide. In yet other embodiments, the volume 134 is under vacuum or a partial vacuum. In some embodiments, at least a portion of the outer wall 135 may be in contact with the surrounding environment 150. The surrounding environment 150 may comprise an ultimate heat sink for the nuclear reactor 101 (FIG. 1A) such as, for example, the Earth (e.g., the ground), the atmosphere (e.g., the air above the ground and in the sky), water, or another heat sink. In some embodiments, at least a portion of the outer wall 135 is buried below a surface of the Earth. In yet other embodiments, at least a portion of the outer wall 135 is exposed to air, such as ambient air. In yet other embodiments, at least a portion of the outer wall 135 is surrounded by an ultimate heat sink comprising water. In yet other embodiments, at least a portion of the outer wall 135 is in contact with an ultimate heat sink via the heat transfer structures 142. As will be described herein, heat may be transferred from the nuclear reactor 101 (FIG. 1A) through the vessel 105 of the nuclear reactor 101 to the inner surface 128 of the inner wall 103. Heat may be transferred from inner surface 128 to the outer surface 130 by conduction and from the outer surface 130 of the inner wall 103 to the inner surface 136 of the outer wall 135 by radiative heat transfer. For example, heat may be transferred from the nuclear reactor core 102 to the vessel 105 by, for example, one or more of (e.g., all of) conductive heat transfer, convective heat transfer, and radiative heat transfer. Heat may be transferred from the vessel 105 (e.g., via a surface interfacing with the inner surface 128 of the inner wall 103 of the heat transfer device 125) to the outer wall 135, such as by, for example, radiative heat transfer. In some embodiments, heat transfer from the inner wall 103 to the outer wall 135 may be by (e.g., substantially by, primarily by, entirely by) radiative heat transfer. Each of the inner wall 103 and the outer wall 135 may independently comprise one or more of steel (e.g., stainless steel, such as austenitic stainless steel (e.g., 304 stainless steel, 316 stainless steel), ferritic stainless steel (e.g., 409 stainless steel, 430 stainless steel), martensitic stainless steel (e.g., 420 stainless steel), duplex stainless steel, precipitation hardened stainless steel (e.g., martensitic 17-4 PH stainless steel)), iron, copper, aluminum, aluminum oxide (Al2O3), alloys of aluminum (e.g., aluminum alloyed with one or more of copper, magnesium, manganese, silicon, tin, and zinc, such as one or more of 1000 series aluminum (alloys with a minimum of 99 weight percent aluminum), 2000 series aluminum (alloys of aluminum and copper), 3000 series aluminum (alloys of aluminum and manganese), 4000 series aluminum (alloys of aluminum and silicon, also referred to as “silumin”), 5000 series aluminum (alloys of aluminum and magnesium), 6000 series aluminum (alloys of aluminum, magnesium, and silicon), 7000 series aluminum (alloys of aluminum and zinc), or 8000 series aluminum), silver, alloys of nickel (e.g., alloys of nickel and chromium (e.g., alloys including nickel, chromium, and one or more of molybdenum, tungsten, and cobalt, such as Inconel® 617, Inconel® 718, alloy 600, alloy X-750), alloys of nickel and manganese), a refractory metal alloy (e.g., alloys of one or more of one or more of molybdenum, niobium, rhenium, tantalum, tungsten, chromium, hafnium, iridium, osmium, ruthenium, titanium, vanadium, and zirconium), cermet materials (composites of a ceramic material (e.g., one or more of tungsten carbide, tungsten nitride, titanium carbide, titanium nitride, silicon carbide, silicon nitride, tantalum carbide, tantalum nitride, niobium carbide, boron carbide) and a metal (e.g., a binder such as one or more of nickel, cobalt, iron, copper, molybdenum), such as one or more of tungsten carbide with one or more of cobalt, nickel, iron, or copper (e.g., WC—Co—Ni), titanium carbide with one or more of cobalt, nickel, iron, or copper (e.g., TiC—Co—Ni), silicon carbide with one or more of cobalt, nickel, iron, copper, or molybdenum), or a carbon-containing material (e.g., a carbon-carbon composite material, such as a carbon fiber reinforced carbon matrix composite material). In some embodiments, the inner wall 103 comprises steel, such as a stainless steel alloy. In other embodiments, the inner wall 103 comprises aluminum, such as one or more of elemental aluminum, aluminum oxide, or an aluminum alloy. In yet other embodiments, the wall 103 comprises silver. In further embodiments, the inner wall 103 comprises an alloy of nickel, a refractory metal alloy, or a cermet material. In additional embodiments, the inner wall 103 comprises a carbon-carbon composite material. In some embodiments, the outer surface 130 of the inner wall 103 may include a coating formulated and configured to facilitate a desired amount of heat transfer from the outer surface 130 of the inner wall 103 to the inner surface 136 of the outer wall 135. By way of non-limiting example, the coating may comprise one or more of paint (e.g., a reflective paint, such as a white paint or a silver paint, an absorptive paint, such as a black paint), an oxide (such as aluminum oxide, titanium oxide, tungsten oxide, nickel oxide, chromium oxide), an epoxy material (e.g., a phenolic epoxy), and a plasma coating (e.g., one or more of chromium carbide, chromium oxide, tungsten carbide-cobalt, aluminum oxide, silicon dioxide, zirconium oxide, boron carbide, molybdenum, nickel, cobalt). In some embodiments, the outer surface 130 may be exposed to ion implantation to create a surface that will radiate and absorb thermal radiation. In some such embodiments, the outer surface 130 may be implanted with one or more of nickel, cobalt, iron, copper, silver, molybdenum, aluminum. In some embodiments, the coating imparts hydrophilic properties to the outer surface 130. In other embodiments, the coating imparts hydrophobic properties to the outer surface 130. In some embodiments, the inner surface 136 of the outer wall 135 comprises one or more of the coating materials described above with reference to the coating of the inner wall 103. In some embodiments, the coating of the outer wall 135 is formulated and configured to facilitate increased absorption of radiative energy emitted from the outer surface 130 of the inner wall 103. The coating on the outer surface 130 may have a thickness within a range from about 0.1 mm to about 5.0 mm, such as from about 0.1 mm to about 0.2 mm, from about 0.2 mm to about 0.4 mm, from about 0.4 mm to about 0.6 mm, from about 0.6 mm to about 0.8 mm, from about 0.8 mm to about 1.0 mm, from about 1.0 mm to about 2.0 mm, from about 2.0 mm to about 3.0 mm, or from about 3.0 mm to about 5.0 mm. In some embodiments, the outer surface 130 of the inner wall 103 may be modified to induce a desired surface roughness of the outer surface 130. By way of non-limiting example, a surface roughness (e.g., an arithmetic mean roughness value (Ra)) of the outer surface 130 may be within a range from about 0.025 μm Ra to about 50.0 μm Ra, such as from about 0.025 μm Ra to about 0.05 μm Ra, from about 0.05 μm Ra to about 0.10 μm Ra, from about 0.10 μm Ra to about 0.25 μm Ra, from about 0.25 μm Ra to about 0.50 μm Ra, from about 0.50 μm Ra to about 1.0 μm Ra, from about 1.0 μm Ra to about 2.0 μm Ra, from about 2.0 μm Ra to about 4.0 μm Ra, from about 4.0 μm Ra to about 6.0 μm Ra, from about 6.0 μm Ra to about 8.0 μm Ra, from about 8.0 μm Ra to about 10.0 μm Ra, from about 10.0 μm Ra to about 20.0 μm Ra, from about 20.0 μm Ra to about 30.0 μm Ra, or from about 30.0 μm Ra to about 50.0 μm Ra. In some embodiments, the surface roughness is from about 0.025 μm Ra to about 1.0 μm Ra. In other embodiments, the surface roughness is from about 1.0 μm Ra to about 50.0 μm Ra. The surface roughness of the outer surface 130 may be selected such that the outer surface 130 exhibits a desired amount of reflectivity and emissivity. The outer surface 130 may include a surface that has been exposed to grinding, turning, polishing, or sand/grit blasting to impart the desired surface roughness. In addition, in some embodiments, the inner surface 136 of the outer wall 135 may similarly be modified to induce a desired surface roughness thereof, such as a surface roughness within a range from about 0.025 μm Ra to about 50.0 μm Ra, by way of non-limiting example, as described above with reference to the outer surface 130. In some embodiments, one or more fins 132 may be coupled to (extend from) the outer surface 130 of the inner wall 103. In some embodiments, the one or more fins 132 extend from the outer surface 130 into the volume 134 between the outer surface 130 of the inner wall 103 and the inner surface 136 of the outer wall 135. In some embodiments, a gap remains between the fins 132 and the inner surface 136 of the outer wall 135. In other words, in some such embodiments, the fins 132 may not extend from the outer surface 130 of the inner wall 103 all the way to the inner surface 136 of the outer wall 135 and the fins 132 may not contact the inner surface 136. The fins 132 may be interspersed with fins 138 extending from the inner surface 136 of the outer wall 135 into the volume 134. In other words, in some embodiments, the fins 132 may be interdigitated with the fins 138. Stated another way, a fin 138 extending from the inner surface 136 may be located between adjacent fins 132 of the inner wall 103. Similarly, a fin 132 extending from the outer surface 130 may be located between adjacent fins 138 of the outer wall 135. In some embodiments, within the volume 134 and along a circumference of the outer surface 130 of the inner wall 103 and the inner surface 136 of the outer wall 135, the respective fins 132, 138 may be spaced such that every other fin 132, 138 is one of the fins 132 and the other of the every other fins 132, 138 is one of the fins 138. Stated in yet another way, within the volume 134, the fins 132 may alternate with the fins 138 and each of the fins 132 may not be directly adjacent one another; rather each of the fins 132 may be spaced from a neighboring one of the fins 132 by one of the fins 138. Similarly, the fins 138 may not be directly adjacent one another and may be spaced from a neighboring one of the fins 138 by one of the fins 132. In some embodiments, an angular spacing between neighboring fins 132 along a circumference of the outer surface 130 of the inner wall 103 may be within a range from about 0.5° to about 180°, such as from about 0.5° to about 1°, from about 1° to about 2°, from about 2° to about 3°, from about 3° to about 4°, from about 4° to about 5°, from about 5° to about 10°, from about 10° to about 15°, from about 15° to about 20°, from about 20° to about 25°, from about 25° to about 30°, from about 30° to about 45°, from about 45° to about 60°, from about 60° to about 90°, or from about 90° to about 180°. Accordingly, an angle between each fin 132 may be from about 0.5° to about 180°. In some embodiments, an angle between each fin 132 is from about 0.5° to about 30°. In some embodiments, the fins 132 are substantially equally spaced from one another. Each fin 138 of the fins 138 may be spaced from each other along a circumference of the inner surface 136 of the outer wall 135 within a range from about 0.5° to about 180°, such as from about 0.5° to about 1°, from about 1° to about 2°, from about 2° to about 3°, from about 3° to about 4°, from about 4° to about 5°, from about 5° to about 10°, from about 10° to about 15°, from about 15° to about 20°, from about 20° to about 25°, from about 25° to about 30°, from about 30° to about 45°, from about 45° to about 60°, from about 60° to about 90°, or from about 90° to about 180°. In some embodiments, an angle between each fin 138 is from about 0.5° to about 30°. In some embodiments, the fins 138 are substantially equally spaced from one another. The fins 132 may be directly coupled to the outer surface 130 of the inner wall 103 by one or more of welding, press fitting, glue, soldering, liquid bonding, or shrink fitting. In other embodiments, the fins 132 are cast with the inner wall 103. Similarly, the fins 138 may be directly coupled to the inner surface 136 of the outer wall 135 by one or more of welding, press fitting, glue, soldering, liquid bonding, or shrink fitting. In other embodiments, the fins 138 are cast with the outer wall 135. The fins 132 may be configured to facilitate selective removal of heat (e.g., decay heat) from the inner wall 103 to the outer wall 135, which is, in turn, transferred to the surrounding environment 150. The fins 132 may be formed of and include one or more of the materials described above with reference to the inner wall 103. The fins 132 may comprise one or more of steel, iron, copper, aluminum, aluminum oxide (Al2O3), alloys of aluminum, alloys of nickel and manganese, a refractory metal alloy, cermet materials, or a carbon-containing material. In some embodiments, the fins 132 comprise steel, such as a stainless steel alloy. In other embodiments, the fins 132 comprise aluminum, such as one or more of elemental aluminum, aluminum oxide, or an aluminum alloy. In yet other embodiments, the fins 132 comprise silver. In further embodiments, the fins 132 comprise an alloy of nickel, a refractory metal alloy, or a cermet material. In additional embodiments, the fins 132 comprises a carbon-carbon composite material. In some embodiments, the fins 132 comprise the same material composition as the outer surface 130 of the inner wall 103. In other embodiments, the fins 132 comprise a different material composition than the outer surface 130. The fins 138 may comprise one or more of the materials described above with reference to the fins 132. In some embodiments, the fins 138 comprise the same material composition as the fins 132. In other embodiments, the fins 138 comprise a different material composition than the fins 132. FIG. 1C is an expanded view of a portion of the heat transfer system 125 illustrating box C of FIG. 1B. With reference to FIG. 1C, an amount of overlap H between the fins 132 extending from the outer surface 130 of the inner wall 103 and the fins 138 extending from the inner surface 136 of the outer wall 135 may be within a range from about 0.1 percent of a length L of the fin 132 to about 99 percent a length L of the fin 132, such as from about 0.1 percent to about 10 percent, from about 10 percent to about 20 percent, from about 20 percent to about 30 percent, from about 30 percent to about 50 percent, from about 50 percent to about 70 percent, from about 70 percent to about 80 percent, from about 80 percent to about 90 percent of the length L of the fin 132, or from about 90 percent to about 99 percent. In some embodiments, the overlap H is selected such that thermal expansion of the components of the heat transfer system 125 results in direct physical thermal contact of the fins 132 with the inner surface 136 of the outer wall 135 when a temperature of the inner wall 103 exceeds a pre-determined threshold. In other embodiments, the overlap H is selected such that direct physical contact of the fins 132 with the inner surface 136 of the outer wall 135 is precluded. In some embodiments, the length L of the fins 132 may be substantially equal to the length of the fins 138. In other embodiments, the length L of the fins 132 is different from the length of the fins 138. In some embodiments, the length L of the fins 132 is substantially uniform. In other embodiments one or more of the fins 132 has a length L different than the length L of at least another of the fins 132. In some embodiments, the length of the fins 138 is substantially uniform. In other embodiments, one or more of the fins 138 has a length different than the length of at least another of the fins 138. In yet other embodiments, the length L of the fins 132 may be different from a length of the fins 138. For example, each of the fins 132 may have a greater length L than each of the fins 138. In yet other embodiments, each of the fins 138 may have a greater length than each of the fins 132. In some embodiments, the overlap H may affect a view factor F130→136 from the outer surface 130 and the fins 132 to the inner surface 136 and the fins 138 (which may also be represented as F130+132→136+138). The view factor F130→136 may be defined as a proportion of radiation that leaves the outer surface 130 of the inner wall 103 and the surfaces of the fins 132 that strikes the surfaces of the fins 138 and the inner surface 136 of the outer wall 135. In some embodiments, the view factor F130→136 may be within a range from about 0.10 to about 0.99, such as from about 0.10 to about 0.20, from about 0.20 to about 0.30, from about 0.30 to about 0.40, from about 0.40 to about 0.50, from about 0.50 to about 0.60, from about 0.60 to about 0.70, from about 0.70 to about 0.80, from about 0.80 to about 0.90, or from about 0.90 to about 0.99. In some embodiments, the length L, the spacing between the fins 132 and the fins 138, the number of the fins 132, 138, and the overlap of the fins 132, 138 may be selected to impart a desired view factor F130→136. In some embodiments, the length L, the spacing between the fins 132 and the fins 138, the number of the fins 132, 138, and the overlap H of the fins 132, 138 may be selected to impose that no physical contact between the inner wall 103 and associated fins 132 and outer wall 135 and associated fins 138 may occur at operating temperatures taking thermal expansion phenomena into account. In some embodiments, the overlap H is selected such that thermal expansion of the components of the heat transfer system 125 results in the fins 132 becoming in physical thermal contact with the inner surface 136 of the outer wall 135 when a temperature of the inner wall 103 exceeds a pre-determined threshold. The fins 132 may increase the surface area from which radiative heat from the outer surface 130 of the inner wall 103 is emitted. Accordingly, the fins 132 increase the surface area through which radiative heat transfer occurs. In addition, the fins 138 may increase the surface area of the inner surface 138 of the outer wall 135 and the surface area by which radiative heat transfer from the inner wall 103 is received. FIG. 1D is a simplified cross-sectional view of a fin 160, which may correspond to one or both of the fins 132, 138. The fin 160 may include a core 162 comprising one or more of the materials described above with reference to the composition of the fins 132, 138, such as one or more of steel, iron, copper, aluminum, aluminum oxide (Al2O3), alloys of aluminum, alloys of nickel and manganese, a refractory metal alloy, cermet materials, or a carbon-containing material. The fin 160 may further optionally comprise a coating material 164 around a perimeter of the core 162. The coating material 164 may be formulated and configured to impart desired thermal transfer properties (e.g., radiative thermal transfer properties) to the fin 160. In some embodiments, the coating material 164 may be selected to impart one or more of a desired emissivity, reflectivity, or absorptivity to the fin 160. The coating material 164 may comprise an exposed surface exhibiting desired emissivity properties. The coating material 164 may comprise one or more of the materials described above with reference to the coating of the outer surface 130. By way of non-limiting example, the coating material 164 may comprise paint, silvered polytetrafluoroethylene (e.g., polytetrafluoroethylene coated with silver), an oxide material, an epoxy material, a plasma coating, or an ion implanted region. The coating material 164 may be hydrophilic or hydrophobic. In some embodiments, the exposed surfaces of the coating material 164 are black. The coating material 164 may exhibit a desired surface roughness to impart a desired reflectivity of the exposed surfaces of the coating material 164. In some embodiments, the exposed surfaces are relatively smooth to facilitate an increased reflectivity of the coating materials 164. In other embodiments, the exposed surfaces of the coating material 164 are relatively rough to facilitate a relatively increased absorption of thermal radiation. The surface roughness of the coating material 164 may be the same as the surface roughness of the outer surface 130 described above. In some embodiments, the surface roughness of the fins 132, 138 is substantially the same as the surface roughness of the outer surface 130. In other embodiments, the surface roughness of the fins 132, 138 is greater (i.e., the fins 132, 138 are more rough) than the surface roughness of the outer surface 130. In yet other embodiments, the surface roughness of the fins 132, 138 is less (i.e., the fins 132, 138 are more smooth) than the surface roughness of the outer surface 130. In some embodiments, the surface roughness Ra of the fins 132 is different from the surface roughness Ra of the fins 138. In some embodiments, the fins 132 have a lower surface roughness Ra (are smoother) than the fins 138. In some such embodiments, the fins 132 may be more reflective than the fins 138. In other embodiments, the fins 132 have a greater surface roughness Ra (are rougher and less smooth) than the fins 138. In some embodiments, the surface roughness Ra may be selected to be less than a wavelength of incident radiation during use and operation of the reactor core 102. In other embodiments, the surface roughness Ra may be selected to be greater than a wavelength of incident radiation during use and operation of the reactor core 102. In some embodiments, the surface roughness is from about 0.025 μm Ra to about 1.0 μm Ra. In other embodiments, the surface roughness is from about 1.0 μm Ra to about 50.0 μm Ra. The coating material 164 may have a thickness T within a range from about, for example, one monolayer to about 5.0 mm, such as from about one monolayer to about 0.5 nm to about 1.0 nm, from about 1.0 nm to about 10 nm, from about 10 nm to about 100 nm, from about 100 nm to about 500 nm, from about 500 nm to about 1.0 μm, from about 1.0 μm to about 10 μm, from about 10 μm to about 100 μm, from about 100 μm to about 500 μm, from about 500 μm to about 1.0 mm, from about 1.0 mm to about 2.0 mm, from about 2.0 mm to about 3.0 mm, or from about 3.0 mm to about 5.0 mm. However, the disclosure is not so limited and the thickness T of the coating material 164 may be different than those described. Referring back to FIG. 1B, the outer surface 140 of the outer wall 135 may be in contact with the environment 150. The outer surface 140 may include heat transfer structures 142 coupled thereto and extending into the environment 150. The heat transfer structures 142 may comprise any material suitable for removing heat from the outer surface 140 to the environment 150 by one or more of conductive heat transfer, convective heat transfer, or radiative heat transfer. By way of non-limiting example, the heat transfer structures 142 may comprise fins, ribbons, plates, cables, thermal radiators, a mesh, an antenna, or another material structure. In some embodiments, the heat transfer structures 142 are in direct contact with the environment 150, such as the Earth (e.g., soil). In other embodiments, the heat transfer structures 142 extend into a heat sink comprising water. The heat transfer structures 142 may transfer heat from the outer wall 135 to the environment 150 by conductive heat transfer. In some embodiments, at least some of the heat transfer structures 142 extend into the atmosphere and are formulated and configured to remove heat from the outer 135 by convective heat transfer (e.g., such as with wind passing adjacent to heat transfer structures 142). The nuclear reactor 101 and the heat transfer system 125 may be structured and arranged such that during normal use and operation of the nuclear reactor 101, heat is not substantially removed from the nuclear reactor 101 (e.g., from the vessel 105) to the environment 150, such as from the vessel 105 to the outer surface 130 of the inner wall 103, from the outer surface 130 of the inner wall 103 to the outer wall 135, and from the outer wall 135 to the environment 150. Rather, during normal use and operation, heat is removed from the nuclear reactor 101 by the circulation of the coolant fluid 114 and the heated coolant fluid 116 (FIG. 1A), that in turn gives off its heat to the working fluid 120 to form the heated working fluid 122 that is, in turn, used to operate the turbine 124 that spins the electric generator 126 which generates electricity. However, during emergency or shut down situations, such as when the nuclear reactor 101 is out of thermal balance (e.g., during periods when circulation of the working fluid 120 has ceased or when the coolant fluid 114, 116 is removing less heat than is generated by the fuel 104 of the reactor core 102 or when the temperature of the nuclear reactor 101 (e.g., the vessel 105) exceeds a predetermined temperature), the nuclear reactor 101 and the heat transfer system 125 may be structured and arranged to facilitate removal of heat from the nuclear reactor 101 to the outer wall 135 and to the environment 150. Since the primary mode of heat transfer from the nuclear reactor 101 to the outer wall 135 is through radiative heat transfer (e.g., radiative heat transfer from the inner wall 103 to the outer wall 135), the rate of heat transfer may increase by the fourth power with increasing temperature, according to the Stephan-Boltzmann Law (i.e., q=σT4·A, wherein q is the rate of heat transfer per unit time (W), σ is the Stephan-Boltzmann Constant, T is the absolute temperature in Kelvin (K) of the outer surface 130 and the fins 132, and A is the area of the emitting body (m2) (e.g., the area of the outer surface 130 and the fins 132)). Accordingly, the outer surface 130 including the fins 132 may be structured and configured to remove heat from the nuclear reactor 101, the rate of which removal increases to the fourth power of the temperature of the outer surface 130 and the fins 132. Accordingly, with increasing temperature of the nuclear reactor 101, the temperature of the outer surface 130 and the fins 132 may exhibit a corresponding increase in temperature and the rate of heat removal from the nuclear reactor 101 may change at nearly the fourth power of the temperature of the outer surface 130 and the fins 132, which may approximate the temperature of the nuclear reactor 101 (depending on a temperature drop between the vessel 105 and the inner wall 103). In some such embodiments, the outer surface 130 and the fins 132 may be sized and shaped such that negligible heat is removed from the nuclear reactor 101 by thermal radiation at temperatures of the nuclear reactor 101 during normal use and operation of the reactor core 102 while sufficient heat is removed from the nuclear reactor 101 during emergency situations where the temperature of the nuclear reactor 101 is increased beyond conventional operating temperatures of the nuclear reactor 101. In some embodiments, the heat transfer system 125 may be configured to be thermally insulative at temperatures less than the normal operating temperature of the nuclear reactor 101, such as at temperatures less than about 900° C., less than about 800° C., less than about 700° C., less than about 600° C., less than about 500° C., or less than about 400° C. The heat transfer system 125 may be thermally conductive at temperatures greater than the normal operating temperature of the nuclear reactor 101. Although FIG. 1B and FIG. 1C have been described and illustrated as including the fins 132 having the overlap H with the fins 138, the disclosure is not so limited. In other embodiments, the fins 132 may not exhibit an overlap with the fins 138. FIG. 1E is a simplified cross-sectional view of a heat transfer system 125′, in accordance with embodiments of the disclosure. The heat transfer system 125′ may be substantially similar to the heat transfer system 125 of FIG. 1B, except that the fins 132 may not exhibit an overlap H with the fins 138. Rather, a gap G may be between an end portion of the fins 132 and an end portion of the fins 138. In other words, the fins 132 may be spaced from the fins 138 by the gap G in the radial direction. In some such embodiments, the heat transfer system 125′ may exhibit thermally insulative properties at higher temperatures compared to heat transfer systems with an overlap H between the fins 132 and the fins 138 while exhibiting thermally conductive properties above a threshold temperature greater than a temperature at which heat transfer systems including an overlap H exhibit thermally conductive properties. In yet other embodiments, and as described above, in some embodiments, a length of the fins 132 may be different than the length of the fins 138. For example, referring to FIG. 1F, in some embodiments, a heat transfer system 125″ may include fins 138 having a length that is greater than a length of the fins 132. In yet other embodiments, the fins 132 may have a length that is greater than the length of the fins 138. Although the fins 132, 138 of FIG. 1B through FIG. 1F have been described and illustrated as having a particular shape, the disclosure is not so limited. FIG. 1G is a simplified cross-sectional view of a heat transfer system 125′″ including fins 132′, 138′ having a torsional shape and including arcuate (e.g., curved) surfaces, in accordance with embodiments of the disclosure. In some embodiments, the curved surfaces of the fins 132′, 138′ may facilitate improved radiative heat transfer from the fins 132′ to the fins 138′. With continued reference to FIG. 1G, in some embodiments, the fins 132, 138 may not overlap with each other. In some such embodiments, a gap may be between an end portion of the fins 132 and an end portion of the fins 138, as described above with reference to FIG. 1F. In some embodiments, the heat transfer system may be disposed around an external reflector or other heat conducting structure disposed around a nuclear reactor. FIG. 1H is a simplified cross-sectional view of a heat transfer system 175 disposed around an external reflector 180 of a nuclear reactor, in accordance with embodiments of the disclosure. The nuclear reactor may include the reactor core 102, a reactor reflector 115 disposed around the reactor core 102, and a vessel 105 disposed around the reactor reflector 115. An external reflector 180 or other heat-conducting structure may be disposed around the vessel 105 and located between the vessel 105 and the heat transfer system 175. In some embodiments, the external reflector 180 exhibits a square or rectangular cross-sectional shape. The heat transfer system 175 may be disposed around the external reflector 180. The heat transfer system 175 may be substantially similar to any of the heat transfer systems 125, 125′, 125″, and 125′″ described above with reference to FIG. 1A through FIG. 1G, except that the heat transfer system 175 may exhibit a square or rectangular shape corresponding to the shape of the external reflector 180. Although FIG. 1A has been described and illustrated as including a nuclear reactor 101 including a reactor core 102 comprising a pebble bed reactor, the disclosure is not so limited. In other embodiments, the reactor core 102 may include other types of reactors, such as, for example, a molten salt reactor (MSR), an ionic salt reactor (ISR), a prismatic block nuclear reactor (e.g., a small modular reactor (SMR)), a heat pipes reactor, a pressurized water reactor (PWR), a light-water reactor (LWR), or another type of nuclear reactor. FIG. 2A is a simplified cross-sectional view of a nuclear reactor 200, in accordance with embodiments of the disclosure. In some embodiments, the nuclear reactor 200 comprises a prismatic nuclear reactor. The nuclear reactor 200 may include a core barrel 210 enclosing a plurality of prismatic fuel blocks 202 (which may also be referred to as “fuel assemblies”) comprising, for example, nuclear fuel materials arranged therein. A reactor vessel 205 may surround the core barrel 210. The fuel assemblies or prismatic fuel blocks 202 may be patterned to surround controlled fuel assemblies or controlled fuel blocks 204 with channels for the insertion of control rods in a hexagonal pattern. In other words, each controlled prismatic fuel block 204 may be surrounded by six prismatic fuel blocks 202. The prismatic fuel blocks 202 and the controlled fuel blocks 204 may collectively form a core of the nuclear reactor 200 and may collectively be referred to herein as the “reactor core” of the nuclear reactor 200 or the “nuclear reactor core.” Each of the prismatic fuel blocks 202 and controlled prismatic fuel blocks 204 comprise one or more coolant channels. Heat generated from the prismatic nuclear fuel blocks 202 and the controlled prismatic fuel blocks 204 may be transferred to a fluid medium within the coolant channels that is, in turn, used to generate electricity. The nuclear reactor 200 further includes replaceable reflector materials 206 arranged within a central portion of the nuclear reactor core 200 and around an outer portion of the annular pattern of the prismatic fuel blocks 202 and the controlled prismatic fuel blocks 204. The replaceable reflector materials 206 may comprise a material formulated and configured to reflect neutrons and reduce or substantially prevent stray neutrons from traveling outside the reactor core of the nuclear reactor 200. The replaceable reflector materials 206 comprise graphite, beryllium, beryllium oxide, or another reflector material. In some embodiments, the replaceable reflector material 206 comprises graphite. A permanent reflector material 208 may be disposed around an inner portion of the core barrel 210. The permanent reflector material 208 may include one or more of the materials described above with reference to the replaceable reflector materials 206. The nuclear reactor 200 may further include control rods comprising, for example, a reactor poison formulated and configured to stop neutron reactions within the fuel of the nuclear reactor 200. In some embodiments, the reactor poison material comprises boron carbide (B4C). In addition, the periphery beyond the nuclear reactor core of the nuclear reactor 200 may include one or more neutron moderators and may further include one or both of a neutron shield and a lead gamma shield. Such materials and structures are known in the art and are, therefore, not described in detail herein. In some embodiments, the nuclear reactor core may be substantially similar to one of the reactor cores described in U.S. Patent Application No. 2018/0226159, titled “Modular Nuclear Reactors, Fuel Structures, and Related Methods,” the entire disclosure of which is hereby incorporated herein by this reference. For example, the nuclear reactor core may include heat pipes that are heated by fuel rods embedded within graphite blocks in a reactor core, as described in U.S. Patent Application No. 2018/0226159. The heat pipes may extend out of the reactor core and into a heat exchanger block. The material heated within the heat pipes (e.g., liquid sodium or liquid potassium that has vaporized) may be cooled to heat a working fluid used to generate electricity, as described in U.S. Patent Application No. 2018/0226159. In some embodiments, the reactor core of U.S. Patent Application No. 2018/0226159 may replace the reactor core 102 of FIG. 1A. In some such embodiments, the reactor core of U.S. Patent Application No. 2018/0226159 may be surrounded by a heat transfer system structured and configured to remove excess heat from the reactor core of U.S. Patent Application No. 2018/0226159 during, for example, situations where the heat removed from the reactor core by the heat exchanger (e.g., the heat exchanger 118 (FIG. 1A)) is less than the heat generated by the reactor core. In some embodiments, the nuclear reactor core of the nuclear reactor 200 and reflectors 206 of FIG. 2A may replace the reactor core 102 and the reflector 115 of FIG. 1B or replace the reactor core 102 and reflector 115 of FIG. 1A. In some such embodiments, the reactor core 200 of FIG. 2A may be surrounded by a heat transfer system structured and configured to remove excess heat from the nuclear reactor 200 during, for example, situations where the heat removed from the nuclear reactor 200 by the coolant fluid 114, 116 and rejected via the heat exchanger (e.g., the heat exchanger 118 (FIG. 1A) is less than the heat generated in the nuclear reactor 200). FIG. 2B is a simplified cross-sectional view of a heat transfer system 250, in accordance with embodiments of the disclosure. The heat transfer system 250 may be substantially the same as any of the heat transfer systems 125, 125′, 125″, 125′″, 175 described above with reference to FIG. 1B through FIG. 1H, except that the heat transfer system 250 is associated with a different nuclear reactor, such as the nuclear reactor 200. The heat transfer system 250 may include fins 232 coupled to an outer surface 230 of an inner wall 203 surrounding the reactor vessel 205. The heat transfer system 250 may include substantially the same structure described above with reference to the heat transfer systems 125, 125′, 125″, 125′″, 175 described above. For example, the heat transfer system 250 may include fins 232 extending from the outer surface 230 of the inner wall 203 and extending into a volume 234 between the outer surface 230 of the inner wall 203 and an inner surface 236 of an outer wall 235. Fins 238 may extend from the inner surface 236 of the outer wall 235 into the volume 234 and may alternate with the fins 232, as described above with reference to the fins 132, 138. An outer surface 240 of the outer wall 235 may be in contact with an environment 150, as described above with reference to the outer surface 140. In addition, one or more heat transfer structures may be coupled to the outer surface 240 of the outer wall 235, as described above with reference to heat transfer structures 142. In some embodiments, the heat transfer systems (e.g., the heat transfer systems 125, 125′, 125″, 125′″, 175, 250) described herein may be sealed at their respective top and bottom portions. For example, the heat transfer systems may be sealed at a bottom portion thereof by a bottom support structure (e.g., a thermally insulating supporting flat structure) and sealed at a top portion thereof by a top cap structure (e.g., an insulating top structure) that may be shaped as an annular slab or as a slab with a shape conforming to the shape of the heat transfer systems. Each of the bottom support structure and the top cap structure may exhibit a shape to conform to respective bottom and top portions of the heat transfer systems. For example, each of the bottom support structure and the top cap structure may exhibit a shape corresponding to the cross-sectional shape of the bottom portion and top portion of the heat transfer systems (e.g., corresponding to the shape of the inner wall 103, the outer wall 135, and the fins 132, 138 therebetween). For example, each of the bottom support structure and the top cap structure may exhibit a shape corresponding to the cross-sectional shape of the bottom portion and top portion of the heat transfer systems (e.g., corresponding to the shape of only the inner wall 103, and only the outer wall 135 if the fins 132, 138 therebetween do not extend the full height of the inner wall 103 and outer wall 135). Each of the bottom support structure and the top cap structure may comprise a thermally insulating refractor material, such as, for example, alumina, calcium silicate materials, kaolin, and zirconia. Accordingly, a heat transfer system may be coupled to a nuclear reactor and structured, sized, and shaped to substantially passively remove fission heat or decay heat from the reactor core by thermal radiation. The heat transfer system may provide cooling to the nuclear reactor when the nuclear reactor is lacking (e.g., without) a circulating fluid. In some embodiments, even if some of the components of the heat transfer system fail (e.g., one or more of the fins 132, 138, 132′, 138′, 232, 238), the heat transfer system may still remove heat from the nuclear reactor by thermal radiation. If the failure involves misalignment of fins and establishment of physical contact between fins attached to the inner wall and fins attached to the outer wall, a conduction path is established and the heat transfer system may still remove heat by radiation and conduction. By way of comparison, failure of a circulation pump or blower, or impediment of a circulation path of a conventional heat transfer system to a heat sink in a conventional reactor may result in insufficient heat removal from the conventional nuclear reactor. Accordingly, the heat transfer systems described herein may exhibit a fail-safe failure mode, meaning that the heat transfer system may continue to effectively remove heat from the nuclear reactor during emergency situations. The surface area of the fins 132, 138, 132′, 138′, 232, 238 and the number of the fins 132, 138, 132′, 138′, 232, 238 (and the associated view factor) may be selected depending on the desired heat removal capacity of the heat transfer system. The heat transfer systems described herein may be configured to facilitate safe removal of heat from reactor cores having a greater power density compared to conventional radial heat paths for conventional nuclear reactors at least partially because the heat transfer systems described herein may exhibit a higher surface area for heat removal due to the total surface area of the fins 132, 138, 132′, 138′, 232, 238. The heat transfer systems described herein may be used on any type of reactor core to provide the reactor core a sufficient heat transfer path to an ultimate heat sink. The heat transfer system may be configured to function as a heat transfer regulator that is configured to shift from exhibiting insulative properties (i.e., an insulator) to exhibiting enhanced radiative heat transfer properties, depending on the temperature of the nuclear reactor and, accordingly, the corresponding temperature of the components of the heat transfer system. In some embodiments, the heat transfer system is configured to remove heat from the nuclear reactor when the temperature of the nuclear reactor exceeds a predetermined temperature and to return to exhibit thermally insulative properties when the temperature of the nuclear reactor is less than the predetermined temperature. Accordingly, the heat transfer system may be configured to operate as a heat transfer valve configured to transition between exhibiting thermally insulative properties to facilitating heat transfer from the nuclear reactor to an ultimate heat sink, depending on the temperature of the nuclear reactor. While embodiments of the disclosure may be susceptible to various modifications and alternative forms, specific embodiments have been shown by way of example in the drawings and have been described in detail herein. However, it should be understood that the disclosure is not limited to the particular forms disclosed. Rather, the disclosure encompasses all modifications, variations, combinations, and alternatives falling within the scope of the disclosure as defined by the following appended claims and their legal equivalents. |
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claims | 1. A riser brace clamp assembly for clamping a riser brace assembly to a jet pump riser pipe in a nuclear reactor, comprising:an upper clamp assembly having an upper clamp and an upper frame, the upper clamp being substantially U-shaped so as to be configured to be secured around the jet pump riser pipe, and having two end portions, each of the two end portions including an opening therein;a lower clamp assembly having a lower clamp and a lower frame, the lower clamp being substantially U-shaped so as to be configured to be secured around the jet pump riser pipe, and having two end portions, each of the two end portions including an opening therein;a plurality of mechanical fasteners for clamping the upper clamp assembly and the lower clamp assembly together,wherein the upper frame includes a respective recess for receiving one of the two end portions of the upper clamp respectively, and the lower frame includes a respective recess for receiving one of the two end portions of the lower clamp respectively, respective recesses and openings being vertically aligned so that at least one of the plurality of mechanical fasteners extends through respective recesses and openings. 2. The riser brace clamp assembly according to claim 1, wherein the plurality of mechanical fasteners are clamping bolts and clamping bolt nuts. 3. The riser brace clamp assembly according to claim 2, wherein the clamping bolt nuts engage a bolt nut latch spring so as to permit the rotation of the bolt nuts in only one direction. 4. The riser brace clamp assembly according to claim 1, wherein the upper clamp further comprises:an adjustable wedge;a jack bolt for adjusting the wedge; anda jack bolt latch spring for preventing the jack bolt to move in one direction. 5. The riser brace clamp assembly according to claim 1, wherein the lower clamp further comprises:an adjustable wedge;a jack bolt for adjusting the wedge; anda jack bolt latch spring for preventing the jack bolt to move in one direction. 6. The riser brace clamp assembly of claim 1, wherein the upper frame further comprises shear pads for preventing deformation and bending of the upper and lower clamps when a clamping force is applied. 7. The riser brace clamp assembly of claim 1, wherein the lower frame further comprises shear pads for preventing deformation and bending of the upper and lower clamps when a clamping force is applied. 8. The riser brace clamp assembly of claim 6, wherein the upper frame further comprises a latch spring for permitting rotation of a clamp bolt nut in only one direction. 9. The riser brace clamp assembly of claim 7, wherein the lower frame further comprises a square counter bore recess which engages with a clamp bolt to prevent rotation of the bolt. 10. The riser brace clamp assembly of claim 1, wherein the upper clamp includes a latch spring for preventing rotation of a bolt nut. |
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047449381 | abstract | An alpha recoil ion-implantation method and apparatus are described which use an alpha-emitting source that is a radioactive parent of the daughter isotope of interest to implant into a suitable substrate the recoil daughter ions resulting from alpha decay of the parent. For example, a .sup.241 Am source in thin layer form is placed next to a substrate such as a solid state track recorder in a vacuum which houses an assembly for rotating opposing disks receiving the alpha-emitting source and the substrate, respectively. Each alpha decay of .sup.241 Am results in a .sup.237 Np ion with enough recoil energy to be implanted into the substrate. Fissionable deposits of .sup.239 Pu, .sup.235 U, and .sup.238 U can also be made by this method and apparatus. |
047132132 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The invention concerns a nuclear reactor plant housed in a steel pressure vessel. More particularly, a plant with a gas cooled, small high temperature reactor, the core of which contains a pile of spherical fuel elements through which a cooling gas flows from bottom to top. A heat utilization system placed in the flow of cooling gas is installed above the small high temperature reactor in a reactor pressure vessel, followed in line preferably by two circulating blowers with at least one decay heat exchanger installed in the steel pressure vessel. 2. Description of the Prior Art A nuclear reactor installation is described in DE P No. 34 35 255.4. In this plant the decay heat exchangers on the primary side are installed immediately following in the direction of flow the main heat exchangers forming the heat utilization system and are traversed constantly by the entire flow of cooling gas. They are arranged under the main heat exchanger. On the secondary side the decay heat exchangers are connected by means of a decay heat removal water circulation loop with a geodesically higher located recooling heat exchanger, which in turn is connected with a further heat sink, such as for example a cooling tower. A plant for the nuclear generation of heat is shown in DE-OS No. 32 12 264, which also uses a small high temperature reactor as the source of energy. This small high temperature reactor is designed so that different heat utilization systems, such as steam generators, split tubular furnaces or He/He heat exchangers may be connected with it. There are no special installations in this plant for the removal of decay heat. SUMMARY OF THE INVENTION It is an object of the present invention to design a nuclear reactor plant housed in a steel pressure vessel, with a gas cooled small high temperature reactor, the core of which contains a pile of spherical fuel elements traversed from bottom to top by a cooling gas, with a heat utilization system arranged in the flow of cooling gas and installed above the small high temperature reactor in the reactor pressure vessel and which is followed in line preferably by two circulating blowers, and with at least one decay heat exchanger installed in the steel pressure vessel so that heat generated may be removed in an intermediate circulation loop and the safe removal of the decay heat from the reactor core is assured. This object is attained by the following characteristics: a. The heat utilization system comprises He/He heat exchanger, in which the primary helium transfers its heat to a secondary helium flow circulating in an intermediate loop; PA0 b. The He/He heat exchanger comprises at least one annular helix coil bundle extending to a hot gas collector chamber located above the reactor core and exposed from below to hot gas; PA0 c. The decay heat exchanger is arranged in the direction of flow immediately following the He/He heat exchanger and is constantly traversed by the entire flow of the cooling gas; PA0 d. The circulating blowers located in the flow of cold gas are connected parallel with respect to each other. The hot gas coming from the small high temperature reactor flows through a roof reflector into the hot gas collector chamber and enters the He/He heat exchanger from below, where it flows around the bundle of tubes while transferring its heat to the secondary helium flowing in the tubes. The cooled primary helium is conducted to the circulating blowers. After passing through the circulating blowers the cold primary helium is returned to the reactor core, which it enters from below. Decay heat is removed by means of the decay heat exchangers located in the primary circulation loop, which in normal operation is exposed on the primary side to the temperature of the cold gas. No special butterfly valves or blowers are required for the decay heat exchangers and none are therefore provided. In case of a failure of the two circulating blowers for the He/He heat exchangers, the decay heat is removed by natural convection. As the He/He heat exchanger is under the same pressure on the primary and the secondary side, it is not endangered by the decay heat removal operation. An alternative solution may be seen in removing the decay heat of a small high temperature reactor combined with a He/He heat exchanger by means of the heat exchangers arranged in the secondary heat circulation loop. Advantageous further developments of the invention are set forth in the dependent claims and the description below of a preferred embodiment. |
049869604 | claims | 1. An upper end fitting for a nuclear fuel assembly, comprising: a. a main body portion; b. a lower plate slidably attached to said main body portion; and c. a substantially hairpin shaped spring positioned between said main body portion and said lower plate whereby said main body portion is resiliently biased away from said lower plate. a. a main body portion which is substantially box shaped having two open opposed sides; b. a lower plate slidably attached to said main body portion; and c. a substantially hairpin shaped spring positioned along each side of said main body portion between said main body portion and said lower plate whereby said main body portion is resiliently biased away from said lower plate. a. a main body portion which is substantially box shaped having two open opposed ends; b. a lower plate provided with a pin at each corner thereof extending perpendicular to the surface of said plate, said main body portion being slidably mounted on said pins; and c. a substantially hairpin shaped spring positioned along each side of said main body portion between said main body portion and said lower plate whereby said main body portion is resiliently biased away from said lower plate. 2. The end fitting of claim 1, wherein one of said springs is positioned along each side of said main body portion. 3. The end fitting of claim 1, wherein said main body portion is substantially box shaped having two open opposed sides. 4. The end fitting of claim 1, wherein the end of said spring against said lower plate is offset from the end of said plate. 5. An upper end fitting for a nuclear fuel assembly, comprising: 6. The end fitting of claim 5, wherein the end of said spring against said lower plate is offset from the end of said plate. 7. An upper end fitting for a nuclear fuel assembly, comprising: 8. The end fitting of claim 7, wherein the end of said spring against said lower plate is offset from the end thereof. 9. The end fitting of claim 7, further comprising means for preventing said main body portion from sliding off said pins. |
description | This application claims priority under 35 U.S.C. § 119 to: U.S. Provisional Patent Application No. 62/781,337, filed on Dec. 18, 2018; U.S. Provisional Patent Application No. 62/784,991, filed on Dec. 26, 2018; and U.S. Provisional Patent Application No. 62/840,216, filed on Apr. 29, 2019. The entire contents of each of the previous applications are incorporated by reference herein. This disclosure relates to radioactive waste repository systems and methods. Hazardous waste, such as radioactive waste, is often placed in long-term, permanent, or semi-permanent storage so as to prevent health issues among a population living near the stored waste. Such hazardous waste storage is often challenging, for example, in terms of storage location identification and surety of containment. For instance, the safe storage of nuclear waste (e.g., spent nuclear fuel, whether from commercial power reactors, test reactors, or even high-grade military waste) is considered to be one of the outstanding challenges of energy technology. Safe storage of the long-lived radioactive waste is a major impediment to the adoption of nuclear power in the United States and around the world. Conventional waste storage methods have emphasized the use of tunnels, and is exemplified by the design of the Yucca Mountain storage facility. Other techniques include boreholes, including vertical boreholes, drilled into crystalline basement rock. Other conventional techniques include forming a tunnel with boreholes emanating from the walls of the tunnel in shallow formations to allow human access. In a general implementation, a hazardous material repository includes a drillhole formed from a terranean surface into a subterranean zone that includes a geologic formation, where the drillhole includes a vertical portion and a non-vertical portion coupled to the vertical portion by a transition portion, the non-vertical portion includes a storage volume for hazardous waste; a casing installed between the geologic formation and the drillhole, the casing including one or more metallic tubular sections; at least one canister positioned in the storage volume of the non-vertical portion of the drillhole, the at least one canister sized to enclose a portion of hazardous material and including an outer housing formed from a non-corrosive metallic material; and a backfill material inserted into the non-vertical portion of the drillhole to fill at least a portion of the storage volume between the at least one canister and the casing. In an aspect combinable with the general implementation, the hazardous material includes radioactive material waste. In another aspect combinable with any one of the previous aspects, the radioactive material waste includes spent nuclear fuel. In another aspect combinable with any one of the previous aspects, the non-corrosive material includes an alloy that includes at least one of nickel, chromium, or molybdenum. In another aspect combinable with any one of the previous aspects, the non-corrosive material includes a nickel-chromium-molybdenum alloy. In another aspect combinable with any one of the previous aspects, the alloy includes Alloy 625. In another aspect combinable with any one of the previous aspects, the nickel-chromium-molybdenum alloy includes a bulk corrosion resistant alloy. In another aspect combinable with any one of the previous aspects, the nickel-chromium-molybdenum alloy includes a clad corrosion resistant metal. In another aspect combinable with any one of the previous aspects, the casing includes a carbon steel alloy. In another aspect combinable with any one of the previous aspects, the backfill includes a slurry pumped from the terranean surface into the non-vertical portion of the drillhole to fill the portion of the storage volume between the at least one canister and the casing. In another aspect combinable with any one of the previous aspects, the slurry includes bentonite. In another aspect combinable with any one of the previous aspects, the slurry includes a bentonite-based slurry. In another aspect combinable with any one of the previous aspects, the storage volume is at an oxidizing environmental state during a first time period that begins at placement of the at least one canister into the storage volume. In another aspect combinable with any one of the previous aspects, the at least one canister includes a corrosion rate of about 2 μm/year during the first time period. In another aspect combinable with any one of the previous aspects, the casing includes a corrosion rate of about 20 μm/year during the first time period. In another aspect combinable with any one of the previous aspects, the first time period is a first time duration from the placement of the at least one canister into the storage volume. In another aspect combinable with any one of the previous aspects, the storage volume is at a first reducing environmental state during a second time period that begins at an end of the first time period. In another aspect combinable with any one of the previous aspects, the at least one canister includes a corrosion rate of about 2 μm/year during the second time period. In another aspect combinable with any one of the previous aspects, the casing includes a corrosion rate of about 4 μm/year during the second time period. In another aspect combinable with any one of the previous aspects, the second time period extends from an end of the first time duration to a second time duration from the placement of the at least one canister into the storage volume. In another aspect combinable with any one of the previous aspects, the storage volume is at a second reducing environmental state during a third time period that begins at an end of the second time period. In another aspect combinable with any one of the previous aspects, the at least one canister includes a corrosion rate of about 1 μm/year during the third time period. In another aspect combinable with any one of the previous aspects, the casing includes a corrosion rate of about 2 μm/year during the third time period. In another aspect combinable with any one of the previous aspects, the third time period extends from an end of the second time duration to a third time duration from the placement of the at least one canister into the storage volume. In another aspect combinable with any one of the previous aspects, the storage volume is at a third reducing environmental state during a fourth time period that begins at an end of the third time period. In another aspect combinable with any one of the previous aspects, the at least one canister includes a corrosion rate of about 0.1 μm/year during the fourth time period. In another aspect combinable with any one of the previous aspects, the casing includes a corrosion rate of about 1 μm/year during the fourth time period. In another aspect combinable with any one of the previous aspects, the fourth time period extends from an end of the third time duration to a fourth time duration from the placement of the at least one canister into the storage volume. In another aspect combinable with any one of the previous aspects, the storage volume is at a fourth reducing environmental state during a fifth time period that begins at an end of the fourth time period. In another aspect combinable with any one of the previous aspects, the at least one canister includes a corrosion rate of about 0.1 μm/year during the fifth time period. In another aspect combinable with any one of the previous aspects, the casing includes a corrosion rate of about 1 μm/year during the fifth time period. In another aspect combinable with any one of the previous aspects, the fifth time period extends from an end of the fourth time duration to a fifth duration from the placement of the at least one canister into the storage volume. In another aspect combinable with any one of the previous aspects, the geologic formation is at a depth in which a hydrostatic pressure at the depth is great enough to prevent boiling of water. In another aspect combinable with any one of the previous aspects, the geologic formation includes pore water that is highly reducing. In another aspect combinable with any one of the previous aspects, the geologic formation includes a rock in which pore waters are anoxic. In another aspect combinable with any one of the previous aspects, the geologic formation includes a fully saturated rock formation. In another aspect combinable with any one of the previous aspects, the non-vertical portion includes a horizontal drillhole. In another aspect combinable with any one of the previous aspects, a thermal load of the hazardous material repository is controlled by spacing of the at least one canister within the storage volume. In another aspect combinable with any one of the previous aspects, a temperatures in the storage volume is controlled by spacing of the at least one canister within the storage volume. In another aspect combinable with any one of the previous aspects, the at least one canister includes a plurality of canisters arranged linearly in the storage volume. In another aspect combinable with any one of the previous aspects, the linear arrangement allows for uniform conditions along the storage volume. In another aspect combinable with any one of the previous aspects, the at least one canister is resistant to corrosion during a heat up and cool down cycle caused by heat from the hazardous waste. In another aspect combinable with any one of the previous aspects, an environment of the storage volume is controlled to reduce corrosivity during the heat up and cool down cycle. In another aspect combinable with any one of the previous aspects, the at least one canister is resistant to corrosion in a high temperature, oxidizing water environment. In another aspect combinable with any one of the previous aspects, the at least one canister includes a protective film by application of a surface treatment or coating. In another aspect combinable with any one of the previous aspects, oxygen in the drillhole is minimized during formation of the hazardous material repository. Another aspect combinable with any one of the previous aspects further includes oxygen scavengers in the drillhole to consume or tie up oxygen. In another aspect combinable with any one of the previous aspects, temperature or pressure or both is controlled in the drillhole to control a formation of metal oxide corrosion products. Another aspect combinable with any one of the previous aspects further includes one or more expansion absorbers. In another aspect combinable with any one of the previous aspects, the one or more expansion absorbers are placed at predetermined locations in the casing. In another aspect combinable with any one of the previous aspects, the non-vertical portion includes an expansion leg at an end of non-vertical portion. In another aspect combinable with any one of the previous aspects, the casing includes an engineered filling to control corrosion. In another aspect combinable with any one of the previous aspects, the engineered filling includes a bentonite-based slurry to control corrosion. In another aspect combinable with any one of the previous aspects, the engineering filling modifies the environment to be mildly alkaline. In another aspect combinable with any one of the previous aspects, the engineering filling reduces oxygen in the drillhole. In another aspect combinable with any one of the previous aspects, the engineering filling mitigates microbial activity. Another aspect combinable with any one of the previous aspects further includes a cementitious material between the casing and geologic formation to control corrosion of the casing. Another aspect combinable with any one of the previous aspects further includes a material between the casing and geologic formation to modify the drillhole to be mildly alkaline. Another aspect combinable with any one of the previous aspects further includes a bentonite-based slurries between the casing and geologic formation to control corrosion of the casing. In another general implementation, a method for forming an engineered barrier system for a hazardous material repository includes forming a drillhole from a terranean surface into a subterranean zone that includes a geologic formation, where the drillhole includes a vertical portion and a non-vertical portion coupled to the vertical portion by a transition portion, the non-vertical portion includes a storage volume for hazardous waste; installing a casing between the geologic formation and the drillhole, the casing including one or more metallic tubular sections; positioning at least one canister in the storage volume of the non-vertical portion of the drillhole, the at least one canister enclosing a portion of hazardous material and including an outer housing formed from a non-corrosive metallic material; and inserting a backfill material into the non-vertical portion of the drillhole to fill at least a portion of the storage volume between the at least one canister and the casing. In an aspect combinable with the general implementation, the hazardous material includes radioactive material waste. In another aspect combinable with any one of the previous aspects, the radioactive material waste includes spent nuclear fuel. In another aspect combinable with any one of the previous aspects, the non-corrosive material includes an alloy that includes at least one of nickel, chromium, or molybdenum. In another aspect combinable with any one of the previous aspects, the alloy includes Alloy 625. In another aspect combinable with any one of the previous aspects, the casing includes a carbon steel alloy. In another aspect combinable with any one of the previous aspects, the backfill includes a slurry pumped from the terranean surface into the non-vertical portion of the drillhole to fill the portion of the storage volume between the at least one canister and the casing. In another aspect combinable with any one of the previous aspects, the slurry includes bentonite. In another aspect combinable with any one of the previous aspects, the storage volume is at an oxidizing environmental state during a first time period that begins at placement of the at least one canister into the storage volume. In another aspect combinable with any one of the previous aspects, the at least one canister includes a corrosion rate of about 2 μm/year during the first time period. In another aspect combinable with any one of the previous aspects, the casing includes a corrosion rate of about 20 μm/year during the first time period. In another aspect combinable with any one of the previous aspects, the first time period is a first time duration from the placement of the at least one canister into the storage volume. In another aspect combinable with any one of the previous aspects, the storage volume is at a first reducing environmental state during a second time period that begins at an end of the first time period. In another aspect combinable with any one of the previous aspects, the at least one canister includes a corrosion rate of about 2 μm/year during the second time period. In another aspect combinable with any one of the previous aspects, the casing includes a corrosion rate of about 4 μm/year during the second time period. In another aspect combinable with any one of the previous aspects, the second time period extends from an end of the first time duration to a second time duration from the placement of the at least one canister into the storage volume. In another aspect combinable with any one of the previous aspects, the storage volume is at a second reducing environmental state during a third time period that begins at an end of the second time period. In another aspect combinable with any one of the previous aspects, the at least one canister includes a corrosion rate of about 1 μm/year during the third time period. In another aspect combinable with any one of the previous aspects, the casing includes a corrosion rate of about 2 μm/year during the third time period. In another aspect combinable with any one of the previous aspects, the third time period extends from an end of the second time duration to a third time duration from the placement of the at least one canister into the storage volume. In another aspect combinable with any one of the previous aspects, the storage volume is at a third reducing environmental state during a fourth time period that begins at an end of the third time period. In another aspect combinable with any one of the previous aspects, the at least one canister includes a corrosion rate of about 0.1 μm/year during the fourth time period. In another aspect combinable with any one of the previous aspects, the casing includes a corrosion rate of about 1 μm/year during the fourth time period. In another aspect combinable with any one of the previous aspects, the fourth time period extends from an end of the third time duration to a fourth time duration from the placement of the at least one canister into the storage volume. In another aspect combinable with any one of the previous aspects, the storage volume is at a fourth reducing environmental state during a fifth time period that begins at an end of the fourth time period. In another aspect combinable with any one of the previous aspects, the at least one canister includes a corrosion rate of about 0.1 μm/year during the fifth time period. In another aspect combinable with any one of the previous aspects, the casing includes a corrosion rate of about 1 μm/year during the fifth time period. In another aspect combinable with any one of the previous aspects, the fifth time period extends from an end of the fourth time duration to a fifth time duration from the placement of the at least one canister into the storage volume. In another aspect combinable with any one of the previous aspects, the geologic formation is at a depth in which a hydrostatic pressure at the depth is great enough to prevent boiling of water. In another aspect combinable with any one of the previous aspects, the geologic formation includes pore water that is highly reducing. In another aspect combinable with any one of the previous aspects, the geologic formation includes a rock in which pore waters are anoxic. In another aspect combinable with any one of the previous aspects, the geologic formation includes a fully saturated rock formation. In another aspect combinable with any one of the previous aspects, the non-vertical portion includes a horizontal drillhole. In another aspect combinable with any one of the previous aspects, a thermal load of the hazardous material repository is controlled by spacing of the at least one canister within the storage volume. In another aspect combinable with any one of the previous aspects, a temperatures in the storage volume is controlled by spacing of the at least one canister within the storage volume. In another aspect combinable with any one of the previous aspects, the at least one canister includes a plurality of canisters arranged linearly in the storage volume. In another aspect combinable with any one of the previous aspects, the linear arrangement allows for uniform conditions along the storage volume. In another aspect combinable with any one of the previous aspects, the at least one canister is resistant to corrosion during a heat up and cool down cycle caused by heat from the hazardous waste. In another aspect combinable with any one of the previous aspects, an environment of the storage volume is controlled to reduce corrosivity during the heat up and cool down cycle. In another aspect combinable with any one of the previous aspects, the at least one canister is resistant to corrosion in a high temperature, oxidizing water environment. In another aspect combinable with any one of the previous aspects, the at least one canister includes a protective film by application of a surface treatment or coating. In another aspect combinable with any one of the previous aspects, oxygen in the drillhole is minimized during formation of the hazardous material repository. Another aspect combinable with any one of the previous aspects further includes oxygen scavengers in the drillhole to consume or tie up oxygen. In another aspect combinable with any one of the previous aspects, temperature or pressure or both is controlled in the drillhole to control a formation of metal oxide corrosion products. Another aspect combinable with any one of the previous aspects further includes one or more expansion absorbers. In another aspect combinable with any one of the previous aspects, the one or more expansion absorbers are placed at predetermined locations in the casing. In another aspect combinable with any one of the previous aspects, the non-vertical portion includes an expansion leg at an end of non-vertical portion. In another aspect combinable with any one of the previous aspects, the casing includes an engineered filling to control corrosion. In another aspect combinable with any one of the previous aspects, the engineered filling includes a bentonite-based slurry to control corrosion. In another aspect combinable with any one of the previous aspects, the engineering filling modifies the environment to be mildly alkaline. In another aspect combinable with any one of the previous aspects, the engineering filling reduces oxygen in the drillhole. In another aspect combinable with any one of the previous aspects, the engineering filling mitigates microbial activity. Another aspect combinable with any one of the previous aspects further includes a cementitious material between the casing and geologic formation to control corrosion of the casing. Another aspect combinable with any one of the previous aspects further includes a material between the casing and geologic formation to modify the drillhole to be mildly alkaline. Another aspect combinable with any one of the previous aspects further includes a bentonite-based slurries between the casing and geologic formation to control corrosion of the casing. Implementations of a hazardous material storage repository according to the present disclosure may include one or more of the following features. For example, a hazardous material storage repository according to the present disclosure may allow for multiple levels of containment of hazardous material within a storage repository located thousands of feet underground, decoupled from any nearby mobile water. A hazardous material storage repository according to the present disclosure may also use proven techniques (e.g., drilling) to create or form a storage area for the hazardous material, in a subterranean zone proven to have fluidly sealed hydrocarbons therein for millions of years. As another example, a hazardous material storage repository according to the present disclosure may provide long-term (e.g., thousands of years) storage for hazardous material (e.g., radioactive waste) in a shale formation that has geologic properties suitable for such storage, including low permeability, thickness, and ductility, among others. In addition, a greater volume of hazardous material may be stored at low cost—relative to conventional storage techniques—due in part to directional drilling techniques that facilitate long horizontal boreholes, often exceeding a mile in length. In addition, rock formations that have geologic properties suitable for such storage may be found in close proximity to sites at which hazardous material may be found or generated, thereby reducing dangers associated with transporting such hazardous material. Implementations of a hazardous material storage repository according to the present disclosure may also include one or more of the following features. Large storage volumes, in turn, allow for the storage of hazardous materials to be emplaced without a need for complex prior treatment, such as concentration or transfer to different forms or canisters. As a further example, in the case of nuclear waste material from a reactor for instance, the waste can be kept in its original pellets, unmodified, or in its original fuel rods, or in its original fuel assemblies, which contain dozens of fuel rods. In another aspect, the hazardous material may be kept in an original holder but a cement or other material is injected into the holder to fill the gaps between the hazardous materials and the structure. For example, if the hazardous material is stored in fuel rods which are, in turn, stored in fuel assemblies, then the spaces between the rods (typically filled with water when inside a nuclear reactor) could be filled with cement or other material to provide yet an additional layer of isolation from the outside world. As yet a further example, secure and low cost storage of hazardous material is facilitated while still permitting retrieval of such material if circumstances deem it advantageous to recover the stored materials. The details of one or more implementations of the subject matter described in this disclosure are set forth in the accompanying drawings and the description below. Other features, aspects, and advantages of the subject matter will become apparent from the description, the drawings, and the claims. FIG. 1 is a schematic illustration of an example implementation of a hazardous material storage repository system 100, e.g., a subterranean location for the long-term (e.g., tens, hundreds, or thousands of years or more), but retrievable, safe and secure storage of hazardous material. For example, this figure illustrates the example hazardous material storage repository system 100 once one or more canisters 126 of hazardous material have been deployed in a subterranean formation 118. As illustrated, the hazardous material storage repository system 100 includes a drillhole 104 formed (e.g., drilled or otherwise) from a terranean surface 102 and through multiple subterranean layers 112, 114, 116, and 118. Although the terranean surface 102 is illustrated as a land surface, terranean surface 102 may be a sub-sea or other underwater surface, such as a lake or an ocean floor or other surface under a body of water. Thus, the present disclosure contemplates that the drillhole 104 may be formed under a body of water from a drilling location on or proximate the body of water. The illustrated drillhole 104 is a directional drillhole in this example of hazardous material storage repository system 100. For instance, the drillhole 104 includes a substantially vertical portion 106 coupled to a radiussed or curved portion 108, which in turn is coupled to a substantially horizontal portion 110. As used in the present disclosure, “substantially” in the context of a drillhole orientation, refers to drillholes that may not be exactly vertical (e.g., exactly perpendicular to the terranean surface 102) or exactly horizontal (e.g., exactly parallel to the terranean surface 102), or exactly inclined at a particular incline angle relative to the terranean surface 102. In other words, vertical drillholes often undulate offset from a true vertical direction, that they might be drilled at an angle that deviates from true vertical, and inclined drillholes often undulate offset from a true incline angle. Further, in some aspects, an inclined drillhole may not have or exhibit an exactly uniform incline (e.g., in degrees) over a length of the drillhole. Instead, the incline of the drillhole may vary over its length (e.g., by 1-5 degrees). As illustrated in this example, the three portions of the drillhole 104—the vertical portion 106, the radiussed portion 108, and the horizontal portion 110—form a continuous drillhole 104 that extends into the Earth. As used in the present disclosure, the drillhole 104 (and drillhole portions described) may also be called wellbores. Thus, as used in the present disclosure, drillhole and wellbore are largely synonymous and refer to bores formed through one or more subterranean formations that are not suitable for human-occupancy (i.e., are too small in diameter for a human to fit therewithin). The illustrated drillhole 104, in this example, has a surface casing 120 positioned and set around the drillhole 104 from the terranean surface 102 into a particular depth in the Earth. For example, the surface casing 120 may be a relatively large-diameter tubular member (or string of members) set (e.g., cemented) around the drillhole 104 in a shallow formation. As used herein, “tubular” may refer to a member that has a circular cross-section, elliptical cross-section, or other shaped cross-section. For example, in this implementation of the hazardous material storage repository system 100, the surface casing 120 extends from the terranean surface through a surface layer 112. The surface layer 112, in this example, is a geologic layer comprised of one or more layered rock formations. In some aspects, the surface layer 112 in this example may or may not include freshwater aquifers, salt water or brine sources, or other sources of mobile water (e.g., water that moves through a geologic formation). In some aspects, the surface casing 120 may isolate the drillhole 104 from such mobile water, and may also provide a hanging location for other casing strings to be installed in the drillhole 104. Further, although not shown, a conductor casing may be set above the surface casing 120 (e.g., between the surface casing 120 and the surface 102 and within the surface layer 112) to prevent drilling fluids from escaping into the surface layer 112. As illustrated, a production casing 122 is positioned and set within the drillhole 104 downhole of the surface casing 120. Although termed a “production” casing, in this example, the casing 122 may or may not have been subject to hydrocarbon production operations. Thus, the casing 122 refers to and includes any form of tubular member that is set (e.g., cemented) in the drillhole 104 downhole of the surface casing 120. In some examples of the hazardous material storage repository system 100, the production casing 122 may begin at an end of the radiussed portion 108 and extend throughout the inclined portion 110. The casing 122 could also extend into the radiussed portion 108 and into the vertical portion 106. As shown, cement 130 is positioned (e.g., pumped) around the casings 120 and 122 in an annulus between the casings 120 and 122 and the drillhole 104. The cement 130, for example, may secure the casings 120 and 122 (and any other casings or liners of the drillhole 104) through the subterranean layers under the terranean surface 102. In some aspects, the cement 130 may be installed along the entire length of the casings (e.g., casings 120 and 122 and any other casings), or the cement 130 could be used along certain portions of the casings if adequate for a particular drillhole 102. The cement 130 can also provide an additional layer of confinement for the hazardous material in canisters 126. The drillhole 104 and associated casings 120 and 122 may be formed with various example dimensions and at various example depths (e.g., true vertical depth, or TVD). For instance, a conductor casing (not shown) may extend down to about 120 feet TVD, with a diameter of between about 28 in. and 60 in. The surface casing 120 may extend down to about 2500 feet TVD, with a diameter of between about 22 in. and 48 in. An intermediate casing (not shown) between the surface casing 120 and production casing 122 may extend down to about 8000 feet TVD, with a diameter of between about 16 in. and 36 in. The production casing 122 may extend inclinedly (e.g., to case the inclined portion 110) with a diameter of between about 11 in. and 22 in. The foregoing dimensions are merely provided as examples and other dimensions (e.g., diameters, TVDs, lengths) are contemplated by the present disclosure. For example, diameters and TVDs may depend on the particular geological composition of one or more of the multiple subterranean layers (112, 114, 116, and 118), particular drilling techniques, as well as a size, shape, or design of a hazardous material canister 126 that contains hazardous material to be deposited in the hazardous material storage repository system 100. In some alternative examples, the production casing 122 (or other casing in the drillhole 104) could be circular in cross-section, elliptical in cross-section, or some other shape. As illustrated, the vertical portion 106 of the drillhole 104 extends through subterranean layers 112, 114, and 116, and, in this example, lands in a subterranean layer 118. As discussed above, the surface layer 112 may or may not include mobile water. In this example, a mobile water layer 114 is below the surface layer 112 (although surface layer 112 may also include one or more sources of mobile water or liquid). For instance, mobile water layer 114 may include one or more sources of mobile water, such as freshwater aquifers, salt water or brine, or other source of mobile water. In this example of hazardous material storage repository system 100, mobile water may be water that moves through a subterranean layer based on a pressure differential across all or a part of the subterranean layer. For example, the mobile water layer 114 may be a permeable geologic formation in which water freely moves (e.g., due to pressure differences or otherwise) within the layer 114. In some aspects, the mobile water layer 114 may be a primary source of human-consumable water in a particular geographic area. Examples of rock formations of which the mobile water layer 114 may be composed include porous sandstones and limestones, among other formations. Other illustrated layers, such as the impermeable layer 116 and the storage layer 118, may include immobile water. Immobile water, in some aspects, is water (e.g., fresh, salt, brine), that is not fit for human or animal consumption, or both. Immobile water, in some aspects, may be water that, by its motion through the layers 116 or 118 (or both), cannot reach the mobile water layer 114, terranean surface 102, or both, within 10,000 years or more (such as to 1,000,000 years). Below the mobile water layer 114, in this example implementation of hazardous material storage repository system 100, is an impermeable layer 116. The impermeable layer 116, in this example, may not allow mobile water to pass through. Thus, relative to the mobile water layer 114, the impermeable layer 116 may have low permeability, e.g., on the order of nanodarcy permeability. Additionally, in this example, the impermeable layer 116 may be a relatively non-ductile (i.e., brittle) geologic formation. One measure of non-ductility is brittleness, which is the ratio of compressive stress to tensile strength. In some examples, the brittleness of the impermeable layer 116 may be between about 20 MPa and 40 MPa. As shown in this example, the impermeable layer 116 is shallower (e.g., closer to the terranean surface 102) than the storage layer 118. In this example rock formations of which the impermeable layer 116 may be composed include, for example, certain kinds of sandstone, mudstone, clay, and slate that exhibit permeability and brittleness properties as described above. In alternative examples, the impermeable layer 116 may be deeper (e.g., further from the terranean surface 102) than the storage layer 118. In such alternative examples, the impermeable layer 116 may be composed of an igneous rock, such as granite. Below the impermeable layer 116 is the storage layer 118. The storage layer 118, in this example, may be chosen as the landing for the horizontal portion 110, which stores the hazardous material, for several reasons. Relative to the impermeable layer 116 or other layers, the storage layer 118 may be thick, e.g., between about 100 and 200 feet of total vertical thickness. Thickness of the storage layer 118 may allow for easier landing and directional drilling, thereby allowing the horizontal portion 110 to be readily emplaced within the storage layer 118 during constructions (e.g., drilling). If formed through an approximate horizontal center of the storage layer 118, the horizontal portion 110 may be surrounded by about 50 to 100 feet of the geologic formation that comprises the storage layer 118. Further, the storage layer 118 may also have only immobile water, e.g., due to a very low permeability of the layer 118 (e.g., on the order of milli- or nanodarcys). In addition, the storage layer 118 may have sufficient ductility, such that a brittleness of the rock formation that comprises the layer 118 is between about 3 MPa and 10 MPa. Examples of rock formations of which the storage layer 118 may be composed include: shale and anhydrite. Further, in some aspects, hazardous material may be stored below the storage layer, even in a permeable formation such as sandstone or limestone, if the storage layer is of sufficient geologic properties to isolate the permeable layer from the mobile water layer 114. In some examples implementations of the hazardous material storage repository system 100, the storage layer 118 (and/or the impermeable layer 116) is composed of shale. Shale, in some examples, may have properties that fit within those described above for the storage layer 118. For example, shale formations may be suitable for a long-term confinement of hazardous material (e.g., in the hazardous material canisters 126), and for their isolation from mobile water layer 114 (e.g., aquifers) and the terranean surface 102. Shale formations may be found relatively deep in the Earth, typically 3000 feet or greater, and placed in isolation below any fresh water aquifers. Other formations may include salt or other impermeable formation layer. Shale formations (or salt or other impermeable formation layers), for instance, may include geologic properties that enhance the long-term (e.g., thousands of years) isolation of material. Such properties, for instance, have been illustrated through the long term storage (e.g., tens of millions of years) of hydrocarbon fluids (e.g., gas, liquid, mixed phase fluid) without escape of substantial fractions of such fluids into surrounding layers (e.g., mobile water layer 114). Indeed, shale has been shown to hold natural gas for millions of years or more, giving it a proven capability for long-term storage of hazardous material. Example shale formations (e.g., Marcellus, Eagle Ford, Barnett, and otherwise) has stratification that contains many redundant sealing layers that have been effective in preventing movement of water, oil, and gas for millions of years, lacks mobile water, and can be expected (e.g., based on geological considerations) to seal hazardous material (e.g., fluids or solids) for thousands of years after deposit. In some aspects, the formation of the storage layer 118 and/or the impermeable layer 116 may form a leakage barrier, or barrier layer to fluid leakage that may be determined, at least in part, by the evidence of the storage capacity of the layer for hydrocarbons or other fluids (e.g., carbon dioxide) for hundreds of years, thousands of years, tens of thousands of years, hundreds of thousands of years, or even millions of years. For example, the barrier layer of the storage layer 118 and/or impermeable layer 116 may be defined by a time constant for leakage of the hazardous material more than 10,000 years (such as between about 10,000 years and 1,000,000 years) based on such evidence of hydrocarbon or other fluid storage. Shale (or salt or other impermeable layer) formations may also be at a suitable depth, e.g., between 3000 and 12,000 feet TVD. Such depths are typically below ground water aquifer (e.g., surface layer 112 and/or mobile water layer 114). Further, the presence of soluble elements in shale, including salt, and the absence of these same elements in aquifer layers, demonstrates a fluid isolation between shale and the aquifer layers. Another particular quality of shale that may advantageously lend itself to hazardous material storage is its clay content, which, in some aspects, provides a measure of ductility greater than that found in other, impermeable rock formations (e.g., impermeable layer 116). For example, shale may be stratified, made up of thinly alternating layers of clays (e.g., between about 20-30% clay by volume) and other minerals. Such a composition may make shale less brittle and, thus less susceptible to fracturing (e.g., naturally or otherwise) as compared to rock formations in the impermeable layer (e.g., dolomite or otherwise). For example, rock formations in the impermeable layer 116 may have suitable permeability for the long term storage of hazardous material, but are too brittle and commonly are fractured. Thus, such formations may not have sufficient sealing qualities (as evidenced through their geologic properties) for the long term storage of hazardous material. The present disclosure contemplates that there may be many other layers between or among the illustrated subterranean layers 112, 114, 116, and 118. For example, there may be repeating patterns (e.g., vertically), of one or more of the mobile water layer 114, impermeable layer 116, and storage layer 118. Further, in some instances, the storage layer 118 may be directly adjacent (e.g., vertically) the mobile water layer 114, i.e., without an intervening impermeable layer 116. In some examples, all or portions of the radiussed drillhole 108 and the horizontal drillhole 110 may be formed below the storage layer 118, such that the storage layer 118 (e.g., shale or other geologic formation with characteristics as described herein) is vertically positioned between the horizontal drillhole 110 and the mobile water layer 114. In this example, the horizontal portion 110 of the drillhole 104 includes a storage area in a distal part of the portion 110 into which hazardous material may be retrievably placed for long-term storage. For example, a work string (e.g., tubing, coiled tubing, wireline, or otherwise) or other downhole conveyance (e.g., tractor) may be moved into the cased drillhole 104 to place one or more (three shown but there may be more or less) hazardous material canisters 126 into long term, but in some aspects, retrievable, storage in the portion 110. Each canister 126 may enclose hazardous material (shown as material 145). Such hazardous material, in some examples, may be biological or chemical waste or other biological or chemical hazardous material. In some examples, the hazardous material may include nuclear material, such as spent nuclear fuel recovered from a nuclear reactor (e.g., commercial power or test reactor) or military nuclear material. Spent nuclear fuel, in the form of nuclear fuel pellets, may be taken from the reactor and not modified. Nuclear fuel pellet are solid, although they can contain and emit a variety of radioactive gases including tritium (13 year half-life), krypton-85 (10.8 year half-life), and carbon dioxide containing C-14 (5730 year half-life). Other hazardous material 145 may include, for example, radioactive liquid, such as radioactive water from a commercial power (or other) reactor. In some aspects, the storage layer 118 should be able to contain any radioactive output (e.g., gases) within the layer 118, even if such output escapes the canisters 126. For example, the storage layer 118 may be selected based on diffusion times of radioactive output through the layer 118. For example, a minimum diffusion time of radioactive output escaping the storage layer 118 may be set at, for example, fifty times a half-life for any particular component of the nuclear fuel pellets. Fifty half-lives as a minimum diffusion time would reduce an amount of radioactive output by a factor of 1×10−15. As another example, setting a minimum diffusion time to thirty half-lives would reduce an amount of radioactive output by a factor of one billion. For example, plutonium-239 is often considered a dangerous waste product in spent nuclear fuel because of its long half-life of 24,100 years. For this isotope, 50 half-lives would be 1.2 million years. Plutonium-239 has low solubility in water, is not volatile, and as a solid. its diffusion time is exceedingly small (e.g., many millions of years) through a matrix of the rock formation that comprises the illustrated storage layer 118 (e.g., shale or other formation). The storage layer 118, for example comprised of shale, may offer the capability to have such isolation times (e.g., millions of years) as shown by the geological history of containing gaseous hydrocarbons (e.g., methane and otherwise) for several million years. In contrast, in conventional nuclear material storage methods, there was a danger that some plutonium might dissolve in a layer that comprised mobile ground water upon confinement escape. In some aspects, the drillhole 104 may be formed for the primary purpose of long-term storage of hazardous materials. In alternative aspects, the drillhole 104 may have been previously formed for the primary purpose of hydrocarbon production (e.g., oil, gas). For example, storage layer 118 may be a hydrocarbon bearing formation from which hydrocarbons were produced into the drillhole 104 and to the terranean surface 102. In some aspects, the storage layer 118 may have been hydraulically fractured prior to hydrocarbon production. Further in some aspects, the production casing 122 may have been perforated prior to hydraulic fracturing. In such aspects, the production casing 122 may be patched (e.g., cemented) to repair any holes made from the perforating process prior to a deposit operation of hazardous material. In addition, any cracks or openings in the cement between the casing and the drillhole can also be filled at that time. As further shown in FIG. 1, a backfill material 140 may be positioned or circulated into the drillhole 104. In this example, the backfill material 140 surrounds the canisters 126 and may have a level that extends uphole to at or near a drillhole seal 134 (e.g., permanent packer, plug, or other seal). In some aspects, the backfill material 140 may absorb radioactive energy (e.g., gamma rays or other energy). In some aspects, the backfill material 140 may have a relatively low thermal conductivity, thereby acting as an insulator between the canisters 126 and the casings. As further shown in FIG. 1, another backfill material 150 may be positioned or placed within one or more of the canisters 126 to surround the hazardous material 145. In some aspects, the backfill material 150 may absorb radioactive energy (e.g., gamma rays or other energy). In some aspects, the backfill material 150 may have a relatively low thermal conductivity, thereby acting as an insulator between the hazardous material 145 and the canister 126. In some aspects, the backfill material 150 may also provide a stiffening attribute to the canister 126, e.g., reducing crushability, deformation, or other damage to the canister 126. In some aspects, one or more of the previously described components of the system 100 may combine to form an engineered barrier of the hazardous waste material repository 100. For example, in some aspects, the engineered barrier is comprised of one, some, or all of the following components: the storage layer 118, the casing 130, the backfill material 140, the canister 126, the backfill material 150, the seal 134, and the hazardous material 145, itself. In some aspects, one or more of the engineered barrier components may act (or be engineered to act) to: prevent or reduce corrosion in the drillhole 104, prevent or reduce escape of the hazardous material 145; reduce or prevent thermal degradation of one or more of the other components; and other safety measures to ensure that the hazardous material 145 does not reach the mobile water layer 114 (or surface layer 112, including the terranean surface 102). FIG. 2 is a schematic illustration of an example implementation of a hazardous material storage repository 200 for radioactive liquid. In some aspects, one or more components of repository 200 may be similar to components described in reference to the hazardous material repository 100. For example, this figure illustrates the example hazardous material storage repository system 200 once (or as) a volume of radioactive liquid 226 that includes hazardous material (e.g., radioactive solid material) 232 is provided to a horizontal portion 210 of a drillhole 204. As illustrated, the hazardous material storage repository system 200 includes the drillhole 204 formed (e.g., drilled or otherwise) from a terranean surface 202 and through multiple subterranean layers 212, 214, 216, and 218. Although the terranean surface 202 is illustrated as a land surface, terranean surface 202 may be a sub-sea or other underwater surface, such as a lake or an ocean floor or other surface under a body of water. Thus, the present disclosure contemplates that the drillhole 204 may be formed under a body of water from a drilling location on or proximate the body of water. The illustrated drillhole 204 is a directional drillhole in this example of hazardous material storage repository system 200. For instance, the drillhole 204 includes a substantially vertical portion 206 coupled to a radiussed or curved portion 208, which in turn is coupled to a substantially horizontal portion 210. As illustrated in this example, the three portions of the drillhole 204—the vertical portion 206, the radiussed portion 208, and the horizontal portion 210—form a continuous drillhole 204 that extends into the Earth. As used in the present disclosure, the drillhole 204 (and drillhole portions described) may also be called wellbores. Thus, as used in the present disclosure, drillhole and wellbore are largely synonymous and refer to bores formed through one or more subterranean formations that are not suitable for human-occupancy (i.e., are too small in diameter for a human to fit therewithin). The illustrated drillhole 204, in this example, has a surface casing 220 positioned and set around the drillhole 204 from the terranean surface 202 into a particular depth in the Earth. The surface casing 220 extends from the terranean surface through a surface layer 212. The surface layer 212, in this example, is a geologic layer comprised of one or more layered rock formations. In some aspects, the surface layer 212 in this example may or may not include freshwater aquifers, salt water or brine sources, or other sources of mobile water (e.g., water that moves through a geologic formation). In some aspects, the surface casing 220 may isolate the drillhole 204 from such mobile water, and may also provide a hanging location for other casing strings to be installed in the drillhole 204. Further, although not shown, a conductor casing may be set above the surface casing 220 (e.g., between the surface casing 220 and the surface 202 and within the surface layer 212) to prevent drilling fluids from escaping into the surface layer 212. As illustrated, a production casing 222 is positioned and set within the drillhole 204 downhole of the surface casing 220. Although termed a “production” casing, in this example, the casing 222 may or may not have been subject to hydrocarbon production operations. Thus, the casing 222 refers to and includes any form of tubular member that is set (e.g., cemented) in the drillhole 204 downhole of the surface casing 220. In some examples of the hazardous material storage repository system 200, the production casing 222 may begin at an end of the radiussed portion 208 and extend throughout the inclined portion 210. The casing 222 could also extend into the radiussed portion 208 and into the vertical portion 206. As shown, cement 230 is positioned (e.g., pumped) around the casings 220 and 222 in an annulus between the casings 220 and 222 and the drillhole 204. The cement 230, for example, may secure the casings 220 and 222 (and any other casings or liners of the drillhole 204) through the subterranean layers under the terranean surface 202. In some aspects, the cement 230 may be installed along the entire length of the casings (e.g., casings 220 and 222 and any other casings), or the cement 230 could be used along certain portions of the casings if adequate for a particular drillhole 202. The cement 230 can also provide an additional layer of confinement for the radioactive liquid 226. The drillhole 204 and associated casings 220 and 222 may be formed with various example dimensions and at various example depths (e.g., true vertical depth, or TVD). As illustrated, the vertical portion 206 of the drillhole 204 extends through subterranean layers 212, 214, and 216, and, in this example, lands in a subterranean layer 218. As discussed above, the surface layer 212 may or may not include mobile water. In this example, a mobile water layer 214 is below the surface layer 212 (although surface layer 212 may also include one or more sources of mobile water or liquid). For instance, mobile water layer 214 may include one or more sources of mobile water, such as freshwater aquifers, salt water or brine, or other source of mobile water. In this example of hazardous material storage repository system 200, mobile water may be water that moves through a subterranean layer based on a pressure differential across all or a part of the subterranean layer. For example, the mobile water layer 214 may be a permeable geologic formation in which water freely moves (e.g., due to pressure differences or otherwise) within the layer 214. In some aspects, the mobile water layer 214 may be a primary source of human-consumable water in a particular geographic area. Examples of rock formations of which the mobile water layer 214 may be composed include porous sandstones and limestones, among other formations. Other illustrated layers, such as the impermeable layer 216 and the storage layer 218, may include immobile water. Immobile water, in some aspects, is water (e.g., fresh, salt, brine), that is not fit for human or animal consumption, or both. Immobile water, in some aspects, may be water that, by its motion through the layers 216 or 218 (or both), cannot reach the mobile water layer 214, terranean surface 202, or both, within 10,000 years or more (such as to 1,000,000 years). Below the mobile water layer 214, in this example implementation of hazardous material storage repository system 200, is an impermeable layer 216. The impermeable layer 216, in this example, may not allow mobile water to pass through. Thus, relative to the mobile water layer 214, the impermeable layer 216 may have low permeability, e.g., on the order of nanodarcy permeability. Additionally, in this example, the impermeable layer 216 may be a relatively non-ductile (i.e., brittle) geologic formation. One measure of non-ductility is brittleness, which is the ratio of compressive stress to tensile strength. In some examples, the brittleness of the impermeable layer 216 may be between about 20 MPa and 40 MPa. As shown in this example, the impermeable layer 216 is shallower (e.g., closer to the terranean surface 202) than the storage layer 218. In this example rock formations of which the impermeable layer 216 may be composed include, for example, certain kinds of sandstone, mudstone, clay, and slate that exhibit permeability and brittleness properties as described above. In alternative examples, the impermeable layer 216 may be deeper (e.g., further from the terranean surface 202) than the storage layer 218. In such alternative examples, the impermeable layer 216 may be composed of an igneous rock, such as granite. Below the impermeable layer 216 is the storage layer 218. The storage layer 218, in this example, may be chosen as the landing for the horizontal portion 210, which stores the hazardous material, for several reasons. Relative to the impermeable layer 216 or other layers, the storage layer 218 may be thick, e.g., between about 100 and 200 feet of total vertical thickness. Thickness of the storage layer 218 may allow for easier landing and directional drilling, thereby allowing the horizontal portion 210 to be readily emplaced within the storage layer 218 during constructions (e.g., drilling). If formed through an approximate horizontal center of the storage layer 218, the horizontal portion 210 may be surrounded by about 50 to 100 feet of the geologic formation that comprises the storage layer 218. Further, the storage layer 218 may also have only immobile water, e.g., due to a very low permeability of the layer 218 (e.g., on the order of milli- or nanodarcys). In addition, the storage layer 218 may have sufficient ductility, such that a brittleness of the rock formation that comprises the layer 218 is between about 3 MPa and 10 MPa. Examples of rock formations of which the storage layer 218 may be composed include: shale and anhydrite. Further, in some aspects, hazardous material may be stored below the storage layer, even in a permeable formation such as sandstone or limestone, if the storage layer is of sufficient geologic properties to isolate the permeable layer from the mobile water layer 214. In some aspects, the formation of the storage layer 218 and/or the impermeable layer 216 may form a leakage barrier, or barrier layer to fluid leakage that may be determined, at least in part, by the evidence of the storage capacity of the layer for hydrocarbons or other fluids (e.g., carbon dioxide) for hundreds of years, thousands of years, tens of thousands of years, hundreds of thousands of years, or even millions of years. For example, the barrier layer of the storage layer 218 and/or impermeable layer 216 may be defined by a time constant for leakage of the hazardous material more than 10,000 years (such as between about 10,000 years and 1,000,000 years) based on such evidence of hydrocarbon or other fluid storage. The present disclosure contemplates that there may be many other layers between or among the illustrated subterranean layers 212, 214, 216, and 218. For example, there may be repeating patterns (e.g., vertically), of one or more of the mobile water layer 214, impermeable layer 216, and storage layer 218. Further, in some instances, the storage layer 218 may be directly adjacent (e.g., vertically) the mobile water layer 214, i.e., without an intervening impermeable layer 216. In some examples, all or portions of the radiussed drillhole 208 and the horizontal drillhole 210 may be formed below the storage layer 218, such that the storage layer 218 (e.g., shale or other geologic formation with characteristics as described herein) is vertically positioned between the horizontal drillhole 210 and the mobile water layer 214. In this example, the horizontal portion 210 of the drillhole 204 includes a storage area in a distal part of the portion 210 into which hazardous material may be retrievably placed for long-term storage. For example, a work string or tubular 224 (e.g., tubing, coiled tubing, wireline, or otherwise) may be moved into the cased drillhole 204 to circulate (e.g., with a pump, not shown) the radioactive liquid 226 into long term (e.g., permanent) storage in the portion 210. In some aspects, the drillhole 204 may be formed for the primary purpose of long-term storage of hazardous materials. In alternative aspects, the drillhole 204 may have been previously formed for the primary purpose of hydrocarbon production (e.g., oil, gas). For example, storage layer 218 may be a hydrocarbon bearing formation from which hydrocarbons were produced into the drillhole 204 and to the terranean surface 202. In some aspects, the storage layer 218 may have been hydraulically fractured prior to hydrocarbon production. Further in some aspects, the production casing 222 may have been perforated prior to hydraulic fracturing. In such aspects, the production casing 222 may be patched (e.g., cemented) to repair any holes made from the perforating process prior to a deposit operation of hazardous material. In addition, any cracks or openings in the cement between the casing and the drillhole can also be filled at that time. As further shown in FIG. 2, a drillhole seal 234 (e.g., permanent packer, plug, or other seal) may be placed in the vertical portion 206 of the drillhole. In some aspects, the tubular 224 may extend through the drillhole seal 234 and, once the radioactive liquid 226 is emplaced, the tubular 224 may be withdrawn and the drillhole seal 234 closed (or another seal put in place) to fluidly isolate the drillhole 204 at the terranean surface 202 from the horizontal portion 210. In some aspects, one or more of the previously described components of the system 200 may combine to form an engineered barrier of the hazardous waste material repository 200. For example, in some aspects, the engineered barrier is comprised of one, some, or all of the following components: the storage layer 218, the casing 230, and the seal 234. In some aspects, one or more of the engineered barrier components may act (or be engineered to act) to: prevent or reduce corrosion in the drillhole 204, prevent or reduce escape of the hazardous material 232; reduce or prevent thermal degradation of one or more of the other components; and other safety measures to ensure that the hazardous material 232 (within the radioactive liquid 226) does not reach the mobile water layer 214 (or surface layer 212, including the terranean surface 202). FIG. 3 is a schematic illustration of another example implementation of a hazardous material storage repository 300 that includes an engineered barrier. In some aspects, one or more components of repository 300 may be similar to components described in reference to the hazardous material repository 100. For example, this figure illustrates an example hazardous material storage repository system 300 after deployment of one or more canisters 326 of hazardous material in a subterranean formation. As illustrated, the hazardous material storage repository system 300 includes a drillhole 304 formed (e.g., drilled or otherwise) from a terranean surface 302 and through multiple subterranean layers 312, 314, and 316. Although the terranean surface 302 is illustrated as a land surface, terranean surface 302 may be a sub-sea or other underwater surface, such as a lake or an ocean floor or other surface under a body of water. Thus, the present disclosure contemplates that the drillhole 304 may be formed under a body of water from a drilling location on or proximate the body of water. The illustrated drillhole 304 is a directional drillhole in this example of hazardous material storage repository system 300. For instance, the drillhole 304 includes a substantially vertical portion 306 coupled to a J-section portion 308, which in turn is coupled to a substantially horizontal portion 310. The J-section portion 308 as shown, has a shape that resembles the bottom portion of the letter “J” and may be shaped similar to a p-trap device used in a plumbing system that is used to prevent gasses from migrating from one side of the bend to the other side of the bend. As used in the present disclosure, “substantially” in the context of a drillhole orientation, refers to drillholes that may not be exactly vertical (e.g., exactly perpendicular to the terranean surface 302) or exactly horizontal (e.g., exactly parallel to the terranean surface 302), or exactly inclined at a particular incline angle relative to the terranean surface 302. In other words, vertical drillholes often undulate offset from a true vertical direction, that they might be drilled at an angle that deviates from true vertical, and horizontal drillholes often undulate offset from exactly horizontal. The J-section portion 308 is an example of an angled drillhole portion that, e.g., may prevent or help prevent migration of hazardous waste (or subterranean liquid in which leaked hazardous waste has been entrained) from the horizontal portion 310 to the vertical portion 306 of the drillhole 304. An “angled drillhole portion,” in this example, is a portion of the drillhole 304 that is angled toward the terranean surface 202 between the vertical portion 306 and the horizontal portion 310 (or the inclined portion 340). As illustrated in this example, the three portions of the drillhole 304—the vertical portion 306, the J-section portion 308, and the substantially horizontal portion 310—form a continuous drillhole 304 that extends into the Earth. As also shown in dashed line in FIG. 3, the J-section portion 308 may be coupled to an inclined portion 340 rather than (or in addition to) the substantially horizontal portion 310 of the drillhole 304. The illustrated drillhole 304, in this example, has a surface casing 320 positioned and set around the drillhole 304 from the terranean surface 302 into a particular depth in the Earth. For example, the surface casing 320 may be a relatively large-diameter tubular member (or string of members) set (e.g., cemented) around the drillhole 304 in a shallow formation. In this implementation of the hazardous material storage repository system 300, the surface casing 320 extends from the terranean surface through a surface layer 312. The surface layer 312, in this example, is a geologic layer comprised of one or more layered rock formations. In some aspects, the surface layer 312 in this example may or may not include freshwater aquifers, salt water or brine sources, or other sources of mobile water (e.g., water that moves through a geologic formation). In some aspects, the surface casing 320 may isolate the drillhole 304 from such mobile water, and may also provide a hanging location for other casing strings to be installed in the drillhole 304. Further, although not shown, a conductor casing may be set above the surface casing 320 (e.g., between the surface casing 320 and the surface 302 and within the surface layer 312) to prevent drilling fluids from escaping into the surface layer 312. As illustrated, a production casing 322 is positioned and set within the drillhole 304 downhole of the surface casing 320. The casing 322 refers to and includes any form of tubular member that is set (e.g., cemented) in the drillhole 304 downhole of the surface casing 320. In some examples of the hazardous material storage repository system 300, the production casing 322 may begin at an end of the J-section portion 308 and extend throughout the substantially horizontal portion 310. The casing 322 could also extend into the J-section portion 308 and into the vertical portion 306. As shown, cement 330 is positioned (e.g., pumped) around the casings 320 and 322 in an annulus between the casings 320 and 322 and the drillhole 304. The cement 330, for example, may secure the casings 320 and 322 (and any other casings or liners of the drillhole 304) through the subterranean layers under the terranean surface 302. In some aspects, the cement 330 may be installed along the entire length of the casings (e.g., casings 320 and 322 and any other casings), or the cement 330 could be used along certain portions of the casings if adequate for a particular drillhole 302. The cement 330 can also provide an additional layer of confinement for the hazardous material in canisters 326. The drillhole 304 and associated casings 320 and 322 may be formed with various example dimensions and at various example depths (e.g., true vertical depth, or TVD). For instance, a conductor casing (not shown) may extend down to about 120 feet TVD, with a diameter of between about 28 in. and 60 in. The surface casing 320 may extend down to about 2500 feet TVD, with a diameter of between about 22 in. and 48 in. An intermediate casing (not shown) between the surface casing 320 and production casing 322 may extend down to about 8000 feet TVD, with a diameter of between about 16 in. and 36 in. The production casing 322 may extend inclinedly (e.g., to case the substantially horizontal portion 310 and/or the inclined portion 340) with a diameter of between about 11 in. and 22 in. The foregoing dimensions are merely provided as examples and other dimensions (e.g., diameters, TVDs, lengths) are contemplated by the present disclosure. As illustrated, the vertical portion 306 of the drillhole 304 extends through subterranean layers 312, 314, and 316, and, in this example, lands in a subterranean layer 319. As discussed above, the surface layer 312 may or may not include mobile water. Subterranean layer 314, which is below the surface layer 312, in this example, is a mobile water layer 314. For instance, mobile water layer 314 may include one or more sources of mobile water, such as freshwater aquifers. salt water or brine, or other source of mobile water. In this example of hazardous material storage repository system 300, mobile water may be water that moves through a subterranean layer based on a pressure differential across all or a part of the subterranean layer. Other illustrated layers, such as the impermeable layer 316 and the storage layer 319, may include immobile water. Immobile water, in some aspects, is water (e.g., fresh, salt, brine), that is not fit for human or animal consumption, or both. Immobile water, in some aspects, may be water that, by its motion through the layers 316 or 319 (or both), cannot reach the mobile water layer 314, terranean surface 302, or both, within 10,000 years or more (such as to 1,000,000 years). Below the mobile water layer 314, in this example implementation of hazardous material storage repository system 300, is an impermeable layer 316. The impermeable layer 316, in this example, may not allow mobile water to pass through. Thus, relative to the mobile water layer 314, the impermeable layer 316 may have low permeability, e.g., on the order of 0.01 millidarcy permeability. Additionally, in this example, the impermeable layer 316 may be a relatively non-ductile (i.e., brittle) geologic formation. One measure of non-ductility is brittleness, which is the ratio of compressive stress to tensile strength. In some examples, the brittleness of the impermeable layer 316 may be between about 20 MPa and 40 MPa. As shown in this example, the impermeable layer 316 is shallower (e.g., closer to the terranean surface 302) than the storage layer 319. In this example rock formations of which the impermeable layer 316 may be composed include, for example, certain kinds of sandstone, mudstone, clay, and slate that exhibit permeability and brittleness properties as described above. In alternative examples, the impermeable layer 316 may be deeper (e.g., further from the terranean surface 302) than the storage layer 319. In such alternative examples, the impermeable layer 316 may be composed of an igneous rock, such as granite. Below the impermeable layer 316 is the storage layer 319. The storage layer 319, in this example, may be chosen as the landing for the substantially horizontal portion 310, which stores the hazardous material, for several reasons. Relative to the impermeable layer 316 or other layers, the storage layer 319 may be thick, e.g., between about 100 and 200 feet of total vertical thickness. Thickness of the storage layer 319 may allow for easier landing and directional drilling, thereby allowing the substantially horizontal portion 310 to be readily emplaced within the storage layer 319 during constructions (e.g., drilling). If formed through an approximate horizontal center of the storage layer 319, the substantially horizontal portion 310 may be surrounded by about 50 to 100 feet of the geologic formation that comprises the storage layer 319. Further, the storage layer 319 may also have only immobile water, e.g., due to a very low permeability of the layer 319 (e.g., on the order of milli- or nanodarcys). In addition, the storage layer 319 may have sufficient ductility, such that a brittleness of the rock formation that comprises the layer 319 is between about 3 MPa and 10 MPa. Examples of rock formations of which the storage layer 319 may be composed include: shale and anhydrite. Further, in some aspects, hazardous material may be stored below the storage layer, even in a permeable formation such as sandstone or limestone, if the storage layer is of sufficient geologic properties to isolate the permeable layer from the mobile water layer 314. In some examples implementations of the hazardous material storage repository system 300, the storage layer 319 (and/or the impermeable layer 316) is composed of shale. In some aspects, the formation of the storage layer 319 and/or the impermeable layer 316 may form a leakage barrier, or barrier layer to fluid leakage that may be determined, at least in part, by the evidence of the storage capacity of the layer for hydrocarbons or other fluids (e.g., carbon dioxide) for hundreds of years, thousands of years, tens of thousands of years, hundreds of thousands of years, or even millions of years. For example, the barrier layer of the storage layer 319 and/or impermeable layer 316 may be defined by a time constant for leakage of the hazardous material of more than 10,000 years (such as between 10,000 years and 1,000,000 years) based on such evidence of hydrocarbon or other fluid storage. The present disclosure contemplates that there may be many other layers between or among the illustrated subterranean layers 312, 314, 316, and 319. For example, there may be repeating patterns (e.g., vertically), of one or more of the mobile water layer 314, impermeable layer 316, and storage layer 319. Further, in some instances, the storage layer 319 may be directly adjacent (e.g., vertically) the mobile water layer 314, i.e., without an intervening impermeable layer 316. In some examples, all or portions of the J-section drillhole 308 and the substantially horizontal portion 310 (and/or the inclined portion 340) may be formed below the storage layer 319, such that the storage layer 319 (e.g., shale or other geologic formation with characteristics as described herein) is vertically positioned between the substantially horizontal portion 310 (and/or the inclined portion 340) and the mobile water layer 314. As shown in this example, the substantially horizontal portion 310 of the drillhole 304 includes a storage area 317 in a distal part of the portion 310 into which hazardous material may be retrievably placed for long-term storage. For example, a work string (e.g., tubing, coiled tubing, wireline, or otherwise) or downhole tractor may be moved into the cased drillhole 304 to place one or more hazardous material canisters 326 into long term, but in some aspects, retrievable, storage in the portion 310. Each canister 326 may enclose hazardous material, such as radioactive material. Examples of radioactive material include spent nuclear fuel and high level waste, e.g., in solid form, as well as radioactive liquid, such as radioactive water. In some aspects, the storage layer 319 should be able to contain any radioactive output (e.g., gases) within the layer 319, even if such output escapes the canisters 326. For example, the storage layer 319 may be selected based on diffusion times of radioactive output through the layer 319. For example, a minimum diffusion time of radioactive output escaping the storage layer 319 may be set at, for example, fifty times a half-life for any particular component of the nuclear fuel pellets. Fifty half-lives as a minimum diffusion time would reduce an amount of radioactive output by a factor of 1×10−15. As another example, setting a minimum diffusion time to thirty half-lives would reduce an amount of radioactive output by a factor of one billion. As further shown in FIG. 3, the storage canisters 326 may be positioned for long term storage in the substantially horizontal portion 310, which, as shown, is coupled to the vertical portion 106 of the drillhole 104 through the J-section portion 308. As illustrated, the J-section portion 308 includes an upwardly directed portion angled toward the terranean surface 302. In some aspects, for example when there is radioactive hazardous material stored in the canisters 326, this inclination of the J-section portion 308 (and inclination of the inclined portion 340, if formed) may provide a further degree of safety and containment to prevent or impede the material, even if leaked from the canister 326, from reaching, e.g., the mobile water layer 314, the vertical portion 306 of the drillhole 304, the terranean surface 302, or a combination thereof. For example, radionuclides of concern in the hazardous material tend to be relatively buoyant or heavy (as compared to other components of the material). Buoyant radionuclides may be the greatest concern for leakage, since heavy elements and molecules tend to sink, and would not diffuse upward towards the terranean surface 302. Krypton gas, and particularly krypton-85, is a buoyant radioactive element that is heavier than air (as are most gases) but much lighter than water. Thus, should krypton-85 be introduced into a water bath, such gas would tend to float upward towards the terranean surface 302. Iodine, on the other hand, is denser than water, and would tend to diffuse downward if introduced into a water bath. By including the J-section portion 308 of the drillhole 304, any such diffusion of radioactive material (e.g., even if leaked from a canister 326 and in the presence of water or other liquid in the drillhole 304 or otherwise) would be directed angularly upward toward the substantially horizontal portion 310, and more specifically, toward a distal end 321 of the substantially horizontal portion 310 and away from the J-section portion 308 (and the vertical portion 306) of the drillhole 304. Thus, leaked hazardous material, even in a diffusible gas form, would not be offered a path (e.g., directly) to the terranean surface 302 (or the mobile water layer 314) through the vertical portion 306 of the drillhole 310. For instance, the leaked hazardous material (especially in gaseous form) would be directed and gathered at the distal end 321 of the drillhole portion 310, or, generally, within the substantially horizontal portion 310 of the drillhole 304. In some aspects, the drillhole 304 may be formed for the primary purpose of long-term storage of hazardous materials. In alternative aspects, the drillhole 304 may have been previously formed for the primary purpose of hydrocarbon production (e.g., oil, gas). For example, storage layer 319 may be a hydrocarbon bearing formation from which hydrocarbons were produced into the drillhole 304 and to the terranean surface 302. In some aspects, the storage layer 319 may have been hydraulically fractured prior to hydrocarbon production. Further in some aspects, the production casing 322 may have been perforated prior to hydraulic fracturing. In such aspects, the production casing 322 may be patched (e.g., cemented) to repair any holes made from the perforating process prior to a deposit operation of hazardous material. In addition, any cracks or openings in the cement between the casing and the drillhole can also be filled at that time. As further shown in FIG. 3, a backfill material 340 may be positioned or circulated into the drillhole 304. In this example, the backfill material 340 surrounds the canisters 326 and may have a level that extends uphole to at or near a drillhole seal 334 (e.g., permanent packer, plug, or other seal). In some aspects, the backfill material 340 may absorb radioactive energy (e.g., gamma rays or other energy). In some aspects, the backfill material 340 may have a relatively low thermal conductivity, thereby acting as an insulator between the canisters 326 and the casings. Another backfill material (such as material 150 shown in FIG. 1) may be positioned or placed within one or more of the canisters 326 to surround the hazardous material within the canisters. In some aspects. such a backfill material may absorb radioactive energy (e.g., gamma rays or other energy). In some aspects, such a backfill material may have a relatively low thermal conductivity, thereby acting as an insulator between the hazardous material and the canister 326. In some aspects, such a backfill material may also provide a stiffening attribute to the canister 326, e.g., reducing crushability, deformation, or other damage to the canister 326. In some aspects, one or more of the previously described components of the system 100 may combine to form an engineered barrier of the hazardous waste material repository 300. For example, in some aspects, the engineered barrier is comprised of one, some, or all of the following components: the storage layer 319, the casing 330, the backfill material 340, the canister 326, the backfill material in canister 326, the seal 334, and the hazardous material within the canister 326, itself. In some aspects, one or more of the engineered barrier components may act (or be engineered to act) to: prevent or reduce corrosion in the drillhole 304, prevent or reduce escape of the hazardous material; reduce or prevent thermal degradation of one or more of the other components; and other safety measures to ensure that the hazardous material does not reach the mobile water layer 314 (or surface layer 312, including the terranean surface 302). FIG. 4 is a schematic illustration of another example implementation of a hazardous material storage repository 400 for radioactive liquid. In some aspects, one or more components of repository 300 may be similar to components described in reference to the hazardous material repositories 200 and 300. For example, this figure illustrates the example hazardous material storage repository system 400 once (or as) a volume of radioactive liquid 426 that includes hazardous material (e.g., radioactive solid material, such as material 232 in FIG. 2) is provided to a horizontal portion 410 of a drillhole 404. As illustrated, the hazardous material storage repository system 400 includes the drillhole 404 formed (e.g., drilled or otherwise) from a terranean surface 402 and through multiple subterranean layers 412, 414, 416, and 419. Although the terranean surface 402 is illustrated as a land surface, terranean surface 402 may be a sub-sea or other underwater surface, such as a lake or an ocean floor or other surface under a body of water. Thus, the present disclosure contemplates that the drillhole 404 may be formed under a body of water from a drilling location on or proximate the body of water. The illustrated drillhole 404 is a directional drillhole in this example of hazardous material storage repository system 400. For instance, the drillhole 404 includes a substantially vertical portion 406 coupled to a J-section portion 408, which in turn is coupled to a substantially horizontal portion 410. The J-section portion 408 as shown, has a shape that resembles the bottom portion of the letter “J” and may be shaped similar to a p-trap device used in a plumbing system that is used to prevent gasses from migrating from one side of the bend to the other side of the bend. The J-section portion 408 is an example of an angled drillhole portion that, e.g., may prevent or help prevent migration of hazardous waste (or subterranean liquid in which leaked hazardous waste has been entrained) from the horizontal portion 410 to the vertical portion 406 of the drillhole 404. An “angled drillhole portion,” in this example, is a portion of the drillhole 404 that is angled toward the terranean surface 402 between the vertical portion 406 and the horizontal portion 410 (or the inclined portion 440). As illustrated in this example, the three portions of the drillhole 404—the vertical portion 406, the J-section portion 408, and the substantially horizontal portion 410—form a continuous drillhole 404 that extends into the Earth. As also shown in dashed line in FIG. 4, the J-section portion 408 may be coupled to an inclined portion 440 rather than (or in addition to) the substantially horizontal portion 410 of the drillhole 404. The illustrated drillhole 404, in this example, has a surface casing 420 positioned and set around the drillhole 404 from the terranean surface 402 into a particular depth in the Earth. For example, the surface casing 420 may be a relatively large-diameter tubular member (or string of members) set (e.g., cemented) around the drillhole 404 in a shallow formation. In this implementation of the hazardous material storage repository system 400, the surface casing 420 extends from the terranean surface through a surface layer 412. The surface layer 412, in this example, is a geologic layer comprised of one or more layered rock formations. In some aspects, the surface layer 412 in this example may or may not include freshwater aquifers, salt water or brine sources, or other sources of mobile water (e.g., water that moves through a geologic formation). In some aspects, the surface casing 420 may isolate the drillhole 404 from such mobile water, and may also provide a hanging location for other casing strings to be installed in the drillhole 404. Further, although not shown, a conductor casing may be set above the surface casing 420 (e.g., between the surface casing 420 and the surface 402 and within the surface layer 412) to prevent drilling fluids from escaping into the surface layer 412. As illustrated, a production casing 422 is positioned and set within the drillhole 404 downhole of the surface casing 420. The casing 422 refers to and includes any form of tubular member that is set (e.g., cemented) in the drillhole 404 downhole of the surface casing 420. In some examples of the hazardous material storage repository system 400, the production casing 422 may begin at an end of the J-section portion 408 and extend throughout the substantially horizontal portion 410. The casing 422 could also extend into the J-section portion 408 and into the vertical portion 406. As shown, cement 430 is positioned (e.g., pumped) around the casings 420 and 422 in an annulus between the casings 420 and 422 and the drillhole 404. The cement 430, for example, may secure the casings 420 and 422 (and any other casings or liners of the drillhole 404) through the subterranean layers under the terranean surface 402. In some aspects, the cement 430 may be installed along the entire length of the casings (e.g., casings 420 and 422 and any other casings), or the cement 430 could be used along certain portions of the casings if adequate for a particular drillhole 402. The cement 430 can also provide an additional layer of confinement for the radioactive liquid 426. As illustrated, the vertical portion 406 of the drillhole 404 extends through subterranean layers 412, 414, and 416, and, in this example, lands in a subterranean layer 419. As discussed above, the surface layer 412 may or may not include mobile water. Subterranean layer 414, which is below the surface layer 412, in this example, is a mobile water layer 414. For instance, mobile water layer 414 may include one or more sources of mobile water, such as freshwater aquifers, salt water or brine, or other source of mobile water. In this example of hazardous material storage repository system 400, mobile water may be water that moves through a subterranean layer based on a pressure differential across all or a part of the subterranean layer. Other illustrated layers, such as the impermeable layer 416 and the storage layer 419, may include immobile water. Immobile water, in some aspects, is water (e.g., fresh, salt, brine), that is not fit for human or animal consumption, or both. Immobile water, in some aspects, may be water that, by its motion through the layers 416 or 419 (or both), cannot reach the mobile water layer 414, terranean surface 402, or both, within 10,000 years or more (such as to 1,000,000 years). Below the mobile water layer 414, in this example implementation of hazardous material storage repository system 400, is an impermeable layer 416. The impermeable layer 416, in this example, may not allow mobile water to pass through. Thus, relative to the mobile water layer 414, the impermeable layer 416 may have low permeability, e.g., on the order of 0.01 millidarcy permeability. Additionally, in this example, the impermeable layer 416 may be a relatively non-ductile (i.e., brittle) geologic formation. One measure of non-ductility is brittleness, which is the ratio of compressive stress to tensile strength. In some examples, the brittleness of the impermeable layer 416 may be between about 20 MPa and 40 MPa. As shown in this example, the impermeable layer 416 is shallower (e.g., closer to the terranean surface 402) than the storage layer 419. In this example rock formations of which the impermeable layer 416 may be composed include, for example, certain kinds of sandstone, mudstone, clay, and slate that exhibit permeability and brittleness properties as described above. In alternative examples, the impermeable layer 416 may be deeper (e.g., further from the terranean surface 402) than the storage layer 419. In such alternative examples, the impermeable layer 416 may be composed of an igneous rock, such as granite. Below the impermeable layer 416 is the storage layer 419. The storage layer 419, in this example, may be chosen as the landing for the substantially horizontal portion 410, which stores the hazardous material, for several reasons. Relative to the impermeable layer 416 or other layers, the storage layer 419 may be thick, e.g., between about 100 and 200 feet of total vertical thickness. Thickness of the storage layer 419 may allow for easier landing and directional drilling, thereby allowing the substantially horizontal portion 410 to be readily emplaced within the storage layer 419 during constructions (e.g., drilling). If formed through an approximate horizontal center of the storage layer 419, the substantially horizontal portion 410 may be surrounded by about 50 to 100 feet of the geologic formation that comprises the storage layer 419. Further, the storage layer 419 may also have only immobile water, e.g., due to a very low permeability of the layer 419 (e.g., on the order of milli- or nanodarcys). In addition, the storage layer 419 may have sufficient ductility, such that a brittleness of the rock formation that comprises the layer 419 is between about 3 MPa and 10 MPa. Examples of rock formations of which the storage layer 419 may be composed include: shale and anhydrite. Further, in some aspects, hazardous material may be stored below the storage layer, even in a permeable formation such as sandstone or limestone, if the storage layer is of sufficient geologic properties to isolate the permeable layer from the mobile water layer 414. In some examples implementations of the hazardous material storage repository system 400, the storage layer 419 (and/or the impermeable layer 416) is composed of shale. In some aspects, the formation of the storage layer 419 and/or the impermeable layer 416 may form a leakage barrier, or barrier layer to fluid leakage that may be determined, at least in part, by the evidence of the storage capacity of the layer for hydrocarbons or other fluids (e.g., carbon dioxide) for hundreds of years, thousands of years, tens of thousands of years, hundreds of thousands of years, or even millions of years. For example, the barrier layer of the storage layer 419 and/or impermeable layer 416 may be defined by a time constant for leakage of the hazardous material of more than 10,000 years (such as between 10,000 years and 1,000,000 years) based on such evidence of hydrocarbon or other fluid storage. The present disclosure contemplates that there may be many other layers between or among the illustrated subterranean layers 412, 414, 416, and 419. For example, there may be repeating patterns (e.g., vertically), of one or more of the mobile water layer 414, impermeable layer 416, and storage layer 419. Further, in some instances, the storage layer 419 may be directly adjacent (e.g., vertically) the mobile water layer 414, i.e., without an intervening impermeable layer 416. In some examples, all or portions of the J-section drillhole 408 and the substantially horizontal portion 410 (and/or the inclined portion 440) may be formed below the storage layer 419, such that the storage layer 419 (e.g., shale or other geologic formation with characteristics as described herein) is vertically positioned between the substantially horizontal portion 410 (and/or the inclined portion 440) and the mobile water layer 414. As shown in this example, the substantially horizontal portion 410 of the drillhole 404 includes a storage area 417 in a distal part 421 of the portion 410 into which hazardous material may be retrievably placed for long-term storage. For example, a work string or tubular 424 (e.g., tubing, coiled tubing, or otherwise) may be moved into the cased drillhole 404 circulate (e.g., with a pump, not shown) the radioactive liquid 426 into long term (e.g., permanent) storage in the portion 410. In some aspects, the storage layer 419 should be able to contain any radioactive output (e.g., gases) within the layer 419. For example, the storage layer 419 may be selected based on diffusion times of radioactive output through the layer 419. For example, a minimum diffusion time of radioactive output escaping the storage layer 419 may be set at, for example, fifty times a half-life for any particular component of the nuclear fuel pellets. Fifty half-lives as a minimum diffusion time would reduce an amount of radioactive output by a factor of 1×10−15. As another example, setting a minimum diffusion time to thirty half-lives would reduce an amount of radioactive output by a factor of one billion. In some aspects, the drillhole 404 may be formed for the primary purpose of long-term storage of hazardous materials. In alternative aspects, the drillhole 404 may have been previously formed for the primary purpose of hydrocarbon production (e.g., oil, gas). For example, storage layer 419 may be a hydrocarbon bearing formation from which hydrocarbons were produced into the drillhole 404 and to the terranean surface 402. In some aspects, the storage layer 419 may have been hydraulically fractured prior to hydrocarbon production. Further in some aspects, the production casing 422 may have been perforated prior to hydraulic fracturing. In such aspects, the production casing 422 may be patched (e.g., cemented) to repair any holes made from the perforating process prior to a deposit operation of hazardous material. In addition, any cracks or openings in the cement between the casing and the drillhole can also be filled at that time. As further shown in FIG. 4, a drillhole seal 434 (e.g., permanent packer, plug, or other seal) may be placed in the vertical portion 406 of the drillhole. In some aspects, the tubular 424 may extend through the drillhole seal 434 and, once the radioactive liquid 426 is emplaced, the tubular 424 may be withdrawn and the drillhole seal 434 closed (or another seal put in place) to fluidly isolate the drillhole 404 at the terranean surface 402 from the horizontal portion 410. In some aspects, one or more of the previously described components of the system 400 may combine to form an engineered barrier of the hazardous waste material repository 400. For example, in some aspects, the engineered barrier is comprised of one, some, or all of the following components: the storage layer 419, the casing 422, the cement 430, and the seal 234. In some aspects, one or more of the engineered barrier components may act (or be engineered to act) to: prevent or reduce corrosion in the drillhole 404, prevent or reduce escape of the hazardous material within the radioactive liquid 426; reduce or prevent thermal degradation of one or more of the other components; and other safety measures to ensure that the hazardous material (within the radioactive liquid 426) does not reach the mobile water layer 414 (or surface layer 412, including the terranean surface 402). In some aspects, one or more of the previously described components of the system 100 may combine to form an engineered barrier of the hazardous waste material repository 400. For example, in some aspects, the engineered barrier is comprised of one, some, or all of the following components: the storage layer 419, the casing 430, the backfill material 440, the canister 426, the backfill material in canister 426, the seal 434, and the hazardous material within the canister 426, itself. In some aspects, one or more of the engineered barrier components may act (or be engineered to act) to: prevent or reduce corrosion in the drillhole 404, prevent or reduce escape of the hazardous material; reduce or prevent thermal degradation of one or more of the other components; and other safety measures to ensure that the hazardous material does not reach the mobile water layer 414 (or surface layer 412, including the terranean surface 402). Turning briefly to FIG. 5A, this figures illustrates a radial cross-section of an example engineered barrier (or engineered barrier system) 500 for a radioactive (e.g., hazardous) material repository, such as, for example hazardous material repository 100. In this example, the engineered barrier system 500 includes a host rock formation 502. The host rock formation 502 (e.g., storage layer 118) may be selected due to, e.g., permeability, ductility, or other geological criteria that allows the formation 502 to fluidly isolate gasses or liquids from other formations. The engineered barrier system 500 also includes a cement 504 and casing 506. The cement 504 binds or helps bind the casing 506 (e.g., a steel casing or otherwise) to the rock formation 502. The example engineered barrier system 500 also includes a backfill material 508 (e.g., backfill material 140) that fills an annular space between the casing 506 and a canister housing 510. The canister housing 510 defines a volume in which hazardous material is enclosed and may be formed, for example, from a corrosion resistant material, such as a metal alloy. A canister backfill 512 is also part of this example engineered barrier system 500. The canister backfill 512 (like the backfill material 508) may, e.g., absorb or partly absorb radioactive energy (e.g., gamma rays) and be a non-conductive thermal material. Lastly, this example engineered barrier system 500 includes the radioactive material 514, which may be processed or formed to reduce or eliminate a possibility of escape or leakage from the canister housing 510. Turning briefly to FIG. 5B, this figure illustrates another view of a portion 550 of engineered barrier system 500. As shown in this example, canister housings 510 are spaced apart in a drillhole formed in the rock formation 502. The casing 506 separates the drillhole from the formation 502. As shown in this example, the canister housings 510 that enclose the radioactive material 514, such that heat from each canister housing 510 (due to thermal output from the radioactive material 514) is spread out along the drillhole. Referring generally to the present disclosure, the example hazardous material storage repository systems (e.g., 100, 200, 300, and 400) may provide for multiple layers of containment as an engineered barrier system to ensure that a hazardous material (e.g., biological, chemical, nuclear) is sealingly stored in an appropriate subterranean layer. In some example implementations, there may be at least twelve layers of containment. In alternative implementations, a fewer or a greater number of containment layers may be employed. First. using spent nuclear fuel as an example hazardous material, the fuel pellets are taken from the reactor and not modified. They may be made from sintered uranium dioxide (UO2), a ceramic, and may remain solid and emit a variety of radioactive gases including tritium (13 year half-life), krypton-85 (10.8 year half-life), and carbon dioxide containing C-14 (5730 year half-life). Unless the pellets are exposed to extremely corrosive conditions or other effects that damage the multiple layers of containment, most of the radioisotopes (including the C-14, tritium or krypton-85) will be contained in the pellets. Second, the fuel pellets are surrounded by the zircaloy tubes of the fuel rods, just as in the reactor. As described, the tubes could be mounted in the original fuel assemblies, or removed from those assemblies for tighter packing. Third, the tubes are placed in the sealed housings of the hazardous material canister. The housing may be a unified structure or multi-panel structure, with the multiple panels (e.g., sides, top, bottom) mechanically fastened (e.g., screws, rivets, welds, and otherwise). Fourth, a material (e.g., solid or fluid) may fill the hazardous material canister to provide a further buffer between the material and the exterior of the canister. Fifth, the hazardous material canister(s) are positioned, in a drillhole that is lined with a steel or other sealing casing that extends, in some examples, throughout the entire drillhole (e.g., a substantially vertical portion, a radiussed portion, and a inclined portion). The casing is cemented in place, providing a relatively smooth surface (e.g., as compared to the drillhole wall) for the hazardous material canister to be moved through, thereby reducing the possibility of a leak or break during deposit or retrieval. Sixth, the cement that holds or helps hold the casing in place, may also provide a sealing layer to contain the hazardous material should it escape the canister. Seventh, the hazardous material canister is stored in a portion of the drillhole (e.g., the inclined portion) that is positioned within a thick (e.g., 100-200 feet) seam of a rock formation that comprises a storage layer. The storage layer may be chosen due at least in part to the geologic properties of the rock formation (e.g., only immobile water, low permeability, thick, appropriate ductility or non-brittleness). For example, in the case of shale as the rock formation of the storage layer, this type of rock may offers a level of containment since it is known that shale has been a seal for hydrocarbon gas for millions of years. The shale may contain brine, but that brine is demonstrably immobile, and not in communication with surface fresh water. Eighth, in some aspects, the rock formation of the storage layer may have other unique geological properties that offer another level of containment. For example, shale rock often contains reactive components, such as iron sulfide, that reduce the likelihood that hazardous materials (e.g., spent nuclear fuel and its radioactive output) can migrate through the storage layer without reacting in ways that reduce the diffusion rate of such output even further. Further, the storage layer may include components, such as clay and organic matter, that typically have extremely low diffusivity. For example, shale may be stratified and composed of thinly alternating layers of clays and other minerals. Such a stratification of a rock formation in the storage layer, such as shale, may offer this additional layer of containment. Ninth, the storage layer may be located deeper than, and under, an impermeable layer, which separates the storage layer (e.g., vertically) from a mobile water layer. Tenth, the storage layer may be selected based on a depth (e.g., 3000 to 12,000 ft.) of such a layer within the subterranean layers. Such depths are typically far below any layers that contain mobile water, and thus, the sheer depth of the storage layer provides an additional layer of containment. Eleventh, example implementations of the hazardous material storage repository system of the present disclosure facilitate monitoring of the stored hazardous material. For example, if monitored data indicates a leak or otherwise of the hazardous material (e.g., change in temperature, radioactivity, or otherwise), or even tampering or intrusion of the canister, the hazardous material canister may be retrieved for repair or inspection. Twelfth, the one or more hazardous material canisters may be retrievable for periodic inspection, conditioning, or repair, as necessary (e.g., with or without monitoring). Thus, any problem with the canisters may be addressed without allowing hazardous material to leak or escape from the canisters unabated. Thirteenth, even if hazardous material escaped from the canisters and no impermeable layer was located between the leaked hazardous material and the terranean surface, the leaked hazardous material may be contained within the drillhole at a location that has no upward path to the surface or to aquifers (e.g., mobile water layers) or to other zones that would be considered hazardous to humans. For example, the location, which may be a dead end of an inclined drillhole, a J-section drillhole, or peaks of a vertically undulating drillhole, may have no direct upward (e.g., toward the surface) path to a vertical portion of the drillhole. Analysis of Engineered Barriers for Radioactive Material Repositories. The present disclosure describes an analysis of engineered barriers (or, engineered barrier systems (EBS)) as part of one or more of the repositories 100, 200, 300 and 400. The present disclosure also describes EBS corrosion performance. EBS corrosion performance is analyzed for disposal of, e.g., spent nuclear fuel and other high-level nuclear waste in deep horizontal drillholes (such as drillhole 104), and aspects of the EBS design are related to corrosion performance. Time periods set duration and environmental conditions to follow the evolution of the environment over 10,000 years once the hazardous material (in the canister) has been emplaced. The first 20 years covers the heat-up to maximum temperature, start of cool-down and transition from moderately oxidizing the anaerobic conditions. An EBS may be designed to make it through the aggressive, initial period and enter the anaerobic period in a condition to survive for tens of thousands of years. In some aspects, the EBS is defined with a corrosion resistant Ni—Cr—Mo alloy as the canisters and carbon steel casing. Time-temperature behavior is from thermal simulation. Results include metal loss and time-to-perforation for canisters and casing along with amounts of hydrogen generated and metal oxide formed for each time period. Time for the first perforation of a canister with 9.5-mm wall is 45,000 years. The steel casing is a barrier between environment inside and outside of the casing for nearly 3,000 years. Volume of hydrogen and metal oxide formed track corrosion rate of the metals. Design considerations relate to favorable aspects of the environment: reducing environment, fully saturated rock, no boiling and no wet-dry-wet cycle. The material, as described, is disposed in horizontal drillholes, e.g., in sedimentary rock that overlies basement rock, although metamorphic rock can also be used. The EBS in this example represents the engineered materials placed within the repository, including the hazardous material form, hazardous material canisters, buffer materials, backfill, and seals. The hazardous material is sealed in canisters that are emplaced along a steel casing in the horizontal drillhole. Canisters are a non-permeable, absolute barrier to radionuclide transport until they are breached. A steel casing aids in canister emplacement, provides structural support, and is a barrier to radionuclide transport until perforated. Corrosion is a risk for perforation of the canister and controls the degradation of casing. Corrosion behaviors of canisters and casing are contributors to the performance of the engineered barrier system and the control of radionuclide transport. For safe and reliable performance, canisters may need to be made of highly corrosion resistant metal for long-term containment of the waste. For the example here, canisters are made of nickel-chromium-molybdenum (Ni—Cr—Mo) alloys, a family of alloys that have excellent corrosion resistance over a wide range of environments. In a reducing environment, the time-to-perforation for a Ni—Cr—Mo canister with a wall thickness 1-cm is 50,000 years. Casing may be made of carbon steel to aid in canister emplacement, provide structural strength, and separate the inside casing environment from the outside casing environment until the casing is perforated. The objectives of this analysis are to analyze the corrosion performance of the EBS for a base case configuration and relate aspects of the EBS design to corrosion performance. The long-term corrosion behavior of canisters and casing from emplacement through 10,000 years was divided into five time periods to track evolution of the drillhole environment. Corrosion rates of metals were set for the drillhole environment of each period. Calculations were made for corrosion reactions of metal and water to form metal oxides and hydrogen. Results include metal loss and time-to-perforation for canisters and casing along with amounts of hydrogen generated and metal oxide formed for each time period and overall. Nuclear waste disposal in deep horizontal drillholes is suitable for a variety of waste types, where canister size is designed to accommodate the specific fuel type. For disposal of cesium/strontium capsules (a form of legacy waste from the United States' nuclear defense programs), the canisters are on the order of 12-cm diameter and 60-cm long. For spent nuclear fuel assemblies, larger diameter and longer canisters each may hold one fuel assembly. The waste is placed within a metal cylinder with an end plate attached, and the other endplate is positioned and sealed. The sealed canister is lowered through the vertical drillhole section and emplaced in the casing along the horizontal disposal section. Corrosion behavior is determined by the combination of corrosion resistance of the alloy and corrosivity of the environment. The mode of corrosion is general corrosion. The environment in fully saturated rock evolves from a transition stage with high temperature, moderately oxidizing waters to ambient temperature, highly reducing waters. The waters are brines and can have multiple dissolved species with a range of concentrations. Chloride brines are typical. Ambient temperature at the horizontal disposal section is higher than at the surface and depends upon the depth. Temperature rises from the residual decay heat emanating from the waste forms and then decreases as the waste decays. Heating to maximum temperature and start of cool down occurs within 5-10 years after canister emplacement. The maximum temperature of the canister wall is 170° C. A 40-year thermal heat-up and cool-down period is followed by a slow cool-down to the ambient temperature of 60° C. at 1 km drillhole depth. An example case analyzed here is for disposal of cesium/strontium capsules. The case is hypothetical, but it is representative of the configuration of the EBS, arrangement of the drillhole and geology of the host rock for an actual disposal project. A short segment of the horizontal drillhole is analyzed. This segment can be scaled up for the full number of capsules for disposal and total length of the disposal section. Each waste capsule is sealed in a corrosion resistant alloy canister. The configuration comprises ten canisters, 0.6-m long×11.4-cm outer diameter×9.5 mm wall thickness. Canisters are emplaced along a steel casing, 12-m long×14-cm inner dimeter×12.5 mm wall thickness. The separation between canisters along the steel casing is 0.6-m. Outer diameter of the casing is 16.5 cm, and drillhole diameter is 21.6 cm. The depth of the horizontal section is at 1 km with ambient conditions of 10 MPa hydrostatic pressure and 60° C. temperature. The metal canisters are made of highly corrosion-resistant Ni—Cr—Mo alloy. The Ni—Cr—Mo alloys have high strength, excellent fabricability along with outstanding corrosion resistance. These alloys are also highly resistant to localized corrosion processes. For analysis, Alloy 625 (UNS N06625) with composition of major elements of Ni-60 w/o, Cr-27 w/o and Mo-6 w/o represents the Ni—Cr—Mo family of alloys. The full composition in weight percent and atomic percent is presented in Table 1 as shown in FIG. 14A. The casing is made of API-5CT L80 Casing Pipe which belongs to a steel grades group of corrosion resistant casing. The composition of L80 steel in weight percent and atomic percent is Fe-87 w/o and Cr-13 w/o and Fe-85 at/o and Cr-15 at/o, respectively. Metal loss and depth of penetration were calculated from corrosion rate and time of exposure. The product of corrosion rate and time of exposure determines the depth of metal penetration and remaining wall thickness. FIG. 5D shows graph 572 that relates casing wall thickness (millimeters, y-axis) over time (years, x-axis). The corrosion rate, time of exposure and surface area of the canisters determine the volume of metal corroded. The density of Alloy 625 is 8.44 g/cm3 and of L80 steel is 7.44 g/cm3. Note that values for corrosion rate in microns per year (μm/year) are equivalent to volume of metal loss (cm3/m2-year) and mols of metal loss (mols/m2-year). Hydrogen gas generation and oxide formation analysis was for metal to metal oxide reactions presented in Table 2 as shown in FIG. 14B. The reactions are for metal reacting with water to form metal oxide and hydrogen gas. Hydrogen generated per mole of metal corroded was based on the mol fraction of elements in the metal and hydrogen generated per mole of each element. For Alloy 625, the mol fractions of iron, chromium, molybdenum and iron were 0.60, 0.27, 0.06, and 0.05 respectively. The hydrogen generated for Alloy 625 was 1.19 cm3 of hydrogen per cm3 of metal corroded. For L80 steel, the mol fractions of iron and chromium were 0.85 and 0.15, respectively. The hydrogen generated for L80 steel was 1.36 cm3 of hydrogen per cm3 of metal corroded. Volumes of hydrogen gas were calculated at Standard Temperature and Pressure (STP). Volumes at STP were then converted for pressure and temperature at the horizontal disposal section. The temperature/pressure factor is 0.016 for pressure of 10 MPa and temperature of 170° C. The volume of hydrogen gas is greatly reduced for conditions in the disposal zone. During the corrosion process, metal oxides are generated. The metal oxides formed from Alloy 625 were NiO, Cr2O3, MO2 and Fe3O4. and oxides for L80 steel were Fe3O4 and Cr2O3. Metal oxide formation per mole of alloy corroded was the sum of mol fraction times oxide formation per mol of each element. Volume expansion was that of oxides formed minus the volume of metal consumed. The oxide formation for Alloy 625 was 2.33 cm3 of oxide per cm3 of metal corroded. The oxide formation for L80 steel was 2.09 cm3 of oxide per cm3 of metal corroded. The Ni—Cr—Mo alloys provide the highest level of corrosion resistance in a wide range of environments. Alloy 625 (UNS N06625) is used for its high strength, excellent fabricability, and outstanding corrosion resistance. It is used for subsea pipelines for conveying sour gas and oil, saltwater pipelines, process pipes in the chemical industry. Extremely corrosive conditions include concentrations of hydrogen sulphide higher than 35%, temperatures reaching 220° C. and well pressure nearing 150 MPa. In the nuclear field, Alloy 625 is used for reactor-core and control-rod components in nuclear reactors. Sea water applications often require high tensile strength and corrosion resistance and Alloy 625 is used as wire rope, propeller blades, submarine auxiliary propulsion motors. It is widely used for high temp, corrosive environments. Passive metals have remarkably low corrosion rates. Measurable corrosion rate for Ni—Cr—Mo alloys is 0.1 μm/year in harsh environments. The corrosion resistance is provided by a self-forming, thin (e.g., nanometers thick) film. For Ni—Cr—Mo alloys, the passive film is a chromium rich oxide. Corrosion performance depends upon the stability/durability of this thin film. For Ni—Cr—Mo alloys, the passive film is not only remarkably stable, but the film is self-healing and reforms rapidly if damaged by mechanical or chemical action. These alloys have great resistance to localized corrosion in harsh environments, so the corrosion mode is general corrosion. The canisters will corrode extremely slowly, and the eventual perforation will be full-wall penetration in patches by general corrosion. For the base case here, the first perforation occurs at 45,000 years. Designation of corrosion rates of the canister material in the environment for each time period is a fundamental aspect the analysis. The corrosion rate and duration of the time period determine the metal penetration, volume of metal corroded, hydrogen generated, and oxide formed. Corrosion rates for Ni—Cr—Mo alloys and carbon steel were gathered from the literature, technical reports, analysis of other repository systems for nuclear waste disposal and relevant industrial applications. The objective was to collect relevant data for the range of conditions and evolution of the environment in the EBS from, e.g., laboratory and field studies in support of nuclear waste repositories in a number of countries. Further, there is information for the corrosion performance of Ni—Cr—Mo alloys and carbon steel in industrial applications. Two major reviews address corrosion issues associated with the storage and disposal of nuclear waste and corrosion performance of nuclear waste containers. Corrosion rates for disposal in deep horizontal drillhole draw upon information in those reviews and their supporting references. The corrosion rates are for general corrosion of the metals in the passive condition. Although, both EBS metals have high localized corrosion resistance in reducing environments, localized corrosion is not within the scope of this paper. The basis for designation of corrosion rates for Ni—Cr—Mo Alloys follow. Extensive studies of the corrosion behavior of Ni—Cr—Mo alloys were carried out in support of the Yucca Mountain repository. Electrochemical methods accurately determine corrosion rates to 0.01 μm/year and lower. For Ni—Cr—Mo alloys in high-temperature, oxidizing, chloride solutions measured corrosion rate in harsh environments, such as high temperature, high chloride aerated solution is 0.1 μm/year and less. These conditions are harsher than those in the first two time periods when the thermal pulse and transient oxidizing conditions occur and much harsher than those in the last three periods with anoxic environments. Expected rates for anaerobic environments are 0.01 μm/year. Designated corrosion rates for canisters in the base case are conservative, e.g., values above the expected values. Carbon steels are widely used for their strength and ease of fabrication. Steel casing will be installed in the drillhole and cemented in place for structural support. L80 steel grades (API-5CT) have greater corrosion resistance than lower grades in H2S and CO2 service. Steel is passive in reducing environments with corrosion rates 1 μm/year or less. Until it is perforated, the casing also separates the environment inside the casing from the environment between the casing and drillhole. Perforation will be full-wall penetration in patches by general corrosion. Corrosion of carbon steel in bentonite has been studied in a number of international nuclear repository programs and for oil field applications. Oxygen accelerates the corrosion of steel, And the level of oxygen is a key determinant of corrosion rates. Steel and bentonite in anaerobic conditions result in passive behavior. For carbon steel overpack in contact with compacted bentonite, corrosion rates on the order of 0.1 μm/year were measured. Corrosion rates in deaerated bentonite slurry were 1 μm/year or less. The corrosion rate measurements of Smart and co-workers appear to strongly and consistently support the suggestion that protective corrosion product films will develop on C-steel under anaerobic conditions and that long-term corrosion rates will be of the order of 0.1-1 μm/year. The basis for corrosion rates of carbon steel follow. Short-term corrosion rates for steel are several μm/year and decrease to 1 μm/year or lower after several years. In alkaline cement pore water, the primary corrosion product is Fe3O4 and corrosion rates are of the order of 0.01 μm/year to 0.1 μm/year. Designated corrosion rates for casing in the base case are considered to be conservative, e.g., values above the expected values. Evolution of the environment is a factor in analysis of corrosion behavior of materials in the EBS, and the composition of waters is an important determinant of corrosion performance. Typical pore waters in the rock are a chloride brine, and chloride concentration can range from dilute to concentrated levels. Other dissolved species can be present. The brines are anoxic and due to this lack of oxygen, the environment is highly reducing. The acidity/alkalinity ranges from near neutral to mildly alkaline, e.g., pH 6-10. The horizontal drillhole is in fully saturated rock. There is a region of “disturbed” rock surrounding the drillhole. This rock has been cracked or otherwise affected by the drilling process. The disturbed zone typically extends beyond the drillhole wall to a distance equivalent to about one drillhole radius. For the base case, the disturbed zone is about 10-cm thick, and undisturbed rock extends outward from there. Process waters from drilling and placement of the casing and/or pore water from the surrounding rock fully saturate the disturbed rock. In typical shale gas/oil operations, the annular space between the casing and rock is filled with cement. Although for waste storage in the horizontal drillholes such cement is may not be necessary, the base case includes cementing of the casing/drillhole annular space. With cement, the waters are moderately alkaline with pH 10-12. Steel corrosion rates in alkaline solution are low and remain so while alkaline species persist. The canister/casing annular space is filled with bentonite-based slurries. Alternate fillings in the EBS can be used. When canisters containing nuclear waste are emplaced, the temperature in the rock initially rises, and then decreases as the short-lived radioisotopes decay and heat is conducted away. The rock is saturated and EBS is filled with liquid and solids with no vapor space. Due to hydrostatic pressure at depth, there is no boiling of water. Corrosion processes are for metal immersed in liquid. The aggressive conditions and complex behaviors of water droplets and films on hot metal surfaces, liquid/vapor interfaces and two-phase liquid/vapor processes do not pertain. Corrosion processes in the EBS modify the environment. Corrosion of the steel casing is the primary process for reduction of residual oxygen in the drillhole. Hydrogen gas generation and metal oxide formation accompany the metal loss by corrosion in anaerobic conditions. Hydrogen generation beyond the solubility of hydrogen in the drillhole and rock waters produces a gas phase within the EBS. The high hydrostatic pressure at the depth of the drillhole greatly reduces the volume of gas. Hydrogen is a nutrient for microbiological activity that can affect the environment. The formation of metal oxides by the corrosion process can also alter the chemistry of the environment and can affect the sorption and transport of species. Volume expansion from metal going to metal oxide increases the pressure in EBS. Temperature is also a characteristic of the environment for corrosion analysis. The horizontal disposal section goes through a heat-up and cool-down cycle. The decay of the radioisotopes generates heat in the fuel pellets, which is transferred to the canisters, the casing, and the rock. The result is that the temperature rises and then decreases as the radioactivity drops and the rock continues to conduct heat away. After emplacement of the canisters, the temperature in the horizontal disposal section rises from ambient to a maximum temperature and then slowly cools down to ambient. The temperature-time behaviors from canister emplacement to 10,000 years or more have been computed by thermal modeling for a number of repository systems. The thermal evolution near heat-generating nuclear waste canisters disposed in horizontal drillholes has been analyzed by numerical simulations. The hypothetical case analyzed here draws representative values from these simulations. Heating to maximum temperature and start of cool down occurs within 5-10 years after canister emplacement. The maximum temperatures are: capsule, 182° C.; canister wall, 170° C.; casing, 165° C.; drillhole wall, 160° C.; and 1-m into the rock, 103° C. Heat up extends a few meters into the host rock. There is a 40-year thermal heat-up and cool-down period followed by a slow cool-down to the ambient temperature. The ambient temperature and pressure are determined by the depth of the horizontal section. For a drillhole at 1 km depth, representative values are temperature of 60° C. and hydrostatic pressure of 10 MPa. There is no boiling of water on metal surfaces within the EBS or the host rock due to hydrostatic pressure at depth. For 10 MPa hydrostatic pressure, the boiling point of water is 310° C. For the analysis, five periods are defined, and conditions described to span time from canister emplacement through 10,000 years. The approach is useful to break the extremely long repository time frame into more manageable zones to track corrosion behavior, evolution of the environment and amount of cumulative damage. The objective is to relate corrosion performance to evolution of the environment over time. Corrosivity of the environment is more aggressive in the early times with the heat-up and cool-down cycle and much more benign in the later years with highly reducing, anoxic, lower temperatures conditions. An initial period from 0-2 years covers an early transition period for change in corrosion resistance of the alloy and evolution of the environment toward reducing conditions. The exposure conditions are moderately oxidizing from oxygen introduced during the drilling, casing installation and canister emplacement. The initial heat-up has begun, and there are fresh metal surfaces on canisters and casing. During a second time period from 2-20 years, the highest canister surface temperature is reached, cool-down begins, and oxygen is consumed moving conditions to a highly reducing environment. The next two periods allow for consideration of temperature on corrosion behavior in highly reducing environment as temperature cools. For the third period from 20-100 years, the temperature has cooled from 120 to 80° C., and for the fourth period from 100-1000 years, the temperature cooled further to less than 80° C. At 1000 years, the temperature had cooled to 60° C. the ambient rock temperature, and remains steady for 10,000 years and beyond. Conditions during the fifth period from 1000 to 10,000 year are 60° C. and a highly reducing environment. Corrosion rates during each period were set based on reported values for corrosion in the relevant environment for that period. A single corrosion rate was assigned for the period and for calculations presumed to remain constant. Amounts of corrosion, hydrogen generation and oxide formation were determined for the corrosion rate and duration of each period. The corrosion rates were set higher during initial periods and lower for periods after 100 years, due to lower corrosivity and temperatures in later periods. The corrosion rates were set at the upper range of reported rates for general corrosion of nickel-chromium-molybdenum alloys and carbon steel for conditions during each period. For Alloy 625, the corrosion rate for the first 20 years was set at 2 μm/year to recognize higher temperatures and an early transition period from fresh metal surfaces and not fully anaerobic conditions. The rate for 20 to 100 years was set at 1 μm/year to recognize temperature above ambient. The baseline corrosion rate after 100 years was set at 0.1 μm/year which is above expected values for the benign environmental conditions that pertain. For carbon steel, the corrosion rate for years 0-2 was set at 20 μm/year for fresh metal surfaces in the moderately oxidizing environment. The rate decreased to 4 μm/year for years 2-20 due to transition toward fully anaerobic conditions. The rate was 2 μm/year for corrosion in high temperature, anoxic brines. After 100 years, the corrosion rate was 1 μm/year as temperature decreased and reached ambient conditions. To demonstrate the impact of corrosion rate on the annual amounts of metal loss, hydrogen gas generation and metal oxide formation were determined for a range of corrosion rates. For each corrosion rates (μm/year), the annual weight loss (g/m2), mols of hydrogen generated (mols/m2), volume of oxide generated (cm3/m2) and volume expansion from oxide formation (cm3/m2) are presented. Data for canisters made of Alloy 625 are presented in Table 3 (as shown in FIG. 14C) for corrosion rates varying from 0.01 to 10 μm/year. Over the range of corrosion rates, the weight loss was from 0.084 to 84 g/m2-year. For Alloy 625, 1.19 mots of hydrogen gas are generated per mol of metal corroded, and gas generation ranged from 0.01 to 11.9 mols hydrogen per year per square meter of canister surface. There are 2.33 mols of metal oxide produced per mole of alloy 625 corroded, and the volume of metal oxide corrosion products ranged from 0.03 to 28 cm3/m2 of canister metal per year. The volume expansion of solids from the formation of these corrosion products ranged from 0.02 to 18 cm3/m2 of canister metal per year The corrosion behavior of the Alloy 625 canisters is presented in Table 4 (as shown in FIG. 14D) for time periods from the time of canister emplacement to 10,000 years. The excellent corrosion resistance of the Ni—Cr—Mo alloy is reflected in the extremely low metal penetration rates. The metal losses are 0.12 mm at 100 years, 0.2 mm at 1000 years and only 1.1 mm after 10,000 years. The starting metal thickness of 9.25 mm is reduced to 9 mm after 1000 years, and wall thickness remains over 8-mm after 10,000. While general corrosion is the mode of corrosion, the advancing corroded surface is not perfectly smooth but has some shallow hills and valleys. Also, mechanical strength of the canister decreases as a result of the metal loss. To account for these, a criterion for time-to-perforation of the canister wall was set to be the time to penetrate 50% of the 9.25 mm wall thickness. On that basis, the first perforation occurs at 45,000 years. Until that time, canister has remained an absolute, non-permeable barrier to the transport of radionuclides, and no waters from outside the canister enter and come in contact with the nuclear waste. After the early years corrosion rates decrease dramatically for three primary reasons. The environment evolves from moderately oxidizing to highly reducing as residual oxygen in the drillhole is consumed and conditions are anoxic. The heat-up and cool-down period is nearly complete, and temperatures are at or near ambient conditions. The corrosion resistance of the alloy has become even greater. A self-forming, protective film on the Ni—Cr—Mo alloy provides the corrosion resistance, and the structure and composition of the air-formed film change on exposure to the environment. The aging film becomes more protective over the first months/years. In oxygen-free, reducing waters, hydrogen gas generation and metal oxide formation are products of the metal corrosion process. The metal oxides formed from Alloy 625 were NiO, Cr2O3, MO2 and Fe3O4. The reactions for metal reacting with water were presented in Table 3. Volume of hydrogen gas was calculated at STP and converted to volume at the pressure and temperature at the depth of the horizontal drillhole. Values for hydrogen gas generation are presented in Table 5 (as shown in FIG. 14E) for each time period and for cumulative gas generation. Cumulative gas generation at (STP) were 11 cm3, 114 cm3, 341 cm3, and 597 cm3 at 2, 20, 100 and 1,000 years, respectively. The rates of gas generated decreased significantly with time. Values for the hydrogen generated per year are presented in FIG. 5C. FIG. 5C shows graph 570 that relates hydrogen generated (cubic centimeters, y-axis) over time (years, x-axis). Rates were highest at 5.7 cm3/year during the initial transition period and then greatly reduced to 0.3 cm3/year after 100 years. Hydrogen is soluble in the drillhole and rock waters, and if the solubility limit is exceeded then hydrogen gas forms. The percent of volume expansion from gas generation within the casing and within the overall drillhole were calculated based on the volume of gas at temperature and pressure of the horizontal disposal section. Volume expansion from hydrogen gas is negligible, 0.001% at 100 years and 0.01% or less for all times less than 1000 years. This is due to the modest amounts of hydrogen generated and a reduction in the volume of gas by a factor of 100 due to hydrostatic pressure at the depth of the drillhole. The amounts of metal oxide per period and cumulative volume expansion from the metal oxide generated are presented in Table 6 as shown in FIG. 14F. The cumulative volume for oxide expansion is the net change in volume from the cm3 of oxide generated minus the cm3 of metal loss. Cumulative volume expansion at 100 and 1000 years was 668 cm3 and 1169 cm3 respectively. Metal oxides formed by the corrosion process build up a porous layer on the metal surface. Metal loss at 1000 years is 0.2 mm, and thickness of the corrosion product layer on top of the protective film at 1000 years is less than 0.5 mm. The volume of metal oxide formed is 2.3 times that of the metal corroded, and volume expansion from metal going to metal oxide increases the pressure in the engineered barrier system. The volume of oxide expansion per year tracked the corrosion metal loss. Values of oxide expansion per year as a function of time after and emplacement of the canisters are presented in Table 7 as shown in FIG. 14G. After the initial transition period of 100 years, the rates have decreased dramatically to 0.6 cm3 per year. The percent volume expansion values from canister corrosion and oxide generation within the casing and within the overall drillhole show expansion at 1000 years is nearly 1% inside the casing and 0.3% for the total drillhole volume. To demonstrate the impact of corrosion rate on the annual amounts of metal loss, hydrogen gas generation and metal oxide formation were determined for a range of corrosion rates. Data for L80 steel casing are presented in Table 8 (as shown in FIG. 14H) for corrosion rates varying from 0.01 to 100 μm/year. Over the range of corrosion rates, the weight loss was from 0.78 to 784 g/m2-year. For L80 steel, 1.36 mols of hydrogen gas are generated per mol of metal corroded, and gas generation ranged from 0.13 to 136 mols hydrogen per year per square meter of canister surface. There are 2.09 mols of metal oxide produced per mole of L80 steel corroded, and the volume of metal oxide corrosion products ranged from 0.21 to 209 cm3/m2 of canister metal per year. The volume expansion of solids from the formation of these corrosion products ranged from 0.11 to 109 cm3/m2 of canister metal per year. If the 12.5 mm thick steel casing corrodes at 10 μm/year, all of the steel would be consumed after 625 years. The corrosion behavior of L80 steel casing is presented in Table 9 (as shown in FIG. 14I) for time periods from the time of canister emplacement to 10,000 years. Corrosion rates are highest during the initial periods while residual oxygen is being consumed and higher temperatures are experienced. After 100 years, the environment is anaerobic, and corrosion proceeds slowly at the 1 μm/year rate. The metal loss at 100 years is 0.27 mm and at 1000 years is 1.2 mm. Remaining casing wall thickness is 12 mm and 10.2 mm at 100 years and 1000 years, respectively. The time to consume all of the steel is 6,078 years. While general corrosion is the mode of corrosion, the advancing corroded surface is not perfectly smooth but has some shallow hills and valleys. Also, mechanical strength of the casing decreases as a result of the metal loss. A criterion for time-to-perforation of the casing wall was set to be the time to penetrate 50% of the 12.5 mm wall thickness. On that basis, the first perforation occurs at nearly 3,000 years. After the early years corrosion rates decrease dramatically. The environment evolves from moderately oxidizing to highly reducing as residual oxygen in the drillhole is consumed and conditions are anoxic. The heat-up and cool-down period is nearly complete, and temperatures are at or near ambient conditions. In oxygen-free, reducing waters, hydrogen gas generation and metal oxide formation are products of the metal corrosion process. The metal oxides formed from L80 steel are Fe3O4 and Cr2O3. The reactions for metal reacting with water are presented in Table 3. Volume of hydrogen gas was calculated at STP and converted to volume at the pressure and temperature at the depth of the horizontal drillhole. Values for hydrogen gas generation for each time period and cumulative gas generation are presented in Table 10 as shown in FIG. 14J. Cumulative gas generations at STP were 636 cm3, 1780 cm3, 3687 cm3, and 16844 cm3 at 2, 20, 100 and 1,000 years, respectively. The rates of gas generated decreased significantly with time. Values for the hydrogen generated per year are presented in FIG. 5E. FIG. 5E shows graph 576 that relates hydrogen generated (cubic centimeters, y-axis) over time (years, x-axis). Rates were highest at 318 cm3/year during the initial transition period and then greatly reduced to 25 cm3/year after 100 years. Hydrogen is soluble in the drillhole and rock waters, and if the solubility limit is exceeded then hydrogen gas forms. The percent of volume expansion from gas generation within the casing and within the overall drillhole were calculated based on the volume of gas at temperature and pressure of the horizontal disposal section. Since, there is reduction in the volume of gas by a factor of 100 due to hydrostatic pressure at the depth of the drillhole, volume expansion from hydrogen gas is negligible, 0.015% at 100 years and 0.07% at 1000 years. Volume metal oxide formed and volume expansion per period are presented in Table 11 as shown in FIG. 14K. Each volume of metal consumed forms 2.09 volume of metal oxides, and the net volume expansion is volume of oxide formed minus volume metal corroded. Results are presented for oxide formation from the inner casing surface, the outer casing surface and total oxide formation within the drillhole. The inner surface area of the casing is 5.36 m2 and outer surface area is 6.32 m2. The volume of metal loss and oxide formation is greater on the outer surface of the casing than on the inner surface due to the surface area. Volume of oxide produced per period increases while the corrosion rate decreases due to longer duration of each subsequent period. The volume expansion per year for each subsequent period decreases. The total volume of oxide formation in the drillhole per year is shown in FIG. 5F. FIG. 5F shows graph 578 that relates a volume of oxide expansion (cubic centimeters, y-axis) over time (years, x-axis). The volume expansion drops from 254 cm3 in the first two years to 51 cm3 in the next 18 years and then continues to decrease for subsequent periods. At 100 and 1000 years, the casing metal loss is 0.27-mm and 1.2-mm, respectively. Presuming a fully dense oxide with no porosity, the oxide formation thicknesses are 0.56-mm and 2.5-mm. The oxide thickness on inner and outer casing surfaces are approximately equal. When the casing is fully consumed, the 12.5-mm thick steel will have produced oxide equivalent to a layer 26-mm thick. Cumulative volume of oxide formed and volume expansion from the oxide are presented in Table 12 (as shown in FIG. 14L) for each period. Total oxide expansion in the drillhole was 3,448 cm3 and 14,858 cm3 at 100 and 1000 years, respectively. Corresponding percent volume expansion was 0.9% and 3.87%. Total expansion after 6078 years when the steel casing was completely consumed is 12%. Volume expansion from metal going to metal oxide increases the pressure in the engineered barrier system. Pressure increases independently inside and outside the casing until the casing is perforated and cumulatively in the drillhole perforation. Volume expansion of solids in the drillhole is equivalent to a volume reduction of waters. The coefficient of compressibility of water is 4.4E-10 l/Pa. One percent volume decrease of waters leads to a pressure increase of 22 MPa. The ultimate goal is to design and construct a nuclear waste disposal system with a robust EBS to meet the long-term requirements for safe and reliable disposal. The analysis of the corrosion behavior of a base case for disposal in deep horizontal drill holes yields information on the performance of canisters and casing over 10,000 years and provides insights to design of the EBS. The analysis focused on the corrosion behavior of canisters and casing, since corrosion is the greatest risk for canister perforation, and the casing is an integral component of the engineered barrier system. In summary, the canisters made of corrosion resistant, Ni—Cr—Mo alloy exhibited excellent performance. Metal loss was only 1.1 mm after 1,000 years and starting metal thickness of 9.25-mm remained over 8-mm thick after 10,000. For a criterion that first perforation of the canister wall is when 50% the wall thickness is consumed, the canister was an absolute barrier to water penetration into and radionuclide egress for 45.000 years. The L80 steel casing was reduced from 12.5 mm to 10.2 mm after 1000 years. The first perforation of casing occurred at nearly 3,000 years, and the time to consume all of the steel was 6,078 years. Until the casing is perforated, it is a barrier between the inner canister/casing environment and the outer casing/drillhole environment. Hydrogen generation and metal oxide formation accompanied the metal corrosion. The amount of hydrogen and metal oxide formed tracked the corrosion rate of the metals. Both were higher for the corrosion of steel casing than for the corrosion resistant alloy canister. The calculated STP volume of hydrogen generated was reduced by a factor of 100 in the drillhole by the hydrostatic pressure at the depth. Molar volume of metal oxides formed is 2.3 and 2.09 times the metal molar volume of Alloy 625 and L80 steel, respectively. The volume expansion from metal oxide formation increases the pressure within the EBS. The hydrogen generation and metal oxide formation can affect the environment, transport, sorption and other processes. The analysis of these affects is beyond the scope of this paper. Two factors for analysis of corrosion performance are the corrosion resistance of the metal and the corrosivity of the environment. The interaction between these two determines the corrosion modes of interest, the corrosion processes and rates, potential failure modes and performance assessment. A special feature of analysis of waste disposal is the extraordinarily long periods of performance with interest extending to 10,000 years and beyond. A useful tool is to define time periods with duration and environmental conditions to follow the evolution of the environment. Here, five periods were defined. The first two, 0-2 and 2-20 years, covered the heat-up to maximum temperature, start of cool-down and transition from moderately oxidizing the anaerobic conditions as residual oxygen was consumed. The remaining periods captured times to cool to 120° C., 80° C. and near ambient 60° C. These temperatures relate to thresholds for corrosion phenomena and availability of industrial experience and laboratory data and analysis. The trajectories of corrosion, hydrogen generation and oxide formation over the 10,000 years are apparent from the results. Annual metal loss, hydrogen generated, and oxide formation dropped dramatically in the first 20 years. This is due primarily to the transition to anaerobic conditions. After this time, the environment is anoxic and remains essentially unchanged for the duration. After 100 years, the environment is has cooled to 80° C. and low corrosion rates prevail for canister and casing. A challenge then is to design a system that makes it through the aggressive, initial period of 100 years and enters the anaerobic period in condition to survive for 10,000's of years. This focuses a 10,000-year analysis on the first 100 years when the thermal pulse has peaked, and temperature has decreased to 80° C. Analysis for 100 years is within the realm of traditional engineering and experience. Thus, there is a consideration of which alloy to select for canisters that will survive 10's of years at high temperature (170° C. in the base case here) in a moderately oxidizing environment with minimal damage. The greatest threat to passive metals for these conditions is susceptibility to localized corrosion processes. Hence, the selection of Ni—Cr—Mo alloy for the canisters. For reducing environments, Ni—Cr—Mo would be considered an overkill and less corrosion resistant metals at lower cost would be selected. However, a highly corrosion resistant alloy is required in lieu of the unprecedented needs for safe and reliable containment, survival through the aggressive transition period and the extraordinarily long time periods to follow. Ni—Cr—Mo alloys have both mechanical strength and corrosion resistance to meet these needs. During the initial 20 to 100 years, aggressive corrosion conditions prevail from the high temperatures and moderately oxidizing environment. A number of passive alloys, e.g., those that form protective passive films, can be passive and have extremely low corrosion rates in this environment. However, the key issues are will the metal remain passive, and if the passive film is damaged will it reform spontaneously. If the passive film does not reform, then the metal is severely damaged and penetration rates are rapid. The Ni—Cr—Mo alloys are designed to have a durable protective (passive) film that is self-forming and will reform quickly if damaged mechanically or chemically. This is what distinguishes them from lesser corrosion resistant alloys. The engineered barrier consists of several components at work in combination to prevent the transport of radionuclides from the EBS to the host rock for the regulated period of time period and beyond. An overall assessment of the EBS starts with the uranium dioxide spent fuel pellets and moves outward to the drillhole surface. The focus here is on corrosion performance of the EBS. The components of interest are the canister, filler between canister and casing, casing, and filler between casing and drill hole. The strategy for corrosion mitigation is to select suitable materials for performance in the evolving environments over 10,000 years. Design considerations for materials and the environment are presented below along with some strategies to further enhance EBS performance. In addition to the admirable corrosion resistance, Ni—Cr—Mo alloys have structural strength to elevated temperatures beyond those for deep isolation disposal. They are available in shapes and sizes required. Canisters can be manufactured by common industrial processes. Welding procedures are standardized as are inspection and quality procedures along a substantial history of industrial applications and extensive materials performance data. There is a wealth of data and analysis documenting the outstanding corrosion behavior of Ni—Cr—Mo alloys in hostile environments. In particular, comprehensive studies were carried out is support of the Yucca Mountain repository under conditions much harsher than those for deep horizontal drillhole disposal. Carbon steel casing has the required structural strength, availability in shapes and sizes, and fabricability. These steels are widely used in the oil field and a broad range of other industrial applications. The casing is made of API-5CT L80 Casing Pipe which belongs to a steel grades group of corrosion resistant casing. It has greater corrosion resistance than plain carbon steels in environments that contain hydrogen sulfide and carbon dioxide. Several international programs for nuclear waste repositories have generated laboratory data and field test results for carbon steels in anaerobic environments. Corrosion rates are documented to be extremely low. A number of design considerations relate to aspects of the environment. In addition to suitable location and favorable geology, site selection and environmental characteristics contribute to the high performance of the EBS. Highly Reducing Environment: After a brief transition period, the initial moderately oxidizing environment is much less corrosive, and environmental conditions thereafter remain steady. Uniformity of environment: Conditions along the horizontal drillhole are uniform. In comparison, vertical segments can go through a variety of layers with different aeration levels. Heterogeneous aeration/deaeration zones can result in localized corrosion or longline corrosion cells. Zones with carbon dioxide or hydrogen sulfide in waters can cause severe corrosion. This is a distinguishing factor for the long life of steel casing in the horizontal segment versus industrial experience for oil and gas wells where high corrosion rates and casing lives of 20-50 years are observed. Fully Saturated Rock: The absence of two-phase gas/liquid solutions eliminates aggressive corrosion processes such as droplets on hot metal surfaces and thin films of moisture in the vapor phase. Full saturation in the rock simplifies the analysis and reduces uncertainty. Hydrostatic pressure at depth: There is no boiling of waters at the canister surface or in the rock due to suppression of boiling by hydrostatic pressure. Unlike conditions for several other repository systems, there is no boiling on metal surfaces or in the rock and no wet-dry-wet cycle for these horizontal drillholes. Environment between canisters and casing: Procedures are designed to minimize oxygen in waters and fluids for drilling, casing installation and canister emplacement. The canister/casing annular space is filled with bentonite-based slurries that are treated to minimize dissolved oxygen. Environment between casing and drillhole: Cement fills this annular space and modulates pore waters to be moderately alkaline. Steel corrosion rates in alkaline solution are low. Alkalinity will not persist for repository times; however, reduction of corrosion is beneficial during the early transition period to anaerobic conditions. Reduction of the thermal period and lower maximum temperatures would shorten exposure times to the harsh environment, lower corrosion rates and decrease risk for localized corrosion. Spacing of canisters along the drillhole allows control of the thermal load and temperatures. Controlled selection and spacing of canisters based on the heat load of contained waste can yield uniform temperatures and avoid hot spots. EBS design can enhance heat transfer to the rock. Means to quicken the transition to reducing conditions include reduction of residual oxygen in the drillhole and use scavenger species to consume or tie up oxygen. The horizontal drill hole configuration and procedures for placement of casing and emplacement of canisters allows the use of “engineered” fillings for the canister/casing and casing/drillhole annular spaces. This provides the opportunity to control the environment within the casing from installation in the drillhole until the casing is perforated. In this analysis, casing perforation was at nearly 3000 years. Bentonite-based slurries or the like are infused with beneficial additives, such as oxygen scavengers, alkalinity modulators and antimicrobial treatments. Space between the casing and drillhole is filled with cementitious filler and it is feasible to include other beneficial additives. The mild alkalinity promotes passivity of the steel at the initial casing installation. The alkalinity and additives moderate corrosion during the retrieval period prior to sealing and during the transition time to anaerobic conditions. Volume expansion from formation of iron corrosion products increases pressure inside and outside of the casing. The pressure could be moderated in the EBS by expansion absorbers, such as empty thin-walled components and expansion zones such as the capsule spacing 509 shown in FIG. 5B) included along the drillhole. Increasing the drillhole diameter increases volume for expansion and reduction of the amount of steel decreases the volume of metal oxide formed. The objectives were to analyze the corrosion performance of the EBS for disposal of nuclear waste in deep horizontal drillholes and to relate aspects of the EBS design to corrosion performance. For analysis, time periods with defined duration and environmental conditions to follow the evolution of the environment over 10,000 years. The first 20 years, covered the heat-up to maximum temperature, start of cool-down and transition from moderately oxidizing the anaerobic conditions. Corrosion, hydrogen generation and oxide formation decreased drastically after this early transition period. The major challenge then is to design a system that makes it through the aggressive, initial period and enters the anaerobic period in condition to survive for 10,000's of years. Canisters of Ni—Cr—Mo meet this challenge extremely low corrosion rates and extraordinary resistance to localized corrosion in high temperature, moderately oxidizing environments. Canisters of Ni—Cr—Mo alloy exhibited excellent performance. Metal loss at 1000 years was 0.2 mm and only 1.1 mm at 10,000 years. Time for the first perforation of a canister with 9.5-mm wall was 45,000 years. The L80 steel casing was reduced from 12.5 mm to 10.2 mm after 1000 years. The first perforation of casing occurred at nearly 3,000 years. The canister remained an absolute, non-permeable barrier to water entry and egress of radionuclides for 10,000's years, and the casing was a barrier between the inner canister/casing environment and the outer casing/drillhole environment for several 1000's years. In design of the EBS, materials are selected for performance to meet repository requirements in the evolving environment over 10,000 years. Several aspects of deep horizontal drillholes are beneficial, the environment is highly reducing after the initial transition period and uniform along the length of the drillhole, the rock is fully saturated and hydrostatic pressure at depth suppresses boiling. In addition, the EBS configuration allows for use of “engineered” fillings for the canister/casing and casing/drillhole annular spaces. Thermal load and temperatures can be controlled by spacing and distribution of canisters based on heat load of the contained waste. The deep horizontal drillhole waste disposal system has favorable attributes that contribute to a strong technical basis for long-term control of radionuclide transport and reduction of uncertainty in the supporting safety case. In part. because the horizontal drillhole disposal system avoids several phenomena and processes that pertain to other repository systems, complicate their analysis and increase uncertainty. Major complexities for analysis avoided by deep horizontal drillholes include: analysis of two-phase gas/liquid processes, determining the effects of boiling on metal surfaces and in surrounding rock, and dealing with a large thermal pulse from decay heat that results in a wet-dry-wet cycle over time. Disposal of Radioactive Liquid in Directional Drillholes. As described with reference to FIGS. 1-4, radioactive liquid, such as radioactive water, may be emplaced (in canisters or not) within a storage area of a hazardous waste repository, such as repositories 100, 200, 300, and 400. For example, at the Fukushima nuclear reactor site in Japan, water flowing underground past the melted radioactive waste became contaminated with tritium, the radioactive isotope of hydrogen. Tritium has a half-life of 12.3 years. This contaminated (e.g., tritiated) water at Fukushima had a total volume of nearly a million cubic meters and is currently stored in large 1000 cubic meter tanks above ground at the site. It has been proposed that the water could be discharged into the sea, but that approach has been strongly opposed by the Japanese public. The Japanese government has also considered placing the tritiated water in a concrete pit at a relatively shallow depth (less than 10 meters below the surface). This underground burial is too shallow to have any substantial advantage from geologic isolation. An alternative method proposed for the disposal of the tritiated water of Fukushima is “geosphere injection.” Three different variations were considered: with no pre-treatment, with dilution, and with separation. The fundamental barrier is that suitable underground formations may not exist, and a regulatory barrier that disposal of radioactive waste in liquid form is prohibited by the Nuclear Regulation Authority of Japan unless severely diluted (by a factor of 70 or more). Example implementations of the present disclosure describe disposing radioactive water in deep, human-unoccupiable, directional drillholes formed into one or more subterranean formations as described with reference to FIGS. 1-4. In some aspects, a subterranean formation (layers 118, 218, 319, and 419 and other suitable formations) into which the deep, directional drillholes are formed contain stagnant water (e.g., stagnant brine). In some aspects, the stagnant water can be shown to be sufficiently old that by the time any of the stagnant water can reach surface water (e.g., potable or human-consumable water from surface water surfaces), the tritium will have decayed to natural levels. Since the half-life of tritium is 12.3 years, that means that after 123 years, the radioactivity is reduced by a factor of (1/2)10=0.001. After 250 years, the level is reduced to (1/2)20=0.000001=1 millionth of its original level. In some aspects, the stagnant water can be tested (e.g., by using radioisotopic methods) to determine whether the age of the stagnant water is sufficiently old (and thus the mobility is sufficiently slow toward a terranean surface). For example, the sufficiency of the age (and/or mobility) of the stagnant water can be determined using measurements of natural tritium and carbon-14 that exist in the brines stored in the proposed disposal subterranean formation. If the presence of these radioisotopes is low compared to the levels in the surface waters, then isolation for required times can be demonstrated. The subterranean formation, based on the successful testing, may be suitable as a hazardous waste repository for the long term (e.g., tens, hundreds, or thousands of years) storage of hazardous waste, such as radioactive water. In the example implementations of hazardous waste repositories 100, 200, 300, and 400, radioactive water can be stored in long horizontal or nearly-horizontal drillholes at depths from a few hundred meters to several kilometers below the Earth's surface. The depth may be chosen as one that through measurements of tritium and/or carbon-14, satisfies the isolation requirements previously described. In alternative implementations, vertical drillholes may be used, although such vertical or nearly vertical drillholes (e.g., with no directional or horizontal portion) may provide less disposal volume per drillhole than can be obtained by directional or horizontal drillholes. As described with reference to FIGS. 1-4. a vertical or nearly vertical access drillhole is drilled from the surface. Before the drillhole reaches the disposal formation, the direction of the drillhole is curved so that when it reaches the disposal formation, the drillhole is horizontal or nearly horizontal. In some aspects, the drillhole is made that can hold an 8-inch diameter canister. An 8″ hole has area of 0.033 square meters=1/30 square meters. For 3 km, the volume of the drillhole is 100 cubic meters. For 1,000,000 cubic meters of tritiated (e.g., radioactive) water, 10,000 such drillholes would be required. The drillhole could be drilled with a larger diameter than 8 inches. If, for example, the drillhole has a 16-inch bore, then only 2,500 drillholes need be drilled. A drillhole with a 16-inch hole might be more cost efficient for disposal than would be an 8-inch well. In some aspects, the vertical portion and curved portion of the drillhole are “access” portions in that no radioactive water is stored in these portions (and only stored in the horizontal drillhole portion). The access portions are used to convey the radioactive liquid (inside of canisters or not) to the horizontal or nearly horizontal disposal drillhole portion. After disposal, the access portions may be sealed. In some example implementations, a single access section (e.g., a single vertical portion) can be used to access several horizontal or nearly horizontal disposal drillhole portions (e.g., several multilateral drillhole portions). In disposal of spent nuclear fuel assemblies, the use of multilateral disposal is sometimes avoided because recovery of the assemblies may be made more difficult. However, if there is no perceived value in the recovery of the waste, then there may be cost-saving advantages to using multilateral disposal drillhole portions for the long-term storage of tritiated (e.g., radioactive) water. As described in FIGS. 1-4, all or portions of the directional drillhole may include a casing (e.g., tubular pipe sections secured into place with cement to the drillhole). Alternatively, all or portions of the directional drillhole may not include any casing or cement. For example, the casing may not provide an advantage if no retrieval of waste is required. Also, the absence of casing and cement may increase a volume available for disposal of the radioactive water. In some aspects, the directional drillhole may first be drained of any brine that has entered it, and then filled with the radioactive water. The access hole would then be sealed (e.g., with a wellbore plug or packer, or cement, bentonite, gravel and rock). In some aspects, there may be multiple seals (e.g., at a junction between the horizontal drillhole portion and access portion, and in the vertical access portion). In some aspects, the radioactive water may be circulated (e.g., pumped) into the directional drillhole portion, which is then sealed. In some aspects, the radioactive water may be mixed with cement, such as the cement used to secure the casing to the drillhole or another batch of cement that is then pumped into the drillhole (or both). Since cement can be made that contains 70% or more water by volume, the required drillhole length at depth could be increased by about 43%. In other aspects, the radioactive water may be mixed with a gel such as sodium polyacrylate, and the gel could then be pumped into the hole. Sodium polyacrylate creates a gel that is more than 99% water, so no additional drilling would be needed. In some aspects, the radioactive water mixed with either gel or cement can be place in sealed canisters (as described with reference to FIGS. 1 and 3), and these can be moved into the horizontal drillhole section. For example, the canisters may be moved into the drillhole with a wireline tractor or by coiled tubing or drillpipe. In some aspects, canisters containing tritiated water (or radioactive water mixed with cement or gel) may be lighter in weight than canisters that contain spent nuclear fuel assemblies. This is because water is much less dense than the uranium dioxide that makes up the spent nuclear fuel assembly. If canisters are used, they can be made of CRA (corrosion-resistant alloy). CRAs include Alloy-22 and Alloy-625, both made of nickel/molybdenum/chrome. There are other CRAs that may be used for the canisters. The canisters could be made of non CRA if the chemistry at depth indicates that corrosion will not breach the canister walls within 125 to 250 years, at which time the tritium will have decayed. However, if other radioisotope such as Sr-90 and Cs-137 are present in the water at significantly high levels, a CRA might be preferred. The half-life of Sr-90 and Cs-137 are both about 30 years, so ten half-lives would possibly require a 300-year CRA. In some aspects, after emplacement of the radioactive water (in canisters or not), the disposed water can be monitored. Monitoring can be done by placing a wire or fiber optic in the access hole and one (or part of the same one) in the disposal drillhole. The wire or fiber can have a radiation sensor at the end or at locations along its length. One possible radiation sensor would be a phosphor coating on a fiber cable. When tritium decays it emits a short-range beta particle (an electron). If that beta particle enters a phosphor, it causes a pulse of light to be emitted. If the tritiated water is placed in canisters, and the sensor is outside, then the presence of light pulses with the magnitude expected for tritium would be an indicator that tritium has escaped the canister. If no canister is used, then the radiation sensor should show a steady rate of light pulses dropping in rate with a half-life of 12.3 years. That drop begins immediately, and in 1 month the drop would be 0.5%, a change that is readily detectable. If canisters are not used, then any additional drop in rate could be due to increased saturation of the gel by brine or a loss of tritium by flow into the rock. The fact that a measurable change is expected from tritium decay helps assure that the monitoring method is functional. The tritium detector could also be used to detect gamma radiation from Cs-137 or beta decay and gamma radiation from Sr-90. These rays would be distinguished from the tritium beta rays by the larger flash of light they would produce. These rates might be too low to be detected by the tritium detector, and if a separate monitor for them is required, a larger scintillation detector connected by a fiber optic cable could be included. This Cs/Sr detector would be surrounded by a metal shield that would prevent the low energy beta rays from tritium to enter can cause the scintillator to flash. Thermal Test Process for Determination of Suitability of a Subterranean Formation as a Hazardous Waste Repository. FIGS. 6A-6D are schematic illustrations of a thermal property testing system 600 for a hazardous material storage repository. For example, the disposal of spent nuclear fuel and high-level radioactive waste in horizontal holes drilled into deep, low-permeable geologic formations using directional drilling technology is described herein (e.g., with reference to FIGS. 1-4). Residual decay heat emanating from these waste forms leads to temperature increases within the drillhole and the surrounding host rock. The spacing of waste canisters and the configuration of the various barrier components within the horizontal drillhole can be designed such that the maximum temperatures remain below limits that are set for each element of the engineered and natural repository system. The present disclosure includes design calculations that examine the thermal evolution around heat-generating waste for a wide range of material properties and disposal configurations. Moreover, the present disclosure describes alternative layouts of a monitoring system to be part of an in situ heater test that helps determine the thermal properties of the as-built repository system. A data-worth analysis is performed to ensure that sufficient information will be collected during the heater test so that subsequent model predictions of the thermal evolution around horizontal deposition holes will reliably estimate the maximum temperatures in the drillhole. The simulations demonstrate that the proposed drillhole disposal strategy can be flexibly designed to ensure dissipation of the heat generated by decaying nuclear waste. The present disclosure thus describes an in situ heater test can provide the relevant data needed to develop a reliable prediction model of repository performance under as-built conditions, thereby providing a determination that a subterranean formation is suitable (or not) as a hazardous waste repository, such as any one of repositories 100, 200, 300, and 400. FIG. 6A illustrates an example thermal property testing system 600 for a hazardous material storage repository. In this example of FIG. 6A, the thermal property testing system includes one or more sensors 638 placed in the drillhole 604 (e.g., within the substantially horizontal portion 610) and communicably coupled to a monitoring control system 646 through a cable 636 (e.g., electrical, optical, hydraulic, or otherwise). A downhole heater 626 is positioned in a substantially horizontal portion 610 of the wellbore 604. The substantially horizontal portion 610 is coupled to a radiussed portion 608, which in turn is coupled to a substantially vertical portion 606 of the drillhole 604, which is formed through subterranean layers 612, 614, 616 and into subterranean layer 618. In this example, the drillhole 604 includes casing portions 620 and 622 (e.g., surface or conductor casing and production casing) which are held in place by cement 630. Although illustrated as within drillhole 602 (e.g., inside of the casings), the sensors 638 may be placed outside of the casings, or even built into the casings before the casings are installed in the drillhole 602. Sensors 638 could also be placed outside the casing (e.g., casings 620 and/or 622), or outside the fluid control casing 634. The downhole heater 626, as shown, is placed within the drillhole 610, and in this example, within a storage area for hazardous waste, such as spent nuclear fuel or other radioactive material. The downhole heater 626 may be emplaced in the drillhole 610 through, e.g., a downhole conveyance (e.g., work string or wireline) or downhole tractor. As shown in this example, the downhole heater 626 is controlled by the cable 636 to provide a controllable amount of heat within the drillhole portion 610. The downhole heater 626 may be, for example, an electric resistance heater, a microwave or laser heater, or a downhole combustion heater. As shown, the sensors 638 may monitor temperature within the drillhole portion 610 (and other places, such as the subterranean layer 618) during operation of the downhole heater 626. Temperature data may be transmitted along the cable 636 to the monitoring control system 646. The monitoring control system 646, in turn, may record the data, determine trends in the data (e.g., rise of temperature and other data). In some aspects, there may be a single sensor 638. In alternative aspects, there may be multiple sensors 638. FIG. 6B shows another example implementation of thermal property testing system 600. In this example, sensors 638 are positioned within a secondary horizontal drillhole 640 that is formed separately from the substantially vertical portion 606. The secondary horizontal drillhole 640 may be an uncased drillhole, through which the cable 636 may extend between the monitoring control system 646 and the sensors 638. In this example, the secondary horizontal drillhole 640 is formed above the substantially horizontal portion 610 but within the storage layer 618. Thus, the sensors 638 may record the temperature data of the storage layer 618. In alternative aspects, the secondary horizontal drillhole 640 may be formed below the storage layer 618, above the storage layer in the impermeable layer 616, or in other layers. Further, although FIG. 6B shows the secondary horizontal drillhole 640 formed from the same substantially vertical portion 606 as the substantially horizontal portion 610, the secondary horizontal drillhole 640 may be formed from a separate vertical drillhole and radiussed drillhole. FIG. 6C shows another example implementation of the thermal property testing system 600. In this example, sensors 638 are positioned within a secondary vertical drillhole 642 that is formed separately from the drillhole 604. The secondary vertical drillhole 642 may be a cased or an uncased drillhole, through which the cable 636 may extend between the monitoring control system 646 and the sensors 638. In this example, the secondary vertical drillhole 642 bottoms out above the substantially horizontal portion 610 but within the storage layer 618. Thus, the sensors 638 may record the temperature data of the storage layer 618. In alternative aspects, the secondary vertical drillhole 640 may bottom out below the storage layer 618, above the storage layer in the impermeable layer 616, or in other layers. Further, although shown placed in the secondary vertical drillhole 642 at a level adjacent the storage layer 618, sensors 638 may be placed anywhere within the secondary vertical drillhole 642. Alternatively, the secondary vertical drillhole 642 may, in some aspects, be constructed prior to drillhole 602, thereby permitting monitoring by installed sensors 638 during construction of the drillhole 602. Also, the monitoring borehole 642 could be sealed to prevent the possibility that material that leaks into borehole 642 would have a path to the terranean surface 602. FIG. 6D shows another example implementation of the thermal property testing system 600. In this example, sensors 638 are positioned within a secondary directional drillhole 644 that is formed separately from the drillhole 604. The secondary directional drillhole 644 may be an uncased drillhole, through which the cable 636 may extend between the monitoring control system 646 and the sensors 638. In this example, the secondary directional drillhole 644 lands adjacent the substantially horizontal portion 610 and within the storage layer 618. Thus, the sensors 638 may record temperature data of the storage layer 618. In alternative aspects, the secondary directional drillhole 644 may land below the storage layer 618, above the storage layer in the impermeable layer 616, or in other layers. Further, although shown placed in the secondary directional drillhole 644 at a level adjacent the storage layer 618, sensors 638 may be placed anywhere within the secondary directional drillhole 644. The example implementations of the thermal property testing system 600 provide a descriptive basis of the components used to determine the suitability of a subterranean formation as a hazardous waste repository from a thermal perspective. An example analysis of the system 600 and operation thereof is provided. For example, the geologic disposal of spent nuclear fuel (SNF) and high-level radioactive waste (HLW) using corrosion-resistant canisters placed in deep, sub-horizontal, small-diameter holes drilled in suitable hydrostratigraphic units that safely and securely isolate the waste from the accessible environment are analyzed. A vertical access hole cased with steel pipe is drilled preferably at or near the site where nuclear waste is currently stored in surface facilities. At the kickoff point (slightly above the targeted repository depth), the drillhole gradually curves until it is nearly horizontal, with a slight upward tilt. The diameter of the drillhole varies from 9 to 30 inches (0.23 to 0.76 m) depending on the waste type and canister dimensions. Canisters containing the waste are lowered into the vertical access hole and pushed into the horizontal disposal section; they are emplaced end-to-end (potentially spaced apart by a separation distance that is one of the design parameters investigated in the current analysis) in a casing that lines the drillhole. The disposal section and vertical access hole are eventually sealed. As for any other geologic disposal concept (such as mined repositories or deep vertical borehole disposal), the performance of the engineered and natural barrier systems must be assessed for the specific repository design and the conditions expected during the regulatory compliance period. Focus here is on the thermal aspects of such an assessment and how they impact design decisions. Nuclear waste releases heat due to the decay of radionuclides, elevating temperatures within the canister. The heat then dissipates into the nearby repository engineered structures and the host formation. Predicting the temperature evolution within the disposal section of the drillhole and the surrounding host rock is necessary as it may alter the properties of the multi-barrier system and potentially lead to driving forces that affect the migration of radionuclides in the near field of the repository. Heat-driven degradation mechanisms may also make the retrievability of the waste canisters more difficult. The maximum temperature and temperature-time profile of components of the engineered barrier system are primary determinants of performance, and specifically, the corrosion performance of corrosion-resistant alloy canisters and steel casing. The maximum allowable temperature which needs to be determined by analyzing the acceptable impact on barrier functions, and which may eventually be set by the regulator is thus an important design variable for a geological repository, because it determines interim storage time as well as canister loading, canister spacing, and the minimum distance between disposal drillholes. All these factors affect the configuration and length—and thus cost—of the drillholes for a given amount of waste. The decay heat is time-dependent and determined by (a) the radionuclide inventory of the waste (itself a function of waste type and—in the case of SNF—initial enrichment and burnup percentage), and (b) the duration of post-reactor cooling. The initial temperature rise and subsequent cooling period are referred to as the heat pulse, which typically lasts a few decades to a few hundred years, until temperatures approach their ambient values prior to waste emplacement. The temperature evolution during the heat pulse has been extensively studied for various disposal systems using both laboratory and field experiments as well as numerical analyses. Large-scale, long-term heater tests for mined repositories in the saturated zone have been conducted in underground rock laboratories dedicated to nuclear waste research. Data collected during these experiments were analyzed using advanced simulators to predict and reproduce the observed thermal, hydrological, geomechanical, and geochemical evolution of various buffer materials and the surrounding formation. These studies reveal the importance of heat generation as it induces coupled thermal-hydrologic (TH) effects. Strong thermal perturbations also affect the geochemical conditions as well as the geomechanical properties and stress state of the repository components, with complex feedback mechanisms to thermal and hydrologic processes. Several heater tests were also conducted and numerically analyzed as part of the Yucca Mountain project. The unsaturated, highly fractured volcanic rocks at Yucca Mountain and the arrangement of waste packages in open disposal drifts lead to conditions that are significantly different from those encountered in repositories that store waste in backfilled deposition holes located in the saturated zone. Since the latter configuration is more akin to that encountered in deep horizontal drillhole disposal, thermal testing and modelling at Yucca Mountain are not discussed further here. Finally, thermal effects arising from the disposal of high-level radioactive waste in vertical boreholes drilled deep into crystalline basement rocks of the continental crust were investigated using semi-analytical and numerical models. Some of these analyses also examined fluid flow induced by thermal expansion of the rocks and the pore fluids, and considered very high temperature cases designed to partially melt and recrystallize the granitic host rock for additional borehole sealing. The concept of disposing nuclear waste in horizontal drillholes has some favorable attributes. In addition to operational advantages, there are a number of beneficial factors. For example, the reducing environment of the fully saturated host rock further prolongs the longevity of the canisters that are made of corrosion-resistant alloy. The linear arrangement of heat-generating nuclear waste in a drillhole makes thermal management considerably less challenging, as will be discussed below. Boiling of water at depth can be avoided, reducing the complexities of multi-phase flow processes. Moreover, relatively minor temperature changes lead to weaker thermal-mechanical stresses, helping to preserve the integrity of the engineered barriers and reducing the disturbance to the host formation. These attributes considerably reduce uncertainties that need to be propagated through performance assessment, and strengthen the technical basis for the safety case. The goal of the present analysis is to examine the impact of (a) design parameters and (b) uncertainty in host-rock thermal properties on temperatures in and around a horizontal disposal drillhole. Response surfaces are generated based on numerical simulations of heat dissipation in such a system. Moreover, sensitivity and data-worth analyses are performed to help design an in situ heater experiment that can reduce the uncertainty in subsequent model predictions. The analyses show that the temperature evolution in a horizontal drillhole containing heat-generating nuclear waste can be managed by adjusting a few design parameters. The thermal properties of the host formation have a dominant influence on the temperature evolution; these properties thus must be determined with sufficiently low estimation uncertainty, which can be accomplished by appropriate drillhole characterization methods and the collection of sensitive data during a short-term heater test. Conceptual and Numerical Model Development Waste emplacement geometry and the configuration of the engineered barriers within the horizontal drillhole need to be designed such that the maximum temperatures remain below certain limits that are set for each component of the engineered and natural repository system. The design calculations presented below are based on numerical simulations. The sophistication of the conceptual model to be developed and the level of detail with which features and processes must be represented are given by the specific purpose of the model, which in this case is to examine the thermal evolution around heat-generating waste canisters for a wide range of material properties and disposal configurations. Such scoping calculations typically have lower requirements regarding fidelity and accuracy than detailed studies in support of the safety case and performance assessment for a nuclear waste repository. Nevertheless, the simplifying assumptions made during model development must be transparent and justified in the context of the ultimate analysis to be done once the fully detailed configuration is known. The assumptions and model choices made for the current general design calculations are described in the following subsections. System Description The deep horizontal drillhole disposal concept targets a variety of waste forms, ranging from nuclear waste from the U.S. defense program to spent nuclear fuel (SNF) assemblies from different reactor types to vitrified high level waste (HLW). While design calculations must accommodate the specifics of each waste type (especially canister geometry and heat output characteristics), the method described here is general and can thus be illustrated using a single waste type. The disposal of capsules that contain primarily short-lived cesium-137 (137Cs) and strontium-90 (90Sr) extracted in the form of cesium chloride (CsCl) and strontium fluoride (SrF2) during the chemical processing of defense fuel are considered here. The capsules, fabricated from 316L stainless steel, are typically 20.775 inches (0.528 m) long, 2.6 inches (0.066 m) in diameter and weigh less than 10 kg. Currently, there are 1,335 cesium and 601 strontium capsules stored underwater at the Hanford Waste Encapsulation and Storage Facility; the present disclosure describes an analysis of the permanent disposal of these capsules in deep horizontal drillholes. The analysis examines a proposal to insert one or several such capsules into a canister made of a corrosion-resistant alloy (e.g., Alloy 625); the canister will have an outer diameter of approximately 4.5 inches (0.114 m). The space between the capsule and canister is filled with an appropriate backfill material (such as quartz sand) for mechanical stability and to provide sufficient conductivity for heat dissipation. The canister is placed in a liner or casing, which has an inner diameter of 5.5 inches (0.140 m). The space between the canister and the casing (and axially between individual canisters) may be filled with drilling fluid, a slurry or a suitable buffer material (such as bentonite). The casing is likely to be cemented into an 8.5 inch (0.216 m) diameter, horizontal drillhole, which is the disposal section of the repository. The disposal section is completed in a host rock that not only exhibits favorable hydrogeological, geochemical and geomechanical properties, but is also protected by low-permeable overlaying strata (such as shales, claystones and mudstones) and has been isolated from surface waters and aquifers for very long times, as demonstrated, for example. by isotopic age determination of the resident brines. While drilling may damage the rock around the hole, the thickness of such a skin or excavation disturbed zone is expected to be small with minor impacts on the rock's thermal properties. It is further assumed that the various components are perfect cylindrical shells that are centered on the drillhole axis. The impact of an off-centered configuration on the temperature distribution has been examined and was determined to be insignificant for the purpose of these scoping calculations. FIG. 5A shows a schematic 500 of the various components in a vertical cross section along and perpendicular to the drillhole axis (i.e., an engineered barrier system). It is assumed that waste capsule spacing is constant, that the heat source is distributed uniformly among and within waste capsules, and that gravity effects can be ignored in and around the horizontal drillhole. Under these conditions, a two-dimensional, radial model can be developed with symmetry planes perpendicular to the drillhole axis at the center of a capsule and in the midpoint between two capsules (e.g., as shown in FIG. 5B). The capsule spacing is an adjustable design parameter. The outer model radius is large enough to avoid boundary effects. The average power output of a Sr and Cs capsule may be 193.2 W and 143.6 W, respectively. The heat output from Sr capsules is substantially more variable than that of Cs capsules, with standard deviations of 101.0 W and 14.1 W, and a maximum output of 504.6 W and 195.4 W, respectively. Nevertheless, the resulting temperature evolution for conduction-dominated heat transfer depends approximately linearly on the heat output, i.e., results calculated for a reference heat generation rate of 100 W per capsule can readily be scaled to capsules with a different initial radioactivity and a different cooling period. As heat generation is directly related to radioactive decay, the time-dependent rate follows the exponential decay curve of the respective isotope, i.e.,QH(t)=QH0·e−λκt, (1) where QH0 is the initial heat generation rate, λκ is the decay constant of isotope κ, which is related to the half-life T1/2 by λκ=(ln 2)/T1/2, and t is time. The half-lives of cesium and strontium are, respectively, 30.17 and 28.79 years. Heat generation will be assigned exclusively to the capsule itself, i.e., no heating of the other components of the drillhole or the host rock due to radiation is considered. This is justified by the fact that 90Sr (and its decay products) undergo a beta-decay, whereby the emitted electron is absorbed within the capsule. For 137Cs, about 22% of the decay energy is released by short-range electrons; the remaining 78% of the energy released by gamma rays is effectively attenuated in CsCl and Alloy 625, with only a very small fraction being deposited in the casing, and virtually none in the host rock. Physical Processes The dissipation of thermal energy in engineered and natural materials is mainly driven by heat conduction, and to a much smaller degree through convection by moving fluids (liquids or gases) and radiative heat transport. Latent heat effects during phase transitions and contributions from changes in the gravity potential also impact the temperature distribution. Many thermal and hydrological processes are strongly coupled, specifically if phase changes occur. Mechanical effects are triggered by thermal stresses, and the geochemistry of pore fluids and the mineral composition of the rocks are affected by temperature. While feedback mechanisms that affect temperature due to chemical reactions and stress changes do exist, they are typically much weaker than coupled thermal-hydrological effects. For the deep horizontal drillhole system of interest, conduction is the dominant heat transfer mechanism. This is undoubtedly the case for the hydraulically impermeable, but thermally highly conductive metals of the engineered barrier system, but also for the porous backfill materials and the host rock, which—by design—are of low permeability and porosity and are located in a low hydraulic gradient environment. Radiative heat transfer is negligible for the expected temperatures and in the absence of large open space, or is included in the experimentally determined thermal conductivity value. The gravitational potential is irrelevant in the horizontal disposal section of the drillhole, and of no significance in the vertical section, specifically in the absence of flow. Latent heat effects are not expected in a deep drillhole repository, where ambient fluid pressures are close to hydrostatic and thus likely above the saturated vapor pressure, preventing boiling even for relatively high temperatures. FIG. 7 includes graph 700 that shows the boiling temperature (in ° C.) as a function of pressure (in bar), which is correlated to depth (in meters, m) assuming a hydrostatic pressure profile. For example, for a repository depth of 1 km, temperatures below 300° C. will not lead to boiling. Finally, latent heat effects due to melting and recrystallization of the host rock are not relevant for the temperature range considered in this study. To avoid the complexity of coupled thermal-hydrological-geochemical processes, which are exacerbated if a steam phase evolves, it is recommended that the maximum allowable temperature in the repository be below the boiling temperature curve shown in FIG. 7. Note that a lower maximum temperature criterion may be advisable for other reasons, such as expansion and associated thermal stresses or undesirable mineralogical alterations of the buffer material or host rock. Avoiding steam also improves the corrosion performance of engineered barrier components, particularly that of canister materials. Having heat conduction identified as the dominant heat transfer mechanism, it is helpful that the material properties appearing as coefficients in the heat conduction equation are known with an acceptable level of uncertainty, as they are likely the most influential parameters for temperature predictions. Heat conduction is a diffusive process governed by a parameter group referred to as thermal diffusivity, (K/ρc), where K is the thermal conductivity, p is the density, and c is the specific heat. These are all effective parameters for the bulk material, which consists of multiple components and phases. While density and specific heat can be calculated reasonably well as the volume average of each of the material's components, the thermal conductivity of a composite porous medium depends to a large degree on the connectivity between its more conductive and more resistive components. The arrangement and contact of particles and the connectivity of fluids in the pore space of a backfill material or geologic formation is complex and prevents an easy calculation of thermal conductivity from the properties of its components, resulting in a considerable range of values even for similar rock types. Nevertheless, effective thermal conductivities can be experimentally determined with good accuracy. The parameters are also temperature dependent, with generally decreasing thermal conductivity and increasing heat capacity as temperature increases, partially compensating each other's influence on thermal diffusivity and thus overall effect on temperature. For the design calculations discussed below, thermal conductivity is isotropic and constant, and good thermal contact is assumed at material interfaces. Note that even a small gap between two materials (filled by a liquid or gas) has the effect of an insulator, which can either be modeled explicitly or accounted for by adjusting the effective thermal conductivity. The thermal properties of water are well known and only weakly dependent on salinity. Should a special drilling fluid, mud, or slurry be used, their thermal (and hydraulic) properties need to be measured and included in the simulations, specifically if no porous backfill material is used or convection within the drillhole or in the formation becomes significant. While fluid flow and associated heat conduction is expected to be a minor contributor to heat transfer, it will be accounted for in the simulations. However, it is assumed that the sole driving force for fluid flow is that triggered by the thermal expansion of the fluids and pore space. Note that thermal pore expansivity partly compensates for fluid expansion, and the resulting pressure change is further mediated by elastic deformation of the pores, which in the model is assumed to depend on the pore pressure rather than effective stress. As discussed above, a time-dependent heat source is specified, which follows the decay curve of the radionuclides in the waste. The heat source is assumed to be uniformly distributed within the volume representing the waste capsule. While the waste is not necessarily uniform, the high thermal conductivity of the capsule is likely to homogenize the temperatures and heat release to the engineered barrier system. It should be noted that while the heat-driven coupled processed outlined above are inherently complex, the horizontal drillhole concept, which promotes heat dissipation, reduces the thermal stresses and thus the challenge to predict their impacts on repository performance. Mathematical and Numerical Model A mathematical model of the physical processes discussed in the previous subsection is implemented in the TOUGH2 numerical simulator, which calculates non-isothermal, multiphase, multicomponent fluid flow in fractured porous media. TOUGH2 solves mass- and energy-balance equations formulated in a general, integral form. A simplified version (assuming single-phase liquid conditions with water being the only component) of the time-dependent energy balance equation can be written for an arbitrary subdomain Vn, which is bounded by the closed surface Γn as: d dt ∫ V n [ ( 1 - ϕ ) ρ s c s T + ϕ ρ w u w ] dV n = ∫ Γ n [ - K ∇ T + hF ] · n d Γ n + ∫ V n qdV n . ( 2 ) The energy accumulation during time interval dt on the left-hand side of Equation (2) contains contributions from the solid and liquid phases, where ϕ is porosity, ρs and ρw are, respectively, the grain and water densities, T is temperature, cs is the solid specific heat, and uw is the specific internal energy of liquid water. The first term on the right-hand side is the heat flux across the volume boundary, which includes conductive and convective components. Here, K is the effective thermal conductivity discussed above, h is the specific enthalpy of liquid water, and n is a normal vector on the surface element dΓn, pointing inward into Vn. The liquid mass flux F is given by Darcy's law, F = ρ w u = - k ρ w μ w ( ∇ P - ρ w g ) , ( 3 ) where u is the Darcy velocity, k is absolute permeability, μw is the dynamic viscosity of liquid water, P is fluid pressure, and g is the vector of gravitational acceleration. All thermophysical fluid properties are a function of pressure and temperature, accurately calculated based on the IAPWS-95 formulation. Finally, the specific source term, q in Equation (2), is proportional to the time-dependent decay heat curve of Equation (1). TOUGH2 uses a finite volume formulation, where space discretization is made directly from the integral form of the governing conservation equations, without converting them into partial differential equations. Time is discretized fully implicitly as a first-order backward finite difference. The resulting coupled, nonlinear algebraic equations (with pressure and temperature in each grid block as the unknown primary variables), are solved simultaneously using Newton-Raphson iterations. The elements of the Jacobian matrix are calculated numerically. At each iteration, the set of linear residual equations is inverted using a preconditioned conjugate gradient solver. All analyses discussed in the following subsections are performed within the iTOUGH2 simulation-optimization framework, which performs forward simulations, solves the inverse problem, and conducts sensitivity, uncertainty, and data-worth analyses. Model Setup The coupled fluid flow and heat transfer processes is simulated within the two-dimensional, radial model domain shown in FIG. 5A. The model domain is discretized into cylindrical shell elements, each with an axial length of 0.5 inches (0.0127 m). The total length of the model domain in axial direction is adjustable between 12.0 inches (0.3048 m) and 84.0 inches (2.1336 m) to accommodate different separation distances between waste capsules. In radial direction, the first 100 shells have a constant thickness of 0.125 inches (0.003175 m) up to a radius of 12.5 inches (0.3175 m), after which the shell thicknesses increase logarithmically until the outer model domain radius of 3600 inches (91.44 m) is reached. The model has a total of 22,008 elements and 43,717 connections between them. Three equations (for the three primary variables pressure, saturation, and temperature) are solved at each point in space. No-flow boundaries are specified at the symmetry planes. At the outer model domain radius, a Dirichlet boundary condition is specified with a pressure of 100 bar and a temperature of 40° C., representative of a horizontal waste disposal section at a depth of 1 km. The same values are used as initial conditions throughout the model domain. As heat transfer is only mildly impacted by the absolute pressure and temperature values, results are reported as temperature changes with respect to the initial temperature of 40° C. A transient simulation for a duration of 30 years is performed with automatic time-step adjustment based on the convergence behavior of the Newton-Raphson iterations. The temperature change is extracted at the center of the waste capsule (X=0.0; as shown in FIG. 5B) and for select radial distances, each representing a component of the engineered barrier system. Response surfaces are created for the maximum temperature change, which is extracted by fitting a polynomial through the three highest points of the discrete time series, and setting its derivative to zero. The key material properties are summarized in Table 13 as shown in FIG. 14M. These are reference material properties that will be adjusted over a considerable range to account for different selections of backfill materials and potential host rocks. In Table 13, the “Range” refers to lower and upper bounds of parameters, defining the range examined by global sensitivity analysis and response surfaces; n/a: not applicable, i.e., parameter is fixed. Also, “backfill” refers to backfill of canister, casing, and annulus; each may consist of a different material, e.g., quartz sand, bentonite, drilling mud, or cement; properties to be selected based on chosen backfill material. Finally, “host rock” refers to various host rocks that can be considered, including sedimentary, magmatic, and metamorphic rocks; properties to be selected based on site-specific host rock. Local and Global Sensitivity Analyses In addition to calculating the temperature evaluation for the reference parameter set of Table 13 and some discrete variants, this analysis also includes extensive local and global sensitivity analyses and a data-worth analysis. Local sensitivity coefficients are needed to calculate composite sensitivity measures, and to calculate estimation and prediction uncertainties. The local sensitivity coefficients are the partial derivatives of an output variable z1 with respect to an input parameter pj, evaluated at the reference parameter set p*: S ij = ∂ z i ∂ p j p * . ( 4 ) Because Sij has units of the model output over the units of the parameter, these sensitivity coefficients cannot be readily compared to each other if inputs and outputs of different types are involved. A scaled, dimensionless local sensitivity coefficient is introduced: S _ ij = S ij · σ p j σ z i , ( 5 ) where σp is the input- or parameter-scaling factor, and σz is the output- or observation-scaling factor. In the context of a sensitivity analysis, σp is the expected parameter variation, and σz denotes the threshold at which a change in the model prediction is considered significant. In the context of a data-worth analysis (as described later), σp is interpreted as the acceptable parameter uncertainty, and σz is the expected mean residual obtained after the inversion, or the acceptable prediction uncertainty of the target predictions. A local sensitivity analysis indicates the relative influence of each of the unknown, uncertain, or variable parameters on the target predictions. which in in this example are the maximum temperatures at specific points within the repository system. However, if the model is nonlinear, the sensitivity coefficients depend on the parameter set, which varies considerably during the early design stages of a project. A global sensitivity analysis method is employed to identify the overall most influential parameters. As any global method, the Morris one-at-a-time (MOAT) elementary effects method examines many parameter combinations within the range of acceptable values. The MOAT method subdivides each axis of the parameter hypercube into r−1 intervals for a total of rn grid points, where n is the number of parameters. A perturbation A is then calculated for each parameter j, Δ j = r 2 ( r - 1 ) · ( p j , ma x - p j , m i n ) . ( 6 ) A random grid point in the parameter space is selected, the model is run, and the performance measure z is evaluated. Then—one at a time and in random order—each parameter pj is perturbed by Δj, the model is run to recalculate z, and the corresponding impact on the output (referred to as elementary effect, EEj) is computed as EE j = z ( p 1 , p 2 , … , p j + Δ j , … , p n ) - z ( p 1 , p 2 , … , p n ) Δ j . ( 7 ) The procedure is repeated for multiple, randomly selected starting points of a path in the parameter space that consists of n+1 simulation runs for the evaluation of the elementary effect in the vicinity of this point. After completion of a number of such paths, the mean and standard deviation of the absolute elementary effects are calculated (denoted by EE and σEE, respectively). The mean assesses the overall influence of the respective parameter on the output; the standard deviation indicates whether the effects are linear and additive or nonlinear, or whether interactions among the parameters are involved. Response surfaces for pairs of the most important design factors identified by the global sensitivity analysis are created. Data-Worth Analysis Finally, a data-worth analysis is performed to help design an experiment in which the key parameters affecting maximum temperatures can be determined with sufficient accuracy. A data-worth analysis identifies and ranks the contribution that each (potential or existing) data point makes to the solution of an inverse problem (e.g., for the estimation of thermal properties) and a subsequent predictive simulation (e.g., of maximum repository temperatures). It is based on sensitivity coefficients, a linear estimation error analysis (to obtain the uncertainty in the estimated parameters given the available data and their uncertainties), and a linear uncertainty propagation analysis (to obtain the prediction uncertainty given uncertainty in the estimated parameters). This analysis denotes n as the number of uncertain parameters that will be estimated based on in discrete measurements, i.e., n is the length of the parameter vector p, and m is the length of the observation vector z. Note that m changes during a data-worth analysis, as individual data points (or entire data sets) are either removed from the reference data set or added as potential observations. The covariance matrix of the estimated parameters, Cpp, is calculated asCpp=(JTCzz−1J)−1. (8) Here, J is the m×n Jacobian matrix, holding the sensitivity coefficients Sij; Czz is the m×m observation covariance matrix, containing the variances σz2 on its diagonal. A linear uncertainty propagation analysis is performed to yield the covariance matrix of the model predictions:C{circumflex over (z)}{circumflex over (z)}=ĴCppĴT. (9) Here, the Jacobian matrix Ĵ holds sensitivity coefficients of the prediction of interest with respect to the parameters p, whose uncertainty is described by Cpp calculated using Equation (8). In a data-worth analysis, the estimation and prediction uncertainty matrices, Cpp and C{circumflex over (z)}{circumflex over (z)}. respectively, are re-evaluated for different calibration data sets. The data worth, ω±k, is then defined as the relative increase in the prediction uncertainty (measured by the trace of C{circumflex over (z)}{circumflex over (z)}) caused by the removal of data, or the relative decrease in the prediction uncertainty gained by the addition of data. Starting with reference data, the uncertainty analyses of Equations (8) and (9) determine whether the estimation or prediction uncertainties are sufficiently low, i.e., acceptable for the intended purpose of the model. If so, the data-worth analysis indicates which data could be removed to arrive at a less complex and less expensive design with minimal impact on the quality of the estimated parameters and without substantially increasing prediction uncertainty. If uncertainties are unacceptably high, the data-worth analysis suggests which potential data could be added to the reference data set to effectively reduce the estimation and prediction uncertainty. Temperature Evolution The temperature evolution at various radial distances from the center of the capsule (representing different components of the system) is shown in the graphs 800(a), 805(a), and 810(a) in FIG. 8 for three different capsule spacings; the simulated temperature distribution three years after the emplacement of heat-generating waste capsules is shown in the right column. Generally, FIG. 8 shows graphs that illustrate evolution of temperature change ((a) graphs) and temperature distribution after 3 years ((b) graphs) for an initial heat release of 100 W per waste capsule with capsule spacings of (graphs 800) 2 ft (0.6096 m); (graphs 805) 4 ft (1.219 m); and (graphs 810) 6 ft (1.829 m). In this example, temperatures are higher if waste capsules are emplaced end-to-end with very little separation distance, and maximum temperature changes are reduced if the capsules are spaced farther apart. However, the cooling effect becomes smaller for larger separation distances, as the heat dissipation regime transitions from cylindrical (FIG. 8, graphs 800(a) and 800(b)) to approximately spherical (FIGS. 8, 810(a) and 810(b)). As a result, only irrelevant benefits regarding maximum temperature can be gained by spacing capsules by more than about 2 m. For an initial heat output of 100 W per capsule, a dense capsule emplacement configuration with a spacing of 2 ft (0.6096 m) leads to maximum temperature increases of about 73° C. for the capsule itself, and about 60° C. at the drillhole wall. Recall that these temperature increases are proportional to the heat output. To avoid boiling in the backfill material between the canister and the casing, the initial heat output must be limited to about 360 W within a drillhole that is at a depth of 1 km at an ambient temperature of 40° C., as inferred from the boiling curve of FIG. 7 and the maximum temperature increase shown in FIG. 8 (graphs 800(a), 805(a), and 810(a)). Note that none of the cesium capsules and only a small fraction of the strontium capsules generate heat in excess of 360 W. These capsules can be stored at the surface for a longer period, or placed in the horizontal drillhole with an appropriately increased separation distance to their neighbors. In general, a slightly broadened emplacement pattern should be used to account for uncertainties in heat output, in ambient temperature and pressure, and in the thermal properties of the various materials, in particular the host rock, whose heat conductivity is most uncertain, most variable, and at the same time most influential, as demonstrated in the following sensitivity analyses. Sensitivity Analyses This analysis includes local and global sensitivity analyses to obtain insights into the system behavior and to identify influential and non-influential parameters. This analysis also mapped out maximum temperature changes over a wide range of the most influential parameters, creating response surfaces as a convenient design tool. For the drillhole disposal concept, heat dissipation is almost exclusively in the radial direction, passing through different materials that are arranged in concentric, cylindrical shells. Because of this configuration, the components are encountered in series, and, consequently, heat flow is controlled by the components of relatively low thermal diffusivity. The metallic elements with a high thermal conductivity and small shell thickness (i.e., the canister and casing) are expected to have an insignificant impact on the spatial and temporal temperature distribution. This is confirmed by a local sensitivity analysis, which is performed for a capsule spacing of 48 inches (1.22 m). A composite sensitivity measure—defined as the sum of the absolute values of the scaled sensitivity coefficients (Equation 5) for each column and row of the sensitivity matrix—is calculated for each thermal parameter (columns of S) and the maximum temperature at select locations (rows of S). This analysis also evaluated the impact of a 10% change in the heat output on the maximum temperature. Table 14 (as shown in FIG. 14N) indicates that the heat conductivity of the host rock is the most influential parameter, followed by the strength of the heat source. In Table 14, the parameter scaling factor for thermal conductivities and heat output are σK=1.0 and σQ=10.0, respectively; subscripts back1 and back2 refer to the backfill material between the capsule and the canister, and between the canister and the casing, respectively; the sensitivity coefficient for heat capacity and material densities are significantly smaller and are thus not tabulated. Also, observations (“Obs.”) of interest are the maximum temperatures encountered during the repository lifetime; Trock×m is the maximum temperature in the rock×m from the drillhole wall; as only temperatures are considered, the observation scaling factor is irrelevant—it is set to σT=1.0. As expected, a 10% change in rock thermal conductivity has about the same impact as a 10% change in heat output. The conductivity of the canister backfill material has some effect on the capsule temperature, but not on the temperatures outside the canister. The thermal properties of the capsule, canister and casing are essentially irrelevant if fabricated of highly conductive material. As heat dissipates in a radially outward direction, the composite sensitivity measures for the observations generally decline with radial distance from the drillhole axis. The thermal conductivity of the host rock has its maximum influence at the drillhole wall, where the observation is collocated within the domain to which the parameter refers. These general insights are quite robust with respect to the somewhat subjective choice of the parameter scaling factor, i.e., even if the uncertainties in thermal conductivity vary between materials, this does not substantially affect the qualitative statements made above. The local sensitivity analysis is contingent on the chosen reference parameter set (e.g., Table 13). Therefore, a Morris global sensitivity analysis is performed to examine the validity of the simple local sensitivity analysis and to examine nonlinearity and interaction effects. The parameters involved in this global sensitivity analysis and their upper and lower bounds (defining the parameter hypercube) are listed in Table 13 above. The 12-dimensional parameter hypercube is subdivided into r−1=5 intervals and examined along np=40 paths, as described in Section 2.6. FIG. 9 shows a graph 900 that is cross-plot between the mean and standard deviation of the absolute elementary effect (EE; Equation 7) of the Morris global sensitivity analysis. The dashed line represents |EE|=2·σEE=σEE/√{square root over (np)} is the standard error of the mean of the elementary effect. All the parameters are below the dashed line, indicating that their non-zero impacts are statistically significant. By far the most influential parameters are the heat output (circle), the host rock's thermal conductivity (diamond), and the capsule spacing (X). With the exception of the host rock's heat capacity (triangle), thermal conductivities (diamonds) are considerably more influential than the heat capacities (triangles) for all other components. Properties that are closer to the drillhole axis (dark colors) are less influential than those further out (light colors), with the exception of the capsule's heat capacity (triangle), which influences the maximum temperature of the waste capsule. The parameters also have considerable non-zero standard deviations, indicating that they exhibit interaction effects. This is expected as the temperature is essentially determined by a weighted harmonic average of all thermal diffusivities. The global sensitivity analysis corroborates the parameter ranking previously obtained by the local, composite sensitivity measures. While the capsule spacing is an adjustable design parameter, and the heat output of the waste capsule is well known, the host rock's heat conductivity is the main parameter that needs to be accurately determined. Any unacceptably high estimation uncertainty in this influential parameter will be propagated to high uncertainties in the predicted maximum repository temperatures. This will be addressed by the data-worth analysis, which helps reduce the estimation uncertainties of the parameters that are most influential on the model prediction of interest. Response Surfaces FIG. 10 (graphs (a) through (e)) shows two-dimensional response surfaces of the maximum temperature increase at select radial locations as a function of thermal conductivity of the host rock and capsule spacing. More specifically, as shown, response surfaces of maximum temperature increase as a function of host-rock thermal conductivity and capsule spacing for a 100 W initial heat output for the following repository components: (graph (a)) Waste capsule; (graph (b)) Canister; (graph (c)) Casing; (graph (d)) Drillhole wall; and (graph (e)) Host rock 1 m from the drillhole wall. Host-rock conductivity is chosen because it is the most influential property that may also vary over a considerable range depending on the rock type and spatial heterogeneity. Capsule spacing is selected as the main design parameter that can be adjusted for effective temperature control. To obtain the actual temperature for a given combination of host-rock thermal conductivity and capsule spacing, the value from the response surface must be multiplied by the heat output factor fH=QH0/100 W, and the result added to the ambient temperature at the depths of the disposal zone. Parameter combinations in the white corners of the response surfaces would lead to boiling if waste capsules with a heat output of 200 W were disposed in a horizontal drillhole at a depth of 1 km. (Recall that thermal criteria other than the boiling temperature may be relevant.) These response surfaces can be directly used to determine an appropriate capsule spacing given the relevant maximum temperature criterion for each of the repository components and the in-situ thermal conductivity of the host rock. FIG. 11 shows the impact of backfill thermal conductivities on the maximum capsule and drillhole wall temperatures. More specifically, FIG. 11 shows response surfaces of maximum temperature increase as a function of casing-backfill and annulus thermal conductivities for a 100 W initial heat output for the following repository components: (graph (a)) Waste capsule; and (graph (b)) Drillhole wall. The backfill between the canister and the casing may be a drilling mud, a slurry, sand, bentonite, cement, or another suitable material; the annulus backfill (between the casing and the drillhole wall) is, e.g., either drilling mud or cement. The lower bound of the thermal conductivity range examined in these response surfaces represents a slurry or accounts for the presence of a fluid-filled gap. Only if backfill conductivities approach these lower bounds does temperature increase slightly relative to the reference case. Note that the temperature ranges in the two response surfaces of FIG. 11 are much smaller compared to those shown in FIG. 10, confirming the lower influence of these two parameters. For thermal conductivities above about 1.5 W m−1 K−1, the sensitivity of the capsule temperature becomes small and essentially disappears for the drillhole wall temperature. Note that increasing thermal conductivities of the backfill materials leads to faster heat dissipation away from the capsule, thus cooling it down, while speeding up the outward propagation of the heat pulse, thus leading to increased maximum temperatures at the drillhole wall. Data-Worth Analysis The purpose of the data-worth analysis is to design an in situ heater test (e.g., as shown in FIGS. 6A-6D) that determines influential thermal properties with sufficient accuracy so that the maximum temperature throughout the drillhole and the adjacent host rock can be predicted with acceptable uncertainty and suitability of the host rock as a hazardous waste repository. The basic idea is to insert a capsule containing an electrical heater into the disposal section of the drillhole, backfill the test section according to the design specifications, then start releasing heat at a controlled wattage. Next, the temperature evolution data are recorded by a distributed temperature sensor (DTS). A DTS system uses a laser backscattering technique to measure temperatures continuously along an optical sensor cable, resulting in data with high spatial and temporal resolution. The temperature data are inverted to determine key properties, specifically the host rock thermal conductivity. Once the thermal properties are identified, the response surfaces of FIG. 10 can be used to determine the appropriate spacing of the actual waste capsules. The data-worth analysis provides quantitative measures that help determine the number and location of the temperature sensors and the duration of the heating experiment. Two models—referred to as the calibration and prediction models—need to be developed and run sequentially. The calibration model simulates the heater test data, whereas the prediction model simulates the maximum temperatures induced by the disposal of heat-generating nuclear waste. The calibration model covers the short duration of the heater test; the prediction model covers the much longer duration of the thermal period. For the reference test, a single heater of the size of an actual waste capsule, heating at a constant output of 200 W for up to 30 days is analyzed. This analysis measures temperature at a DTS sensor attached to the casing. These temperature data can be matched by the calibration model with an average residual of 1° C. This standard deviation is chosen larger than the measurement accuracy of DTS of about 0.1° C. to account for model simplifications. Should such measurements be insufficient, a temperature measurement at the surface of the heater, as well as potential DTS sensors attached to the drillhole wall can be considered. Moreover, this analysis includes some prior information about the thermal conductivities and heat capacities, reflecting independent property measurements on engineered materials (metals and backfill) as well as retrieved drill core samples or cutting fragments of the host rock. However, there is not reliance on this information to be very accurate; it is mainly used to stabilize the solution of the notional inverse problem (Equation 8). Uncertainty in the heater output is also considered by estimating it during the inversion, with a standard deviation of 20° C. assigned to its prior information value. FIG. 12 shows a graph 1200 that shows the temperature increase and the data-worth metric as a function of heating time. The temperature increase exceeds 20° C. after less than 2 days, and reaches 40° C. after 30 days of heating, with only slightly higher temperatures at the heater compared to the drillhole wall. The dimensionless data-worth metric measures the relative reduction in uncertainty of the predicted maximum repository temperatures as data are added. Data worth increases sharply during the initial days of heating. At later times the data worth, which accounts for parameter correlations and redundancies of closely spaced data points, approaches a constant value. This indicates that the information content of the DTS data initially grows quickly, but is reduced to a constant rate as the heater test is prolonged. Accurately measuring temperatures at early times is most beneficial, with later data providing additional, albeit less important information. The test can be terminated once the acceptable prediction uncertainty is achieved. In some aspects, no significant benefits can be gained by moving the DTS from the casing to the drillhole wall or towards the heater. Installation of the DTS fiber-optic cable by clamping it to the outside of the casing is not only most practical, but also desirable as it avoids interference of the cable with waste emplacement operations. Thus in some aspects, focus is only on the DTS data collected along the casing, discarding the use of additional sensors. The results of the notional inversion are first discussed, which is simply the evaluation of Equation (8) with the assumption that the match to the (still non-existent) data is consistent with the prior observation covariance matrix, Czz. The resulting covariance matrix of the estimated parameters, Cpp, reveals that performing the heater test mainly helps determine the thermal conductivity of the host rock. Table 15 (as shown in FIG. 14O) shows the estimation and prediction uncertainties for different testing durations. Without conducting an in situ heater test, the uncertainty in the predicted maximum temperature is high. For example, on the 95% confidence level, the prediction of the maximum temperature change at the drillhole wall would read ΔTwall=34±30° C. Even granted that the normality and linearity assumptions underlying the uncertainty analysis of Equation (9) are violated, this large uncertainty renders the prediction essentially not very useful. In Table 15, the “Prediction” refers to uncertainty of predicted maximum component temperature during repository life time. Performing a one-day long heater test, the estimation uncertainty of the most influential parameter, the host rock thermal conductivity, is reduced from its prior value of 1.0 to less than 0.3 W m−1 K−1. An even lower uncertainty can be achieved if the heater output is controlled accurately, a result of the fact that these two parameters are strongly correlated. Combined with uncertainty reductions in the other parameters that are concurrently estimated with Krock leads to considerably improved temperature predictions. Specifically, the maximum temperature at the drillhole wall now reads ΔTwall=34±6° C. Whether such a prediction uncertainty is acceptable depends on its use for repository design and performance assessment. The uncertainties can be further reduced by prolonging the heater test, albeit with diminishing added value for the later times. If testing lasts for 10 days or longer, the uncertainty of the model-predicted maximum temperature experienced by the host rock at the drillhole wall is less than 1° C. The temperature evolution in the disposal section of a horizontal drillhole was simulated for a wide range of thermal properties of engineered and natural materials. For example, this analysis specifically examined the maximum temperatures encountered during the thermal pulse period at selected locations within the drillhole and the near-field host rock. The sensitivity analyses indicate that the key factors affecting maximum temperatures are the thermal conductivity of the host rock, the spacing between waste capsules, and the wattage of the heat-generating waste. The global sensitivity analysis demonstrate that the identification of the most influential parameters is robust even if the reference property values are uncertain or variable over a wide range. Of these three influential parameters, only the heat conductivity of the host rock cannot be adjusted and needs to be determined for in situ conditions at the selected disposal site. Should its value turn out to be too low (leading to excessive temperatures in the repository), a different layer needs to be chosen, or an altogether different site explored. The heat output of the waste can be partly controlled by extending the post-reactor cooling period. Finally, the spacing of waste within the drillhole is the main, readily adjustable design parameter used for thermal management of the repository. The thermal properties of the backfill material have a much smaller impact on the maximum temperatures. Temperatures increase somewhat if backfill conductivities approach very small values. Such small values may only occur if a relatively wide, fluid-filled gap develops over the entire circumference of the canister, casing, or drillhole wall, acting as an insulator. Despite this possibility, it can be recommended that a suitable backfill material should be selected mainly based on its ability to fulfill a specific barrier function rather than because of its thermal properties. This analysis also examined the possibility to perform an in situ heater test to determine the thermal performance of the as-built repository system. From an operational point of view, the proposed heater test is well integrated into the site development and characterization process. After completion of the drillhole, the heater (which has the same dimensions as the waste capsule) is pushed to the end of the disposal section, testing the integrity of the drillhole and the absence of obstructions, confirming that emplacement of actual waste capsules is possible. The short testing section is then instrumented and backfilled according to the design specifications, testing the corresponding procedures. Heating and data collection begins. While the heater experiment is running, the entire drillhole is available for logging, characterization and disposal preparation. Temperature data are analyzed in real time by performing inversions using a calibration model that will be set up in advance. Once sufficient data are collected such that the site-specific thermal properties are determined with the desired level of accuracy, the heater test can be terminated, and waste emplacement may commence. If there are indications of considerable heterogeneities along the drillhole, the heater test may be repeated at selected locations. Finally, the DTS sensors can be used to observe the thermal evolution along the disposal section as part of performance confirmation monitoring. The design of the heater experiment (including the way power is supplied) should be further optimized and tested in a pilot drillhole. If the main goal is only to determine the host rock's thermal conductivity under in situ conditions (i.e., without testing the thermal performance of the as-built engineered barrier system), a less intrusive approach using a combination of DTS and a borehole-length electrical resistance heater (a system referred to as distributed thermal perturbation sensor) could be considered. The maximum temperature expected within a horizontal drillhole and the surrounding host formation is an important factor that mainly affects our ability to robustly predict repository performance. These temperatures need to be simulated with acceptably low prediction uncertainty in order to provide a defensible basis for the demonstration that they are below regulatory thermal limits. While such thermal limits are not discussed or proposed in this study, it is recommended that repository temperatures remain below the boiling temperature under in situ conditions at all times to avoid the significant complexities arising from phase changes and the related coupled processes. In general, the linear arrangement of waste capsules or spent nuclear fuel assemblies in a long horizontal drillhole leads to relatively large specific surface areas available for heat dissipation. Thermal management for a drillhole repository is thus less challenging compared to that in other repository concepts, where relatively large volumes of heat-generating waste are densely packed in mined caverns or large-diameter deposition holes. For a moderate capsule spacing of about 2 m, thermal interference is very small; denser loading of the disposal section of the drillhole can be justified using the design approach outlined here. The simulations show that waste spacing is a very effective design parameter to manage temperatures in the disposal section of the drillhole. The design calculations presented here were done for the disposal of relatively small, but thermally hot cesium and strontium capsules. The maximum temperatures for a cesium capsule, which typically generates about 100 W at the time it is emplaced in the drillhole, are less than 100° C. above the ambient temperature, i.e., far below the in situ boiling temperature. Only a small fraction of the strontium capsules have high enough heat output to considerably raise the temperatures, but these cases can be readily managed by increasing the capsule spacing. The thermal maximum is reached after less than 10 years, i.e., a time much shorter than that predicted for a large, mined repository. Note that deep nuclear waste isolation in horizontal drillholes is considered feasible also for other waste forms, specifically SNF assemblies. The thermal analyses discussed in this paper need to be adapted for the specific geometry and heat output of these other waste forms. While predominantly conductive heat transfer is appropriately captured by focusing on the local behavior in a short section of the repository, other processes (e.g., corrosion gas migration, regional fluid flow, radionuclide transport) may require that the entire drillhole (including the vertical access section) be modeled. Nevertheless, an overall approach similar to the one presented here can be used to examine such processes in support of repository design, uncertainty quantification, and performance assessment. Parameters that are both influential and uncertain need to be carefully assessed prior to performing design calculations and uncertainty analyses. This analysis demonstrates through a data-worth analysis that a short-term in situ heater test is a viable approach to robustly identify the key factors affecting the temperature evolution in the repository. The main conceptual idea is to run a test that (a) uses the as-built configuration under in situ conditions (thus testing the actual disposal system), (b) examines the system at the actual scale (thus avoiding the need for upscaling), (c) perturbs the system using thermal stresses (thus invoking the relevant process), and (d) collects (temperature) data that are identical to the prediction variable of interest (thus avoiding the need for indirect inferences). A well-designed heater test, which can readily be integrated into the operation of a horizontal drillhole waste repository, is an effective, defensible way to obtain confidence in the thermal system behavior, and to improve the ability to make robust predictions about the suitability of a rock formation as a hazardous waste repository. FIG. 13 is a schematic illustration of a controller (or control system) 1300 according to the present disclosure. For example, the controller 1300 can be used for the operations described previously, for example as or as part of the heating and monitoring control system 646. For example, the controller 1300 may be communicably coupled with, or as a part of, a hazardous material repository as described herein. The controller 1300 is intended to include various forms of digital computers, such as printed circuit boards (PCB), processors, digital circuitry, or otherwise that is part of a vehicle. Additionally the system can include portable storage media, such as, Universal Serial Bus (USB) flash drives. For example, the USB flash drives may store operating systems and other applications. The USB flash drives can include input/output components, such as a wireless transmitter or USB connector that may be inserted into a USB port of another computing device. The controller 1300 includes a processor 1310, a memory 1320, a storage device 1330, and an input/output device 1340. Each of the components 1310, 1320, 1330, and 1340 are interconnected using a system bus 1350. The processor 1310 is capable of processing instructions for execution within the controller 1300. The processor may be designed using any of a number of architectures. For example, the processor 1310 may be a CISC (Complex Instruction Set Computers) processor, a RISC (Reduced Instruction Set Computer) processor, or a MISC (Minimal Instruction Set Computer) processor. In one implementation, the processor 1310 is a single-threaded processor. In another implementation, the processor 1310 is a multi-threaded processor. The processor 1310 is capable of processing instructions stored in the memory 1320 or on the storage device 1330 to display graphical information for a user interface on the input/output device 1340. The memory 1320 stores information within the controller 1300. In one implementation, the memory 1320 is a computer-readable medium. In one implementation, the memory 1320 is a volatile memory unit. In another implementation, the memory 1320 is a non-volatile memory unit. The storage device 1330 is capable of providing mass storage for the controller 1300. In one implementation, the storage device 1330 is a computer-readable medium. In various different implementations, the storage device 1330 may be a floppy disk device, a hard disk device, an optical disk device, a tape device, flash memory, a solid state device (SSD), or a combination thereof. The input/output device 1340 provides input/output operations for the controller 1300. In one implementation, the input/output device 1340 includes a keyboard and/or pointing device. In another implementation, the input/output device 1340 includes a display unit for displaying graphical user interfaces. The features described can be implemented in digital electronic circuitry, or in computer hardware, firmware, software, or in combinations of them. The apparatus can be implemented in a computer program product tangibly embodied in an information carrier, for example, in a machine-readable storage device for execution by a programmable processor; and method steps can be performed by a programmable processor executing a program of instructions to perform functions of the described implementations by operating on input data and generating output. The described features can be implemented advantageously in one or more computer programs that are executable on a programmable system including at least one programmable processor coupled to receive data and instructions from, and to transmit data and instructions to, a data storage system, at least one input device, and at least one output device. A computer program is a set of instructions that can be used, directly or indirectly, in a computer to perform a certain activity or bring about a certain result. A computer program can be written in any form of programming language, including compiled or interpreted languages, and it can be deployed in any form, including as a stand-alone program or as a module, component, subroutine, or other unit suitable for use in a computing environment. Suitable processors for the execution of a program of instructions include, by way of example, both general and special purpose microprocessors, and the sole processor or one of multiple processors of any kind of computer. Generally, a processor will receive instructions and data from a read-only memory or a random access memory or both. The essential elements of a computer are a processor for executing instructions and one or more memories for storing instructions and data. Generally, a computer will also include, or be operatively coupled to communicate with, one or more mass storage devices for storing data files; such devices include magnetic disks, such as internal hard disks and removable disks; magneto-optical disks; and optical disks. Storage devices suitable for tangibly embodying computer program instructions and data include all forms of non-volatile memory, including by way of example semiconductor memory devices, such as EPROM, EEPROM, solid state drives (SSDs), and flash memory devices; magnetic disks such as internal hard disks and removable disks; magneto-optical disks; and CD-ROM and DVD-ROM disks. The processor and the memory can be supplemented by, or incorporated in, ASICs (application-specific integrated circuits). To provide for interaction with a user, the features can be implemented on a computer having a display device such as a CRT (cathode ray tube) or LCD (liquid crystal display) or LED (light-emitting diode) monitor for displaying information to the user and a keyboard and a pointing device such as a mouse or a trackball by which the user can provide input to the computer. Additionally, such activities can be implemented via touchscreen flat-panel displays and other appropriate mechanisms. The features can be implemented in a control system that includes a back-end component, such as a data server, or that includes a middleware component, such as an application server or an Internet server, or that includes a front-end component, such as a client computer having a graphical user interface or an Internet browser, or any combination of them. The components of the system can be connected by any form or medium of digital data communication such as a communication network. Examples of communication networks include a local area network (“LAN”), a wide area network (“WAN”), peer-to-peer networks (having ad-hoc or static members), grid computing infrastructures, and the Internet. A first example implementation according to the present disclosure includes a method for testing a hazardous waste repository site. The method includes running a heating unit into a drillhole that is formed from a terranean surface into or under a subterranean zone that includes a rock formation. The drillhole includes a vertical portion formed from the terranean surface and a non-vertical portion coupled to the vertical portion that is formed in or under the subterranean zone. The non-vertical portion includes a hazardous material repository portion configured to store one or more canisters configured to enclose hazardous material. The method further includes positioning the heating unit in the hazardous material repository portion of the non-vertical portion of the drillhole; operating the heating unit to generate heat in the hazardous material repository portion for a specified time duration; monitoring a temperature in or near the hazardous material repository portion of the drillhole during the specified time duration; and based on the monitored temperature, determining one or more thermal properties of at least one of the rock formation or the hazardous material repository portion of the drillhole. In an aspect combinable with the first example implementation, the rock formation includes at least one of a sedimentary, an igneous, or a metamorphic rock formation, such as at least one of shale, claystones, or mudstones. In another aspect combinable with any of the previous aspects of the first example implementation, the hazardous material includes nuclear material waste. In another aspect combinable with any of the previous aspects of the first example implementation, the nuclear waste includes spent nuclear fuel. In another aspect combinable with any of the previous aspects of the first example implementation, the heating unit is a similar size of one of the one or more canisters. In another aspect combinable with any of the previous aspects of the first example implementation, the heating unit is configured to output between 50 and 500 watts of heat for the time duration. In another aspect combinable with any of the previous aspects of the first example implementation, monitoring the temperature includes measuring the temperature with a temperature sensor positioned on a casing positioned in the drillhole. In another aspect combinable with any of the previous aspects of the first example implementation, the temperature sensor includes a distributed temperature sensor (DTS) that includes an fiber-optical sensor cable. In another aspect combinable with any of the previous aspects of the first example implementation, the one or more thermal properties includes thermal diffusivity of the rock formation. Another aspect combinable with any of the previous aspects of the first example implementation further includes aggregating a plurality of the monitored temperatures during the time duration. Another aspect combinable with any of the previous aspects of the first example implementation further includes determining a change of temperature in or near the hazardous material repository portion of the drillhole based on the plurality of the monitored temperatures during the time duration. Another aspect combinable with any of the previous aspects of the first example implementation further includes calculating a data-worth metric based on the plurality of the monitored temperatures during the time duration, the data-worth metric including a relative reduction in uncertainty of a relevant performance measure such as maximum temperature predicted in or near the hazardous material repository portion of the drillhole. Another aspect combinable with any of the previous aspects of the first example implementation further includes determining an uncertainty of the thermal diffusivity of the rock formation based on the temperature monitored in or near the hazardous material repository portion of the drillhole during the time duration. Another aspect combinable with any of the previous aspects of the first example implementation further includes stopping operation of the heating unit based on at least one of the uncertainty of the thermal diffusivity of the rock formation or the uncertainty of the predicted performance measure being less than a threshold value. Another aspect combinable with any of the previous aspects of the first example implementation further includes determining a portion of the one or more thermal properties of at least one of the rock formation or the hazardous material repository portion of the drillhole based on the determined thermal diffusivity of the rock formation. In another aspect combinable with any of the previous aspects of the first example implementation, the portion of the one or more thermal properties of at least one of the rock formation or the hazardous material repository portion of the drillhole includes at least one of a spacing distance between adjacent canisters placed in the hazardous material repository portion; a maximum allowable heat output of the hazardous material; a minimum allowable range of a thermal diffusivity of a backfill material positioned in the one or more canisters; a minimum allowable range of a thermal diffusivity of a backfill material positioned in drillhole; a minimum allowable range of a thermal diffusivity of the casing; or a minimum allowable range of a thermal diffusivity of a canister material. Another aspect combinable with any of the previous aspects of the first example implementation further includes determining a suitability of the hazardous material repository portion of the non-vertical portion of the drillhole formed in or under the subterranean zone based on the determined thermal diffusivity of the rock formation. Another aspect combinable with any of the previous aspects of the first example implementation further includes, based on the determined suitability, moving the one or more canisters from the terranean surface, through the vertical portion of the drillhole, through the non-vertical portion of the drillhole, and into the hazardous material repository portion. Another aspect combinable with any of the previous aspects of the first example implementation further includes filling at least a portion of the drillhole with a backfill material. Another aspect combinable with any of the previous aspects of the first example implementation further includes removing the heating unit from the drillhole prior to moving the one or more canisters through the vertical portion of the drillhole. A second example implementation according to the present disclosure includes a thermal testing system for a hazardous waste repository configured to perform the operations of the first example implementation and all of the aspects of the first example implementation. A third example implementation includes a method that includes forming a vertical access drillhole from a terranean surface toward a subterranean zone that includes a hazardous waste repository; forming at least one curved access drillhole from the vertical access drillhole toward or into the subterranean zone; forming a horizontal drillhole into the subterranean zone from the at least one curved access drillhole, the horizontal drillhole including at least a portion of the hazardous waste repository; moving radioactive water from the terranean surface, through the vertical access drillhole and the at least one curved access drillhole, and into the hazardous waste repository of the horizontal drillhole; and installing at least one seal within at least one of the vertical access drillhole or the curved access drillhole. In an aspect combinable with the third example implementation, the radioactive water includes a radioactive material. In another aspect combinable with any of the previous aspects of the third example implementation, the radioactive material includes at least one of tritium, cesium, or strontium. In another aspect combinable with any of the previous aspects of the third example implementation, moving the radioactive water includes pumping the radioactive water into the hazardous waste repository of the horizontal drillhole. Another aspect combinable with any of the previous aspects of the third example implementation further includes, prior to pumping the radioactive water into the hazardous waste repository of the horizontal drillhole, mixing the radioactive water with a cementitious material or a gel. Another aspect combinable with any of the previous aspects of the third example implementation further includes enclosing the radioactive water into one or more hazardous waste canisters. In another aspect combinable with any of the previous aspects of the third example implementation, moving the radioactive water into the hazardous waste repository of the horizontal drillhole includes moving the one or more hazardous waste canisters into the hazardous waste repository of the horizontal drillhole. Another aspect combinable with any of the previous aspects of the third example implementation further includes, prior to enclosing the radioactive water into one or more hazardous waste canisters, mixing the radioactive water with a cementitious material or a gel. In another aspect combinable with any of the previous aspects of the third example implementation, the one or more hazardous waste canisters include a corrosion-resistant alloy. Another aspect combinable with any of the previous aspects of the third example implementation further includes installing a casing in the horizontal drillhole; and securing the casing in the horizontal drillhole with cement. Another aspect combinable with any of the previous aspects of the third example implementation further includes mixing at least a portion of the radioactive water with a cementitious aggregate to form the cement; and circulating the formed cement between the casing and the horizontal drillhole. Another aspect combinable with any of the previous aspects of the third example implementation further includes monitoring the radioactive water stored in the hazardous waste repository of the horizontal drillhole. In another aspect combinable with any of the previous aspects of the third example implementation, monitoring includes measuring an amount of beta or gamma radiation in the horizontal drillhole near the stored radioactive water. In a fourth example implementation, a hazardous waste repository includes a vertical access drillhole formed from a terranean surface toward a subterranean zone that includes a hazardous waste repository; at least one curved access drillhole formed from the vertical access drillhole toward or into the subterranean zone; and a horizontal drillhole formed into the subterranean zone from the at least one curved access drillhole. The horizontal drillhole includes at least a portion of the hazardous waste repository configured to enclose a volume of radioactive water moved from the terranean surface. The horizontal drillhole is formed through the vertical access drillhole and the at least one curved access drillhole, and into the hazardous waste repository of the horizontal drillhole. The repository also includes at least one seal installed within at least one of the vertical access drillhole or the curved access drillhole. In an aspect combinable with the fourth example implementation the radioactive water includes a radioactive material. In another aspect combinable with any of the previous aspects of the fourth example implementation, the radioactive material includes at least one of tritium, cesium, or strontium. In another aspect combinable with any of the previous aspects of the fourth example implementation, the volume of radioactive water is pumped into the hazardous waste repository of the horizontal drillhole. In another aspect combinable with any of the previous aspects of the fourth example implementation, the volume of radioactive water is mixed with a cementitious material or a gel. In another aspect combinable with any of the previous aspects of the fourth example implementation, the volume of radioactive water is enclosed in one or more hazardous waste canisters. In another aspect combinable with any of the previous aspects of the fourth example implementation, the one or more hazardous waste canisters are stored in the hazardous waste repository of the horizontal drillhole. In another aspect combinable with any of the previous aspects of the fourth example implementation. the mixed volume of radioactive water and cementitious material or gel is enclosed within the one or more hazardous waste canisters. In another aspect combinable with any of the previous aspects of the fourth example implementation, the one or more hazardous waste canisters include a corrosion-resistant alloy. Another aspect combinable with any of the previous aspects of the fourth example implementation further includes a casing installed in the horizontal drillhole; and cement that secures the casing in the horizontal drillhole. In another aspect combinable with any of the previous aspects of the fourth example implementation, the cement includes at least a portion of radioactive water mixed with a cementitious aggregate. Another aspect combinable with any of the previous aspects of the fourth example implementation further includes a downhole monitoring system configured to monitor the radioactive water stored in the hazardous waste repository of the horizontal drillhole. In another aspect combinable with any of the previous aspects of the fourth example implementation, the monitoring system is configured to measure an amount of beta or gamma radiation in the horizontal drillhole near the stored radioactive water. In another aspect combinable with any of the previous aspects of the fourth example implementation, the monitoring system includes a wire or fiber optic cable positioned in at least a portion of the horizontal drillhole to measure the amount of beta or gamma radiation. A number of implementations have been described. Nevertheless, it will be understood that various modifications may be made without departing from the spirit and scope of the disclosure. For example, example operations, methods, or processes described herein may include more steps or fewer steps than those described. Further, the steps in such example operations, methods, or processes may be performed in different successions than that described or illustrated in the figures. Accordingly, other implementations are within the scope of the following claims. |
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description | Priority is claimed as a divisional application to U.S. patent application Ser. No. 12/774,944 (now U.S. Pat. No. 8,798,224), filed May 6, 2010, which claims priority to U.S. Provisional Patent Application Ser. No. 61/175,899, filed May 6, 2009. The aforementioned priority applications are incorporated herein by reference in their entirety as if set forth in full. The present invention relates generally to apparatus, systems and methods for storing and/or transporting high level radioactive waste, and specifically to such apparatus, systems and methods that utilize a ventilated vertical overpack that allows natural convection cooling of the high level radioactive waste, which can be spent nuclear fuel (“SNF”) in certain instances. In the operation of nuclear reactors, it is customary to remove fuel assemblies after their energy has been depleted down to a predetermined level. Upon removal, this SNF is still highly radioactive and produces considerable heat, requiring that great care be taken in its packaging, transporting, and storing. In order to protect the environment from radiation exposure, SNF is first placed in a canister, which is typically a hermetically sealed canister that creates a confinement boundary about the SNF. The loaded canister is then transported and stored in a large cylindrical container called a cask. Generally, a transfer cask is used to transport spent nuclear fuel from location to location while a storage cask is used to store SNF for a determined period of time. In a typical nuclear power plant, an open empty canister is first placed in an open transfer cask. The transfer cask and empty canister are then submerged in a pool of water. SNF is loaded into the canister while the canister and transfer cask remain submerged in the pool of water. Once the canister is fully loaded with SNF, a lid is placed atop the canister while in the pool. The transfer cask and canister are then removed from the pool of water. Once out of the water, the lid of the canister is welded to the canister body and a cask lid is then installed on the transfer cask. The canister is then dewatered and backfilled with an inert gas. The transfer cask (which is holding the loaded canister) is then transported to a location where a storage cask is located. The loaded canister is then transferred from the transfer cask to the storage cask for long term storage. During transfer of the canister from the transfer cask to the storage cask, it is imperative that the loaded canister is not exposed to the environment. One type of storage cask is a ventilated vertical overpack (“VVO”). A VVO is a massive structure made principally from steel and concrete and is used to store a canister loaded with spent nuclear fuel. Traditional VVOs stand above ground and are typically cylindrical in shape and are extremely heavy, often weighing over 150 tons and having a height greater than 16 feet. VVOs typically have a flat bottom, a cylindrical body having a cavity to receive a canister of SNF, and a removable top lid. In using a VVO to store SNF, a canister loaded with SNF is placed in the cavity of the cylindrical body of the VVO. Because the SNF is still producing a considerable amount of heat when it is placed in the VVO for storage, it is necessary that this heat energy have a means to escape from the VVO cavity. This heat energy is removed from the outside surface of the canister by ventilating the VVO cavity. In ventilating the VVO cavity, cool air enters the VVO chamber through bottom ventilation ducts, flows upward past the loaded canister as it is warmed from the heat emanating from the canister, and exits the VVO at an elevated temperature through top ventilation ducts. Such VVOs do not require the use of equipment to force the air flow through the VVO. Rather, these VVOs are passive cooling systems as they use the natural air flow induced by the heated air to rise within the VVO (also know as the chimney effect). While it is necessary that the VVO cavity be vented so that heat can escape from the canister, it is also imperative that the VVO provide adequate radiation shielding and that the SNF not be directly exposed to the external environment. The inlet duct located near the bottom of the overpack is a particularly vulnerable source of radiation exposure to security and surveillance personnel who, in order to monitor the loaded VVOs, must place themselves in close vicinity of the ducts for short durations. Therefore, when a typical VVO is used to store a canister of SNF in its internal cavity, the canister is supported in the cavity so that the bottom surface of the canister is higher than the top of inlet ventilation ducts. This is often accomplished by providing support blocks on the floor of the cavity. By positioning the bottom surface of the canister above the inlet ventilation ducts, a line of sight does not exist from the canister to the external atmosphere through the inlet ventilation ducts, thus eliminating the danger of radiation shine out of inlet ventilation ducts. However, as discussed below, positioning a canister in the cavity of a VVO so that the bottom surface of the canister is above the top of the inlet ventilation ducts creates two issues: (1) a potential cooling problem during a “smart flood” condition; and (2) an increased height of the VVO. Subpart K of 10 C.F.R. § 72 provides for a “general certification” of casks for on-site storage of SNF. A number of casks have been licensed by the United States Nuclear Regulatory Committee (“U.S.N.R.C.”) and are listed in subpart L of 10 C.F.R. § 72. These casks are certified to store a whole class of SNF (including SNF coming from pressurized water reactors (PWRs) or boiling water reactors (BWRs)). Unfortunately, reactors burn fuel in a wide variety of lengths. For example, PWRs in the U.S. presently burn fuel as short as 146″ (e.g., Ft. Calhoun) and as long as 198″ (e.g., South Texas). A general certified cask has been licensed in one or two fixed lengths (models) by the U.S.N.R.C. However, if the SNF is too long to fit in a licensed cask, then the cask simply cannot be used. Moreover, if the SNF is too short, then axial spacers are used to fill the open space in the storage cells to limit the movement of SNF in the axial direction. Thus, most casks and canisters used in the on-site storage of SNF have significant open spaces in their storage cells. This condition is particularly undesirable for VVOs because of the adverse consequence to the occupational dose to the plant personnel and cost (because of physical modifications forced on the plant), as set forth below. First, the dose received by the workers performing the loading operations is directly influenced by the amount of shielding material per unit length in the body of the cask. The total quantity of shielding that can be installed in a transfer cask is governed by the lifting capacity of the plant's cask crane. A longer than necessary transfer cask means less shielding per unit length installed in the cask which in turn results in increased dose to the workers. In VVOs, the VVO is often loaded inside the plant's truck bay by stacking the transfer cask over the VVO. Minimizing the height of the VVO's body is essential to allow the VVO to be moved out through the plant's truck bay (typically, a roll-up door) after the canister is installed therein. The loaded VVO is typically moved out across the roll-up door without its lid, and the lid is then installed on it immediately after the VVO body clears the door. Therefore, a key objective in the storage VVO design is to minimize the height of VVO body. In another variation, the transfer cask itself is taken outside through the plant's truck bay and carried over to a pit where the transfer of the canister to the VVO takes place. In this case, the height of the transfer cask must be short enough to clear the plant's roll-up door to avoid the need to shorten the transfer cask (or alternatively, to increase the height of the roll-up door). Shortening the transfer cask is not always possible. The present invention, in one aspect, is a ventilated overpack having specially designed inlet ducts that allow a canister loaded with SNF (or other high level radioactive waste) to be positioned within the overpack so that a bottom end of the canister is below a top of the inlet ducts while still preventing radiation from escaping through the inlet ducts. This aspect of the present invention allows the overpack to be designed with a minimized height because the canister does not have to be supported in a raised position above the inlet ducts within the cavity of the overpack. Thus, it is possible for the height of the cavity of the overpack to be approximately equal to the height of the canister, with the addition of the necessary tolerances for thermal growth effects and to provide for an adequate ventilation space above the canister. When the canister is supported within the overpack cavity so that the bottom end of the canister is below the top end of the inlet ducts, the canister is protected from over-heating during a “smart flood” condition because a substantial portion of the canister will become submerged in the flood water prior to the incoming air flow from the inlet duct being choked off. Moreover, the design and arrangement of inlet ducts of the inventive overpack result in the cooling air flow within the overpack to not be significantly impacted by high wind conditions exterior to the overpack. In one embodiment, the invention can be an apparatus for transporting and/or storing high level radioactive waste comprising: an overpack body having an outer surface and an inner surface forming an internal cavity about a longitudinal axis; a base enclosing a bottom end of the cavity; a plurality of inlet ducts in a bottom of the overpack body, each of the inlet ducts extending from an opening in the outer surface of the overpack body to an opening in the inner surface of the overpack body so as to form a passageway from an external atmosphere to a bottom portion of the cavity; a columnar structure located within each of the inlet ducts, the columnar structures dividing each of the passageways of the inlet ducts into first and second channels that converge at the first and second openings, wherein for each inlet duct a line of sight does not exist between the opening in the inner surface of the overpack body and the opening in the outer surface of the overpack body; a lid enclosing a top end of the cavity; and a plurality of outlet ducts, each of the outlet ducts forming a passageway from a top portion of the cavity to the external atmosphere. In another embodiment, the invention is an apparatus for transporting and/or storing high level radioactive waste comprising: a cylindrical radiation shielding body forming an internal cavity and having a vertical axis; a base enclosing a bottom end of the cavity; a plurality of inlet ducts in a bottom of the radiation shielding body, each of the inlet ducts forming a horizontal passageway from an external atmosphere to a bottom portion of the cavity; a radiation shielding structure located within each of the inlet ducts that divides the horizontal passageway of the inlet duct into at least first and second horizontally adjacent portions and blocks a line of sight from existing from the cavity to the external atmosphere through the inlet duct; a radiation shielding lid enclosing a top end of the cavity; and a plurality of outlet ducts, each of the outlet ducts forming a passageway from a top portion of the cavity to the external atmosphere. In another aspect, the invention is directed to a method of utilizing a general license obtained for two different ventilated vertical overpacks to manufacture a third ventilated vertical overpack that is covered by the general license without filing an application for certification of the third ventilated vertical overpack. In one embodiment, the invention can be a method of manufacturing a licensed ventilated vertical overpack without filing an application for certification comprising: designing a first ventilated vertical overpack comprising: a first cavity for receiving a first canister containing high level radioactive waste, the first cavity having a first horizontal cross section and a first height; a first ventilation system for facilitating natural convection cooling of the first canister within the first cavity, the first ventilation system comprising a first plurality of inlet vents for introducing cool air into a bottom of the first cavity and a first plurality of outlet vents for allowing heated air to escape from a top of the first cavity; and wherein the first ventilated vertical overpack is designed to withstand an inertial load resulting from a postulated tip-over event so as to maintain the integrity of the first canister within the cavity; designing a second ventilated vertical overpack comprising: a second cavity for receiving a second canister containing high level radioactive waste, the second cavity having a second horizontal cross section that is the same as the first horizontal cross section and a second height that is less than the first height; a second ventilation system for facilitating natural convective cooling of the second canister within the second cavity, the second ventilation system comprising a second plurality of inlet vents for introducing cool air into a bottom of the second cavity and a second plurality of outlet vents for allowing heated air to escape from a top of the second cavity, wherein the second plurality of inlet vents have the same configuration as the first plurality of inlet vents and the second plurality of outlet vents have the same configuration as the first plurality of outlet vents; and wherein the second ventilated vertical overpack is designed to achieve a heat rejection capacity; obtaining a license from a regulatory agency for the first and second ventilated vertical overpacks; manufacturing a third ventilated vertical overpack comprising: a third cavity for receiving a third canister containing high level radioactive waste, the third cavity having a third horizontal cross section that is the same as the first and second horizontal cross sections and a third height that is less than the first height and greater than the second height; a third ventilation system for facilitating natural convective cooling of the third canister within the third cavity, the third ventilation system comprising a third plurality of inlet vents for introducing cool air into a bottom of the third cavity and a third plurality of outlet vents for allowing heated air to escape from a top of the third cavity, wherein the third plurality of inlet vents have the same configuration as the first and second plurality of inlet vents, and the third plurality of outlet vents have the same configuration as the first and second plurality of outlet vents; and wherein the third ventilated vertical overpack is automatically covered by the license without filing a new application for certification with the regulatory agency. In another embodiment, the invention can be a method of manufacturing a licensed ventilated vertical overpack without filing an application for certification comprising: designing a first ventilated vertical overpack having a first cavity for receiving a first canister containing high level radioactive waste and having a structural configuration that can withstand an inertial load resulting from a postulated tip-over event so as to maintain the integrity of the first canister within the cavity, the first cavity having a first height that corresponds to a height of the first canister; designing a second ventilated vertical overpack having a second cavity for receiving a second canister containing high level radioactive waste and an inlet and outlet duct configuration for facilitating natural convective cooling of the second canister that achieves a heat rejection capacity, the second cavity having a second height that corresponds to a height of the second canister, the first height being greater than the second height; obtaining a license from a regulatory agency for the first and second ventilated vertical overpacks; manufacturing a third ventilated vertical overpack comprising: a third cavity for receiving a third canister containing high level radioactive waste, the third cavity having a third height that corresponds to a height of the third canister, the third height being greater than the second height and less than the first height; a structural configuration that is the same as the structural configuration of the first ventilated vertical overpack; and an inlet and outlet duct configuration for facilitating natural convective cooling of the third canister that is the same as the inlet and outlet duct configuration of the second ventilated vertical overpack; and wherein the first, second and third cavities have the same horizontal cross-sections and the first, second and third canisters have the same horizontal cross-sections; wherein the third ventilated vertical overpack is automatically covered by the license without filing a new application for certification with the regulatory agency. Referring to FIGS. 1-4 concurrently, a ventilated vertical overpack (“VVO”) 1000 according to an embodiment of the present invention is illustrated. The VVO 1000 is a vertical, ventilated, dry, SNF storage system that is fully compatible with 100 ton and 125 ton transfer casks for spent fuel canister transfer operations. The VVO 1000 can, of course, be modified and/or designed to be compatible with any size or style of transfer cask. Moreover, while the VVO 1000 is discussed herein as being used to store SNF, it is to be understood that the invention is not so limited and that, in certain circumstances, the VVO 1000 can be used to transport SNF from location to location if desired. Moreover, the VVO 1000 can be used in combination with any other type of high level radioactive waste. The VVO 1000 is designed to accept a canister for storage at an Independent Spent Fuel Storage Installation (“ISFSI”). All canister types engineered for the dry storage of SNF can be stored in the VVO 1000. Suitable canisters include multi-purpose canisters (“MPCs”) and, in certain instances, can include thermally conductive casks that are hermetically sealed for the dry storage of high level radioactive waste. Typically, such canisters comprise a honeycomb basket 250, or other structure, to accommodate a plurality of SNF rods in spaced relation. An example of an MPC that is particularly suited for use in the VVO 1000 is disclosed in U.S. Pat. No. 5,898,747 to Krishna Singh, issued Apr. 27, 1999, the entirety of which is hereby incorporated by reference. The VVO 1000 comprises two major parts: (1) a dual-walled cylindrical overpack body 100 which comprises a set of inlet ducts 150 at or near its bottom extremity and an integrally welded baseplate 130; and (2) a removable top lid 500 equipped with radially symmetric outlet vents 550. The overpack body 100 forms an internal cylindrical storage cavity 10 of sufficient height and diameter for housing an MPC 200 fully therein. As discussed in greater detail below, the VVO 1000 is designed so that the internal cavity 10 has a minimized height that corresponds to a height of the MPC 200 which is to be stored therein. Moreover, the cavity 10 preferably has a horizontal (i.e., transverse to the axis A-A) cross-section that is sized to accommodate only a single MPC 200. The overpack body 100 extends from a bottom end 101 to a top end 102. The base plate 130 is connected to the bottom end 101 of the overpack body 100 so as to enclose the bottom end of the cavity 10. An annular plate 140 is connected to the top end 102 of the overpack body 100. The annular plate 140 is ring-like structure while the base plate 130 is thick solid disk-like plate. The base plate 130 hermetically encloses the bottom end 101 of the overpack body 100 (and the storage cavity 10) and forms a floor for the storage cavity 10. If desired, an array of radial plate-type gussets 112 may be welled to the inner surface 121 of an inner shell 120 and a top surface 131 of the base plate 130. In such an embodiment, when the MPC 200 is fully loaded into the cavity 10, the MPC 200 will rest atop the gussets 112. The gussets 112 have top edges that are tapered downward toward the vertical central axis A-A. Thus, the gussets 112 guide the MPC 200 during loading and help situate the MPC 200 in a coaxial disposition with the central vertical axis A-A of the VVO 1000. In certain embodiments, the MPC 200 may not rest on the gussets 112 but rather may rest directly on the top surface 131 of the base plate 130. In such an embodiment, the gussets 112 may still be provided to not only act as guides for properly aligning the MPC 200 within the cavity 10 during loading but also to act as spacers for maintaining the MPC 200 in the desired alignment within the cavity 10 during storage. By virtue of its geometry, the overpack body 100 is a rugged, heavy-walled cylindrical vessel. The main structural function of the overpack body is provided by its carbon steel components while the main radiation shielding function is provided by an annular plain concrete mass 115. The plain concrete mass 115 of the overpack body 100 is enclosed by concentrically arranged cylindrical steel shells 110, 120, the thick steel baseplate 130, and the top steel annular plate 140. A set of four equispaced steel radial connector plates 111 are connected to and join the inner and outer shells 110, 120 together, thereby defining a fixed width annular space between the inner and outer shells 120, 110 in which the plain concrete mass 115 is poured. The plain concrete mass 115 between the inner and outer steel shells 120, 110 is specified to provide the necessary shielding properties (dry density) and compressive strength for the VVO 1000. The principal function of the concrete mass 115 is to provide shielding against gamma and neutron radiation. However, the concrete mass 115 also helps enhance the performance of the VVO 1000 in other respects as well. For example, the massive bulk of the concrete mass 115 imparts a large thermal inertia to the VVO 1000, allowing it to moderate the rise in temperature of the VVO 1000 under hypothetical conditions when all ventilation passages 150, 550 are assumed to be blocked. The case of a postulated fire accident at an ISFSI is another example where the high thermal inertia characteristics of the concrete mass 115 of the VVO 1000 controls the temperature of the MPC 200. Although the annular concrete mass 115 in the overpack body 100 is not a structural member, it does act as an elastic/plastic filler of the inter-shell space. Four threaded steel anchor blocks (not illustrated) are also provided at the top of the overpack body 100 for lifting. The anchor blocks are integrally welded to the radial plates 111, which join the inner and outer shells 120, 110. The four anchor blocks are located at 90° angular spacings around the circumference of the top of the overpack body 100. While the cylindrical body 100 has a generally circular horizontal cross-section, the invention is not so limited. As used herein, the term “cylindrical” includes any type of prismatic tubular structure that forms a cavity therein. As such, the overpack body can have a rectangular, circular, triangular, irregular or other polygonal horizontal cross-section. Additionally, the term “concentric” includes arrangements that are non-coaxial and the term “annular” includes varying width. The overpack body 100 comprises a plurality of specially designed inlet vents 150. The inlet vents 150 are located at a bottom of the overpack body 100 and allow cool air to enter the VVO 1000. The inlet vents 150 are positioned about the circumference of overpack body 100 in a radially symmetric and spaced-apart arrangement. The structure, arrangement and function of the inlet vents 150 will be described in much greater detail below with respect to FIGS. 4-6 and 10. Referring now to FIGS. 1-4 and 7 concurrently, the overpack lid 500 is a weldment of steel plates 510 filled with a plain concrete mass 515 that provides neutron and gamma attenuation to minimize skyshine. The lid 500 is secured to a top end 101 of the overpack body 100 by a plurality of bolts 501 that extend through bolt holes 502 formed into a lid flange 503. When secured to the overpack body 100, surface contact between the lid 500 and the overpack body 100 forms a lid-to-body interface. The lid 500 is preferably non-fixedly secured to the body 100 and encloses the top end of the storage cavity 10 formed by the overpack body 100. The top lid 500 further comprises a radial ring plate 505 welded to a bottom surface 504 of the lid 500 which provides additional shielding against the laterally directed photons emanating from the MPC 200 and/or the annular space 50 (best shown in FIG. 9) formed between the outer surface 201 of the MPC 200 and the inner surface 121 of the inner shell 120. The ring plate 505 also assists in locating the top lid 500 in a coaxial disposition along axis A-A of the VVO 1000 through its interaction with the annular ring 140. When the lid 500 is secured to the overpack body 100, the outer edge of the ring plate 505 of the lid 500 abuts the inner edge of the annular plate 140 of the overpack body 100. A third function of the radial ring 501 is to prevent the lid 500 from sliding across the top surface of the overpack body 100 during a postulated tipover event defined as a non-mechanistic event for the VVO 1000. As mentioned above, the lid 500 comprises a plurality of outlet vents 550 that allow heated air within the storage cavity 10 of the VVO 1000 to escape. The outlet vents 550 form passageways through the lid 500 that extend from openings 551 in the bottom surface 504 of the lid 500 to openings 552 in the peripheral surface 506 of the lid 500. While the outlet ducts 550 form L-shaped passageways in the exemplified embodiment, any other tortuous or curved path can be used so long as a clear line of sight does not exist from external to the VVO 1000 into the cavity 10 through the inlet ducts 550. The outlet vents 550 are positioned about the circumference of the lid 500 in a radially symmetric and spaced-apart arrangement. The outlet ducts 550 terminate in openings 552 that are narrow in height but axi-symmetric in the circumferential extent. The narrow vertical dimensions of the outlet ducts 550 helps to efficiently block the leakage of radiation. It should be noted, however, that while the outlet vents 550 are preferably located within the lid 500 in the exemplified embodiment, the outlet vents 550 can be located within the overpack body 100 in alternative embodiments, for example at a top thereof. Referring briefly to FIG. 10, the purpose of the inlet vents 150 and the outlet vents 550 is to facilitate the passive cooling of an MPC 200 located within the cavity 10 of the VVO 1000 through natural convection/ventilation. In FIG. 10, the flow of air is represented by the heavy black arrows 3, 5, 7. The VVO 1000 is free of forced cooling equipment, such as blowers and closed-loop cooling systems. Instead, the VVO 1000 utilizes the natural phenomena of rising warmed air, i.e., the chimney effect, to effectuate the necessary circulation of air about the MPC 200 stored in the storage cavity 10. More specifically, the upward flowing air 5 (which is heated from the MPC 200) within the annular space 50 that is formed between the inner surface 121 of the overpack body 100 and the outer surface 201 of the MPC 200 draws cool ambient air 3 into the storage cavity 10 through inlet ducts 150 by creating a siphoning effect at the inlet ducts 150. The rising warm air 5 exits the outlet vents 550 as heated air 7. The rate of air flow through the VVO 1000 is governed by the quantity of heat produced in the MPC 200, the greater the heat generation rate, the greater the air upflow rate. To maximize the cooling effect that the ventilating air stream 3, 5, 7 has on the MPC 200 within the VVO 1000, the hydraulic resistance in the air flow path is minimized to the extent possible. Towards that end, the VVO 1000 comprises eight inlet ducts 150 (shown in FIG. 6). Of course, more or less inlet ducts 150 can be used as desired. In one preferred embodiment, at least six inlet ducts 150 are used. Each inlet duct 150 is narrow and tall and has an internally refractive contour (shown in FIG. 6) so as to minimize radiation streaming while optimizing the size of the airflow passages. The curved shape of the inlet ducts 150 also helps minimize hydraulic pressure loss. The structure of the inlet ducts 150 will be described below in much greater detail with respect to FIGS. 4-6. Referring back to FIGS. 1-4 and 7 concurrently, in order to decrease the amount of radiation scattered to the environment, an array of duct photon attenuators (DPAs) may be installed in the inlet and/or outlet ducts 150, 550. An example of a suitable DPA is disclosed in U.S. Pat. No. 6,519,307, the entirety of which is hereby incorporated by reference. The DPAs scatter any radiation streaming through the ducts 150, 550, thereby significantly decreasing the local dose rates around the ducts 150, 550. The configuration of the DPAs is such that the increase in the resistance to air flow in the air inlet ducts 150 and outlet ducts 550 is minimized. The inlet ducts 150 permit the MPC 200 to be positioned directly atop the top surface 131 of the base plate 130 of the VVO 1000 if desired, thus minimizing the overall height of the cavity 10 that is necessary to house the MPC 200. Naturally, the height of the overpack body 100 is also minimized. Minimizing the height of the overpack body 100 is a crucial ALARA-friendly design feature for those sites where the Egress Bays in their Fuel Buildings have low overhead openings in their roll-up doors. To this extent, the height of the storage cavity 10 in the VVO 1000 is set equal to the height of the MPC 200 plus a fixed amount to account for thermal growth effects and to provide for adequate ventilation space above the MPC 200, as set forth in Table 1 below. TABLE 1OPTIMIZED MPC, TRANSFER CASK, AND VVO HEIGHT DATAFOR A SPECIFIC UNIRRADIATED FUEL LENGTH, lMPC Cavity Height, cl + Δ1MPC Height (including top lid), hc + 11.75″VVO Cavity HeightH + 3.5″Overpack Body Body Height (height from theH + 0.5″bottom end to the top end of the overpackbody)Transfer Cask Cavity Heighth + 1″Transfer Cask Height (loaded over the pad)h + 27″Transfer Cask Total HeightH + 6.5″1Δ shall be selected as 1.5″ < Δ < 2″ so that c is an integral multiple of ½ inch (add 1.5″ to the fuel length and round up to the nearest ½″ or full inch). As can be seen from Table 1, the first step in the height minimization plan is to minimize the height of the MPCs 200. The MPC cavity height, c, is customized for each plant (based on its fuel) so that there is no unnecessary (wasted) space. The MPC 200 can be placed directly on the base plate 130 such that the bottom region of the MPC 200 is level with the inlet ducts 150 because radiation emanating from the MPC 200 is not allowed to escape through the specially shaped inlet ducts 150 due to: (1) the inlet ducts 150 having a narrow width and being curved in shape so as to wrap around a columnar structure 155 made of alloy steel or steel (or a combination of steel and concrete); (2) the configuration of the inlet ducts 150 is such that that there is no clear line of sight from inside the cavity 10 to the exterior environment; and (3) there is enough steel and/or concrete in the path of any radiation emanating from the MPC 200 to de-energize it to acceptable levels. The columnar structure 155 is configured to be cylindrical so as to be internally refractive, but it can also be of rectangular, elliptical, or other prismatic cross-sections to fulfill the essence of the above design features. With the radiation streaming problem at the inlet ducts 150 solved, the top 102 of the overpack body 100 can be as little as ½″ higher than the top surface 202 of the MPC 200. Table 1 above gives typical exemplary dimensions but, of course, is not limiting of the present invention. Finally, with reference to FIG. 4, to protect the concrete mass 115 of the VVO 1000 from excessive temperature rise due to radiant heat from the MPC 200, a thin cylindrical liner 160 of insulating material, can be positioned concentric with the inner shell 120. This insulating liner 140 is slightly smaller in diameter than the inner shell 120. The liner acts as a “heat shield” and can be hung from top impact absorbers 165 or can be connected directly to the inner shell 120 or another structure. The insulating layer 140 can be constructed of, without limitation, blankets of alumina-silica fire clay (Kaowool Blanket), oxides of alimuna and silica (Kaowool S Blanket), alumina-silica-zirconia fiber (Cerablanket), and alumina-silica-chromia (Cerachrome Blanket). The underside of the overpack lid 500 may also include a liner of insulating material if desired. The top impact absorbers 165 are connected to the inner surface 121 of the inner shell 120 in a circumferentially spaced apart arrangement at or near the top end of the cavity 10. Similarly, bottom impact absorbers 166 are connected to the inner surface 121 of the inner shell 120 in a circumferentially spaced apart arrangement at or near the bottom end of the cavity 10. The top and bottom impact absorbers 165, 166 are designed to absorb kinetic energy to protect the MPC 200 during an impactive collision (such as a non-mechanistic tip-over scenario). In the exemplified embodiment, the top and bottom impact absorbers 165, 166 are hollow tube like structures but can be plate structures if desired. The impact absorbers 165, 166 serve as the designated locations of impact with the MPC lid 210 and the base plate 220 of the MPC 200 in case the VVO 1000 tips over. The impact absorbers 165, 166 are thin steel members sized to serve as impact attenuators by crushing (or buckling) against the solid MPC lid 210 and the solid MPC base 220 during an impactive collision (such as a non-mechanistic tip-over scenario). Referring now to FIGS. 4-6 concurrently, the details of the inlet ducts 150 will be discussed in detail. Generally, each of the inlet ducts 150 extends from an opening 151 in the outer surface 112 of the overpack body 100 (which in the exemplified embodiment is also the outer surface of the outer shell 110) to an opening 152 in the inner surface 121 of the overpack body 100 (which in the exemplified embodiment is also the inner surface of the inner shell 120). Each of the inlet ducts 150 forms a passageway 153 from an atmosphere external to the VVO 1000 to a bottom portion of the cavity 10 so that cool air can enter the cavity 10. A columnar structure 155 is located within each of the inlet ducts 150. Each of the columnar structures 155 extend along their own longitudinal axis B-B. In the exemplified embodiment, the longitudinal axes B-B of the columnar structures 155 are substantially parallel with the central vertical axis A-A of the VVO 1000. Thought of another way, the longitudinal axes B-B extend in the load bearing direction of the overpack body 100. Of course, the invention will not be so limited in all embodiments and the longitudinal axes B-B of the columnar structures 155 may be oriented in a different manner if desired. The columnar structures 155 are formed by a combination of steel plates 156, 157 and concrete 115. The plates 157 are cylindrical in shape and bound the outer circumferences of the columnar structures 155, thereby forming the outer surfaces of the columnar structures 155. The plates 156 are flat plates that are thicker than the plates 157 and are centrally positioned within the columnar structures 155 so as to extend along the axes B-B. The plates 156 provide structural integrity to the columnar structures 155 (similar to rebar) and also add additional gamma shielding to the columnar structures 155. The columnar structures 155 have a transverse cross-section that is circular in shape. However, the invention is not so limited and the columnar structures 155 can have a transverse cross-section of any prismatic shape. The columnar structures 155 divide each of the passageways 153 of the inlet ducts 150 into a first channel 153A and a second channel 153B. For each inlet duct 150, the first and second channels 153A, 1538 converge at both openings 151, 152, thereby collectively surrounding the entire circumference of the outer surface of the columnar structure 155. Thought of another way, for each inlet duct 150, the first and second channels 153A, 53B collectively circumferentially surround the longitudinal axes B-B of the columnar structures 155, forming a circular (or other prismatic) passageway contained within the walls of the overpack body 100. Importantly, for each inlet duct 150, a line of sight does not exist between the opening 152 in the inner surface 121 of the overpack body 100 and the opening 151 in the outer surface 112 of the overpack body 100. This is because the columnar structures 155 block such a line-of-sight and provide the required radiation shielding, thereby preventing radiation shine into the environment via the inlet ducts 150. As such, the MPC 200 can be positioned within the cavity 10 so as to be horizontally and vertically aligned with the inlet ducts 150 without radiation escaping into the external environment (see FIGS. 8-9). Stated conceptually, for each inlet duct 150, the opening 152 in the inner surface 121 of the overpack body 100 is aligned with the opening 151 in the outer surface 112 of the overpack body 100 so that: (i) a first reference plane D-D that is perpendicular to the longitudinal axis A-A of the overpack body 100 intersects both the opening 152 in the inner surface 121 of the overpack body 100 and the opening 151 in the outer surface 112 of the overpack body 100; and (ii) a second reference plane C-C that is parallel with and includes the longitudinal axis A-A of the overpack body 100 intersects both the opening 152 in the inner surface 121 of the overpack body 100 and the opening 151 in the outer surface 112 of the overpack body 100. When an MPC 200 is positioned in the cavity 10 as shown in FIGS. 8-9, the MPC 200 is also intersected by the reference plane C-C and the reference plane D-D. The inlet vents 150 (and thus the first and second channels 153A, B) are lined with steel. For each inlet duct 160, the steel liner includes the cylindrical plate 157 of the columnar structure 155, two arcuate wall plates 158, an annular roof plate 159, and the base plate 130. All connections between these plates can be effectuated by welding. As can best be seen in FIGS. 5 and 6, the width of the first and second channels 153A, B is defined by a gap located between the cylindrical plate 157 of the columnar structure 155 and the two arcuate plates 158. Preferably, the cylindrical plate 157 of the columnar structure 155 and the two arcuate plates 158 are arranged in a concentric and evenly spaced-apart manner so that the first and second channels 153A, B have a constant width. Most preferably, the first and second channels 153A, B are curved so as to reduce hydraulic pressure loss. Finally, it is also preferred that the inlet ducts 150 have a height that is at least three times that of its width. Referring now to FIGS. 8-11 concurrently, the benefits achieved by the special design of the inlet ducts 150 with respect to MPC 200 storage will be discussed. During use of the VVO 1000, an MPC 200 is positioned within the cavity 10. An annular gap 50 exists between the outer surface 201 of the MPC 200 and the inner surface 121 of the overpack body 100 The annular gap 50 creates a passageway along the outer surface 201 of the MPC 200 that spatially connects the inlet vents 150 to the outlet vents 550 so that cool air 3 can enter VVO 1000 via the inlet vents 150, be heated within the annular space 50 so as to become warm air 5 that rises within the annular space 50, and exit the VVO 1000 via the outlet vents 550. The MPC 200 is supported within the cavity 10 so that the bottom surface of the MPC 200 rests directly atop the top surface 131 of the base plate 130. This is made possible because the inlet ducts 150 are shaped so as not to allow radiation to shine therethrough because a clear line-of-sight does not exist from the cavity 10 to the atmosphere outside of the VVO 1000 through the inlet ducts 150. Thus, the cavity 10 (and as a result the overpack body 100) can be made as short as possible and substantially correspond to the height of the MPC 200, as discussed above with respect to Table 1. Additionally, positioning the MPC 200 in the cavity 10 so that the bottom surface of the MPC 200 is below the top of the opening 152 of the inlet vents 150 ensures adequate MPC cooling during a “smart flood condition.” A “smart flood” is one that floods the cavity 10 so that the water level is just high enough to completely block airflow though the inlet ducts 150. In other words, the water level is just even with the top of the openings 152 of the inlet ducts 150. Because the bottom surface of the MPC 200 is situated at a height that is below the top of the openings 152 of the inlet ducts 150, the bottom of the MPC 200 will be in contact with (i.e. submerged in) the water during a “smart flood” condition. Because the heat removal efficacy of water is over 100 times that of air, a wet bottom is all that is needed to effectively remove heat and keep the MPC 200 cool. The MPC cooling action effectively changes from ventilation air-cooling to evaporative water cooling. Additionally, as shown in FIG. 11, the MPC 200 is particularly suited for “smart-flood” cooling because the MPC 200 is designed to achieve an internal natural thermopshion cyclical flow. Thus, in a smart-flood,” the thermosiphon flow in the MPC 200 will circulate the internal gas so that the hot gas is circulated to the top of the MPC where its heat can be effectively removed. As mentioned above, the design discussed above for the VVO 1000 allows the VVO 1000 to be constructed so that the height of the cavity 10 (and thus the VVO 1000) is minimized to the extent possible to accommodate an MPC 200 that, in turn, corresponds in height to the length of the SNF assemblies at issue. It has been further discovered that because the MPC 200 does not have to be positioned above the inlet ducts 150, the same configuration of inlet ducts 150 can be used for any and all VVOs 1000, irrespective of the height of the MPC 200 to be positioned therein. Additionally, it has been further discovered that if the outer horizontal cross-section of the MPC 200 and the inner horizontal cross-section of the VVO 1000 are also kept constant, that it is possible to manufacture VVOs 1000 of variable heights under a single N.R.C. (or other regulatory agency) license without having to obtain a new license, so long as a taller and shorter version of the VVO 1000 has already been licensed. Licensing of the shorter VVO 1000 is necessary because the shorter a VVO 1000 is, the less effective the heat rejection capacity of that VVO's natural ventilation system becomes. This is because decreasing the height of the MPC 200 results in a decreased upward flow of air within the annular space 50, thereby reducing the ventilation of the MPC 200. Licensing of the taller VVO 1000 is necessary because the taller a VVO 1000 is, the more susceptible it becomes to inertial loading resulting from a postulated tip-over event that would destroy the integrity of the MPC 200 within the cavity 10. Stated simply, assuming that the ventilation system of the taller and shorter VVOs are held constant, if the shorter VVO meets the required heat rejection capacity, it can be assumed that all taller VVOs will also meet the required heat rejection capacity. Similarly, assuming that the structural configuration of the taller and shorter VVOs are held constant, if the taller VVO can withstand an inertial load resulting from a postulated tip-over event and maintain the integrity of the MPC within its cavity, it can be assumed that all shorter VVOs will also withstand the inertial load resulting from the postulated tip-over event and maintain the integrity of the MPC within its cavity. As used herein, the structural configuration of two VVOs are held constant if the structural components and arrangements remain the same, with exception of the height of the shells 110, 120 and possibly the diameter of the outer shell 110. Thus, in on embodiment, the invention is directed to a method of designing embodiments of the VVO 1000 so that its height is variable and greater than the plant's fuel length by a certain fixed amount. Thus, VVOs 1000 of varying heights can be manufactured under a single U.S.N.R.C. license and be suitable to store SNF in an optimized configuration at all nuclear plants in the world. An embodiment of the present invention will now be described in relation to VVO 1000 discussed above with the addition to suffixes “A-C” to distinguish between the tall version of the VVO 1000A the short version of the VVO 1000B, and the intermediate version of the VVO 1000C respectively. According to one embodiment of the present invention, a VVO 1000A having a first cavity 10A for receiving a first MPC 200A containing high level radioactive waste is designed. This first VVO 1000A comprises a structural configuration that can withstand an inertial load resulting from a postulated tip-over event of the VVO 1000A so as to maintain the integrity of the first MPC 200A within the cavity. The first cavity 100A has a first height H1 that corresponds to the height of the first MPC 200A as discussed above in relation to Table 1. A second VVO 1000B having a second cavity 10B for receiving a second MPC 200B containing high level radioactive waste is then be designed. The second VVO 1000B comprises a configuration of inlet and outlet ducts 150B, 550B for facilitating natural convective cooling of the second MPC 200B that achieves a required heat rejection capacity. The second cavity 10B has a second height H2 that corresponds to the height of the second MPC 200B as discussed above in relation to Table 1. The first height H1 is greater than the second height H2. The designs of the first and second VVOs 1000A, 1000B are then submitted to the appropriate regulatory agency, such as the U.S.N.R.C., for licensing. A license is obtained from the regulatory agency for the first and second VVOs 1000A, 1000B. After the licenses are obtained, a third VVO 1000C comprising a third cavity 10C for receiving a third MPC 200C containing high level radioactive waste is manufactured. The third cavity 10C has a third height H3 that corresponds to a height of the third MPC 200C as discussed above in relation to Table 1. The third height H3 is greater than the second height H2 and less than the first height H1. The VVO 1000C is manufactured to have a structural configuration that is the same as the structural configuration of the first VVO 1000A and a configuration of inlet and outlet ducts 150C, 550 for facilitating natural convective cooling of the third MPC 200C that is the same as the configuration of the inlet and outlet ducts 150B, 550B of the second VVO 1000B. The first, second and third cavities 10A, 10B, 10C all have the same horizontal cross-sections and the first, second and third MPCs 200A, 200B, 200C all have the same outer horizontal cross-sections. Thus, the third VVO 1000C will automatically be covered by the license granted for the VVOs 1000A and 1000B without filing a new application for certification with the regulatory agency. In the example above, the taller VVO 1000A may also be designed to comprise a configuration of inlet and outlet ducts 150A, 550A for facilitating natural convective cooling of the second MPC 200B that achieves a required heat rejection capacity. The configuration of inlet and outlet ducts 150A, 550A may be the same as the configuration of inlet and outlet ducts 150B, 550B of the shorter VVO 1000B. Similarly, the shorter VVO 1000B may also be designed to comprise a structural configuration that can withstand an inertial load resulting from a postulated tip-over event of the VVO 1000B so as to maintain the integrity of the first MPC 200B within the cavity 10B. The structural configuration of the VVO 1000B may be the same as the structural configuration of the VVO 1000A. 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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abstract | The method and system for control of oxygen concentration in the coolant of a reactor plant including a reactor, coolant in the reactor, gas system, mass-exchange apparatus, disperser and an oxygen sensor in the coolant have been disclosed. The method includes the following steps implemented by the system: estimation of the oxygen concentration; comparison of the oxygen concentration with the permissible value; if the oxygen concentration is reduced, comparison of the reduction value and\or rate with the corresponding threshold value; if the reduction value and\or rate of oxygen concentration is below the threshold value, activation of the mass-exchange apparatus; if the reduction value and/or rate of oxygen concentration is above the corresponding threshold value, supply of oxygen-containing gas from the gas system to the near-coolant space and/or activation of the disperser. Technical result: improvement of controllability of oxygen concentration in coolant, enhancement of safety and extension of reactor plant operating life. |
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053207866 | description | The present invention is more particularly described in the following examples which are intended as illustrative only, since numerous modifications and variations will be apparent to those skilled in the art. EXAMPLE A A cold pressed uranium dicarbide pellet was prepared as follows. Uranium dioxide was converted to uranium dicarbide as the initial material by reaction of uranium dioxide with graphite at temperatures above about 1800.degree. C. in accordance with the chemical equation: EQU UO.sub.(2+y) + (2+y+x)C .fwdarw. UC.sub.x + (2+y)CO. UO.sub.2 powder and graphite powder were hand blended and then placed into a ball mill for about 16 hours. The mixture was then sieved through a fine 250 mesh. The mixture of fine particles was then pressed into 1.91 centimeter (cm) diameter briquettes. The briquettes were placed onto graphite trays and entered into a high temperature oven for reaction by carbothermic reduction. The reduction process included drawing a vacuum of 5 .times. 10.sup.-5 Torr in the furnace and heating at temperatures between 1875.degree. C. and 2150.degree. C. for about 4 hours. The furnace was held at the elevated temperature for at least four hours to allow sufficient time for carbon to remove the oxygen as CO and form the particular carbide depending upon the amount of carbon added. The furnace was then rapidly cooled to room temperature. The resultant UC.sub.x briquettes were crushed and then passed through a 16-mesh screen. A binder (0.3 percent by weight polyethylene glycol) and a lubricant (0.3 percent by weight stearic acid) were added in powder form to the crushed material to hold the particles together after subsequent pressing and to protect the die press from the abrasive uranium dicarbide particles. About 2.0 grams (g) of the UC.sub.x powder, including the binder and lubricant additives, was placed into a die cavity and about 110 MPa pressure was applied to form the initial cold pressed uranium dicarbide pellet. EXAMPLE 1 Pellets having an initial composition of UC.sub.1.95 were placed on tantalum, tungsten and graphite trays respectively. The pellets were then heated to 2100.degree. C. under an argon atmosphere for 12 hours to sinter the compositions. The pellets were then analyzed to determine the density as a function of theoretical density and to determine final pellet composition. Pellets sintered upon the tantalum or tungsten trays had a density of above 97 percent theoretical density and a composition of UC.sub.1.88. In comparison, pellets sintered upon a graphite tray had a density of about 75 percent theoretical density and a composition of UC.sub.1.95. The results of this example demonstrate the advantages of sintering uranium dicarbide compositions while in contact with carbon accepting materials such as tantalum and tungsten whereby high theoretical densities are obtained. Although, the present invention has been described with reference to specific details, it is not intended that such details should be regarded as limitations upon the scope of the invention, except as and to the extent that they are included in the accompanying claims. EXAMPLE 2 Discs having an initial diameter of 1.0 inch, a thickness of 0.150 inches, and a composition of UC.sub.1.95 were placed between two tungsten plates. The discs were then heated to 2100.degree. C. under an argon atmosphere for 12 hours to sinter to high density. The discs were then analyzed to determine the density as a function of theoretical density and to determine final disc composition. The discs sintered without cracking to a density above 97 percent theoretical density and a composition near UC.sub.1.88. The results of this example demonstrate the flexibility of this technique to fabricate high density large diameter discs. |
047708450 | claims | 1. In a self-actuated reactor shutdown system utilizing a thermionic means responsive to at least coolant temperature and utilizing reactor coolant differential pressure for controlling the location of a neutron absorber element with respect to a reactor core region and for effecting release of the absorber element, the improvement comprising: electromagnetic means for controlling reactor coolant flow and coolant pressure differential across the absorber element, and thermionic means responsive to at least reactor coolant over-temperature conditions and operatively connected to said electromagnetic means for short-circuiting said electromagnetic means causing a change in coolant flow and a resulting decrease of the differential pressure across the absorber element allowing the element to drop into the reactor core region by gravitational force. 2. The improvement of claim 1, additionally including means for restoring the absorber element and a member of the electromagnetic means to a ready position wherein the absorber element is retained exterior of the reactor core region and the electromagnetic means allows flow of reactor coolant past said absorber element which establishes a differential pressure across the element which retains said element exterior of said core region, whereafter said restoring means can be removed from said absorber element and said electromagnetic means. 3. The improvement of claim 2, wherein said restoring means comprises a retriever rod extending through said absorber element and said electromagnetic means and provided with members positioned in spaced relation to contact said absorber element and said member of said electromagnetic means, and means for moving said retriever rod whereby said absorber element and said member of said electromagnetic means can be moved to the ready position. 4. The improvement of claim 1, wherein said electromagnetic means comprises an electromagnetic coil operatively connected to a power source, and a magnetically attracted slide valve. 5. The improvement of claim 4, wherein said electromagnetic coil is connected to a power source via a control circuit, and wherein said thermionic means is connected in said control circuit so as to be electrically in parallel with said electromagnetic coil. 6. The improvement of claim 5, wherein said thermionic means is a thermionic diode. 7. The improvement of claim 6, wherein said thermionic diode is provided with a uranium blanket extending therearound. 8. The improvement of claim 1, wherein said thermionic means consists of a uranium blanketed thermionic diode connected electrically in parallel with said electromagnetic means, whereby heating of said thermionic diode due to at least reactor coolant over-temperature conditions causes material therein to change state thereby effecting a short-circuiting of said electromagnetic means. 9. The improvement of claim 8, wherein said uranium-blanketed thermionic diode consists of a sealed container having therein an emitter, a collector plate positioned within and spaced from said emitter, a uranium blanket positioned around the exterior of said emitter, electrical leads operatively extending through said container and connected to said emitter and said collector plate, and a quantity of thermionic material located within said collector plate, whereby heating of said diode by at least reactor coolant over-temperature conditions causes said thermionic material to change state from high-electrical resistance to low-electrical resistance, thereby conducting available current away from said electromagnetic means effecting a short-circuiting thereof. 10. A self-actuating reactor shutdown system comprising: a longitudinally extending casing adapted to extend through a reactor core region and having a coolant inlet at one end and a control assembly positioned in the opposite end, a neutron absorber element having coolant flow passageways therethrough located in said casing intermediate said coolant inlet and said control assembly, an absorber up-stop secured to an inner surface of said casing adjacent said control assembly, said control assembly including a housing having a chamber adjacent said up-stop and a coolant passage connecting said chamber with a coolant outlet, an electromagnetically actuated slide valve positioned in said chamber, an electromagnetic coil located adjacent said chamber and adapted to be connected to an associated power source via a control circuit, and a thermionic means connected electrically to said electromagnetic coil, whereupon under normal reactor operating conditions said absorber element hydrostatically is retained against said up-stop in a ready position exterior of the core region by the pressure differential across the absorber element created by coolant flow, and upon loss of sufficient coolant flow to reduce the pressure differential below a specified amount the absorber element drops into the core region by gravitational force, and upon an over-power and/or coolant over-temperature condition the thermionic means causing short-circuiting of said electromagnetic coil causing said slide valve to cover said coolant passage thereby reducing coolant flow past said absorber element resulting in an equalization of the differential pressure across said absorber element and allowing said element to drop into the core region by gravitational force. 11. The system of claim 10, additionally including kinetic energy absorbing means positioned in said casing adjacent said coolant inlet end thereof to retard the fall of said absorber element. 12. The system of claim 10, additionally including retriever means for returning said absorber element into abutment with said up-stop and returning said slide valve to a position for magnetic attraction thereof by said electromagnetic coil. 13. The system of claim 10, wherein said thermionic means consists of a thermionic diode. 14. The system of claim 13, wherein said thermionic diode incorporates a uranium blanket. 15. The system of claim 10, wherein said thermionic means consists of a uranium blanketed thermionic diode containing thermionic material which upon heating to a specified temperature changes from high resistance to low resistance thereby effecting a short-circuiting of said electromagnetic coil causing loss of sufficient magnetic attraction of said slide valve allowing said slide valve to move by gravitational force to block flow through said coolant passage. 16. The system of claim 10, additionally including a plurality of position detection coils positioned on said casing for determining the location of said absorber element. |
abstract | A concentrated irradiation type radiotherapy apparatus comprises a radiation source, a multi-channeled radiation detector, a rotating mechanism, an image reconstruction unit, a multi-leaf collimator disposed between the radiation source and the subject to trim the radioactive rays in arbitrary shapes and including a plurality of first leaves and a plurality of second leaves each disposed to be individually movable forwards/backwards and each having a strip shape and in which types of the first leaves are different from those of the second leaves. |
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055263844 | summary | FIELD OF THE INVENTION The invention relates to a fuelling machine for fuel assemblies for the core of a nuclear reactor. BACKGROUND OF THE INVENTION Nuclear reactors such as pressurized water nuclear reactors include a vessel which contains the core of the nuclear reactor which consists of fuel assemblies and through which, in service, the cooling fluid of the reactor flows. The core of the reactor consists of fuel assemblies which are generally of a right prismatic shape, which rest on a core support plate via their lower part or bottom nozzle and which are placed in a vertical arrangement. The fuel assemblies are juxtaposed and constitute a dense arrangement in which each of the fuel assemblies is in contact with adjacent assemblies, in a lattice arrangement, via its nozzles and its spacer grids. The fissile fuel material contained in the fuel assemblies is progressively consumed in the nuclear reactor in service, so that the fuel assemblies are progressively depleted in fissile fuel material and consequently undergo a form of wear. It is therefore necessary periodically to carry out refuelling operations of the core of the nuclear reactor. These operations, which require shutdown and cooling of the nuclear reactor, are generally carried out on a fraction of the core of the reactor, so as to optimize the use of fuel. Furthermore, when first commissioning a nuclear reactor, it is necessary to fuel the core with new fuel assemblies which constitute the first charge of the nuclear reactor. The operations of fuelling or refuelling a nuclear reactor are carried out under water, with the vessel head being dismounted, from the upper level of the cavity of the reactor, in the bottom of which the reactor pit opens out. In order to carry out fuelling and refuelling operations, use is made of a fuel assembly lifting and handling machine, called a fuelling machine, which includes horizontal guide means arranged above the upper level of the cavity and a carriage mounted movably on the guide means, in at least two directions of a horizontal plane, so as to be capable of placing fuel assembly gripping and lifting means in line with each of the fuel assembly positions in the core of the nuclear reactor. The fuel assembly gripping and lifting means include a tubular external shaft of generally cylindrical shape, fastened on the carriage with its axis vertical, and a cylindrical internal mast in a coaxial arrangement and mounted movably in the axial direction, inside the external shaft. In order to fit the fuel assemblies in the core of the reactor where these assemblies are placed contiguously in a lattice, it is necessary to provide extremely precise handling and lifting means. In particular, the internal mast of the fuelling machine must make it possible to displace the fuel assemblies along a perfectly defined vertical axial direction. For this purpose, the mobile internal mast of the fuelling machine generally includes longitudinal slideways which interact, during displacement of the internal mast, with sets of rollers having axes perpendicular to the axial direction of the external shaft and of the internal mast and which are arranged in a plurality of groups aligned in directions parallel to the axis of the external shaft and of the mobile internal mast. The displacements in the axial direction, i.e. in the vertical direction, of the mobile mast for carrying out the core refuelling operations are of very high amplitude and must be carried out with very high precision with respect to the alignment of the direction of displacement of the mobile mast with the axes of the assembly positions in the core of the reactor being fuelled. The guide groups for the mobile mast which are carried by the external shaft are very long and must therefore have perfectly straight alignment axes which are perfectly defined in orientation and in position. Before start-up of the fuelling machine, or even during use, it may be necessary to check and adjust the alignment of the mobile mast guide means carried by the external shaft. These adjustments may be required, in particular, by loss of adjustment of the roller groups of the guide devices or by deformation of the external shaft, for example due to an impact. This operation, which is carried out in the reactor building, in proximity to the reactor pit, requires scaffolding to be installed in order to be able to access the various roller positions along the external shaft. Adjustment is carried out by means of eccentrics which are arranged at the level of each roller support which is accessible through an opening provided in the external shaft, at the level of the roller support. Such an operation, which includes mounting and dismounting scaffolding, may require a length of the order of 48 hours, during which the fuelling machine is unavailable, which commensurately lengthens the total time of the shutdown of the reactor for refuelling. In addition, the operations of adjusting the guide elements of the mobile mast of the fuelling machine inside the cavity of the reactor require this part of the cavity to be emptied, after fitting a gate. SUMMARY OF THE INVENTION The object of the invention is therefore to provide a fuelling machine for fuel assemblies for the core of a nuclear reactor, inside a reactor vessel which is open at its upper end, including horizontal guide means arranged above the vessel of the reactor, a carriage mounted movably on the guide means, a tubular external shaft fastened on the carriage with its axis vertical, a cylindrical internal mast mounted in a coaxial arrangement and so that it can move in the axial direction inside the external shaft, by virtue of guide elements, and including means of attachment of a fuel assembly in the vertical position and means of displacing the internal mast in the axial direction of the external shaft, which means are carried by the carriage, this fuelling machine making it possible to adjust alignment and straightness of the guide elements of the mobile mast, in a simple manner and outside the cavity of the nuclear reactor, without lengthening the shutdown time of the reactor for fuelling or refuelling. For this purpose, the guide elements for the internal mast include at least two straight beams provided with means for fastening in an axial direction on the exterior surface of the external shaft, on each of which beams at least two sets of rollers which can rotate about an axis perpendicular to the axial direction of the external shaft and of the internal mast are mounted, spaced apart in the axial direction, the external shaft including a through opening for passage of each of the sets of rollers of each of the beams, inside the external shaft in order to guide the internal mast. |
claims | 1. A method for implementing Bragg-diffraction leveraged modulation of X-ray pulses using MicroElectroMechanical systems (MEMS) based diffractive optics comprising:providing an oscillating crystalline MEMS device;providing an incident pulse train of X-ray synchrotron radiation on the oscillating crystalline MEMS device; andselecting pulses with Bragg-diffraction leveraged modulation of the incident pulse train of X-ray synchrotron radiation and generating a controllable time-window of the selected pulses-using the oscillating crystalline MEMS device and diffracting X-ray pulses during an oscillation cycle of the oscillating crystalline MEMS device when the incident pulse train of X-ray synchrotron radiation has an angle of incidence equal to a Bragg angle θB for the oscillating crystalline MEMS device; a width of the controllable time-window determined by an angular velocity of the oscillating crystalline MEMS device; andproviding an angle of incidence equal to said Bragg angle θB for the oscillating crystalline MEMS device for isolating the selected pulses, and angularly separating each of the selected pulses. 2. The method as recited in claim 1, wherein providing incident X-ray radiation on the oscillating crystalline MEMS device includes providing X-ray pulses from a synchrotron radiation source. 3. The method as recited in claim 1 wherein providing an oscillating crystalline MEMS device includes providing a controllably oscillated crystalline MEMS device by providing a selected oscillation frequency. 4. The method as recited in claim 3 includes changing said controllable time-window of selected pulses by providing said selected oscillation frequency. 5. The method as recited in claim 3 wherein providing said selected oscillation frequency includes providing a pair of comb-drive actuators together with respective torsional flexures for driving an X-ray diffractive crystal. 6. The method as recited in claim 1 includes providing a selected oscillation frequency for the oscillating crystalline MEMS device for isolating the selected pulses, and angularly separating the selected pulses. 7. The method as recited in claim 1 includes providing an angle of incidence equal to a Bragg angle θB for the oscillating crystalline MEMS device and providing a selected oscillation frequency for the oscillating crystalline MEMS device for separating a pulse from an X-ray pulse-train and diffracting X-ray pulses during said oscillation cycle of the oscillating crystalline MEMS device when the incident X-ray radiation has said angle of incidence equal to said Bragg angle θB for the oscillating crystalline MEMS device. 8. The method as recited in claim 1 wherein providing said oscillating crystalline MEMS device includes fabricating said oscillating crystalline MEMS device using a Silicon-On-Insulator (SOI) wafer for providing a single-crystal-silicon, and removing a substrate beneath the single-crystal-silicon. 9. The method as recited in claim 8 wherein fabricating said oscillating crystalline MEMS device includes providing a pair of torsional flexures coupled to single-crystal-silicon and anchored to the substrate. 10. The method as recited in claim 9 includes providing a respective pair of comb-drive actuators coupled to the pair of torsional flexures. 11. The method as recited in claim 10 includes providing said comb-drive actuators with inter-digitated capacitors (IDCs). 12. An apparatus for implementing Bragg-diffraction leveraged modulation of X-ray pulses using MicroElectroMechanical systems (MEMS) based diffractive optics comprising:an oscillating crystalline MEMS device;an X-ray source providing an incident pulse train of X-ray synchrotron radiation on the oscillating crystalline MEMS device; andsaid oscillating crystalline MEMS device selecting pulses with Bragg-diffraction leveraged modulation of the incident pulse train of X-ray synchrotron radiation and generating a controllable time-window of the selected pulses-and diffracting X-ray pulses during an oscillation cycle of the oscillating crystalline MEMS device when the incident pulse train of X-ray synchrotron radiation has an angle of incidence equal to a Bragg angle θB for the oscillating crystalline MEMS device; a width of the controllable time-window determined by an angular velocity of the oscillating crystalline MEMS device; andsaid oscillating crystalline MEMS device provides Bragg-diffraction leveraged modulation of X-ray pulses including isolating a pulse, and angularly separating individual pulses from an X-ray pulse-train and diffracting X-ray pulses during the oscillation cycle of the oscillating crystalline MEMS device when the incident X-ray radiation has said angle of incidence equal to said Bragg angle θB for the oscillating crystalline MEMS device. 13. The apparatus as recited in claim 12 wherein said oscillating crystalline MEMS device includes a Silicon-On-Insulator (SOI) wafer including a single-crystal-silicon forming an X-ray diffractive crystal, a substrate being removed below the single-crystal-silicon. 14. The apparatus as recited in claim 13 wherein said oscillating crystalline MEMS device includes a respective pair of torsional flexures coupled to said single-crystal-silicon and said substrate. 15. The apparatus as recited in claim 14 wherein said oscillating crystalline MEMS device includes a respective pair of comb-drive actuators coupled to the pair of torsional flexures, said comb-drive actuators including inter-digitated capacitors (IDCs). 16. The apparatus as recited in claim 12 wherein said X-ray source includes a synchrotron radiation source providing X-ray pulses to said oscillating crystalline MEMS device. |
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description | The present invention relates to a container. More particularly this invention concerns a container for holding and transporting radioactive waste such as spent fuel rods. A container for holding radioactive waste, particularly for holding spent fuel elements, typically has a side wall, a floor connected to the side wall, and at least one cover. The invention further relates to a container assembly of a canister and a transport and/or storage container. Containers of the above-described type are known from practice in diverse variants. For instance, canisters for the transfer of spent fuel elements are known in particular. These canisters are loaded under water with the spent fuel elements and then sealed with a cover. These sealed canisters are then transferred to a transport and/or storage container. Due to the vertical handling of a canister upon insertion into the transport and/or storage container, the load attachment point must be on the upper side of the canister cover and be designed to be loaded by the canister mass, fuel element support basket, and fuel elements. For this reason, a multilayer, massive force-transmitting weld is provided between the cover and side wall of the canister as a load-bearing component of such a canister. In addition, a weld ring is also provided as a seal weld. The production and reproducibility of welds must be ensured by an appropriate inspection. Different national regulations exist for this purpose. In countries in which the canisters cannot be used for interim storage or permanent storage, the welds on the canisters must be ultimately reopened and the fuel elements transferred. As a result, the effort and expense associated with the measures described are substantial. Furthermore, it is known to provide transport and/or storage containers first with a primary cover and then with a secondary cover. After loading a container, the primary cover is fixed in place by screws with interposition of metal gaskets. The screws must dissipate the inertial forces acting on the cover that are caused by the mass of the cover, the mass of the support basket, and the mass of the fuel elements. In order to ensure the sealing and compression of the metal gaskets, elastic loading of the screws is required, and the screws are subjected to bending stress in addition to the purely tensile load. Due to the high loaded masses and accelerations, a relatively large number of cover screws is required. The number of cover screws is structurally limited due to the installation space required for the screw heads, so the load-bearing cross section of the screws is limited. It will readily be understood that, as the number of screws increases, the effort and expense involved in servicing increases. The known measures have inherently proven their worth. Nevertheless, they are relatively costly. It is therefore an object of the present invention to provide an improved container for holding radioactive waste. Another object is the provision of such an improved container for holding radioactive waste that overcomes the above-given disadvantages, in particular that enables the container to be sealed in a simple and inexpensive manner while nonetheless providing an optimal sealing function. Another object of the invention is to provide a corresponding container assembly. A container for holding radioactive waste has according to the invention a side wall, a floor connected to a lower end of the side wall, and a cover. A set of side-wall formations is provided at an upper end of the side wall and on an inner surface of the side wall, and a set of cover-edge formations is distributed around an outer edge of the cover and fittable with the side-wall formations. Thus, as a result of the interfitting of cover-edge formations with the side-wall formations, the cover can be or is fixedly connected to the side wall without welds. It therefore lies within the scope of the invention for the cover to be fixedly connected to the formations of the side wall as a result of the interfitting of its complementary formations when the container is in the closed state. It also lies within the scope of the invention for a support basket for holding spent fuel elements to be in the interior of the container according to the invention. Advantageously, the floor and the side wall are integrally connected to one another. According to another alternative of the invention, the floor and the side wall of the container are interconnected by at least one weld. According to a very preferred embodiment of the invention, the container according to the invention is a canister loaded with the fuel elements and is then introduced into a transport and/or storage container. As a rule, the canister is loaded with the fuel elements under water and sealed with the cover. The canister is then transferred by a transfer container to the transport and/or storage container and introduced there into the transport and/or storage container. According to another embodiment of the invention, the container is a transport and/or storage container and the cover is then the primary cover of the transport and/or storage container, the primary cover being equipped with the complementary formations. According to the invention, a form-fitting connection is established between the cover and the side wall, particularly without welds. At the same time, it lies within the scope of the invention for the cover to be moved from an open position into a locking position through rotation. Advantageously, the rotation can also be reversed from the locking position to the open position. Accordingly, the container can also be easily opened again. It is recommended that the rotation of the cover from the open position to the locked position and vice versa occur over the smallest possible angle. In order to achieve the form-fitting connection between the cover and the side wall, at least three, at least four, more at least five, and very at least six formations are advantageously distributed in the interior of the container around the inner surface of the side wall. One especially recommended embodiment of the invention is characterized in that at least eight, advantageously at least ten, and at least twelve formations are in the interior of the container so as to be distributed around the inner surface of the side wall. Advantageously, at least three, at least four, more at least five, and very at least six complementary formations are distributed around the outer edge of the cover. It is especially preferred in the context of the invention for at least eight, at least ten, and especially at least twelve complementary formations to be provided in such a manner as to be distributed around the outer edge of the cover. It lies within the scope of the invention for the number of formations on the inner surface of the side wall to correspond to the number of complementary formations on the outer edge of the cover. One very preferred embodiment of the invention is characterized in that the side-wall formations and/or the complementary cover-edge formations are formed as bumps, more particularly as projections. These projections can either be welded on or machined from the corresponding container material. It is advantageous for either the side-wall formations or the complementary cover-edge formations to be projections. It has been found to be advantageous in the context of the invention if the formations that are distributed in the interior of the container around the inner surface thereof are bumps, more particularly projections. The projections are then either welded to the upper cover-side edge region of the side wall or machined from the side wall material. It is recommended that the projections have a rectangular or substantially rectangular shape when viewed from above. According to a design variant, the projections are rectangular in shape with rounded corners. It lies within the scope of the invention if such a projection can be fitted positively in a groove described below and, particularly, in a portion of such a groove. One especially recommended embodiment of the invention is characterized in that the side-wall formations and/or the complementary cover-edge formations are formed as grooves. It is advantageous for either the side-wall formations or the complementary cover-edge formations to be grooves. It has been found to be advantageous in the context of the invention if the complementary formations that are distributed around the outer edge of the cover are grooves. A preferred embodiment is characterized in that a groove has at least two groove portions that extend perpendicular or substantially perpendicular to one another. Advantageously, at least one groove or one portion of each groove is parallel or substantially parallel to the longitudinal axis L of the container, and at least one additional groove or one additional groove portion is perpendicular or substantially perpendicular to the longitudinal axis L of the container. An L-shaped or substantially L-shaped configuration of the grooves is especially preferred in the context of the invention. One groove portion is advantageously oriented parallel or substantially parallel to the container axis L, and the other groove portion is perpendicular or substantially perpendicular to the axis L of the container. It is especially recommended in the context of the invention that a groove or a groove portion be a tangential. “Tangential groove portion” refers here to a groove portion that, given a cover having a round or circular cross section, would form a tangent or approximately a tangent in relation to the round or circular cover if it were extended. Advantageously, the groove portion oriented perpendicular or substantially perpendicular to the longitudinal axis L of the container is a tangential groove portion. Incidentally, the expressions “perpendicular to the longitudinal axis L of the container” and “parallel to the longitudinal axis L of the container” refer particularly to the locking position of the container with cover already inserted. It is recommended that the groove portion parallel to the longitudinal axis L of the container be open toward the underside of the cover, so that as described below these groove portions can be pushed onto the projections projecting from the interior of the side wall. Preferably, both the groove portion oriented parallel to the longitudinal axis L of the container and the L-portion oriented perpendicular to the longitudinal axis L of the container are upwardly closed, in which case a groove abutment edge or groove abutment surface is realized on the upper side of the cover. A recommended embodiment of the invention is characterized in that a plurality of complementary formations in the form of grooves, at least eight, more at least ten, and very at least twelve grooves, are distributed around the outer edge of the cover. It is recommended that in this embodiment the side wall have a plurality of formations in the form of projections, particularly in the form of rectangular projections, that are distributed around its inner surface. Advantageously, at least eight, at least ten, and more at least twelve formations or projections are provided on the inner surface of the side wall. It lies within the scope of the invention for the number of formations on the side wall to correspond to the number of complementary formations on the cover. Advantageously, the formations on the inner surface of the side wall are distributed uniformly around the inner surface of the side wall, specifically at equal or substantially equal distances from one another. It is recommended that the complementary formations that are provided on the outer edge of the cover be distributed uniformly around the outer edge of the cover, specifically and at equal or substantially equal distances from one another. It was already noted that the complementary formations that are distributed around the outer edge of the cover have L-shaped grooves and that, advantageously, the groove portion is oriented parallel to the longitudinal axis L of the container, it being recommended that this groove portion be open toward the underside of the container. In order to seal the container with the cover, the cover is inserted into the side wall with the understanding that the grooves that are distributed around the outer edge of the cover, particularly the groove portions that are distributed around the outer edge of the cover and oriented so as to be parallel to the longitudinal axis L of the container, are each pushed onto a projection provided in the interior of the container. Each of the projections comes to rest against the upper-side groove abutment edge or groove abutment surface of the parallel groove portions of the cover. When the cover is twisted, each of the projections that are present on the inner surface of the side wall then comes into engagement with the groove portion perpendicular to the longitudinal axis L of the container. When the webs are in engagement with, more particularly in full engagement with this latter groove portion, the cover is in its locking position. Preferably, the number of formations and complementary formations is selected such that the torsion angle during the movement of the cover from the opening position to the locking position is as small as possible and less than 25°, more less than 20°, and very less than 15°. It lies within the scope of the invention for the projections, particularly the projections on the inner surface of the side wall, to abut against a groove abutment edge or groove abutment surface of a groove, of an L-shaped groove, on the upper side of the container. The cover is supported in the axial direction or in the direction of the longitudinal axis L of the container of the container in a functionally reliable manner on the container, thus enabling a load attachment point to be effectively realized on the cover. It is recommended that the groove abutment edges or groove abutment surfaces be provided on both groove portions of the L-shaped grooves. One highly recommended embodiment of the invention is characterized in that the grooves, the grooves on the outer edge of the cover, each have at least one groove region whose groove base on the bottom side of the container is angled toward the interior of the container and the bottom of the container. It is recommended that, in the case of L-shaped grooves, at least the groove portion oriented so as to be perpendicular to the longitudinal axis L of the container has a groove lower face on the bottom side of the container angled toward the interior of the container and the floor. It lies within the scope of the invention for the projections on the inner surface of the side wall to have complementary angles. This configuration with the inclined surfaces ensures that, if the container or the cover is lifted when in the locking position, the forces acting radially on the side wall can be reduced, thus preventing spreading of the side wall. According to the invention, a form-fitting connection occurs between the cover and side wall. It lies within the scope of the invention for sealing measures for a functionally reliable seal between the cover and side wall to be additionally implemented. A preferred embodiment of the invention is characterized in that, when the container is in the locking position, at least one sealing groove that runs around the periphery of the container is provided on the upper side of the container between the cover and the side wall. It is recommended that the sealing groove be between the upper edge of the side wall and the outer edge of the cover, the cover, more particularly the outer edge of the cover, having a recess on the upper side. It has proven advantageous if the lower face of this recess is frustoconically angled toward the side wall. Advantageously, at least one seal, more particularly at least one sealing ring, is fitted in or inserted into the sealing groove. It lies within the scope of the invention for the at least one seal to extend annularly around the periphery of the container. It is recommended that the at least one seal, more particularly at least one seal ring, be a metal gasket and/or an elastomer seal. In the context of the embodiment in which the container is a canister that can be transferred to a transport and/or storage container, at least one elastomer seal can first be inserted into the sealing groove after the canister is loaded with the fuel elements under water. During further handling, the elastomer seal can be removed, and then at least one metal gasket can be advantageously used for the seal between cover and side wall. If the cover is a primary cover of a transport and/or storage container, at least one metal gasket is used as a seal. According to an especially recommended embodiment of the invention, a metal gasket used in the context the invention has a core ring made of at least one metal and at least one jacket enclosing the core ring made of at least one metal. Advantageously, such a metal gasket or such a metal sealing ring is provided with a core ring made of at least one metal, a nickel alloy or a nickel-based alloy, and with an inner jacket made of at least one metal and at least one outer jacket mounted thereon made of at least one metal. The inner jacket is advantageously made of steel and of stainless steel. One embodiment of the invention is characterized in that the outer jacket consists or substantially consists of a metal from the group aluminum, silver, gold. Such a metal seal or such a metal sealing ring has proven to be very especially useful in combination with the form-fitting measures according to the invention. It lies within the scope of the invention for at least one compression element, particularly at least one compression ring to be placed onto the cover when the container is in the locked state or in the locking position. Advantageously, the compression element or the compression ring compresses the at least one seal in the sealing groove. This compression results in an outstanding sealing function. According to a preferred design variant of the invention, the compression element or the compression ring is fixed in place by screwing to the container, particularly to the cover. Advantageously, the compression element or the compression ring runs around the periphery of the container or cover. According to the recommended embodiment of the invention, the compression element or the compression ring has a ridge on the cover side of the container that engages in the sealing groove in order to compress the seal. It lies within the scope of the invention for this ridge to run around the periphery of the container. Recommendably, the ridge has a frustoconical lower face angled upward and toward the side wall. With the angled face provided on the groove-type recess base of the groove-type recess or sealing groove, this angling enables an especially effective compression of the seal and hence a very functionally reliable radial seal between cover and side wall to be achieved. The invention also relates to a container assembly comprising a container according to the invention, the container being a canister loaded with spent fuel elements and received in a transport and/or storage container that can be or is sealed with at least one primary cover and can be or is advantageously additionally sealed with at least one secondary cover fitted over the primary cover. The canister is advantageously loaded under water with the spent fuel elements, the fuel assemblies being received by a support basket in the canister. The canister is then transferred by a transfer container to the transport and/or storage container and introduced there into the transport and/or storage container. The transport and/or storage container is then sealed with the primary cover and the secondary cover and fed to interim or permanent storage. According to another embodiment of the invention, the container according to the invention is already a transport and/or storage container, in which case the cover is the primary cover of the transport and/or storage container. Advantageously, the primary cover is fixed in a form-fitting manner to the side wall of the transport and/or storage container by the measures according to the invention. After the form-fitting positioning of the primary cover, the transport and/or storage container is advantageously sealed by an additional secondary cover. The secondary cover is fixed to the container by screws. The invention is based on the discovery that the container according to the invention can be sealed in a simple, inexpensive, and functionally reliable manner by the form-fitting measures according to the invention without welds. If the container according to the invention is a canister for transferring the spent fuel elements, a simple closure mechanism can be realized without a load-bearing weld that nevertheless easily ensures optimal bearing capacity of the canister during the corresponding manipulations. The cover provides an effective load attachment point even with larger masses to be transported and under greater loads. Due to the lack of a load-bearing weld, a costly country-specific process qualification of the welding process by experts is unnecessary. It should be emphasized that the container or the canister can be easily sealed and also reopened without any difficulty. In principle, any closing and opening of the container is readily possible. This is advantageous particularly if the canister is not intended for interim storage or permanent storage. The inventively designed cover can still be used without restriction when loading a canister under water. In fact, production is simplified, production time is shortened, and the risk during production is reduced. It should also be emphasized that, in comparison to the conventional canisters for the implementation of the measures according to the invention, no substantial alterations of the canister body are required. If the container is a transport and/or storage container and the cover is the primary cover of this container, elaborate and time-consuming screw attachment of the primary cover using cover screws can be dispensed with. Thus, no dissipation of inertial forces via cover screws is required. The load-bearing cross sections can be increased substantially when implementing the measures according to the invention as opposed to screws. Apart from that, the outer diameter of the primary cover, and hence the outer diameter of the secondary cover as well, can be reduced. This has the advantage that a larger residual wall thickness of the container body or side wall in the cover region is possible. This increases stability, among other things, in the event of a lateral impact to the container. The use of a seal in the groove between cover and side wall is also especially advantageous in a container according to the invention. It is possible here to replace the seal without disassembling the cover. The compression state of the seal is advantageously independent of the locking of the cover. By virtue of the radial seal in the measures according to the invention, no impairment of tightness occurs as a result of transverse displacements. As a result, the measures according to the invention provide an alternative canister closure or container closure that on the one hand offers the same advantages as the closures known hitherto from the prior art, in particular, mechanical stability and leak-tightness, but on the other hand has additional substantial advantages over them. As seen in drawing, a container 1 according to the invention for holding radioactive waste, in this embodiment for holding spent fuel elements 2, has a cylindrically tubular side wall 3, a flat planar floor 4 connected to a lower edge of the side wall 3, and a cover 5 sealing the open upper end of the side wall 3. The side wall 3 is centered on a normally upright axis L, and the floor 4 and cover 5 are axially spaced and extend perpendicular to the axis L. In this embodiment a support basket 11 for holding the spent fuel elements 2 is provided in the interior 10 of the container 1. The floor 4 and the side wall 3 may be unitarily connected to one another. According to the invention, the side wall 3 has an annular array of formations 7 inside the container in its upper cover-side edge region 6 that are distributed around the inner surface of the side wall 3, and the cover 5 is provided on its outer edge with a complementary array of radially outwardly directed formations 9 that are angularly uniformly distributed around its outer edge. In the locking position of the container 1, the cover 5 is fixedly attached to the side wall 3 due to the interfitting of the cover-edge formations 9 with the formations 7 of the side wall 3. According to the invention, however, the connection between cover 5 and side wall 3 is free of welds. A plurality of the side-wall formations 7, in this embodiment at least twelve side-wall formations 7, are arrayed around the inner surface of the side wall 3. An identical plurality of complementary cover-edge formations 9, at least twelve complementary cover-edge formations 9, are arrayed around the inner periphery of the cover 5. In this embodiment the side-wall formations 7 that are inside the container on the inner surface of the side wall 3 project radially inward, and are rectangular when viewed from above and from radially inside. The complementary cover-edge formations 9 that are distributed on the outer edge 8 of the cover 5 around the outer surface thereof are L-shaped grooves that each have a portion 20 extending parallel to a longitudinal axis L of the container as well as a portion 21 perpendicular thereto or perpendicular to the longitudinal axis L of the container. These portions 21 extend tangentially on the outer edge of the cover 5. The portions 20 parallel to the longitudinal axis L of the container open downward on the edge of the cover 5, so that the formations/projections 7 can be inserted upward into them. In order to move the container 1, more particularly the cover 5, from an opening position to the locking position, the projections 7 of the side wall 3 engage in the formations/grooves 9 of the cover 5. Advantageously, the cover 5 is first fitted into the upper end of the side wall 3 so that the projections 7 engage axially upward into the vertical portions 20 of the grooves 9 until the projections 7 abut downwardly facing upper edges 15 of the portions 21 of the grooves 9. The cover 5 is then rotated relative to the side wall 3 so that the projections 7 are trapped in the groove portions 21 that extend perpendicular to the longitudinal axis L of the container. In this position, the projections 7 abut against the groove abutment edges 15 on the upper side of the container, so that the cover 5 is held captive on the container 1 axially of the container 1. Since the angular depth of the tangential groove portions 21 past the downwardly open vertical groove portions 20 is roughly equal to the angular dimension of the rectangular projections/formations 7, there is considerable load-bearing capacity. Especially in this embodiment as shown in FIG. 3 the portions 21 of the grooves 9 have upwardly directed lower faces 16 that are angled downward toward the interior 10 of the container and the floor 4. It lies within the scope of the invention for the projections 7 to have complementarily angled lower faces. Each of the projections 7 also has a complementarily angled face. The engagement of the projections 7 in the groove portions 21 prevents the side wall 3 from spreading as a result of loads during vertical handling of the container 1. In fact an upwardly directed vertical force applied to the cover 5 will pull in the upper edge of the side wall 3. It can be seen particularly in FIGS. 3 and 4 that a sealing groove 17 runs around the periphery of the container 1 on the upper end of the container 1 between the cover 5 and the upper edge of the container 3 when the container 1 is in the locking position. The cover 5 has an annular upwardly open recess 25 that runs around its edge, more particularly where its outer edge and upper surface meet to form this sealing groove 17. At least one seal 18 is in the sealing groove 17, in this embodiment is a seal ring or O-ring. Advantageously, this seal 18 is a metal gasket. In this embodiment this recess 17 of the cover 5 has a frustoconical and upwardly directed lower face 26 angled downward and outward toward the side wall 3. The seal 18, more particularly the sealing ring, advantageously rests against this base face 26. A compression ring 19 sits on top of the cover 5 when the container 1 is in the locking position to compresses the seal 18 in the groove 17. The compression ring 19 is fixed to the cover 5 by screws. It also lies within the scope of the invention for the compression ring 19 extends around the edge of the container 1 and the cover 5. The compression ring 19 has a downwardly projecting annular ridge 27 on its outer edge that advantageously comes into direct contact with the seal 18. In this embodiment, this annular ridge 27 of the compression ring 19 has a frustoconical lower face 28 on its underside that is angled upward and outward toward the side wall 3, oppositely to the groove face 26. By virtue of this arrangement, an effective compression of the seal 18 takes place and an optimal radial sealing is achieved because the seal 18 is compressed axially by the faces 26 and 28 and forced radially outward into good contact with the inner face of the side wall 3. It is also recommended means be provided between the compression ring 19 and the side wall 3 advantageously in the form of fitted keys or wedges or similar elements not shown in the figures that fit into complementary grooves of the compression ring 19 on the one hand and of the side wall 3 on the other hand. This means prevents rotation of the ring 19 relative to the cover 5 and/or side wall 3. FIG. 4 shows the seal 18 used according to a preferred embodiment that here is a metal gasket. This metal gasket has a core ring 22 made of at least one metal, a nickel alloy or a nickel-based alloy. Alternately, the seal 18 has a tubular inner jacket 23 that consists at least substantially of stainless steel as well as a tubular outer jacket 24 that in this embodiment consists or substantially consists of aluminum. This outer jacket 24 could also be a coating of a noble metal such as silver or gold. |
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047754940 | summary | BACKGROUND OF THE INVENTION The disposal of hazardous and radioactive waste materials is of extreme importance. Federal and state laws and requirements covering such disposals are particularly severe and stringent due to the dangers to plant and animal life if the desired standards are not met and the hazardous or radioactive materials become exposed to the environment. Because of the potential dangers, the U.S. Nuclear Regulatory Commission has not only identified the hazardous and radioactive materials to date, which list is continually being amended and updated, but has set forth specific standards and requirements for protecting the environment against such waste materials. The resulting laws and regulations are set forth in 10 CFR, particularly sections 1-199. Other regulations relating to transportation, packaging, labeling and identifying hazardous and radioactive materials are also found in 40 CFR 1-799 and 49 CFR 100-177. Other publications which relate to classifying, indexing and discussing radioactive and hazardous waste materials include DOE/LLW-14T publication "Waste Classification, A Proposed Methodology For Classifying Low-Level Radioactive Waste", December 1982, DOE/LLW-17T, "Survey Of Chemical And Radiological Indexes Evaluating Toxicity", March 1983, FW-874, "Hazardous Waste Land Treatment", April 1983 and FW-872 "Guide To The Disposal Of Chemically Stabilized and Solidified Waste", September 1982. It is the common practice to process liquid hazardous or radioactive materials by adding adsorbents in an attempt to enhance handling and transportation, as well as eventual storage thereof. The materials that have been used heretofore include diatomaceous earth, vermiculite or expanded mica such as zonolite and krolite, portland and gypsum cements, as well as clay materials such calcium bentonites. A problem with such materials is that only a relatively small amount of liquid can be absorbed or otherwise treated with less than satisfactory results. For example, liquid materials are desirably transported and disposed of in 55 gallon drums. However, it has been found with the use of these adsorbents, solid compositions cannot be achieved or if temporarily achieved, liquid separation occurs during transportation or storage. Any separated or free-standing liquids are especially undesirable because of the potential danger of leakage from a ruptured or opened container. It is to the substantial elimination of such problems that the present invention is directed. SUMMARY OF THE INVENTION An improved method of treating hazardous and radioactive liquid waste materials comprises placing the materials in a container, such as a 55 gallon drum, and slowly adding sodium montmorillonite until the mixture has substantially solidified. The resulting composition may be handled, transported and stored under a variety of conditions for extended periods of time without evidence of liquid separation or deterioration. These and other advantages as well as the specific sodium montmorillonite compositions used in the invention will be more particularly described in the following detailed description. |
description | This application is the 35 U.S.C. §371 national stage of PCT application PCT/IB2012/051904, filed Apr. 17, 2012 which claims priority to Italian Patent Application No. BS2011A000061, dated Apr. 26, 2011, both of which are incorporated by reference in their entirety. The present invention relates to a device for sterilising thin wall containers, in particular flexible containers, such as those for containing dense fluids especially foods, such as creams, yoghurt, honey, fruit juices, medicines and the like. In the food industry, the sterilisation of containers is extremely important for preventing infections and preserving the food contained therein correctly. Sometimes, chemical sterilisation is performed, during which the container is washed with disinfectants, such as hydrogen peroxide, and then dried, before being sent for subsequent filling operations. However, chemical sterilisation has some drawbacks such as for instance the presence of residues of the chemical disinfectant in the dry container or the presence of areas which have not been disinfected on account of complicated or irregular geometries of the container. Such drawback is particularly felt in the field of thin-walled flexible containers. Electron beam sterilisation is becoming increasingly widespread. Initially, the performance of electron beam sterilisation was restricted to specialised centres, which the containers to be treated had to be sent to and from which the sterilised containers were picked up, with a considerable increase in transport costs and logistics. In such centres high powered (500 kV-10 mV) electron cannons were usually utilised, with all the relative consequences on operator safety and environmental pollution. Recently, electron beam sterilisation is becoming increasingly popular, thanks to the creation of particularly compact electron cannons functioning efficiently even at low voltages (80-150 kV). Such cannons permit the performance of electron beam sterilisation directly in the container production plant, with notable economic savings. The purpose of the present invention is to make a low voltage electron beam sterilisation device particularly suitable for treating thin-walled, flexible containers. According to the drawings, reference numeral 1 globally denotes a low voltage electron beam sterilisation device for thin-walled, flexible containers. In particular, the device 1 is suitable for sterilising containers C consisting of a body B formed of two B′, B″ or more walls of flexible film, facing one another and joined, for example welded along the edges, if required with gusset side walls G, and provided with a straw A in rigid material, fitted in a section of the edge of the body B, usually between the side walls. The straw A projects from the body along a straw axis and may be coupled to a cap, also in rigid material. When the flexible container is just made and empty, the body is particularly thin (FIG. 2), while it appears to bulge when filled (FIG. 1). For the empty containers C, a height H is defined along the straw axis, a width W transversal to the height H, and a thickness T of the thin wall. One embodiment example of such containers is shown in documents EP-A1-1538105 and U.S. Pat. No. D-552,483, in the Applicant's name; one embodiment example of a straw with cap is shown in the document WO-A1-2008-050361, also in the Applicant's name. The sterilisation device 1 comprises a sterilisation group 20, in which the sterilisation by means of low voltage electron beams takes place, an input unit 40 for the introducing the containers to be treated in the sterilisation group 20. and an output unit 60 for the exit of the treated containers from the sterilisation group 20. Said input unit 40 and said output unit 60 also form an obstacle to the leakage of radioactive emissions, and in particular of the X-rays produced by the electron beam inside the sterilisation group 20, at its input and output. The input unit 40 comprises an outer casing 42 provided with a front wall 44a which an entrance 46a is made in for the entrance of the containers C to be treated, coming from machines upstream of the sterilisation device 1. Similarly, the outer casing 42 is provided with a back wall 44b which an exit 46b is made in for the entrance of the containers C to be treated in the sterilisation group 20. The outer casing 42 further comprises a bottom 48 and a cover, preferably fitted with a removable lid 52 for access to the inside of the casing. The outer casing 42 further comprises side walls 54, preferably polygonal-shaped. The input unit 40 further comprises a swivelling body 56, housed at least partially in the compartment inside the casing 42, swivelling on command in alternate directions or always in the same direction, around a rotation axis K. To such purpose, the input unit 40 is fitted with drive means, such as an electric motor 59, preferably of the brushless type connected to the swivelling body 56, preferably positioned under the bottom 48 of the casing 42. For example, the motor 59 is connected to the swivelling body 56 by means of a shaft 60 which crosses the bottom 48 of the outer casing. The swivelling body 56 has at least one loading seat suitable for receiving at least one container to be treated. In particular, the loading seat can be aligned with the entrance 46 of the casing 42 to enable the loading of containers to be treated. Preferably the swivelling body 56 has two loading seats 58, for example diametrically opposite each other, which can be alternately aligned with the entrance 46a. In addition, by rotation on command, the loading seat 58 can be aligned with the exit 46b of the casing for dropping the containers to be treated into the sterilisation group 20. Preferably when one loading seat is aligned with the exit 46b, another loading seat is aligned with the entrance 46a, so that the loading of containers to be treated from the entrance 46a can take place simultaneously with the emptying of other containers from the exit 46b. The swivelling body 56 further presents filling portions, outside the loading seats, which occupy the operating area of the swivelling body rotating in the casing, so as to prevent or limit as far as possible the presence of leaks from inside the sterilisation group 20 outwards through the input unit 40. For example, the swivelling body 56 is a solid cylindrical body in which loadings seats 58 are made, generally diametrically opposite each other, having a radial extension from the periphery towards the inside of the cylindrical body, so as to determine solid portions 62 of operating space in the form of cylindrical segments, which brush the surface of the inner compartment of the casing 42. The containers to be treated pass into the sterilisation group 20 from the input unit 40. To such purpose, the sterilisation device 1 comprises extraction means, suitable for extracting the containers housed in the loading seat 58 facing the exit 46b of the input unit 40. According to a preferred embodiment, the extraction means comprise an extraction guide, for example formed of an extraction rail 140, which extends in an input direction X. For example, the input direction X comes out of the loading seat 58 when aligned with the exit 46b, and preferably, is a rectilinear direction. The containers are suspended from the extraction rail 140 by means of the respective straws A and the extraction means are suitable for pushing the straws along the input direction X, in a direction of advancement IN. To such purpose, for example, the extraction means comprise a pair of pushers provided with fingers 142 positioned on both sides of the extraction rail 140, spaced out along the input direction X, staggered along said input direction. Preferably the pushers are driven by brushless motors. The pushers are suitable for moving the fingers 142 in translation along the input direction X in the direction of advancement IN so as to engage and push the straws, for example corresponding to a group of containers. In addition, the pushers are provided with a lifting and return movement, which is implemented when a pusher has completed its forward stroke, so as to pass over the row of hanging containers and go back, to extract further containers housed in the loading seat 58 of the loading unit 40 and make them advance. In short, according to the embodiment illustrated said extraction means perform a “jumping” movement in which the fingers of the two pushers alternate in pushing each group of containers along the input direction X. According to a preferred embodiment, the device 1 comprises a pre-sterilisation chamber 50, shielded from the leakage of X-rays, extending in an input direction X, positioned downstream of the loading unit 40 and upstream of the sterilisation group 20. For example, the extraction means are housed in the pre-sterilisation chamber. The sterilisation group 20 comprise an outer casing 22 which internally defines a sterilisation chamber, comprising a main chamber 24. In particular, a side wall 26 of the outer casing 22 presents the access for the containers coming from the input unit 40, moved by the extraction means. The main chamber 24 extends mainly in a first sterilisation direction Y, preferably rectilinear. Preferably the first sterilisation direction Y is inclined in relation to the input direction X, preferably orthogonal to it. In addition, the pre-sterilisation chamber 50 helps to prevent the leakage of X-rays from the input unit 40. The sterilisation group 20 further comprises at least one electron cannon suitable for emitting a cloud of electrons to sterilise the containers. For an electron cannon, an emission cone of the electronic cloud and an emission axis which defines such emission cone is defined. For example, the group 20 comprises two electron cannons 28a, 28b, having respective emission axes E1, E2. Preferably the emission axes E1, E2 lie on the same horizontal plane or on planes parallel to the horizontal plane. The cannons 28a, 28b are positioned in sequence in the first sterilisation direction Y, facing each other in the main chamber 24, positioned opposite one another. The cannons 28a, 28b are positioned in sequence in the sense that the containers are struck first by the electronic cloud emitted by the first cannon and then by the electronic cloud emitted by the second cannon, or, in a transition zone only, they are simultaneously subject to the electronic cloud of the previous cannon and to the cloud of electrons of the next cannon. The sterilisation group 20 further comprises support means of the containers along the main chamber 24 suitable for supporting said containers in the first sterilisation direction Y. For example, said support means comprise a first rail 30 which extends in the first sterilisation direction Y, which the containers are hung from, for example by means of the straw A. The sterilisation group 20 further comprises first means of advancing the containers suitable for moving the containers in the first sterilisation direction. To change direction from the input direction X to the first sterilisation direction Y, the first extraction means act in conjunction with the means of advancement. In particular, for example, the extraction means comprise a drum cam, positioned at the end of the pair of rails 140, suitable for pushing the containers into the main chamber 24, where they engage with the first means of advancement. For example, said first means of advancement comprise a rotor element 32 which extends with its axis along the first sterilisation direction Y, rotatable so as to push the containers in said direction Y. For example, the rotor element 32 surmounts the first rail 30 and engages the top of the straw A of the container C, projecting from the first rail 30, to push the container in the first sterilisation direction Y, Preferably, in addition, the sterilisation group 20 further comprises guide means suitable for guiding the body of the containers in the first sterilisation direction Y. For example, said guide means comprise a pair of threadlike guides 34 extending in the first sterilisation direction Y, positioned below the first rail 30 and spaced out, so that the bodies B of the containers C are placed between them. Given the closeness of the guides 34, the oscillation of the containers is thereby limited or prevented. The sterilisation chamber of the sterilisation group 20 further comprises a secondary chamber 36 which extends in a second sterilisation direction Z, incident to the first sterilisation direction Y, for example, preferably, orthogonal to it. The rotor element 32 pushes the containers in the first sterilisation direction Y, as far as the end of the first rail 30, where the first means of advancement co-operate with the second means of advancement to change the direction of advancement from the first sterilisation direction Y to the second sterilisation direction Z. The second means of advancement are suitable for moving the containers in the second sterilisation direction. In addition, the sterilisation group 20 comprises 10 second support means suitable for supporting the containers in the second sterilisation direction Z. For example, the second support means comprise a second rail 150 which extends from the zone where the first rail 30 ends, in the second sterilisation direction Z, and is suitable for hanging the containers C from, for example by means of the straw A. In the transition zone between the first rail 30 and the second rail 150, the second means of advancement comprise a pusher 152 suitable for moving in alternated translatory movement in the second sterilisation direction Z. The movement of the pusher 152 is synchronised with the movement of the rotor 32, for example by means of a rotating cam 154, connected by a belt or chain to the rotor element 32. When the container C is left by the rotor element 32 at the end of the first rail 30, the second pusher 152 pushes said container along the second rail 102. The containers hanging from the second rail 150 proceed in the second sterilisation direction Z as a result of the subsequent insertion of a further container in the tail, that is to say so that the subsequent container pushes the tail of the containers before it forward in the direction of advancement. According to one embodiment variation (not shown), the second means of advancement are suitable for separately engaging sets of containers, to make them advance in the second sterilisation direction Z; For example, said second means of advancement are structurally and functionally similar to said first means of advancement. The sterilisation group 20 further comprises a further electron cannon 28c, such as a third cannon 28c positioned along the second sterilisation direction Z, having an emission axis E3 incident to the horizontal plane, preferably orthogonal to it. In other words, the third cannon is positioned so that the emission axis E3 is parallel to the axis of the straw A of the container being treated. It is therefore clear that the first sterilisation direction Y and the second sterilisation direction Z together define a sterilisation path along the sterilisation chamber. The sterilisation device 1 further comprises an output unit 60, joined to the secondary chamber 36, downstream of the third cannon 28c in relation the direction of advancement of the containers being treated. Preferably the output unit 60 is functionally and structurally similar to the input unit 40. The treated containers pass from the sterilisation group 20 and in particular from its secondary chamber 36 to the output unit 60 and from this outside the device 1. The sterilisation device comprises loading means suitable for loading a predefined number of containers C from the secondary chamber 36 to the loading seat of the input unit. Preferably, said loading means comprise a pusher with a “jumping” movement similar to that described above. According to a preferred embodiment, the device 1 comprises a post-sterilisation chamber 55, shielded from the leakage of X-rays, extending in the second sterilisation direction Z, positioned downstream of the sterilisation group and upstream of the output unit 60. For example, the loading means are housed in the post-sterilisation chamber. The post-sterilisation chamber also helps to prevent the leakage of X-rays from the output unit 60. The containers C1 entering the input unit 40 are aligned one after another with the walls facing each other; the containers C1 are facing in columns that is. The containers C1 are moved so that one or more containers C2 are housed in the loading seat 58 of the input unit 40. When the number of containers housed in the loading seat reaches a predefined number, the swivelling body 56 rotates to bring the loading seat 58 containing the containers C2 to align with the exit 46b and, preferably, the other loading seat 58, empty, to align with the entrance 46a for a further load. The containers C3 contained in the loading seat 58 aligned with the exit 46b are moved so as to proceed along the input direction X, to then be deviated along the first sterilisation direction Y, In particular, the containers C4 cross the main chamber 20 aligned one behind the other in the first sterilisation direction Y, that is so to be facing in rows. In the first sterilisation direction Y, the containers C4 undergo a first sterilisation by the first and second cannon 28a, 28b, the emission axes of which E1, E2 lie substantially on a horizontal plane. In the first sterilisation direction Y, the containers C4 are positioned in rows, so that the walls B′. B″ of the body B are in front of the emission cone of the electron cannons 28a, 28b. In other words, the bodies B of the containers C4 are substantially coplanar and lie on a single plane in relation to which the emission axes E1, E2 are incident. Such arrangement permits an excellent sterilisation of the walls B′, B″ of the body B of the containers C4 and, if necessary, of the gusset sides G where provided. Downstream of the second cannon 28b, the containers being treated are deviated from the first sterilisation direction Y to the second sterilisation direction Z and at the same time moved along it so to enter and pass through the secondary chamber 36. In particular, the containers C5 cross the secondary chamber 36 aligned one behind another with the walls facing each other; that is to say the containers C5 are facing in columns. In the second sterilisation direction Z, the containers C5 undergo a second sterilisation by the third electron cannon 28c, the emission axis of which E3 is incident, in particular orthogonal to the horizontal plane. In the second sterilisation direction Z, the containers C5 are positioned in columns, so that the axis of the straw A of the container is substantially parallel to the emission axis of the emission cone of the third cannon 28c. Such arrangement permits an excellent sterilisation of the straw A of the container C. The containers C5 are moved so that one or more containers C6 are housed in the loading seat of the output unit 60. When the number of containers housed in the loading seat reaches a predefined number, the swivelling body rotates to bring the loading seat containing the containers C6 to align with the exit and, preferably, the other loading seat, empty, to align with the entrance for a further load. The containers C7 contained in the loading seat aligned with the exit are moved to the outside of the sterilisation device. The change of direction in the advancement of the containers during sterilisation makes it possible to optimise the exposition times of the parts of these to the electron beams. In particular, in the first sterilisation direction Y, the containers C4 are facing in rows and, given the arrangement of the electron cannons 28a, 28b, the sterilisation is particularly effective on the walls B′, B″ of the body B, which are particularly thin. The width dimension W of the containers C determines the transit time under the emission cones of the cannons 28a 28b of said containers. In the second sterilisation direction Z, the containers C5 are facing in columns and, given the arrangement of the electron cannon 28c, the sterilisation is particularly effective inside the straw A. The thickness dimension T of the containers C determines the transit time under the emission cone of the cannon 28c of said containers. Since the width W of the containers C is much greater than the thickness T of the same (the reason for which they are called thin walled containers), the exposition to the first sterilisation is much less than the exposition to the second sterilisation, with respect to the relative needs, since the walls of the container need a less prolonged sterilisation than the straw. In general, along a sterilisation path, along which a first reference plane Pr1 and a further reference plane Pr3 are defined, separate from each other and incident to said sterilisation path, the first electron cannon 28a has an emission axis E1 lying on the first reference plane Pr1 and the further electron cannon 28c has an emission axis E3 lying on the further reference plane Pr3. The emission axis E3 of the further cannon 28c is inclined in relation to the sterilisation path, unlike the emission axis E1 of the first cannon 28a. The container C can thereby be sterilised in different directions, to optimise the action of the electron clouds on different areas of the container. Preferably In addition, a sterilisation system 100 comprises the sterilisation device, a support structure 102, which the device is placed on, for example raised off the ground, and an outer casing 104, which the device is contained in, suitable for acting as a screen to the leakage of radiation. The outer casing 104 is composed of side walls in lead, which screen any radiation. The casing 104 is fitted with a pair of seats 108, made in the side wall, at the point of the input unit and the output unit. In the closed configuration, said units project outside the casing 104, to permit access of the containers to be sterilised or to allow the exit of the sterilised containers. Preferably in addition, the structure 102 comprises a plurality of pillars 106, on which the casing 104 slides, to lower onto the device 1 and to contain it, or to lift up to permit servicing operations for example. Innovatively, the low voltage electron beam sterilisation device according to the present invention is particularly suitable for treating thin-walled, flexible containers. Advantageously, the device according to the invention is utilisable directly incorporated into the production plant since it greatly limits, in compliance with existing legislation, the leakage of radiation. In particular, the design of the device according to the present invention follows the practical design rule known as the “3 bounce rule” according to which, any ray emitted by the cannons must bounce of the inner wall of the casing three times before getting out of the casing, thereby practically giving a zero value to the energy possessed. in addition, Advantageously, it optimises the exposition times of the containers to the sterilisation, differentiating the exposure times according to the needs of the parts to be sterilised. It is clear that a person skilled in the art may make modifications to the device described above. For example, according to one embodiment variation, the electron cannon for the second sterilisation is on a plane with the cannon for the first sterilisation, while after the first sterilisation the containers are made to rotate in such a way as to position themselves with the straw parallel to the emission axis of the cannon for the second sterilisation. According to a further embodiment variation, the input unit and/or output unit have more than two loading seats, for example angularly equidistant. According to yet a further embodiment, the first sterilisation is performed by a single electron cannon. According to yet a further embodiment, the second sterilisation is performed by two or more electron cannons. Furthermore, according to one embodiment variation, the swivelling body of the input unit and/or output unit performs a rotation of 90° to bring the loaded containers into alignment for exit. According to a further embodiment variation, the motor for driving the swivelling body is connected to it by a cinematic chain. Such variations are also included within the sphere of protection as defined by the following claims. |
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045335133 | description | In the drawings, the numeral 1 designates a concrete body with a cavity which is formed substantially as a solid of revolution about a vertical axis of rotation 2. The cavity comprises two partial spaces, namely a pressure chamber 3, which can be sealed in a pressure-tight manner by means of a cover 4, and a partial space 5 axially outside the pressure chamber 3. The partial space 5 is limited in an axial direction by means of a substantially transversely extending limiting surface 6 which is made in the concrete body axially outside the cover 4. The pressure chamber 3 has an upwardly facing circular opening which, along its periphery, is provided with a flat steel ring 7 which is attached to the concrete body 1. The steel ring 7 is provided with a sealing means which includes a hollow steel torus 8. The steel torus is slotted along its entire circumference in such a way that an annular gap is formed. One edge of the gap is welded to a steel ring 9 which is welded to the steel ring 7, whereas the other edge of the gap is welded to a metallic sealing ring 10 in such a way that--when the cover 4 is removed or relieved--a gap is formed between the sealing ring 10 and the steel ring 7. An annular sealing surface of the cover 4 is pressed against the sealing ring 10 by means of a plurality of compressive-force transmitting elements 11, which are arranged between the cover 4 and the above-mentioned limiting surface 6 provided in the concrete body 1. The cover 4 is made of pre-stressed concrete and formed with a lower, circular-cylindrical, solid portion, the axial dimension of which is designated C, and with an upper rectangular portion, the axial dimension of which is designated S. At the pressures and the pressure chamber volumes at which a structure according to the invention can suitably be applied, S is at least 30% greater than C, whereas C lies within the range 0.8-3 m. In the example, shown in the drawings the dimensions C and S are 2 m and 3.6 m, respectively. The upper rectangular portion of the cover 4 contains a cup-shaped, substantially circular-cylindrical space, which is divided into seven substantially parallel-epipedic, upwardly open spaces 12 by means of six vertical, mutually parallel force-transmitting walls 13. These walls are each provided with a door opening 14, which together with two outer doors 15 and 16 make the spaces 12 accessible to personnel. Each force-transmitting wall 13 has a horizontally upwardly facing force-transmitting surface on which a plurality of force-transmitting elements 11 are evenly distributed. In addition, such elements are evenly arranged in a ring along the edge of the above-mentioned cup-shaped space. Each compressive-force transmitting element 11 simply consists of a solid or hollow steel body. Alternatively, each element 11 may comprise two threaded parts, whereby a correct axial dimension can be adjusted by screwing one part into the other, which can be made manually and/or by providing the elements 11 individually or in groups with remote-controlled drive members. When sealing the pressure chamber 3 by means of the cover 4, the force-transmitting elements 11 are unloaded or only very weakly loaded as long as no over-pressure prevails in the pressure chamber. Upon pressurizing the pressure chamber, the cover 4 can lift somewhat without any deteriorating effect on the sealing function of the sealing device consisting of the components 7, 8, 9, 10. Impermissible lifting of the cover 4 is prevented by the elements 11, whereby the total force transmitted by these elements to the transversal inner limiting surface 6 of the concrete body increases with increasing pressure in the pressure chamber 3. This force results in the concrete body 1 being loaded with great, axially directed tensile forces. In view of these forces, the concrete body 1 is provided with a plurality of elongated clamping loops 17, which are substantially arranged in parallel with an axial plane through the line IV--IV on FIG. 5. Each clamping loop 17 comprises a bundle of pre-stressed, loop-formed steel rods. The clamping loops 17 are each arranged in a correspondingly shaped channel in the concrete body 1, the channels being each provided with a metallic, thin-walled lining tube embedded in the concrete. Each clamping loop 17 comprises a first straight portion 18 arranged axially outside the limiting surface 6, said portion 18 extending transversally from an outer limiting surface in the concrete body 1 to an axial plane through the line IX--IX on FIG. 5, a second portion 19 connected to the portion 18 and approximately forming a circular arc of 90.degree., an axially running third portion 20 connected to said second portion, a semicircular fourth portion 21 connected to the portion 20, an axially running fifth portion 22 connected to said fourth portion, a sixth portion 23 connected to the portion 22 and approximately forming a circular arc of 90.degree., and, finally, a straight, transversally extending seventh portion 24 connected to the portion 23 and extending to an outer limiting surface of said concrete body. Further, the concrete body 1 comprises a plurality of U-shaped bundes 25 of U-shaped, pre-stressed steel yokes, each bundle being arranged in a U-shaped channel, each bundle 25 lying in a vertical plane which is parallel to an axial plane through the line IX--IX on FIG. 5. In addition, the concrete body 1 includes a plurality of horizontal, substantially circular clamping loops 26, each of which comprises a bundle of pre-stressed, correspondingly formed steel loops. Each clamping loop 26 is arranged in a correspondingly formed channel arranged in the concrete body 1 and provided with a lining tube. Stressing of the elongated clamping loops 17, the U-shaped bundles 25 and the circular clamping loops 26 does not take place until the concrete of the concrete body 1 has solidified and hardened for several weeks. The ends are then secured to metallic anchor plates 17' and 25' and 26', respectively, arranged at the outer limiting surfaces of the concrete body. The partial space 5 is connected to two horizontal transport tunnels 27 for the cover 4, which are formed in the concrete body 1. Alternatively, the concrete body 1 can be formed with one such transport tunnel only. Each transport tunnel 27 is limited in the upward direction by a plane tunnel roof 28, which lies in the same horizontal plane as the above-mentioned horizontal limiting surface 6, and in the downward direction by a tunnel floor 29 which lies on a level with the upwardly facing surface of the plane steel ring 7. In the lateral direction, each transport tunnel 27 is limited by two confronting wall surfaces. The projections of these surfaces in the direction of the tunnel are linear and coincide mainly with the corresponding projections of the side walls 28 of the partial space 5. Since the concrete body 1 immediately above the partial space 5 has a portion, whose outer horizontal dimension in a vertical plane along the line IX--IX is somewhat smaller than the corresponding horizontal dimension below the partial space 5, the extension of each tunnel floor 29 in the direction of the tunnel is somewhat greater than the corresponding extension of the roof and walls of the transport tunnel 27. There is no distinct transition between the partial space 5 and the transport tunnels 27. In the following each tunnel 27 is regarded as extending from a vertical plane through the nearest vertical surface of the cover 4, which means that the dimensions G on FIG. 8 indicate the horizontal dimension of each tunnel floor 29. Two straight transport rails of steel, 30 and 31, are recessed in a horizontal, upwardly directed concrete surface which comprises the two tunnel floors 29 and a lower horizontal limiting surface for the partial space 5. The cover 4 is provided with four wheel stands 33 which are evenly distributed on the two transport rails 30 and 31. Each wheel stand 33 comprises a plurality of wheels 34, which are arranged to be able to roll on the corresponding transport rail. In each wheel stand 33 the wheels 34 are arranged with their wheel axles fixed to a body 35 having U-shaped cross-section, which body is guided by means of vertically directed guiding means provided in the wheel stand 33. The body 35 is mechanically connected to the pistons of a plurality of hydraulic cylinders 36, which are fixed to the cover 4. The pressure chamber 3 contains a nuclear reactor core 39 as illustrated diagrammatically in FIGS. 2 and 3 of the drawings, and is provided with a plurality of conduits (not shown in the drawings) for steam and/or fluid introduced from the outside of the concrete body 1. When the cover 4 is to be removed, the pressure in the pressure chamber 3 is first reduced, whereafter the compressive-force transmitting elements 11 are removed or adjusted to reduced axial dimension. Thereafter the pressure in the hydraulic cylinders 36 is increased to such an extent that a gap arises between the sealing ring 10 and the corresponding sealing surface in the cover 4, whereupon a horizontal force is applied on the cover 4. Thereby the wheels 30 roll on the rails 30 and 31, and the cover is removed through one of the two transport tunnels 27 and to a transport track arranged outside the concrete body. The transport track has two outer transport rails 30' and 31', which are arranged in alignment with the transport rails 30 and 31, respectively. An enclosing means according to the invention is especially well fitted to be used as a pressure vessel in a nuclear reactor of the type disclosed in U.K. patent application GB No. 2098786 A. The means described above is only one of a number of feasible embodiments of the invention. Thus, it is also possible--to a larger or smaller extent--to replace the described clamping loops arranged in channels by reinforcement bars, which in pre-stressed condition are cast into the concrete body 1. Further, the wheels 34 can be replaced by a number of sliding feet. |
050229737 | abstract | A solvent extraction column in which a disperse phase in the form of electrostatically charged droplets contacts a fluid continuous phase. The disperse phase is received in a number of trays having radial nozzles, so that the disperse phase is discharged from the nozzles as charged droplets transverse to the general direction of flow in the column when an electric potential is applied between the trays and the column. |
description | This application is a continuation-in-part of U.S. application Ser. No. 12/333,300, filed Dec. 11, 2008, which is a continuation-in-part of U.S. application Ser. No. 11/441,999, filed May 26, 2006 and a continuation-in-part of U.S. application Ser. No. 11/736,032, filed Apr. 17, 2007, now U.S. Pat. No. 7,466,085. Not Applicable 1. Field of Invention This invention relates to a method and apparatus for producing of radiopharmaceuticals. 2. Description of the Related Art Cyclotrons are used to generate high energy charged particle beams for purposes such as nuclear physics research and medical treatments. One area where cyclotrons have found particular utility is in the generation of radiopharmaceuticals, also known as biomarkers, for medical diagnosis by such techniques as positron emission tomography (PET). A conventional cyclotron involves a substantial investment, both in monetary and building resources. An example of one of the more compact conventional cyclotrons used for radiopharmaceutical production is the Eclipse RD developed by the company founded by the present inventor and now produced by Siemens. The self-shielded version of the Eclipse RD can be installed in a facility without a shielded vault. The minimum room size for housing the Eclipse RD is 7.31 m×7.01 m×3 m (24 ft×23 ft×10 ft). To support the approximately 29 300 kg (64 400 lbs) installed weight of a self-shielded Eclipse RD, the cyclotron room includes a concrete pad with a minimum thickness of 36 cm (14 in). In addition to a large size and weight, the power requirements often involve a dedicated and substantial electrical power system. The minimum electrical service required for the Eclipse RD is a 208 (±5%) VAC, 150 A, 3-phase service. Thus, medical facilities have a need for biomarkers, but the monetary, structural, and power requirements of conventional cyclotrons have historically made it impracticable for most hospitals and other medical facilities to produce biomarkers on-site. The half-life of clinically important positron-emitting isotopes, i.e., radionuclides, relative to the time required to process a radiopharmaceutical is a significant factor in biomarker generation. The large linear dimensions of the reaction vessel in radiochemical synthesis systems commonly used in biomarker generators result in a small ratio of surface area-to-volume and effectively limit the heat transfer and mass transport rates and lengthens processing time. The four primary PET radionuclides, fluorine-18, carbon-11, nitrogren-13, and oxygen-15, have short half-lives (approximately 110 min, 20 min, 10 min, and 2 min, respectively). Consider the case of the production of [18F]2-fluoro-2-deoxy-D-glucose, commonly referred to as [18F]FDG. Converting nucleophilic fluorine-18 ([18F]F−) into [18F]FDG requires up to 45 min using one of the larger conventional radiochemical synthesis systems, such as the Explora FDG4 radiochemistry module, originally developed by a company founded by the present inventor and now produced by Siemens. The processing time is significant with respect to the half-life of the radioisotope. Accordingly, the production yield fraction of a biomarker of a conventional radiopharmaceutical synthesis system is far from ideal, often limited to a range of approximately 50% to 60% of the component substances. For the Explora FDG4, the processing time fraction is approximately 40% of the half-life of the [18F]F− radioisotope. Corrected to the end of bombardment, the Explora FDG4 has an yield fraction of approximately 65%. The limitations of the larger conventional radiochemical synthesis systems are even more evident when preparing biomarkers that are labeled with the radioisotopes having shorter half-lives. A conventional radiopharmaceutical synthesis system is designed to process a significant quantity of radioactivity. For example, the Explora FDG4 accepts up to 333 GBq (9000 mCi) of [18F]F−. During bombardment, a significant percentage of the newly generated radioisotope decays back to its original target state requiring extended bombardment times to produce a sufficient quantity of the radioisotope for use in a conventional radiopharmaceutical synthesis system. For example, the production of approximately 90 GBq (2400 mCi) of [18F]F− requires a bombardment time of approximately 120 min using the Eclipse RD cyclotron. Even with efficient distribution networks, the short half-lives and low yields require production of a significantly greater amount of the biomarker than is actually needed for the intended use. In contrast, the radioactivity of a unit dose of a biomarker administered to a particular class of patient or subject for medical imaging is considerable smaller, generally ranging from 0.185 GBq to 0.555 GBq (5 mCi to 15 mCi) for human children and adults and from 3.7 MBq to 7.4 MBq (100 μCi to 200 μCi) for mice. Recent advancements have led to the development of smaller reaction systems using microreaction or microfluidic technology. By reducing the linear dimensions of the reaction vessel used in the radiochemical synthesis system, the ratio of surface area-to-volume and, consequently, heat transfer and mass transport rates increases. The smaller size of the reaction vessels lends itself to replication allowing multiple reaction vessels to be placed in parallel to simultaneously process the biomarker. In addition to faster processing times and reduced space requirements, these smaller reaction systems require less energy. In the radiopharmaceutical area, a 2005 article discusses production of 0.064 GBq (1.74 mCi) of [18F]FDG, a quantity sufficient for several positron emission tomography (PET) imaging studies on mice, using an integrated microfluidic circuit as proof of principle for automated multistep synthesis at the nanogram to microgram scale. Chung-Cheng Lee, et al., Multistep Synthesis of a Radiolabeled Imaging Probe Using Integrated Microfluidics, Science, Vol. 310, no. 5755, (Dec. 16, 2005), pp. 1793, 1796. The authors conclude that their chemical reaction circuit design should eventually yield sufficiently large quantities (i.e., >100 mCi) of [18F]FDG to produce multiple doses for use in PET imaging of humans. The commercially available NanoTek Microfluidic Synthesis System distributed by Advion BioSciences, Inc., can synthesize [18F]FDG 35 times faster than with conventional macrochemistry, which clearly represents a significant improvement in radiopharmaceutical processing time. However, such level of advancement has not been seen with the cyclotrons producing the radioisotopes used in radiopharmaceutical synthesis. However, such level of advancement has not been seen with the cyclotrons producing the radioisotopes used in radiopharmaceutical synthesis. A conventional cyclotron used in the production of radioisotopes for synthesizing radiopharmaceuticals has significant power requirements. Typically, a conventional cyclotron for radiopharmaceutical production generates a beam of charged particles having an average energy in the range of 11 MeV to 18 MeV, a beam power in the range of 1.40 kW and 2.16 kW, and a beam current of approximately 120 μA. The weight of an electromagnet of such a conventional cyclotron for radiopharmaceutical production typically ranges between 10 tons and 20 tons. The Eclipse RD is an 11 MeV negative-ion cyclotron producing up to two particle beams each with a 40 μA beam current. The major power consuming components of a cyclotron are typically the magnet system power supply, the RF system amplifier, the ion source transformer, the vacuum system cryopump compressor, and the water system. Of these, the magnet system power supply and the RF system amplifier are the most significant. The operating power consumption of the Eclipse RD is specified at 35 kW. The standby power consumption of the Eclipse RD is specified at less than 7 kW. The magnet system of the Eclipse RD produces a mean field of 1.2 T using 3 kW of power. The RF system of the Eclipse RD has a maximum amplifier power of 10 kW. The ion source system of the Eclipse RD is specified for a maximum H− current of 2 mA. FIG. 1 is a representative illustration of an array of dees in a conventional cyclotron. For simplicity, only two dees 12 are illustrated. However, there are typically four or more dees used. Cyclotrons having fewer dees require more turns in the ion acceleration path, a higher acceleration voltage, or both to energize the ions to the desired level. The dees 12 are positioned in the valley of a large electromagnet and enclosed in a vacuum tank. During operation of the cyclotron, an ion source 81 continuously generates ions 19 through the addition or subtraction of electrons from a source substance. As the ions 19 are introduced into the cyclotron at the center of the array of dees 12, they are exposed a strong magnetic field generated by opposing magnet poles 11 situated above and below the array of dees 12. A radio frequency (RF) oscillator applies a high frequency, high voltage signal to each of the dees 12 causing the charge of the electric potential developed across each of the dees 12 to alternate at a high frequency. Neighboring dees are given opposite charges such that ions 19 entering the gap between neighboring dees 12 see a like charge on the dee behind them and an opposite charge on the dee ahead of them, which results in acceleration (i.e., increasing the energy) of the ions 19. With each energy gain, the orbital radius of the ions 19 increases. The result is a stream of ions 19 following an outwardly spiraling path. The ions 19 ultimately exit the cyclotron as a particle beam 40 directed at a target 89. FIG. 2 illustrates an exploded view of selected components of a representative conventional two-pole cyclotron using the concept of sector-focusing to constrain the vertical dimension of the accelerated particle beam. The cyclotron includes upper and lower yokes 54 that cooperatively engage when assembled to define an acceleration chamber and opposing upper and lower magnet poles 11. Each magnet pole 11 includes two wedge-shaped pole tips 32, commonly referred to as “hills” where the magnetic flux 58 is mostly concentrated. The recesses between the hills 32 are commonly referred to as “valleys” 34 where the gap between the magnet poles 11 is wider. As a consequence of the wider gap between the magnets poles 11, the magnetic flux density in the valleys 34 is reduced compared to the magnetic flux density in the hills 32. A dee 12 is located in each open space defined by the corresponding upper and lower valleys 34. Vertical focusing of the beam is enhanced by a large hill field-to-valley field. A higher ratio indicates stronger magnetic forces, which tends to confine the beam closer to the median plane of the cyclotron. In principle, a tighter confinement allows reduction of the gap between the magnet poles without increasing the danger of the beam striking the pole faces of the magnet. For a given amount of flux, a magnet with a smaller gap between the magnet poles requires less electrical power for excitation than a magnet with a larger gap between the magnet poles. Once the ions are extracted from the cyclotron and are no longer under the influence of the magnet poles 11, a beam tube 92 directs the particle beam 40 through a collimator 96, which refines the profile of the particle beam 40 for irradiation of the target substance 100 contained in the target 89. An unfortunate by-product of radioisotope production is the generation of potentially harmful radiation. The radiation generated as a result of operating a cyclotron is attenuated to acceptable levels by a shielding system, several variants of which are well known in the prior art. At the extraction point of a positive ion cyclotron, interaction between the positive ions 19p and the extraction blocks 102 used to induce the positive ions 19p to exit the cyclotron generate prompt high-energy gamma radiation and neutron radiation, a byproduct of nuclear reactions that produce radioisotopes. At the target 89, the nuclear reaction that occurs as the particle beam 40 irradiates the target substance 100 contained therein to produce the desired radioisotope generates prompt high-energy gamma radiation and neutron radiation. Additionally, residual radiation is indirectly generated by the nuclear reaction that yields the radioisotope. During the nuclear reaction, neutrons are ejected from the target substance and when they strike an interior surface of the cyclotron, gamma radiation is generated. Finally, direct bombardment of components such as the collimator 96 and the target window 98 by the particle beam 40 generates induced high-energy gamma radiation. Thus, a cyclotron must be housed in a shielded vault or be self-shielded. Although commonly composed of layers of exotic and costly materials, shielding systems only can attenuate radiation; they cannot absorb all of the gamma radiation or other ionizing radiation. Following irradiation by the cyclotron, the target substance is commonly transferred to a radioisotope processing system. Such radioisotope processing systems are numerous and varied and are well known in the prior art. The radioisotope processing system prepares the radioisotope for the tagging or labeling of molecules of interest to enhance the efficiency and yield of the radiopharmaceutical synthesis processes. For example, the radioisotope processing system may extract undesirable molecules, such as excess water or metals to concentrate or purify the target substance. An improved biomarker generator and a method suitable for efficiently producing short lived radiopharmaceuticals in quantities on the order of a unit dose is described in detail herein and illustrated in the accompanying figures. The improved biomarker generator includes a particle accelerator and a radiopharmaceutical micro-synthesis system. The micro-accelerator of the improved biomarker generator is optimized for producing radioisotopes useful in synthesizing radiopharmaceuticals in quantities on the order of one unit dose allowing for significant reductions in size, power requirements, and weight when compared to conventional radiopharmaceutical cyclotrons. The radiopharmaceutical micro-synthesis system of the improved biomarker generator is a small volume chemical synthesis system comprising a microreactor and/or a microfluidic chip and optimized for synthesizing the radiopharmaceutical in quantities on the order of one unit dose allowing for significant reductions in the quantity of radioisotope required and the processing time when compared to conventional radiopharmaceutical processing systems. The improved biomarker generator includes a small, low-power particle accelerator (hereinafter “micro-accelerator”) for producing approximately 1 unit dose of a radioisotope that is chemically bonded (e.g., covalently bonded or ionically bonded) to a specific molecule. The micro-accelerator produces per run a maximum quantity of radioisotope that is approximately equal to the quantity of radioisotope required by the radiopharmaceutical micro-synthesis system to synthesize a unit dose of biomarker. The micro-accelerator takes advantage of various novel features, either independently or in combination to reduce size, weight, and power requirements and consumption. The features of the micro-accelerator described allow production of a radioisotope with a maximum radioactivity of approximately 2.59 GBq (70 mCi) using a particle beam with an average energy in the range of 5 MeV to 18 MeV or in various sub-ranges thereof and a maximum beam power in the range of 50 W to 200 W. One feature of the micro-accelerator is the use of permanent magnets to contain the ions during acceleration and eliminate the electromagnetic coils of the common to conventional radiopharmaceutical cyclotrons. Each of the permanent magnets and the dees are wedge-shaped and arranged into a substantially circular array. A series of collimator channels in selected dees initially direct the path of the ions introduced at the center of the array. After exiting the series of collimator channels, the ions travel through the main channels of the dees until the desired energy level is achieved. The permanent magnet cyclotron provides substantial improvements with respect to cost, reliability, size, weight, infrastructure requirements, and power requirements compared to conventional radiopharmaceutical cyclotrons. Another feature of the micro-accelerator is the use of an improved radio frequency (RF) system powered by a rectified RF power supply. A rectified input supplies a high voltage transformer to supply power to the RF oscillator. The RF signal produced by the RF system is high peak-to-peak voltage at the resonant frequency of the RF oscillator enveloped by the line voltage frequency. The charged particles are only accelerated during a portion of the line voltage cycle. The resulting RF power supply compensates for reduced activity by increasing the current. A still further feature of the micro-accelerator is the use of an internal target cyclotron where the target is located within the magnetic field and the particle beam irradiates the internal target while still within the magnetic field. This allows the magnet system to assist in containing harmful radiation related to the nuclear reaction that converts the target substance into a radioisotope and eliminates a major source of radiation inherent in a conventional positive-ion cyclotron. As a result, the micro-accelerator can take advantage of the benefits without a significant disadvantage normally associated with a positive particle beam. Beams of positively-charged particles generally are more stable than beams of negatively-charged particle because the reduced likelihood of losing an electron at the high velocities that charged particles experience in a cyclotron. Losing an electron usually causes the charged particle to strike an interior surface of the cyclotron and generate additional radiation. Minimizing the production of excess radiation reduces the amount of shielding required. Additionally, a positive ion cyclotron requires significantly less vacuum pumping equipment. Reducing the amount of shielding and vacuum pumping equipment reduces the size, weight, cost, complexity, power requirements, and power consumption of the cyclotron. Through the use of microreactors and microfluidic chips, which have fast processing times and offer precise control over the various stages of a chemical process, the radiopharmaceutical micro-synthesis system provides a significant reduction in processing time that directly reduces the quantity of the radioisotope required to synthesize the desired biomarker. The method for producing a radiopharmaceutical using the improved biomarker generator calls for providing a micro-accelerator, producing charged particles, accelerating the charged particles, and forming a particle beam to irradiate a target substance and produce a radioisotope. The improved biomarker generator allows operation using a volume of the target substance that is unusually small in the area of radiopharmaceutical production. After irradiation, the radioisotope and at least one reagent are transferred to the radiopharmaceutical micro-synthesis system. The radioisotope undergoes processing as necessary. Ultimately, the radiopharmaceutical micro-synthesis system combines the radioisotope with the reagent or reagents to synthesize the biomarker. The system includes a radiopharmaceutical micro-synthesis system having at least one microreactor and/or microfluidic chip. Using the unit or precursory unit dose of the radioisotope and at least one reagent, the radiopharmaceutical micro-synthesis system synthesizes on the order of a unit dose of a biomarker. Chemical synthesis using microreactors or microfluidic chips (or both) is significantly more efficient than chemical synthesis using conventional macroscale chemical synthesis technology. Yields are higher and reaction times are shorter, thereby significantly reducing the quantity of radioisotope required in synthesizing a unit dose of biomarker. Accordingly, because the micro-accelerator only produces relatively small quantities of radioisotope per production run, the maximum beam power of the micro-accelerator is approximately two to three orders of magnitude less than the beam power of a conventional particle accelerator. As a direct result of this dramatic reduction in maximum beam power, the micro-accelerator is significantly smaller and lighter than a conventional particle accelerator, has less stringent infrastructure requirements, and requires far less electricity. Additionally, many of the components of the small, low-power accelerator are less costly and less sophisticated, such as the magnet, magnet coil, vacuum pumps, and power supply, including the RF oscillator. The synergy that results from combining the micro-accelerator and the radiopharmaceutical micro-synthesis system having at least one microreactor and/or microfluidic chip cannot be overstated. This combination, which is the essence of the improved biomarker generator, provides for the production of approximately one unit dose of radioisotope in conjunction with the nearly on-demand synthesis of one unit dose of a biomarker. The improved biomarker generator is an economical alternative that makes in-house biomarker generation at the imaging site a viable option even for small regional hospitals. An improved biomarker generator and a method suitable for efficiently producing short lived radiopharmaceuticals in quantities on the order of a unit dose is described in detail herein and illustrated in the accompanying figures. The improved biomarker generator includes a particle accelerator and a radiopharmaceutical micro-synthesis system. The micro-accelerator of the improved biomarker generator is optimized for producing radioisotopes useful in synthesizing radiopharmaceuticals in quantities on the order of one unit dose allowing for significant reductions in size, power requirements, and weight when compared to conventional radiopharmaceutical cyclotrons. The radiopharmaceutical micro-synthesis system of the improved biomarker generator is a small volume chemical synthesis system comprising a microreactor and/or a microfluidic chip and optimized for synthesizing the radiopharmaceutical in quantities on the order of one unit dose allowing for significant reductions in the quantity of radioisotope required and the processing time when compared to conventional radiopharmaceutical processing systems. As used herein, “microreactors” and “microfluidic chips” refer broadly small volume reaction systems including microscale, nanoscale, and picoscale systems. As used herein, the term “radiopharmaceutical” encompasses any organic or inorganic compound comprising a covalently-attached radioisotope (e.g., 2-deoxy-2-[18F]fluoro-D-glucose ([18F]FDG)), any inorganic radioactive ionic solution (e.g., Na[18F]F ionic solution), or any radioactive gas (e.g., [11C]CO2), particularly including radioactive molecular imaging probes intended for administration to a patient or subject (e.g., by inhalation, ingestion, or intravenous injection) for imaging purposes. Such probes are also referred to in the art as radiotracers and radioligands and, more generically, as radiochemicals. The terms “patient” and “subject” refer to any human or animal subject, particularly including all mammals. A “unit dose” refers to the quantity of radioactivity that is administered for medical imaging to a particular class of patient or subject. A unit dose of the radiopharmaceutical necessarily comprises a unit dose of a radioisotope. As previously discussed, conventional radiopharmaceutical production focuses on generating a large amount of the radioisotope, typically on the order of Curies, in recognition of the significant radioactive decay that occurs during the relatively long time that the radioisotope undergoes processing and distribution. The improved biomarker generator of the present invention departs significantly from the established practice in that it is engineered to produce a per run maximum amount of radioisotope on the order of tens of millicuries. The micro-accelerator produces a maximum of approximately 2.59 GBq (70 mCi) of the desired radioisotope per production run. A particle accelerator producing a radioisotope on this scale requires significantly less beam power than conventional particle accelerators used for radiopharmaceutical production. The micro-accelerator generates a particle beam having a maximum beam power of 200 W. In various embodiments, the micro-accelerator generates a particle beam having a maximum beam power of approximately 200 W, 175 W, 150 W, 125 W, 100 W, 75 W, or 50 W. As a direct result of the dramatic reduction in maximum beam power, the micro-accelerator is significantly smaller and lighter than a conventional cyclotrons used in radiopharmaceutical production and requires less electricity. Many of the components of the micro-accelerator are less costly and less sophisticated compared to conventional cyclotrons used in radiopharmaceutical production. FIGS. 3 illustrates one embodiment of a selected portion of a micro-accelerator in the form of a cyclotron using permanent magnets 10a (hereinafter a “permanent magnet cyclotron”) with the upper and lower platforms in an open configuration. FIG. 4 omits the upper platform to provide an unobstructed view of the components in the lower platform. FIG. 5 is a cross-sectional view of the micro-accelerator of FIG. 3 shown with the upper and lower platforms 29 in a closed configuration. Each of the upper and lower platforms 29 defines a cavity 31 on the interior side thereof, such that when the upper and lower platforms 29 are engaged, the cavities 31 define an acceleration chamber 27. A plurality of permanent magnets 20 are arranged in a circular array in the cavities of each of the upper and lower platforms 29 to form the magnet poles. Each permanent magnet 20 carried by the upper platform forms an opposing pair with the corresponding permanent magnet 20 carried by the lower platform. The valleys between the respective pairs of permanent magnets 20 are occupied by a plurality of dees 45, with one dee being disposed in each valley. A centrally located ion injection opening 33 is defined through the upper and lower platforms 29 allowing the ion source 82 to generate ions at the center of the circular array of dees 45 and permanent magnets 20. As shown in FIG. 5, the micro-accelerator includes an RF system 44 in electrical communication with each of the dees 45 via a plurality of through-openings defined by the lower platform. A dee support 46 attached to each dee 45 extends through a corresponding through-opening and electrically connects the attached dee to the RF system 44. Each of the permanent magnets 20 and the dees 45 are wedge-shaped. Each permanent magnet 20 has a first end positioned proximate to the center of the array and a second positioned proximate to the periphery of the array. Likewise, each dee 45 has a first end positioned proximate to the center of the array and an second end positioned proximate to the periphery of the array. Each of the dees 45 defines a main channel 14 through which ions travel as they are accelerated. When the dees 45 are disposed with the valleys, the faces of the permanent magnet pole tips are disposed in substantially the same plane as the side of the of the corresponding horizontal member of the dees that define the main channel 14. In the illustrated embodiment, the horizontal inner surfaces of the dees are substantially co-planar with the corresponding pole faces of the magnet pairs. When the upper and lower platforms 29 are engaged, a magnet gap is defined between corresponding permanent magnets 20 of the upper and lower platforms 29. Accordingly, the entire channel has a substantially homogeneous height, which provides an unobstructed flight path for the ions being accelerated therein. The upper and lower platforms 29 are supported by a plurality of legs 37. In the illustrated embodiment and best viewed in FIG. 5, each leg 37 is defined by the body of a pneumatic or hydraulic cylinder 38. The lower platform defines a plurality of through openings 35 for slidably receiving a piston rod 39 of each of the cylinders 38. The distal end 42 of each piston rod 39 is connected to the upper platform. Thus, engagement of the upper and lower platforms 29 is accomplished by retraction of the piston rods 39 into the respective cylinders 38. Separation of the upper and lower platforms 29 is accomplished by extending the piston rods 39 from within the cylinders 38. While this construction is disclosed, it will be understood that other configurations are contemplated as well. FIG. 6 is a sectional top plan view of the permanent magnet cyclotron 10a showing the ion flight path 60. A series of collimator channels 13a, 13b, 13c are used to initially direct the path of the ions introduced at the center of the array. Each collimator channel 13a, 13b, 13c defines an outlet into the gap between corresponding permanent magnets 20 carried by the upper and lower platforms 29. In the illustrated embodiment, a first collimator channel 13a accepts ions introduced at the center of the array that are excited to a desired initial energy. Ions exiting the first collimator channel 13a travel along a generally arcuate course across the interposed hill and enter the second collimator channel 13b. Similarly, ions exiting the second collimator channel 13b travel across the interposed hill and enter the third collimator channel 13c. The first, second and third collimator channels 13a, 13b, 13c are configured to define the first revolution of the ions during acceleration. Ions that lack the desired initial energy level are rejected by not allowing such ions to enter the first collimator channel 13a. After exiting the third collimator channel 13c, the ions travel through the main channels 14 defined by each of the dees 45 until the desired energy level is achieved. The permanent magnet cyclotron 10a provides substantial improvements with respect to cost and reliability when compared to conventional cyclotrons producing particle beams with energies of 10 MeV or less using electromagnets or superconducting magnets. Because the permanent magnet cyclotron 10a allows for the exclusion of the electromagnetic coils of the common to conventional radiopharmaceutical cyclotrons, the volume and weight are significantly reduced. In one embodiment, the volume and weight of the micro-accelerator are 40% of the volume and weight of conventional radiopharmaceutical cyclotrons, with a corresponding minimum equipment cost savings of approximately 25% of the equipment cost of conventional radiopharmaceutical cyclotrons. Additionally, eliminating the electric power needed to excite the electromagnet coils in a conventional cyclotron magnet significantly reduces the power requirements and realizes a significant savings in energy usage. The power requirements are further reduced as a result of the lower acceleration voltage of 8 MeV to 10 MeV or less applied to the dees. As a result of these improvements, the reliability of the permanent magnet cyclotron 10a is enhanced as compared to conventional radiopharmaceutical cyclotrons. As a result of the smaller size and lighter weight, more facilities are capable of operating the present invention, especially in situations where space is of concern. Further, because of the ultimately reduced purchase and operating costs, the permanent magnet cyclotron 10a is also more affordable. While the permanent magnet cyclotron 10a is presently not practical for higher acceleration voltages due to the increased magnetic field requirements of the permanent magnets, such embodiments are not excluded from the spirit of the present invention. FIG. 7 is a block diagram of an improved RF system used in one embodiment of the micro-accelerator (hereinafter the “improved RF cyclotron”). The improved RF system includes a rectifier circuit 220 that accepts line voltage and produces a rectified voltage signal. The rectifier circuit 220 is a full wave rectifier incorporating two or more diodes, such as a dual diode rectifier. In one embodiment, the rectified voltage signal is the positive portion of the line voltage. The rectified voltage signal supplies the input of a high voltage step-up transformer 222 capable of supplying a high voltage and high current RF supply signal. In one embodiment, the step-up transformer is an autotransformer producing an output voltage of 30 kV at the line voltage frequency, e.g., 60 Hz. The RF oscillator 224 uses the RF supply signal to produce an RF signal at a selected frequency based on the resonance frequency of the RF oscillator 224 and having a peak-to-peak voltage corresponding to the peak voltage of the RF supply signal. The resonance frequency and the peak-to-peak voltage are selected to accelerate the charged particles to a selected energy level. The resulting RF signal drives the polarity of the dees to accelerate the charged particles. However, acceleration of positively charged particles occurs only during the positive portion of the 60 Hz cycle. By applying full wave rectification, the acceleration periods occur twice as often. For the production of radioisotopes useful in positron emission tomography imaging, only small amounts of radioactivity are necessary. By increasing the beam current, the improved RF cyclotron compensates for having acceleration during only a small portion of the 60 Hz cycle. In the illustrated embodiment, the resonance frequency of the RF oscillator is 72 MHz producing an RF signal having a frequency of 72 MHz with a maximum peak-to-peak voltage of 30 kV enveloped in the 60 Hz line voltage frequency. To facilitate low-power operation, the ion source of one embodiment of the micro-accelerator is optimized for positive ion production. Beams of positively-charged particles generally are more stable than beams of negatively-charged particle because the reduced likelihood of losing an electron at the high velocities that charged particles experience in a cyclotron. Losing an electron usually causes the charged particle to strike an interior surface of the cyclotron and generate additional radiation. Minimizing the production of excess radiation reduces the amount of shielding required. Additionally, a positive ion cyclotron requires significantly less vacuum pumping equipment. Reducing the amount of shielding and vacuum pumping equipment reduces the size, weight, cost, complexity, power requirements, and power consumption of the cyclotron. In one embodiment, the ion source is optimized for proton (H+) production. In an alternate embodiment, the ion source is optimized for deuteron (2H+) production. In another embodiment, ion source is optimized for alpha particle (He2+) production. FIG. 8 illustrates one embodiment of the micro-accelerator 10b in the form of a positive ion cyclotron (hereinafter “internal target cyclotron”) where the target 183 (hereinafter “internal target”) is located within the magnetic field. In this embodiment, the positive ion particle beam 184 irradiates the internal target 183 while still within the magnetic field 182 produced by the opposing magnet poles 186, 188. Consequently, the magnet system assists in containing harmful radiation related to the nuclear reaction that converts the target substance into a radioisotope. The internal target 183 eliminates a major source of radiation inherent in a conventional positive-ion cyclotron by eliminating the need for the conventional extraction blocks. In their absence, much less harmful radiation is generated. Thus, the internal target 183 eliminates a considerable disadvantage for positive-ion cyclotrons. A reduction in harmful radiation generation translates into a reduction in the amount of shielding and the associated benefits discussed above. In the illustrated embodiment, the internal target 183 includes a stainless steel tube 192 that conducts the target substance. The stainless steel tube 192 has a target section centered in the path that the particle beam 184 travels following the final increment of acceleration. The longitudinal axis of the target section is substantially parallel to the magnetic field 182 generated by the magnet system and substantially perpendicular to the electric field generated by the RF system. The remainder of the stainless steel tube 192 is selectively shaped and positioned such that it does not otherwise obstruct the path followed by the particle beam 184 during or following its acceleration. The internal target 183 defines an opening 196 that is positioned in a path of the particle beam 184. A target window 198, which comprises a very thin layer of a foil such as aluminum, seals the opening 196 and prevents the target substance from escaping. Also, a pair of valves 200 control the flow of the target substance and hold a selected volume of the target solution in place for irradiation by the particle beam 184. The diameter of the stainless steel tube 192 varies depending on the configuration of the internal target cyclotron 10b. Generally, the diameter is less than or equal to the increase in the orbital radius of the charged particles over one orbit, which in this embodiment is approximately four millimeters. Thus, in one embodiment, the diameter of the stainless steel tube 192 is approximately four millimeters. Because the charged particles gain a predetermined fixed quantity of energy that is manifested by an incremental fixed increase in the orbital radius of the beam, the charged particles do not interact with the stainless steel tube 192 prior to the final increment of acceleration, which would result in an undesirable situation that reduces the efficiency of the particle beam 184. The micro-accelerator is designed to produce a particle beam in which the charged particles have an average energy sufficient to overcome the binding energy of the target isotope. In the area of radiopharmaceutical production, the minimum effective average energy of the charged particles is 5 MeV. Higher average particle energies result in more efficient radioisotope production and shorter production times. The micro-accelerator 112 produces a particle beam of charged particles with an average energy in the range of 5 MeV to 18 MeV. In one embodiment, the charged particles have an average energy in the range of 5 MeV to 10 MeV. In another embodiment, the charged particles have an average energy in the range of 7 MeV to 10 MeV. In another embodiment of the micro-accelerator 112, the charged particles have an average energy in the range of 8 MeV to 10 MeV. In yet another embodiment of the micro-accelerator 112, the charged particles have an average energy in the range of 7 MeV to 18 MeV. In more specific embodiments of the micro-accelerator 112, the charged particles are protons, deuterons, or alpha particles with an average energy in the range of 5 MeV to 18 MeV, 5 MeV to 10 MeV, 7 MeV to 10 MeV, 8 MeV to 10 MeV, or 7 MeV to 18 MeV. In a further embodiment, the micro-accelerator 112 generates a particle beam with a beam current of approximately 1 μA consisting essentially of protons having an energy of approximately 7 MeV, the particle beam having beam power of approximately 7 W and being collimated to a diameter of approximately 1 mm. At lower average particle energies, fewer charged particles will be successful in destabilizing the target isotope and production time increases. As production time increases to a point that it is significant with respect to the half-life of the radioisotope, some of the radioisotope that has been produced will decay. The quantities of the radioisotope for which the micro-accelerator is designed are small enough to be practicable even when the ratio of production to decay is small. The various embodiments of the micro-accelerator are limited to producing a radioisotope with a maximum radioactivity of approximately 2.59 GBq (70 mCi) per production run. In one embodiment, the micro-accelerator produces a maximum of approximately 0.666 GBq (18 mCi) of fluorine-18 per production run. In another embodiment, the micro-accelerator produces a maximum of approximately 0.185 GBq (5 mCi) of fluorine-18 per production run. In yet another embodiment, the micro-accelerator produces a maximum of approximately 1.11 GBq (30 mCi) of carbon-11 per production run. In further embodiment, the micro-accelerator produces a maximum of approximately 1.48 GBq (40 mCi) of nitrogen-13 per production run. In still further embodiment, the micro-accelerator produces a maximum of approximately 2.22 GBq (60 mCi) of oxygen-15 per production run. Such embodiments of the micro-accelerator are flexible in that they can provide an adequate quantity of radioisotope for each of various classes of patients and subjects that undergo PET imaging. The improved biomarker generator of the present invention may be embodied in many different forms. The permanent magnet cyclotron 10a, the improved RF cyclotron, and the internal target cyclotron 10b are examples of suitable components for use in a particle accelerator optimized as a micro-accelerator. Moreover, the various features of the permanent magnet cyclotron 10a, the improved RF cyclotron, and the internal target cyclotron 10b can be mixed and matched in a single micro-accelerator. Thus, one embodiment of the micro-accelerator is a combination of the permanent magnet cyclotron 10a with the internal target 183 of the internal target cyclotron 10b. Another embodiment of the micro-accelerator is a combination of the permanent magnet cyclotron 10a with the improved RF system. Yet another embodiment of the micro-accelerator is a combination of the internal target cyclotron 10b with the improved RF system. A still further embodiment is the combination of the permanent magnet cyclotron 10a with the improved RF system and the internal target 183 of the internal target cyclotron 10b. Variations in the overall architecture of the micro-accelerator and the radiopharmaceutical micro-synthesis system are contemplated. For example, one embodiment, the micro-accelerator is a two-pole cyclotron. In another embodiment, the micro-accelerator is a four-pole cyclotron. Using a four-pole cyclotron may be advantageous in certain applications, because a four-pole cyclotron accelerates charged particles more quickly than a two-pole cyclotron using an equivalent accelerating voltage. The micro-accelerator described herein emphasizes the generation of a positively-charged particle beam; however, the acceleration of negatively-charged particles is necessary for certain applications and is considered within the scope of the present invention. The micro-accelerator described herein emphasizes the use of permanent magnets; however, the use of small electromagnets (weighing up to approximately 3 tons) is not outside the scope of the present invention for certain applications where a higher beam power is required. While the foregoing discussion emphasizes the use of a micro-accelerator, other types of particle accelerators may be used for production of the particle beam. Acceptable alternatives for the cyclotron include linear accelerators, radiofrequency quadrupole accelerators, and tandem accelerators. The production quantities, the ion source types, and the particle beam energies, ranges, diameters, particles, and powers apply to the various embodiments and modifications of the micro-accelerators. FIG. 9 illustrates one embodiment of the improved biomarker generator including a micro-accelerator 112 and a radiopharmaceutical micro-synthesis system 114, which as previous indicated incorporates at least one of a microreactor and microfluidic chip. As part of the complete improved biomarker generator, the radiopharmaceutical micro-synthesis system 114 will necessarily be configured to process the quantity of the radioisotope produced by the micro-accelerator 112. Microreactors and microfluidic chips typically perform their respective functions in less than 15 min, some in less than 2 min. This significant reduction in processing time directly allows a reduction in the quantity of the radioisotope required to synthesis the desired biomarker. A microfluidic chip exercises digital control over variables such as the duration of the various stages of a chemical process, which leads to a well-defined and narrow distribution of residence times. Such control also enables extremely precise control over flow patterns within the microfluidic chip. The use of a microfluidic chip facilitates the automation of multiple, parallel, and/or sequential chemical processes. FIG. 10 is a flow diagram of one embodiment of the method for producing a radiopharmaceutical using the improved biomarker generator. The method calls for providing a micro-accelerator, producing charged particles, accelerating the charged particles, and forming a particle beam to irradiate a target substance and produce a radioisotope. As an example, in the production of no-carrier-added fluorine-18, a particle beam of protons bombards the target substance of [18O]water. The protons in the particle beam interact with the oxygen-18 isotope in the [18O]water molecules . The improved biomarker generator allows operation using a volume of the target substance that is unusually small in the area of radiopharmaceutical production. A sufficient quantity of a fluorine-18 can be produced using a [18O]water target substance with a volume of approximately 1 mL because the maximum mass of the radioisotope required to produce a unit dose of a radiopharmaceutical is on the order of nanograms. The internal target 183 discussed above is particularly well-suited for handling target substance volumes on this scale. While this example contemplates the use of a liquid target substance, one skilled in the art will recognize that certain methods of producing a radioisotope or radiolabeled precursor require an internal target that can accommodate a gaseous or solid target substance. Further, while the example given contemplates the production of fluorine-18, the internal target may be modified to enable the production of other radioisotopes or radiolabeled precursors, including [11C]CO2 and [11C]CH4, both of which are widely used in research. Such embodiments are considered to be within the scope and spirit of the present invention. After irradiation, the radioisotope and at least one reagent are transferred to the radiopharmaceutical micro-synthesis system 114. The radioisotope undergoes processing such as concentration, as necessary. Ultimately, the radiopharmaceutical micro-synthesis system 114 combines the radioisotope with the reagent to synthesize the biomarker. In this context, a reagent is a substance used in synthesizing the biomarker because of the chemical or biological activity of the substance. Examples of a reagent include a solvent, a catalyst, an inhibitor, a biomolecule, and a reactive precursor. A reactive precursor is an organic or inorganic non-radioactive molecule that, in synthesizing a biomarker or other radiopharmaceutical, is reacted with a radioisotope, typically by nucleophilic substitution, electrophilic substitution, or ion exchange. The chemical nature of the reactive precursor varies and depends on the physiological process that has been selected for imaging. Exemplary organic reactive precursors include sugars, amino acids, proteins, nucleosides, nucleotides, small molecule pharmaceuticals, and derivatives thereof. Synthesis refers to the production of the biomarker by the union of chemical elements, groups, or simpler compounds, or by the degradation of a complex compound, or both. Synthesis, therefore, includes any tagging or labeling reactions involving the radioisotope and any processes (e.g., concentration, evaporation, distillation, enrichment, neutralization, and purification) used in producing the biomarker or in processing the target substance for use in synthesizing the biomarker. The latter is especially important in instances where (1) the volume of the target substance is too great to be manipulated efficiently within some of the internal structures of the radiopharmaceutical micro-synthesis system and/or (2) the concentration of the radioisotope in the target substance is lower than is necessary to optimize the synthesis reaction(s) that yield the biomarker. Accordingly, one embodiment of the radiopharmaceutical micro-synthesis system incorporates integrated separation components providing the ability to concentrate the radioisotope. Examples of suitable separation components include ion-exchange resins, semi-permeable membranes, or nanofibers. Such separations via semi-permeable membranes usually are driven by a chemical gradient or electrochemical gradient. Another example of processing the target substance includes solvent exchange. Continuing the example from above, the concentration of fluorine-18 obtained from a proton bombardment of [18O]water is usually below 1 ppm. This dilute solution needs to be concentrated to approximately 100 ppm in order to optimize the kinetics of the biomarker synthesis reactions. This processing occurs in the radiopharmaceutical micro-synthesis system 114. The improved biomarker generator enables the small scale in-situ production of a radioisotope and synthesis of biomarkers. Thus, the micro-accelerator 112 produces a sufficient quantity of the radioisotope for the radiopharmaceutical micro-synthesis system 114 to synthesize of the biomarker on the order of a unit dose of the biomarker. In one embodiment, the micro-accelerator 112 generates the radioisotope in a quantity on the order of a unit dose. In another embodiment, the micro-accelerator 112 generates the radioisotope in a quantity on the order of a precursory unit dose of the radioisotope. A precursory unit dose of the radioisotope is a dose of radioisotope that, after decaying for a length of time approximately equal to the time required to synthesize the biomarker, yields a quantity of biomarker having a quantity of radioactivity approximately equal to the unit dose appropriate for the particular class of patient or subject undergoing PET. For example, if the radiochemical synthesis system requires 20 min to synthesize a unit dose of a biomarker comprising carbon-11 (t1/2=20 min), the precursory unit dose of the carbon-11 radioisotope has an radioactivity equal to approximately 200% times the radioactivity of a unit dose of the biomarker in order to compensate for the radioactive decay. Similarly, if the radiopharmaceutical micro-synthesis system requires 4 min to synthesize a unit dose of a biomarker labeled with oxygen-15 (t1/2=2 min), the precursory unit dose of the oxygen-15 radioisotope has an radioactivity equal to approximately 400% times the radioactivity of a unit dose of the biomarker in order to compensate for the radioactive decay. In some instances, the precursory unit dose of the radioisotope may be used to compensate for a radiopharmaceutical micro-synthesis system having a yield fraction that is significantly less than 100% of the radioactivity supplied. Further, the precursory unit dose may be used to compensate for radioactive decay during the time required in administering the biomarker to the patient or subject. One skilled in the art will recognize that the synthesis of a biomarker comprising a positron-emitting radioisotope should be completed within approximately the two half-lives of the radioisotope immediately following the production of the unit or precursory unit dose to avoid the significant increase in inefficiency that would otherwise result. Although the foregoing description emphasizes the production of biomarkers labeled with fluorine-18, such as [18F]FDG, the radiopharmaceutical micro-synthesis system is flexible and may be used to synthesize biomarkers labeled with other radioisotopes, such as carbon-11, nitrogen-13, or oxygen-15. Further, the improved biomarker generator discussed herein is flexible enough to produce quantities on the order of a unit dose of biomarkers that are labeled with radioisotopes that do not emit positrons or for producing small doses of radiopharmaceuticals other than biomarkers. One skilled in the art will recognize also that the radiopharmaceutical micro-synthesis system may comprise parallel circuits, enabling simultaneous production of unit doses of a variety of biomarkers. Finally, one skilled in the art will recognize that the improved biomarker generator may be engineered to produce unit doses of biomarker on a frequent basis. From the foregoing description, it will be recognized by those skilled in the art that an improved biomarker generator has been provided. The improved biomarker generator described herein allows for the nearly on-demand production of a biomarker in a quantity on the order of one unit dose. Because the half-lives of the radioisotopes most suitable for safe molecular imaging of a living organism are very short, nearly on-demand production of unit doses of biomarkers presents a significant advancement for both clinical medicine and biomedical research. The reduced size, weight, and cost, the reduced infrastructure (power and structural) requirements, and the improved reliability of the micro-accelerator coupled with the speed and overall efficiency of the radiopharmaceutical micro-synthesis system make in-house biomarker generation a viable option even for small regional hospitals. The various embodiments of the micro-accelerator generate the magnetic field using permanent magnets, move the target into the magnetic field allowing the magnet system to help contain radiation generated during radioisotope production, incorporate the improved RF system described herein, and use combinations of these features to provide the aforementioned improvements over conventional cyclotrons used in radiopharmaceutical production. While the present invention has been illustrated by description of several embodiments and while the illustrative embodiments have been described in considerable detail, it is not the intention of the applicant to restrict or in any way limit the scope of the appended claims to such detail. Additional advantages and modifications will readily appear to those skilled in the art. The invention in its broader aspects is therefore not limited to the specific details, representative apparatus and methods, and illustrative examples shown and described. Accordingly, departures may be made from such details without departing from the spirit or scope of applicant's general inventive concept. |
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claims | 1. A method for generating metastable technetium-99 and molybdenum-99 comprising:accelerating deuterons;bombarding a first target material comprising molybdenum-98 by the accelerated deuterons;generating molybdenum-99 and metastable technetium-99 in the first target material;capturing neutrons escaping from the first target material in a second target material comprising molybdenum-98 surrounding, at least in part, the first target material;generating molybdenum-99 and metastable technetium-99 in the second target material; andseparating molybdenum-99 and metastable technetium-99 from the first and second target materials. 2. The method of claim 1, further comprising:passing the neutrons through a hydrogenous material between the first, inner target material and the second, outer target material, prior to being captured by the second, outer target material. 3. The method of claim 1, comprising:sequentially bombarding the first target material by the accelerated deuterons at a plurality of locations. 4. The method of claim 3, comprising selectively sequentially bombarding the first target material at a plurality of locations by:deflecting the accelerated deuterons by a magnetic field to the plurality of locations. 5. The method of claim 3, comprising sequentially bombarding the first target material at a plurality of locations by rotating the target. 6. The method of claim 1, comprising accelerating the deuterons by a cyclotron. 7. The method of claim 1, comprising:separating technetium-99 and molybdenum-99 from the first and second target materials by chromatography. 8. A method for generating metastable technetium-99 and molybdenum-99 comprising:accelerating deuterons;selectively, sequentially deflecting the accelerated deuterons by a magnetic field to bombard a target material comprising molybdenum-98 by the accelerated deuterons at respective different target material locations, at respective different times;generating molybdenum-99 and metastable technetium-99 in the target material; andseparating the generated molybdenum-99 and the generated metastable technetium-99 from the target material by:a first column containing resin with relatively higher retention of molybdenum-99 and relatively lower retention of metastable technetium-99, anda second column containing resin with relatively higher retention of metastable technetium-99 and relatively lower retention of molybdenum-99. |
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abstract | Examples of synchronized parallel tile computation techniques for large area lithography simulation are disclosed herein for solving tile boundary issues. An exemplary method for integrated circuit (IC) fabrication comprises receiving an IC design layout, partitioning the IC design layout into a plurality of tiles, performing a simulated imaging process on the plurality of tiles, generating a modified IC design layout by combining final synchronized image values from the plurality of tiles, and providing the modified IC design layout for fabricating a mask. Performing the simulated imaging process comprises executing a plurality of imaging steps on each of the plurality of tiles. Executing each of the plurality of imaging steps comprises synchronizing image values from the plurality of tiles via data exchange between neighboring tiles. |
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050229737 | claims | 1. A method of electrostatically enhanced solvent extraction in which a dispersed phase interacts with a continuous liquid phase flowing through a column in counter-current relationship, wherein the improvement comprises, locating interacting means in the column so as to catch the dispersed phase flowing in the column and so as to define a space at a side of the interacting means between the side and the side of the column and through which space the continuous phase flows, and providing an electric field to cause the discharge of charged droplets of the dispersed phase sideways from the interacting means through the space toward the side of the column. 2. A method as claimed in claim 1, wherein the dispersed phase passes through a plurality of said interacting means and electric fields. 3. A method as claimed in claim 2, wherein alternate said electric fields are of opposite polarity. 4. A method as claimed in claim 2, wherein all said electric fields are of the same polarity. 5. A method as claimed in claim 1, wherein the droplets are discharged in a direction substantially normal to the general direction of flow of the dispersed phase. 6. A method as claimed in claim 1, wherein the electric field is at least 1.5 kv/cm. 7. Apparatus for electrostatically enhanced solvent extraction comprising a column in which a continuous phase is arranged to flow therethrough in counter-current relationship with a dispersed phase, wherein the improvement comprises, means for catching a dispersed phase flowing through the column and for interacting the dispersed phase with the continuous phase, the interacting means being located in the column to define a space at a side thereof between the side and the side of the column and through which space the continuous phase is arranged to flow, and means for providing an electric field to cause the discharge of charged droplets of the dispersed phase sideways from the interacting means through the space towards the side of the column. 8. Apparatus for electrostatically enhanced solvent extraction, the apparatus comprising a column for the counter-current flow of a continuous phase and a dispersed phase therethrough, one or a plurality of inlets for the introduction of the dispersed phase into the column to contact the continuous phase, at least one receptacle at an intermediate location in the column for catching the dispersed phase flowing through the column, the receptacle having side walls, an open top, and a closed base, the receptacle being located to define a space at the side thereof between the side of the receptacle and the column and through which space the continuous phase is arranged to flow, the receptacle having a plurality of discharge ports extending through the side walls and arranged to discharge the dispersed phase through the side walls and through the space towards the side of the column, and means for applying an electric potential to the receptacle to cause the discharge of charged droplets of the dispersed phase through the discharge ports. 9. Apparatus as claimed in claim 8 wherein the space defined is substantially annular. 10. Apparatus as claimed in claim 8, wherein the base of the receptacle is inwardly dished. 11. Apparatus as claimed in claim 8, wherein a collector is disposed below the receptacle, the collector having radially disposed riser ports for the upward flow of the continuous phase therethrough, and a central outlet for the downward flow of dispersed phase therethrough. 12. Apparatus as claimed in claim 11, wherein the collector has a side wall that extends to at least the same height in the column as the discharge ports, and the electric potential is between the discharge ports and the side wall. 13. Apparatus as claimed in claim 11, wherein the base of the collector is of frusto-conical form to bias the flow of the dispersed phase towards the central outlet, and bias the flow of the continuous phase towards the ports. 14. Apparatus as claimed in claim 11, wherein a plurality of receptacles and collectors are located in series in the cell. 15. Apparatus as claimed in claim 14, wherein the electrical potential means is arranged to apply opposite polarity electric potentials to alternate receptacles. 16. Apparatus as claimed in claim 14, wherein the electrical potential means is arranged to apply the same polarity electric potential to all the receptacles. 17. Apparatus as claimed in claim 8, wherein the electric potential means is adapted to apply at least 1.5 kv/cm. |
summary | ||
claims | 1. A plug device for jet pumps installed in nuclear power plant vessels, comprising: two covers (1-2) that form two independent plug units, one fitted on the even pump, which incorporates two plugs (6), and the other one on the odd pump of each jet pump assembly, which incorporates three plugs (6), both mounted on a common base (3) at the end of respective arms (4), articulated in a central area, wherein the three plugs (6) located on the nozzles of the odd pump are blind, and the two plugs (6) located on the nozzles of the even pump have respective fluid inlets (11) communicating through a channeled arm (13) towards an outlet (14), enabling fluid circulation during decontamination tasks, whereas three nozzles of the odd pump of each jet pump assembly are sealed; the arms (4) being actuated by mechanical or hydropneumatic means, applying the corresponding plug on the nozzles of the jet pump once the base (3) of the plug is supported on a mixer part of the jet pump (10), through a guide (12) centering the plug device therein, and in respective prolongations of the base (3) consisting of a lower reaction (8) serving as support both when introducing the plug in the pump and when demounting same, and in an upper reaction (9) serving to allow the closure of the plug against the nozzles of the jet pump which are located on the plug. 2. The device of claim 1, in which the covers (1) forming each plug unit are mounted in the arm (4) in a pivoting manner on a central point (7), which allows self-alignment of a plane of the plugs (6) with a plane of the nozzles to be blocked. 3. The device of claim 1, in which the covers carrying plugs are attached or are a single cover, incorporating the five plugs (6) in a suitable arrangement and are actuated by means of a single arm (4) and clamping screw (5). 4. The device of claim 1, further comprising an adjusting screw (15) arranged between the arm (4) and the guide (12) at the opening of the mixer of the jet pump which allows assuring a parallelism of a sealing surface of the plug with a plane of the nozzles to be plugged of the jet pump. |
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048662844 | summary | FIELD OF THE INVENTION The invention relates to an irradiation device having a base which is pivotably connected by means of an arm to a housing in which radiation sources are accommodated and in which a radiation exit side is present, the housing, the arm and the base being collapsible to a compact unit. An irradiation device of this type is known from European Patent Application No. 0,106,395 laid open to public inspection. BACKGROUND OF THE INVENTION The known irradiation device for irradiating the human body with ultraviolet radiation has a housing accommodating radiation sources such as high-pressure lamps with reflectors arranged behind them. The radiation leaves the housing via the radiation exit side (which is possibly provided with a transparent plate). The housing is pivotably connected to the wheeled base via a telescopic arm. When using the device the base is placed, for example alongside a bed on which a person to be irradiated is lying. The housing is preferably placed above the body approximately halfway its feet and head. The longitudinal direction of the housing is then generally positioned in a direction perpendicular to the axis of the body. For a satisfactory irradiation of the entire body it is necessary that the said reflectors have a special shape. However, it has been found that the differences in the radiation intensity to which the different parts of the body are exposed are relatively large. SUMMARY OF THE INVENTION The invention has for its object to provide a compact irradiation device with which a homogeneous irradiation of the entire human body is obtained. According to the invention an irradiation device of the type described in the opening paragraph is therefore characterized in that the housing has two juxtaposed elongated parts accommodating radiation sources, which parts are pivotable with respect to each other about an axis at right angles to the plane through the radiation exit side. In the operating condition of the device the two parts of the housing can easily be folded out by a user. Then a homogeneous irradiation throughout the length of the body is possible. The advantage of the device according to the invention is that the dimensions of each of the parts of the housing are relatively small as compared with the housing of the known device. The number of radiation sources is the same. The two parts of the housing are preferably pivotable (for example with respect to a coupling member located between the two parts) by means of pivots about axes which are at right angles to the radiation exit side of the housing. Right angles is herein understood to mean perpendicular with a deviation of up to approximately 30.degree. in specific embodiments. In these embodiments the construction of arm and base has been chosen to be such that the device is collapsible in a simple manner. Each of the two parts of the housing accommodates a plurality of radiation sources behind which reflectors are arranged. The parts of the housing have such a shape that, in the folded state, they are accommodated in recesses in the base together with the arm which is pivotably connected to the base. A preferred embodiment of the irradiation device according to the invention is characterized in that the said parts are pivotable up to a position in which their longitudinal axes are in alignment. The advantage of such an embodiment is that if the base of the irradiation device is placed alongside a bed, the housing above this bed assumes such a position that an optimum homogeneous irradiation of a person lying on the bed is obtained throughout the length of his body. The two parts are actually each pivotable through 90.degree. with respect to their initial position. The device then has a stable position. The pivotal points are preferably in the form of pivots with known arresting means being provided. In a further embodiment of the device the two parts of the housing are pivotably secured to one end of an intermediate arm whose other end is connected to the arm secured to the base. The latter arm is preferably in the form of a telescopic arm. The said intermediate arm is in turn also pivotably secured to the telescopic arm. The housing can then easily be placed over a person to be irradiated. The intermediate arm is preferably provided with a coupling member to which the two parts of the housing are pivotably secured. The device is then more easily collapsible. This is notably the case if the intermediate arm (which connects the two parts of the housing to the said telescopic arm) is in the form of a combination of two elongated parallel bars whose ends remote front housing are pivotably connected to a short pivotal bar which is pivotably secured to the top of the arm connected to the base. In a special embodiment this bar is bridged by a gas spring whose ends are pivotably secured to the arm and one of the parallel bars, respectively. The forces on the arm, the pivotal bar and the elongated bars are absorbed by means of the gas spring. |
053902211 | summary | In a boiling water nuclear reactor fuel bundle, a debris catching arrangement is disclosed for incorporation within a flow plenum up stream of the rod supporting grid of the lower tie plate assembly. The disclosed debris catching designs include the two phase separation of the heavier debris from the lighter transporting water by flow direction (momentum) change with the debris directed and detoured to a trapping structure. This allows for partial flow bypass around the trapping structure, to eliminate potential flow blockage concerns. BACKGROUND OF THE INVENTION Boiling water nuclear reactors operate for many years. Commencing with their initial construction and through their service lives, these reactors may accumulate debris in their closed circulation moderator systems. This debris can become an operating hazard if the debris is allowed to enter into the fuel bundle containing core region having the heat generating fuel rods. In order to understand this problem, a summary of reactor construction as it relates to the accumulation of debris needs first to be given. Thereafter, fuel bundle construction will be set forth. Emphasis will be given to the need to preserve substantially unchanged the regions of pressure drop along the flow paths. Thereafter, the effects caused by debris entering into the fuel rod region of the fuel bundles will be summarized. Boiling water nuclear reactor construction can be simply summarized for purposes of understanding the debris entrainment problem. Such nuclear reactors are provided with a large, central core. Liquid water coolant/moderator flow enters the core from the bottom and exits the core as a water steam mixture from the top. The core includes many side-by-side fuel bundles. Water is introduced into each fuel bundle through a fuel bundle support casting from a high pressure plenum which is situated below the core. Water passes in a distributed flow through the individual fuel bundles, is heated to generate steam, and exits the upper portion of the core as a two phase water steam mixture from which the steam is extracted for the generation of electricity. The core support castings and fuel bundle inlets are a source of pressure loss in the circulation of water through the core. This pressure loss assures the substantially even distribution of flow across the individual fuel bundles of the reactor core. When it is remembered that there are as many as 750 individual fuel bundles in a reactor core, it can be appreciated that assurance of the uniformity of flow distribution is important. To interfere with the pressure drop of the fuel bundle flow paths could effect the overall distribution of coolant/moderator within the fuel bundles of the reactor core. Having set forth the construction of the boiling water nuclear reactor in so far as is appropriate, attention can now be directed to the construction of the fuel bundles themselves. The fuel bundles for a boiling water nuclear reactor include a fuel rod supporting lower tie plate assembly, which lower tie plate assembly is a cast structure. The lower tie plate assembly includes at its lowest point a downward protruding bail covering an inlet nozzle. This inlet nozzle includes entry to an enlarged flow volume or tie plate plenum within the lower tie plate. At the upper end of the flow volume, there is located a rod supporting grid. The rod supporting grid has two purposes. First, the rod supporting grid provides the mechanical support connection for the weight of the individual fuel rods to be transmitted through the entire lower tie plate to the fuel support casting. Secondly, the rod supporting grid provides a flow path for liquid water moderator into the fuel bundle for passage between the side-by-side supported fuel rods. Above the lower tie plate, each fuel bundle includes a matrix of upstanding fuel rods--sealed tubes each containing fissionable material which when undergoing nuclear reaction produce the power generating steam. The matrix of upstanding fuel rods includes at the upper end a so-called upper tie plate. This upper tie plate holds at least some of the fuel rods in vertical side-by-side alignment. Some of the fuel rods are attached to both the upper and lower tie plates. Usually, there are included between the upper and lower tie plates water rods for improvement of the water moderator to fuel ratio, particularly in the upper, highest void fraction region of the fuel bundle. Fuel bundles also include about seven fuel rod spacers at varying elevations along the length of the fuel bundle. These spacers are required because the fuel rods are long (about 160 inches) and slender (about 0.4 to 0.5 inches in diameter), and could come into abrading contact under the dynamics of fluid flow and nuclear power generation within the fuel bundles. The spacers provide appropriate restraints for each fuel rod at their respective elevations and thus prevent abrading contact between the fuel rods and maintain the fuel rods at uniform spacing relative to one another along the length of the fuel bundle for optimum performance. As will hereafter be developed, these spacers are sites where debris can be trapped and damage the fuel rods. Each fuel bundle is surrounded by a channel. This channel causes water flowing between the tie plates to be restricted to only one bundle in an isolated flow path between the tie plates. The channel also serves to separate the steam generating flow path through the fuel bundles from the surrounding core bypass region, this region being utilized for the penetration of the control rods. The water in the bypass region also provides neutron moderation. In the operation of a boiling water nuclear reactor, it is important to understand that the maintenance of the originally designed flow distribution is important. Specifically, from the lower (high pressure) plenum inlet to the core to the outlet from the core of the steam and water mixture through the upper tie plates of the fuel bundles, about 20 pounds per square inch (psi) of pressure drop is encountered at typical 100% power/100% flow operating conditions. About 7 to 8 psi of this pressure drop occurs through the fuel support casting. This pressure drop is mainly to assure the uniform distribution of coolant/moderator flow through the many fuel bundles making up the core of the reactor and is related to the prevention of operating instabilities within the reactor at certain operating states of the reactor. At the lower tie plate of each fuel bundle, from the inlet nozzle into the flow volume and through the fuel rod supporting grid, about 1 to 11/2 psi pressure drop occurs which contributes to establishing flow distribution between the individual fuel rods of each fuel bundle. Finally, through the fuel bundle itself--from the lower supporting grid to the exit at the upper tie plate--about 11 psi of pressure drop usually occurs. When new fuel bundles are introduced into a reactor core, their total pressure drop should be preserved. Otherwise, the coolant/moderator flow distribution could be compromised. Having summarized the construction and operation of a boiling water nuclear reactor, the problem of debris resident within the closed circulation moderator system of the reactor can now be understood. Typically debris within boiling water nuclear reactors can include extraneous materials left over from reactor construction. Further, corrosion during the reactor lifetime also liberates debris. Finally, and during the numerous outages and repairs, further debris accumulates. It will therefore be understood that nuclear reactors constitute closed circulation systems that essentially accumulate debris with increasing age. It has been discovered that a particularly vexing and usual place for the deposition of debris is in the fuel bundles between the fuel rods particularly in the vicinity of the fuel rod spacers. It will be remembered that each fuel rod is surrounded by the spacer at the particular elevation of the spacer. Debris particles tend to lodge between the spacer structure and the fuel rods and often dynamically vibrate with the coolant/moderator flow in abrading contact to the sealed cladding of the fuel rods. Such flow induced vibration within the reactor, can and has both damaged--as by fretting--as well as ruptured the cladding of the fuel rods. If a sufficient number of cladding ruptures occurs, plant shutdown could be necessary. It is to be understood that modern nuclear plants have both redundancy and many safety systems designed to counteract anticipated operating casualties, such as fuel rods becoming punctured by debris. Such failures are not catastrophic. However, in almost all cases they result in the plant operating at less than optimum efficiency. Thus, it is highly desirable to reduce the incidence of debris damage to fuel rods. It will be further understood that to a certain extent the rod supporting grid of the fuel bundle acts as a strainer. Debris exceeding the dimension of grid passageways cannot pass through to the fuel bundles. However, it has been found that debris--especially debris with "sail areas"--such as metal shavings, wire and the like--move past the lower rod supporting grid and can become lodged between the fuel rods and spacers. SUMMARY OF THE PRIOR ART Prior art attempts at the placement of devices for preventing debris from entering into the regions of the fuel rods have included alteration of the grid support structure of the lower tie plate assembly. In Nylund U.S. Pat. No. 5,100,611 issued Mar. 31, 1992, an alteration to the grid structure is disclosed. This alteration includes placing the required through holes of the grid structure with flow channel parts that have center lines that are non-collinear. Because these flow channels are part of the fuel rod supporting grid, the size of the through holes is necessarily large to preserve the rod supporting grid strength and the area over which the holes are distributed is only co-extensive to the lower tie plate assembly at the supporting grid. Attempts to screen debris have been made in pressurized water reactors. In Bryan U.S. Pat. No. 4,664,880 issued May 12, 1987 a wire mesh debris trap is utilized at the bottom of a pressurized water reactor fuel bundle. In Rylatt U.S. Pat. No. 4,678,627 issued Jul. 7, 1987, this structure is modified to include a debris retaining trap. These pressurized water reactor fuel bundles constitute open structures and lack the channel confined flow path between the lower high pressure plenum through the fuel support casting and the upper and lower tie plates of the fuel bundle common to boiling water nuclear reactors. The channel structure, required in boiling water nuclear reactor construction, is wholly absent in pressurized water reactor construction. Since flow can occur between adjacent fuel bundles in a pressurized water reactor along the entire length of the fuel bundles, the placement of the disclosed screens and traps does not occur within a confined flow path. In one prior art debris catching device, the lower tie plate is modified with serpentine path--almost in the form of a chevron. Overlying this construction there are placed rod supporting bars so that the weight of the rods does not crush the underlying serpentine path. SUMMARY OF THE INVENTION In a boiling water nuclear reactor fuel bundle, debris catching arrangements are disclosed for incorporation within the flow plenum upstream or below the rod supporting grid of the lower tie plate assembly. The device is preferably placed within the lower tie plate flow plenum between the fuel bundle inlet orifice and the rod supporting grid structure supporting the fuel rods; alternate placement can include any inlet channel upstream of the fuel rods including the fuel support casting. The disclosed debris catching designs includes strainer structures defining spatially separated straining or obstructing layers imparting to the fluid in the plenum a circuitous flow path. This circuitous flow path causes the two phase separation of the heavier debris from the lighter transporting water by flow direction (momentum) change with the debris directed and detoured to a trapping structure. Further, a strainer structure is provided in the plenum that does not constitute a continuum of fine structure across the strained plenum which might become clogged to the extend that overall flow is restricted; spatial separation exists in between the disclosed strainer structures. The straining structure is positioned so that eventual trapping of the debris occurs upon cessation of flow so that with removal of the plenum from the reactor, such as removal of the fuel bundle, the debris is likewise removed. Embodiments are disclosed which include swirling deflection, cone deflection, and strainer structure deflection of debris. |
041359704 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENTS In FIG. 2, a nuclear reactor 60 is, for example, a fast breeder reactor comprising a reactor container 62, a core 64 disposed within the reactor 60, a rotating plug 66 for closing the upper open end of the reactor container 62, and the other associated means, not shown. A liquid metal coolant 68, such as liquid sodium, for capturing heat evolved at the core 64 is circulated in the reactor in a direction indicated by arrows. A pipe 76 extends through the rotating plug 66 in an air-tight fashion and has one end extending into the coolant 68 and the other end connected through a valve 70 to a first inlet 74 of a gas-liquid separator 72. The gas-liquid separator 72 comprises a body 77 having the inlet 74 for introducing the coolant from the reactor 60 into the body 77, an outlet 80 provided in the bottom of the body 77 to discharge the coolant 78 in the body 77, and cover gas inlet and outlet 82 and 84 oppositely provided in the upper side walls of the body 77. The outlet 80 of the separator 72 is connected through a pump means 86 and flow meter 88 to a coolant inlet 91 of an iodine 135.sub.I) adsorpting device 90. Outside of the adsorption device 90 are disposed a cooling means 92 for cooling the coolant through the adsorption device 90 to a temperature of below 200.degree. C. and a heating means for heating the coolant through the adsorption device 90 to a temperature of above 300.degree. C. An absorption member 94 made of an iodine adsorbing material such as stainless steel, nickel or alumina is disposed within the adsorbing device 90. The adsorption member 94 is so constructed that it has a wide contact area without blocking a free flow of the coolant through the adsorption device 90. For example, it is constructed of a mesh-like member. A coolant outlet 96 of the adsorption device 90 is connected to a pipe 100 which extends through a valve 98 and plug covering 66 into the coolant 68 in the reactor 60. The outlet 96 of the adsorption device 90 is also connected through a valve 102 to the coolant inlet 74 of the separator 72. A vapor trap 108, flow meter 110, radioactibity selector 112, pump 114, valve 116, gaseous fission product trap 118 and a valve 120 are arranged between the cover gass inlet 82 and the cover gass outlet 84. Between the pump 114 and the valve 116 is connected a pipe 122 leading to a cover gas treating system, not shown. A valve 124 is mounted over the pipe 122. A pipe 126 has one end connected to a cover gas supply system, not shown, and the other end connected to the inlet 82 of the separator 72. The pipe 126 is also connected partway to the pipe 106 leading to the valve 120 and has a valve 128. The valves 70, 98, 102, 116, 120, 124 and 128 and pumps 86 and 114 are all of an electromagnetic type and are controlled by corresponding output signals from a control device 130. The detection system of this invention will now be described below. Suppose that the cover gas is sealed in the separator 72 and that each valve is closed with each pump in the inoperative state. When in this case the valves 70 and 98 are opened by the control device 130 and the pump 86 and cooling means 92 are operated, the coolant 68 is raised from the reactor 60 and circulated through the valve 70, separator 72, pump 86, flow meter 88, iodine absorption device 90 and valve 98 into the reactor 60. Since the coolant 78 introduced into the separator 72 is hot (usually above 400.degree. C.), when 135m.sub.Xe produced by a 135.sub.I decay is included in the coolant 78 in the separator 78, 135m.sub.Xe is transferred to the inert cover gas in the separator 72. The 135m.sub.Xe -purged coolant is introduced from the separator 72 through the pump 86 and flow meter 88 into the iodine absorption device 90. The adsorbing member 94 constituting the iodine adsorption device 90 cools the coolant introduced into the adsorption device 90 to a temperature of below 200.degree. C. by means of the cooling means 92. Since the mesh-like adsorbing member 94 has a high cooling and high heating effect due to its large contact area, the coolant passed through the adsorbing member 94 is cooled to a temperature (below 200.degree. C.) substantially equal to the temperature (below 200.degree. C.) of the adsorbing member 94. When the coolant is so cooled at the adsorbing member 94, most of 135.sub.I in the coolant is adsorbed on the surface of the adsorbing member 94. Even if the adsorbed 135.sub.I is decayed into 135m.sub.Xe, it remains adsorbed on the adsorbing member 94 under a temperature of below 200.degree. C. The coolant through the adsorbing member 94 is passed through the pipe 100 into the reactor 60. When a predetermined amount of coolant sufficient to attain a saturation adsorption state at the adsorption member 94 is passed through the adsorption member 94, the valves 70 and 98 are closed, the valves 124 and 128 are opened, and the pump 86 and cooling means 92 stop their operation. Thus, the sampling of the coolant 68 from the reactor 60 is stopped to cause a coolant flow through the separator 72, pump means 86, flow meter 88 and adsorption device 90 to be stopped. On the other hand, the 135m.sub.Xe -entrained cover gas in the separator 72 is conducted to the cover gas treating system for treatment and a fresh cover gas is supplied through the pipe 126 into the separator 72. During the treatment of the cover gass the valve 102 may be opened and the pump 86 be operated to cause the coolant in the separator 72 and adsorbing device 90 to be circulated through the pump means 86 and flow meter 88. After the passage of a predetermined time period sufficient to permit the cover gas to be cleaned, the valves 128 and 124 are closed by the control means 130, the valve 102 opened, and the pump 86 and cooling means operated. When the adsorption member 94 of the iodine adsorption device 90 is heated by the heating means 93 to a temperature of above 300.degree. C., the coolant passed through the adsorption member 94 is heated to a temperature (above 300.degree. C.) equal to the temperature of the absorbing member 94 for the same reasons as explained in connection with the cooling of the coolant. When the coolant is heated to above 300.degree. C., most of 135m.sub.Xe absorbed onto the absorbing member 94 is moved into the cover gas in the separator 72. This will be evident from an experimental data, indicating a relation of the temperature of the circulating coolent to the percentage of movement of 135m.sub.Xe into the cover gas in the separator 72 (see FIG. 3). It will be apparent from FIG. 3 that when the coolant is at a temperature of below 200.degree. C., the Xenon 135m.sub.Xe remains absorbed with almost negligible movement of 135m.sub.Xe into the cover gas and that when the coolant is heated to a temperature of above 300.degree. C. a 90% 135m.sub.Xe movement is affected into the cover gas in the separator 72. When, therefore, the coolant is circulated under a temperature of above 300.degree. C. for a predetermined time, most of 135m.sub.Xe produced by the decay of 135.sub.Xe mixed in the sampled coolant is moved into the cover gas in the separator 72. Then, the valves 116 and 120 are opened and the pump 114 is operated. The 135m.sub.Xe -entrained cover gas is introduced through the vapor trap 108 and flow meter 110 into the radioactivity detector 112 where the intensity of radioactivity of 135m.sub.Xe is detected. Then, the 135m.sub.Xe -entrained coolant is introduced through the pump 114 and valve 116 into the gaseous nuclear fission product trap 118 where 135m.sub.Xe in the coolant is trapped. The 135m.sub.Xe -purged coolant is returned into the separator 72. When such an operation is completed, the valves 70 and 98 are opened by the control device 58, the valve 102 closed, the heating means stopped, the cooling means operated, and the coolant 68 again sampled from the reactor 60. The failure of the nuclear fuel rod is detected from the intensity variation of radioactivity of 135m.sub.Xe as obtained from such a series of operations. Generally, the amount of adsorption, A, of 135.sub.I in the coolant which is adsorbed by the adsorption device 90 can be expressed by: EQU A = KSC where K: distribution coefficient of 135.sub.I PA1 S: surface area of the adsorption member PA1 C: density of 135.sub.I in the coolant The distribution coefficient K of 135.sub.I is the function of temperature and is increased in valve with a decrease in temperature. The distribution coefficient K of 135.sub.I was examined using liquid sodium as a coolant and a mesh-like stainless steel structure as an adsorbing member 94, the result of which is as follows: EQU Log K = -5.6 + 3 .times. 10.sup.3 [1/T(K)] with C representing the density of 135.sub.I and the surface area S of the adsorbing member 94 it follows that for a liquid sodium temperature of 200.degree. C., 5500cm.sup.3 of liquid sodium needs to be passed through the adsorption member 94 to reach a saturation absorption amount and for 160.degree. C., 21000cm.sup.3 of liquid sodium. Since the system of this invention is equivalent to the sampling of a large amount of coolant from the reactor and detection of the concentration 135.sub.I in the coolant, a detection accuracy can be enhanced. If the above-mentioned mesh-like structure is used as the iodine adsorption member 94, the surface area of 1000cm.sup.2 can be attained by a small capacity of the order of about 200cm.sup.3, making the detection system compact as a whole. FIG. 4 shows a nuclear fuel rod failure detection system according to another embodiment of this invention. This embodiment is substantially similar to the embodiment in FIG. 2 except that a delayed-neutron/gaseous fission product detecting system 160 is provided between the atomic reactor 60 and the 135.sub.Xe radioactivity counting system. FIG. 4 shows the delayed-neutron/gaseous fission product detecting system 160 alone, to which the following explanation is directed. In FIG. 4 a gas-liquid separator 162 is provided having substatially the same construction as in the system in FIG. 2. The separator 162 has a coolant inlet 164 connected through a flow meter 166 and pump 167 over a pipe 76 to the reactor 60 and a coolant outlet 168 connected by a pipe 76 to the gas-liquid separator 72. Between the pipe 76 and the pipe 100 is disposed a pipe 169 by way of which the coolant flows into the pipe 100. Between a cover gas inlet 172 and a cover gas outlet 174 in the separator 162 is disposed a pipe 176 for a cover gas passage 1, on which are mounted a vapor trap 178, flow meter 180, radioactivity detector 182, pump 184, valve 186, gaseous fission product trap 188 and valve 190 in this order. Between the pump 184 and the valve 186 is connected a pipe 192 leading to a cover gas treating system (not shown) through a valve 194. Between the valve 190 and the cover gas inlet 174 is connected a pipe 196 which communicates with the pipe 176 at the valve 190 side. The pipe 196 is also connected through a valve 198 to a cover gas supply system, not shown. A neutron shielding wall member 200 is located between the separator 162 and the reactor 60, and a neutron detector 202 between the wall member 200 and the separator 162. The valves 170, 186, 190, 194 and 198 as well as the pumps 167 and 184 are electromagnetically operated by the output signals from the control device 130. The delayed-neutron/nuclear fission product detection system of this invention will be operated as follows. Suppose that a cover gas is sealed in the separator 162 and that each valve is closed with each pump in the in operative state. When the valves 70 and 98 (FIG. 2) are opened and the pump 167 is operated in synchronism with the operation of the pump 86 and cooling means 92, the coolant is sampled from the reactor 60 into the separator 162. Since the neutron detector 202 is disposed in the neighborhood of the separator 162, if any fission product emitting a delayed neutron is included in the coolant, the delayed neutron is detected at the neutron detector 202. Any gaseous fission product, if present in the sampled coolant, is moved into a cover gas in the upper space of the separator 162 during its stay at the separator 162. When the valves 116 and 120 are opened to permit the pump 114 to be operated, the valves 186 and 190 are opened in synchronism therewith and the pump 184 is operated, causing the cover gas to flow through the vapor trap 178, flow meter 180, radioactivity detector 182, pump 184, valve 186, gaseous fission product trap 188 and valve 190. The gaseous fission product in the cover gas has its radioactivity counted at the radioactivity detector 182. Then, the coolant present in the separator 162 is introduced into the separator 72 shown in FIG. 2. The subsequent operation is the same as that explained in connection with FIG. 2 and further explanation is omitted for gravity. In this embodiment the occurence, mode and extent of a failure of the nuclear fuel rod are eventually detected, collectively taking into consideration the output signals from the delayed-neutron detector 202, gaseous fission product radioactivity detector 182 and 135m.sub.Xe radioactivity detector 112. That is, signal detection only at the detector 182 means that pinholes occur in the nuclear fuel rod. When a signal is detected from the detector 112, this means that the meat of the nuclear fuel is exposed or released. When a signal is detected from the detector 202, this means that the meat of the nuclear fuel is exposed or released to a considerable extent. The extent of the fuel rod failure can be estimated from the intensity of signal. In this embodiment, not only can the damage of the fuel rod be detected in high sensitivity, but also a proper treatment can be effected as well. The systems as shown in FIGS. 2 and 4 may be provided in large numbers with respect to the reactor and any idle time required for coolant something be eliminated by operating these systems on a time showing basis. A detection accuracy can be enhanced by varying the sampling position of an intrareactor coolant for each detection time. Where no share detection response is required, if redioactivity is counted by moving into the cover gas 135.sub.Xe produced by 135.sub.I decay, 133m.sub.Xe and 133.sub.Xe by 133.sub.I decay, 131m.sub.Xe by 131.sub.I decay etc., detection accuracy is further enhanced. Although in the above-mentioned system liquid sodium is used as a coolant, this invention is not restricted thereto. For example, NaK, or a Na-K alloy, may be likewise used with respect to the reactor. |
claims | 1. A radiation shielding method comprising:installing a hollow container at a predetermined portion of an object to be shielded;feeding fluid into the container via a feeding hose;supplying a shielding material to the feeding hose and transporting and filling a granular shielding material into the container by the fluid;extracting the shielding material filled in the container from the container together with fluid discharged to outside of the container via a returning hose, while feeding fluid into the container via a feeding hose, in a state that the shielding material is filled in the container; andrecovering the shielding material from the fluid, whereinthe extracting the shielding material filled in the container from the container includes injecting the fluid, which is fed from the feeding hose to the returning hose, from an injection port formed in a tubular shaped injecting nozzle toward an inlet formed in a tubular shaped fetching member,the fetching member connected to the returning hose in the container,the inlet having a larger diameter than the injection port, andthe inlet arranged so as to face the injection port. 2. The radiation shielding method according to claim 1, wherein at the feedingfluid into the container, liquid is used as the fluid and the liquid is filled in the container via the feeding hose. 3. The radiation shielding method according to claim 1, wherein the container is mounted on the object to be shielded at all times. 4. A radiation shielding device comprising:a hollow container installed at a predetermined portion of an object to be shielded;a fluid feeding unit that feeds fluid into the container via a feeding hose;a shielding-material supply unit that supplies a granular shielding material to the feeding hose;a shielding-material extracting unit that circulates the shielding material filled in the container together with fluid discharged to outside of the container via a returning hose, while feeding fluid into the container via a feeding hose; anda shielding-material recovering unit that recovers the shielding material from the fluid, whereinthe shielding-material extracting unit includes an injection nozzle having a tubular shape and an injection port that injects the fluid fed into the container, and a fetching member having a tubular shape and an inlet for fetching the shielding material together with fluid discharged from the container, which are provided in the container, and an injection port of the injection nozzle is arranged toward the inlet of the fetching member,the inlet has a larger aperture than the injection port, andthe inlet is disposed so as to face the injection port. 5. The radiation shielding device according to claim 4, wherein the shielding-material extracting unit includes a switching unit that switches a feeding direction of fluid in a reverse flow mode of the fluid. 6. The radiation shielding device according to claim 4, wherein the hose for circulating the shielding material together with fluid between the shielding-material supply unit and the container is made to be transparent. 7. The radiation shielding device according to claim 4, wherein the hose for circulating the shielding material together with fluid between the shielding-material recovering unit and the container is made to be transparent. 8. The radiation shielding device according to claim 4, wherein water is used as the fluid, and a pellet containing tungsten is used as the shielding material. |
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description | This application is a national filing of PCT application Serial No. PCT/IB2013/054003, filed May 16, 2013, published as WO 2013/182928 A1 on Dec. 12, 2013, which claims the benefit of U.S. provisional application Ser. No. 61/655,602 filed Jun. 5, 2012, which is incorporated herein by reference. The following relates generally to x-ray computed tomography. It finds particular application in conjunction with scanner calibration and image artifact compensation, and will be described with particular reference thereto. However, it will be understood that it also finds application in other usage scenarios and is not necessarily limited to the aforementioned application. X-ray computed tomography systems such as cone-beam computed tomography (CBCT), 3D rotational angiography (3DRA), x-ray CT (XCT), interventional x-ray, C-arm and the like, emit x-rays and detect the emitted x-rays after passing through a subject in order to reconstruct images. Air calibration or rotational gain calibration projection images are typically collected at each of a plurality of gantry positions without a subject and stored for the uniformity correction for the corresponding position. The uniformity correction data stored for each gantry position is used in image reconstruction. The air calibration determines x-ray attenuation and intensity changes not caused by the subject, but by elements of the scanner and other sources. The air projection image of a theoretical, ideal imaging system at each gantry angle is a uniform blank image. In practice, the air projection images include non-uniformities from attenuating structures in the beam path, non-uniform illumination by the x-ray source, non-uniform detector sensitivity, and the like. When the patient is imaged these non-uniformities are superimposed on the absorption profile of the patient. The non-uniformities are compensated by normalizing the patient projection image at each gantry angle with the air projection image at the same gantry angle to produce corrected patient projection images. The corrected projection images from a plurality of gantry angles around the patient can be reconstructed into a 3D image. As CT systems evolve with more open systems such as C-arm systems and simpler, less rigid gantries, a change in the non-uniformities (and the respective air projection images) both within one acquisition and between different acquisitions can be observed. The changes are not necessarily reproducible. Some changes of air projection images in the open systems are attributable to elements which move relative to each other. For example, an element located on one arm can move different than an element on another arm. A source at one end can move different than a detector at the other end. With system wear, arm movement, accidental impact, thermal expansion/contraction, and other environmental factors, individual elements even on the same arm can move relative to each other. For example, even though an anti-scatter grid is fixed to a detector, a tilt in the detector causes a change of the shadows induced by the lamellae of the grid. The differing changes in position can occur with each movement, which can lead to uncompensated image artifacts and inaccurately reconstructed absorption coefficients when imaging a subject. The air calibrations which correct for attenuation and intensity changes due to scanner elements may not remain valid from a time of generating the air calibration to a time of imaging the patient. The following discloses a new and improved system and method of tomographic image calibration which addresses the above referenced issues, and others. In accordance with one aspect, an x-ray computed tomography system includes a gantry, a plurality of elements, and one or more processors. The gantry moves to different orientations and generates x-ray data which includes image projection data at a plurality of the orientations. The plurality of elements connect to the gantry and cause x-ray attenuation of the generated projection data. The one or more processors are programmed to receive the generated x-ray data and decompose the received image projection data into indications of relative positions of the plurality of elements at different orientations of the gantry. In accordance with another aspect, a method of x-ray computed tomography calibration includes receiving x-ray data which includes image projection data at each of a plurality of gantry orientations around an imaging region. The received image projection data is decomposed to derive relative positions of a plurality of elements at one or more gantry orientations, each of the elements causing x-ray attenuation attributable to the elements in the received image projection data. A correction of measured attenuation is generated based on the relative positions of the plurality of elements. In accordance with another aspect, an x-ray computed tomography system includes a rotatable gantry, a memory, a decomposition unit, and a correction unit. The rotatable gantry carries elements which include an x-ray source, an x-ray filter, a shutter/collimator, an x-ray detector, and an anti-scatter grid, and the gantry moves to different orientations. The memory stores attenuation contributions attributable to each of the elements. The decomposition unit decomposes air scan projection images at the different orientations into relative positions of each of the elements. The correction unit adjusts the correction of attenuation to projection image data based on the relative positions of the each of the elements. One advantage is artifact reduction. Another advantage resides in dynamic artifact compensation which dynamically adjusts during an imaging session. Another advantage resides in a post processing technique for correcting artifacts due to scanner element movement. Another advantage resides in more accurately reconstructed absorption coefficients. Another advantage resides in the incorporation into existing systems and procedures. Another advantage resides in flexibility in adapting to existing and new more open CT gantry designs. Still further advantages will be appreciated to those of ordinary skill in the art upon reading and understanding the following detailed description. With reference to FIG. 1, a typical flat panel x-ray computed tomography system includes an x-ray source 2, an x-ray filter 4, a shutter/collimator 6, an x-ray detector 8 and an anti-scatter grid 10. The x-ray source 2 such as an x-ray tube anode emits x-rays. The x-ray filter 4 includes a beam shaper or filtration unit which filters the x-rays. The shutter/collimator 6 defines the extent of the beam of x-rays which pass through the subject 12 and impact the x-ray detector 8. The x-ray source 2, the x-ray filter 4 and the shutter/collimator 6 are typically located on one arm or at the end of an arm of the flat-panel x-ray computed tomography system. However, other geometries, such as a ring, and the like are also contemplated. After passing through the subject 12 the x-rays pass through an anti-scatter grid 10 and are detected by an x-ray detector 8. The x-ray detector 8 and anti-scatter grid 10 are typically located opposite the x-ray source 2, the x-ray filter 4 and the shutter/collimator 6 such as on another arm or the other end of the arm of the flat-panel x-ray computed tomography system. The shutter/collimator 6 typically limits the cross section of the x-ray beams to the cross section of the x-ray detector 8 or to an anatomical region of interest to limit the patient's exposure to x-rays. The x-ray detector 8 detects the x-rays passed through the subject 12 in the field of view. The x-ray detector 8 typically includes an array of detector elements which detect x-rays in areas each corresponding to a pixel. The anti-scatter grid 10 such as an assembly of lamellae or plates, typically perpendicular to the detector surface, limits the impact of scatter in images. With reference to FIG. 2, a projection image at one gantry position of a typical air calibration scan is shown. The air calibration scan measures at each detector element or pixel, the intensity of the x-ray received from the x-ray source 2. The air calibration projection images are generated for a number of gantry orientations. The projection image of FIG. 2 shows a darken area on the right where the x-ray filter 4 is thinnest and the x-rays are most intense. The light area to the left indicates the thickest portion of the x-ray filter 4 where the x-rays are the least intense. Although difficult to discern, the air calibration image has a series of evenly spaced thin white lines where the lamellae of the anti-scatter grid block 10 the x-rays and cast shadows on the x-ray detector 8. The air projection image is representative of an off-center x-ray detector 8 such as FIG. 1. An air calibration projection image from a system with a symmetric x-ray detector 8 would show light areas at both ends with the intense area centered. With reference to FIGS. 3 and 4, a C-arm embodiment of the system 14 is shown. The system includes a gantry 15, which in this example includes a “C” shaped arm 16. The system includes elements 18 disposed at opposite ends of the C-arm 16. The system elements 18 include the x-ray source 2, the x-ray filter 4, the shutter/collimator 6 disposed at one end, and the x-ray detector 8 and anti-scatter grid 10 disposed at the opposite end. The C-arm 16 is attached to a horizontal arm 20 which has a pivot 22. A drive (not visible) rotates the C-arm 16 along a trajectory 24 around an axis of the pivot to move the x-ray source 2 and x-ray detector 8 assemblies typically by 360° around an imaging area between opposite ends of the C-arm 16. The region of the patient to be imaged is supported on a patient table or support in the imaging area. The C-arm 16 is mounted in a slide 25 in the horizontal arm 20 which carries a drive (not visible) for moving the C-arm 16 along a trajectory 26 to selectively image the subject over about 180° of projection directions. Calibration information is obtained by processing one or more calibration acquisitions. The calibration information is used to correct images acquired during the scan of a subject such as in the creation of tomographic cross-sectional images. The system elements 18 generate and detect x-rays which pass through the imaging area. The x-rays detected by the x-ray detector 8 are communicated to a decomposition unit 28 connected via circuitry in the gantry 15. The decomposition unit 28 can be embodied by one or more processors. During the air calibration scan, x-ray data is received by the system elements 18 and transmitted to the decomposition unit 28. The decomposition unit 28 uses air scan acquisitions and optionally processing results from geometric phantom acquisitions to decompose a selection of projection images into relative positions of each of the elements 18 for the different gantry orientations. The relative positions are based on ideal positions in a system 14 with no deformation or misalignment compared to their target position obtained from design information. A reference unit 29 stores and maintains the reference images and other data such as design information, system maintenance information, and the like. The system 14 includes a display device 30 and at least one input device 32. A healthcare practitioner can control the operation of the system 14 through the input device 32. The display device 30 displays the images, menus, panels, and user controls and includes one or more of a LCD display, an LED display, a plasma display, a projection display, a touch screen display, and the like. The display device 30 and the input device 32 can operate as part of a computer such as a desktop computer, a laptop, a tablet, a mobile computing device, a smartphone, and the like. The input device 32 can be a keyboard, a mouse, a microphone, and the like. The system 14 can include a storage device 34 such as memory, disk, network attached storage and the like. Reference scans including projection images are stored and maintained by the reference unit or memory 29. During a scan of a subject, x-ray projection data is received by the x-ray detector 8 and transmitted to the correction unit 38. One or more sensors 36 provide data with the x-ray projection data such as operating temperatures, strain measurements, gantry positional measurements, wear measurements and the like. Optionally, such a sensor can be implemented by an analysis of the x-ray projection data to determine and update positional measurements. The measurement unit 37 receives the measurement data and determines the relative positions of each of the elements 18 based on the measured data. Other data can be included with the x-ray projection data from the reference unit 29 such as expected mechanical drift based on history of the system 14 or system type, engineering specifications, manufacturer based reference scans, and the like. A correction unit 38 generates a correction for each position for each of the system elements 18. The correction unit 38 can store the generated corrections in the storage device 34 or calculate them in real time or retroactively. The corrections for each gantry orientation and each element 18 can include overlays or vector translations for each element 18 or portion of an element 18 typically expressed in detector pixels, intensity adjustments, and the like. The combined corrections form a uniformity correction. A reconstruction unit 39 reconstructs images using the received x-ray data which includes projection data, sensor data, and the like from the decomposition unit 28 and corrected by the correction unit 38. The uniformity correction can be generated as an entire or relative adjustment by the correction unit 38. During image reconstruction, the reconstruction unit 39 uses a uniformity correction based on the combined corrections of each element 18 for the different gantry orientations from the correction unit 38. The various units are suitably embodied by an electronic data processing device, such as the electronic processor or electronic processing device of the decomposition unit 28, or by a network-based server computer operatively connected with the system 14 by a network, or so forth. Moreover, the disclosed calibration techniques are suitably implemented as a non-transitory storage medium storing instructions (e.g., software) readable by an electronic data processing device and executable by the electronic data processing device to perform the disclosed calibration techniques. FIG. 4 diagrammatically illustrates one embodiment of image projection data used to decompose system element motion. Acquired air scan calibration projection images such as shown in FIG. 2 are decomposed into calibration projection images for each element such as a filter air scan projection image 40, a shutter air scan projection image 42, and an anti-scatter grid air scan projection image 44. The decomposition makes use of known image processing techniques, taking known properties of the system elements 18 into account, such as their spatial scale or repeated spatial patterns. Alternatively, separate air scan projection images of the system elements 18 can be created as part of the initial manufacturing process. The separate calibration images can then be updated from the decomposition of air scan calibration projection images, e.g. daily, before each patient, etc. In an example image, the air filter scan projection image 40 shows the non-uniform nature of the x-ray filter 4. The illustrated x-ray filter 4 is asymmetric and shows a greater intensity on the right which tapers to the left and tapers most strongly to the lower left. An example shutter air scan projection image 42 shows the intensity greatest in the center. Although not readily visible to the normal eye, inconsistencies in the edges of the shutter/collimator 6, if present, are revealed in the image. The anti-scatter grid air scan projection image 44 shows uniformly spaced lines where the lamellae or grid cast shadows on the x-ray detector 8. Shifting of the lamellae relative to the x-ray detector 8 or shifting of the x-ray source 2 relative to the lamellae the lines. Moreover, if the shift causes misalignment of the lamellae with the x-ray source 2, the lines get wider and the overall throughput of radiation through the scatter grid 10 is reduced non-uniformly over the x-ray detector 8. In the decomposition, each element 18 is separated or decomposed using sensor information for its current position. For example, when this sensor is implemented using image analysis, then a least squares error minimization can be used to determine the relative placement and orientation of the lamellae based on the individual pixels values of an initial air scan calibration projection images and/or the known geometry of the scatter grid 10 and a subsequent air scan at different gantry positions. The multi-layer decomposition measures the detected position of the anti-scatter grid based, for example, on the lines and shadows in the air scan calibration projection image. Similar decomposition is performed for the x-ray filter 4 and shutter/collimator 6. FIG. 5 diagrammatically illustrates an embodiment of the system 14 and decomposed system element motion. A geometric phantom 45 is used to calibrate the positions of some elements 18 such as the x-ray source 2 and x-ray detector 8 elements, e.g., relative to an isocenter of the scanner. The information from a geometric phantom 45 can be used to further refine or correct the relative positions of each element 18. For example, an anti-scatter grid 10 is firmly attached to the x-ray detector 8, but the anti-scatter grid 10 can change its location relative to the focal spot of the x-ray source 2. In another example, the x-ray filter 4 or beam shaper is attached to the tube or the x-ray source housing with mechanics of known degrees of freedom and positioning accuracy. The movement of each element 18 relative to a center can be determined from the data such as the air scan calibration projection images, geometric calibration projection images, imaging of a subject, sensors, etc. For example, image features from imaging of a subject can be measured that capture the position of the individual elements 18. Artifact effects such as due to the anti-scatter grid lamellae, the beam shaper profile, or the collimator edges can be removed from the image. The decomposition can be shown graphically for each element 18 with the y-axis as the deviation from the ideal center or offset, and the x-axis representing the gantry rotation angle determined from a variety of sources such as positional sensors, geometric phantoms, and the like. Note: source graph 46, filter graph 48, collimator/shutter graph 50, anti-scatter grid graph 52, and detector graph 54 of FIG. 5. With reference again the FIG. 4, before imaging a patient or in regular service intervals, the air calibration scan and the geometric calibration scan are conducted. Air and geometric projection images are generated at each angular step, performing the geometric calibration scan first and the air calibration scan after removing the geometric calibration phantom from the x-ray system. The air projection images at each angular step or a subset of angular steps are stored in a memory 34. Alternatively, ideal air projection images of each system element are generated using system knowledge and image simulation methods. Geometric calibration information, i.e., the positions of the x-ray source and detector relative to an isocenter for each requested gantry position, are derived from the geometric calibration scan and stored as well. For each angular step, one or more reference images representing one or more system elements are selected from the ideal air projection images, the subset of air projection images stored in memory 34, or the air projection image acquired at this angular step. The decomposition unit 28 generates calibration data. The decomposition unit 28 uses image processing analysis methods to determine the combination of air projection images for each system element and geometric transformations of these air projection images that best represent the air projection image acquired at each angular step. The determined combination of air projection images and geometric transformations are used to generate calibration data. The calibration data can be stored as a set of one or more air calibration projection images for each system element as shown in FIG. 4 together with the geometric transformations for each angular step. The determined parameters of the geometric transformation and their change over the course of a gantry rotation are represented by the graphs in FIG. 5. To correct projection data from the acquisition of a subject, the correction unit 38 selects the air projection images for each system element and gantry position, executes the geometric transformations determined by the decomposition unit 28 during calibration, and performs the uniformity correction of the subject projection images with these correction images. The geometric transformation parameters can be obtained during calibration and updated during subject imaging using data obtained from external sensors or the subject imaging itself. The differences between geometric transformation parameters during calibration and during the acquisition of a subject can be determined using other measurements. The element positions are a function of the reference measurement and the difference between the reference image and the images at each gantry orientation. For example, an increase in operating temperature of the arm may cause expansion which causes a difference in relative movement of an element. System wear on the lateral track may change the relative position of elements depending on the weights of elements on each arm. System wear can be considered in the differences between the reference image and the images at each orientation or can include measured system wear from sensors, operational time tables and the like. The positional adjustments to the reference image at each orientation can be stored in memory. The correction unit 38 or processor receives the shift or positional change information from the decomposition unit 28, combines the attenuation corrections attributable to each element, e.g. the inverse of images with adjustments based on current measurements. The correction unit adjusts the reference uniformity correction accordingly and stores it in a memory. When the patient is scanned, projection images are generated at one or more gantry orientations. The air scan correction corresponding to the gantry orientation is retrieved from the memory by the reference unit or recalculated by the correction unit using the current information from the measurement unit. The correction can be improved by determining relative shifts of system components from the projection images of the patient scan and using those shifts to generate more accurate correction images. The correction improvement uses known image processing methods which take the known geometrical characteristics of the system components into account. The uniformity correction projection images can be displayed on the display unit 30 or stored in a memory or data storage such as a Picture Archiving and Communication System (PACS), Radiology Information System, and the like. The reconstruction unit or processor 39 reconstructs the projection images into one or more images such as slice images, 2D images, 3D images, digital reconstructed radiographs, and the like. In FIG. 6, an embodiment of the system is flowcharted. In a step 60, x-ray calibration data is received. The x-ray calibration data can include an air scan calibration projection images, and geometric scan calibration projection images received by the decomposition unit. The x-ray calibration data can include data from one or more sensors such as strain gauges, temperature sensors, positional sensors, and the like received by the measurement unit. The x-ray calibration data can include system wear effects, temperature effects, gantry orientations, expected mechanical drift, system specifications, and manufacturing scans stored and maintained by the reference unit. The x-ray calibration data is decomposed in a step 62 using a multi-layer decomposition, which generates projection images specific to each element. For example, the air scan projection image of FIG. 2 can be decomposed to generate the projection images of FIG. 4 specific to each element. The projection images of each element can be stored as reference projection images or as updates to existing reference projection images. In a step 64, the projection images for each element from the decomposition are combined with other x-ray calibration data to determine the reference positions of each element. The reference positions for each element can be represented as graphs such as FIG. 5. The determined reference positions can include translation and rotation for each position in the range of motion of the gantry. The changes in the positions of each element with gantry angle can be stored in the storage device. In a step 66, subject x-ray data is received. The subject x-ray data includes image projection data with a subject received by the decomposition unit. The subject x-ray data can include data from one or more sensors such as strain gauges, temperature sensors, positional sensors, and the like received by the measurement unit. The subject x-ray image projection data is decomposed using a multi-layer decomposition in a step 68 similar to the decomposition of the calibration x-ray data. The decomposition generates projection images specific to each element. The system 14 uses the decomposition projection images, subject x-ray data, and reference information from the reference unit 29 such as the reference position of each element to determine the actual position of each element in a step 70. For example, x-ray data of a subject or image regions adjacent to the subject can be decomposed into effects by individual elements based on prior reference scans from the reference unit 29 and current measurements from the measurement unit 37 or based on image-processing methods. Using the lamellae shadows as an example, the lines can be used to compute a relative difference between the estimated lamellae shadow pattern and the actual pattern. The same comparison can be performed for each element to yield a set of variances for the plurality of elements 18. The actual position of each element can be used to update the reference position in a step 72 or recorded to further analysis on the performance of the system. The comparison of the reference scan adjusted by the measurements can be further adjusted based on decomposition of a projection of a current projection image which includes the subject. A uniformity correction is generated in a step 74 from the actual position or the reference position of each element. The correction can be dynamically updated during the imaging process such as using the set of variances dynamically with x-ray data of a subject or with retrieval of reference relative positions from the reference unit adjusted with measurements from the measurement unit. The correction can include an intensity adjustment or uniformity correction value. In one embodiment, the corrections are constructed using an overlay for each element which includes a relative adjustment in intensity by each element for a volume location. In a step 76, reconstruction of a projection image or images with the subject is performed using the subject x-ray data modified with the generated correction. The correction corrects the attenuation used to reconstruct a projection image or images of a subject by correcting for non-uniformity effects in the measurement. The reconstruction reconstructs images such as the 2D projection images into a 3D volume image. Slice images, surface rendering images and the like derived from the 3D volume image can be displayed on a display device and/or stored in a data storage or memory. A decision step reflects the operation of steps such that calibration scans are periodically performed such as before each patient, daily, weekly, monthly, etc. Even with performing a calibration scan before scanning each subject, one or more scans with the subject or subjects can occur between calibration scans. It is to be appreciated that in connection with the particular illustrative embodiments presented herein certain structural and/or function features are described as being incorporated in defined elements and/or components. However, it is contemplated that these features may, to the same or similar benefit, also likewise be incorporated in other elements and/or components where appropriate. It is also to be appreciated that different aspects of the exemplary embodiments may be selectively employed as appropriate to achieve other alternate embodiments suited for desired applications, the other alternate embodiments thereby realizing the respective advantages of the aspects incorporated therein. It is also to be appreciated that particular elements or components described herein may have their functionality suitably implemented via hardware, software, firmware or a combination thereof. Additionally, it is to be appreciated that certain elements described herein as incorporated together may under suitable circumstances be stand-alone elements or otherwise divided. Similarly, a plurality of particular functions described as being carried out by one particular element may be carried out by a plurality of distinct elements acting independently to carry out individual functions, or certain individual functions may be split-up and carried out by a plurality of distinct elements acting in concert. Alternately, some elements or components otherwise described and/or shown herein as distinct from one another may be physically or functionally combined where appropriate. In short, the present specification has been set forth with reference to preferred embodiments. Obviously, modifications and alterations will occur to others upon reading and understanding the present specification. It is intended that the invention be construed as including all such modifications and alterations insofar as they come within the scope of the appended claims or the equivalents thereof. That is to say, it will be appreciated that various of the above-disclosed and other features and functions, or alternatives thereof, may be desirably combined into many other different systems or applications, and also that various presently unforeseen or unanticipated alternatives, modifications, variations or improvements therein may be subsequently made by those skilled in the art which are similarly intended to be encompassed by the following claims. |
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046541853 | summary | BACKGROUND OF THE INVENTION The present invention relates to an improved nuclear reactor having a calandria in the upper portion thereof, and to a deep beam head adapted for use on a nuclear reactor to seal the top of the reactor pressure vessel. In nuclear reactors, a core is supported in a pressure vessel and a coolant, typically water at critical temperature and pressure is circulated upwardly through the core and outwardly from the pressure vessel to provide power. The coolant flows upwardly through the core and then transversely to the outlets. The reactor also includes control rod assemblies and water displacement assemblies, which are movable into and out of the core to control the same. The use of a calandria structure in the upper region of the pressure vessel has been proposed, the calandria having an upper support plate, lower support plate and a series of hollow members therebetween, with the rod drive mechanisms operating through the hollow members. The coolant passes upwardly into the calandria through apertures in the lower support plate and is directed transversely about the hollow members to the outlet nozzles, while the control rods are protected by their shielding with the walls of the hollow members. Such a system is described in copending application Ser. No. 490,099 filed Apr. 29, 1983 in the names of Luciano Veronesi, et al., entitled "Nuclear Reactor", which application is assigned to the assignee of the present invention, and which application is incorporated by reference herein. In the improved reactor described in said pending application, a yoke member carrying a plurality of control rods is movable in the space between the calandria and the core. In such systems, however, a dome-shaped head or cover for the pressure vessel is used with an open area present between the calandria and the dome-shaped head. This large area of unused space between the calandria and the dome-shaped head causes certain technical difficulties that are overcome by use of the present invention. SUMMARY OF THE INVENTION A deep beam head for use in sealing the top opening of a nuclear reactor pressure vessel and an improved nuclear reactor have a calandria disposed in the upper portion of the reactor vessel, with the upper horizontal support plate of the calandria forming a sealing plate resting on the outer pressure resistant wall of the pressure vessel. A ring member is provided about the upper periphery of the sealing plate and is secured to the top of the pressure resistant wall. A plurality of reinforcing plates extend across the top of the sealing plate and are spaced from each other and the ends thereof secured to the ring, while a plurality of spaced transverse cross-members extend between and are secured to the reinforcing members. The ring member is preferably also secured to the sealing plate such that the ring member, sealing plate and calandria may be removed as a single unit from the pressure vessel to provide access to the upper internals of the reactor. |
048851248 | summary | BACKGROUND OF THE INVENTION The present invention relates to the removal of obstructions which develop in thimbles installed in nuclear power reactors, and relates more particularly to the safe removal of such obstructions while the reactor continues in normal operation. During the operation of a nuclear power reactor, reactions within the core generate a neutron flux the spatial distribution of which must be mapped in order to insure safe and efficient operation of the reactor. For this purpose, reactors are provided with an array of tubes, termed thimbles, which extend into the reactor core and which are closed at their lower ends so that their interiors are isolated from the pressure existing within the reactor and can thus be maintained at normal atmospheric pressure. These thimbles define paths for flux detectors which can be advanced along the length of each thimble in order to produce neutron flux measurements at various points along the length of each thimble. During reactor operation, it frequently occurs that obstructions develop in these thimbles due, for example, to the accumulation of lubricating material at points where the thimbles are provided with bends. While complete removal of the material forming such obstructions requires reactor shutdown, there is a demand in the industry for a procedure which can at least temporarily clear an obstruction, without requiring reactor shutdown, so that the requisite number of thimbles can continue to be used for the flux mapping operation. A known procedure for achieving this result, which will be described below, has been found to impose stresses on the fitting which secures the thimble to the reactor and has on occasion weakened the fitting to such an extent that the high pressure at the interior of the reactor has caused the thimble to be ejected from the pressure vessel, with an accompanying escape of radioactive coolant in the form of steam and/or water. SUMMARY OF THE INVENTION It is a primary object of the present invention to remove obstructions which develop in thimbles installed in nuclear power reactors without requiring reactor shutdown and without creating the danger of thimble ejection. A more specific object of the invention is to introduce an obstruction removal device into a thimble while reliably preventing ejection of the thimble from the reactor pressure vessel. The above and other objects are achieved, according to the present invention, in the operation of a power reactor having a pressure vessel containing a core and a plurality of thimbles extending into the core to provide passages for a flux detector, a movable support member located outside of the pressure vessel and carrying a plurality of tubular members each detachably connected to a respective thimble, and transfer means normally coupled to the tubular members for introducing a flux detector into a selected thimble via a respective tubular member, by a method for removing obstructions from a selected thimble while the reactor remains in operation, which method comprises: disconnecting at least the respective tubular member from the transfer means; PA1 introducing an obstruction removal device into the respective tubular member; and PA1 advancing the obstruction removal device through the respective tubular member and into the selected thimble to the location of the obstruction. |
description | This application claims the benefit of Chinese Application No. 200510009074.2 filed Feb. 17, 2005. The present invention relates to a filter and an X-ray imaging apparatus, and more particularly to a filter that adjusts energy spectra of X-ray, and an X-ray imaging apparatus provided with the filter. An X-ray imaging apparatus irradiates X-ray to a subject by adjusting energy spectra of the X-ray with a filter. The filter is provided in a collimator box attached to an X-ray tube. In order to obtain desired spectra, the filter can be used by switching plural filter plates attached to a rotating disc (e.g., see patent document 1). [Patent Document 1] Japanese Published Unexamined Patent Application No. HEI11-76219 (p. 2, FIG. 1) Spectra can desirably be adjusted in a wide range with close attention, but in the construction in which the filter plates are switched by the rotating disc, a four-step adjustment is about all this construction can provide. Therefore, widening the adjustment range causes a rough step, while providing more steps narrows the adjustment range. When a multi-step adjustment is made possible in the rotating disc system anyway, the rotating disc to which a great number of filters are attached is increased in size, thus unrealistic. Therefore, an object of the present invention is to realize a filter that can make a fine spectrum adjustment in a wide rage and can be miniaturized, and to realize an X-ray imaging apparatus provided with the filter. The invention in one aspect for solving the above-mentioned problem is a filter comprising plural filter plates that can form a layer crossing X-ray and adjusting means that adjusts a combination of filter plates forming the layer by individually moving the plural filter plates so as to come in and out the X-ray passing space. The invention in another aspect for solving the above-mentioned problem is an X-ray imaging apparatus for imaging a subject by X-ray via a filter, wherein the filter comprises plural filter plates that can form a layer crossing X-ray and adjusting means that adjusts a combination of filter plates forming the layer by individually moving the plural filter plates so as to come in and out the X-ray passing space. It is preferable that the plural filter plates are formed such that the thickness of each filter plate is successively doubled with the thinnest filter plate defined as a reference, from the viewpoint of performing a thickness adjustment in which the thickness of the thinnest filter plate is rendered to be the minimum step. From the viewpoint of interleaving the filter plates from both sides, it is preferable that the adjusting means has a pair of moving in/out mechanisms for moving the plural filter plates in and out from both sides of the X-ray passing space. One of the mechanisms moves a first set made up of every other filter plate from the plural filter plates. The other of the pair of mechanisms moves a second set of filter plates made of the alternate group of every other filter plate of the remaining of the plural filter plates. From the viewpoint of simplifying the construction, it is preferable that the moving in/out mechanism moves the filter plates in and out by a reciprocating movement of a link based upon the rotation of a plate cam. From the viewpoint of reducing components, it is preferable that plural plate cams present at the same side with respect to the X-ray passing space have a common rotation axis. It is preferable that the moving in/out mechanism moves the filter plates in and out by a swing movement of an arm driven by a motor for facilitating the miniaturization. It is preferable that the moving in/out mechanism moves the filter plates in and out by using a reciprocating movement of a movable section of an electromagnetic solenoid for facilitating linear movement. It is preferable that the moving in/out mechanism moves the filter plates in and out by using a reciprocating movement of a movable section of an air cylinder for facilitating linear movement. It is preferable that the moving in/out mechanism moves the filter plates in and out by using a reciprocating movement of a movable section of a hydraulic cylinder for facilitating linear movement. According to the present invention, the filter has plural filter plates that can form a layer crossing X-ray and adjusting means that adjusts a combination of filter plates forming the layer by individually moving the plural filter plates so as to come in and out the X-ray passing space, thereby being capable of realizing a filter that can provide a fine spectrum adjustment and can be miniaturized, and an X-ray imaging apparatus provided with the filter. Further objects and advantages of the present invention will be apparent from the following description of the preferred embodiments of the invention as illustrated in the accompanying drawings. A best mode for carrying out the invention will be explained hereinbelow in detail with reference to the drawings. It should be noted that the invention is not limited to the best mode for carrying out the invention. FIG. 1 shows a schematic construction of an X-ray imaging apparatus. This device is one example of the embodiment of the invention. The construction of this device represents one example of the X-ray imaging apparatus according to the embodiment of the present invention. As shown in FIG. 1, this device has an X-ray irradiating device 10, X-ray detecting device 20 and an operator console 30. The X-ray irradiating device 10 and the X-ray detecting device oppose to each other via a subject 40. The X-ray irradiating device 10 has an X-ray tube 12 and a collimator box 14. A filter 16 and a collimator 18 are accommodated in the collimator box 14. The filter 16 is one example of the embodiment of the present invention. The construction of this filter represents one example of the embodiment relating to the filter according to the present invention. X-ray emitted from the X-ray tube 12 whose energy spectra are adjusted by the filter 16 is irradiated to the subject 40 through an opening of the collimator 18. The filter 16 can make the energy spectra variable. The collimator 18 has the opening that is variable. The X-ray passing through the subject 40 is detected by the X-ray detecting device 20 to be inputted to the operator console 30. The operator console 30 reconstructs the radioscopic image of the subject based upon an inputted signal. The reconstructed radioscopic image is displayed on a display 32 of the operator console 30. The operator console 30 further controls the X-ray irradiating device 10. The control of the X-ray irradiating device 10 by the operator console 30 includes the control of the filter 16 and the control of the collimator 18. It should be noted that the filter 16 and the collimator 18 can manually be adjusted according to need. The filter 16 will be explained. FIG. 2 shows a principle view of the filter 16. As shown in this figure, the filter 16 is composed of plural filter plates 161, 162 . . . 16n. The filter plates 161, 162 . . . 16n are one example of the filter plates in the present invention. These filter plates 161, 162 . . . 16n form a layer crossing the X-ray. Each filter plate 16i (i: 1, 2 . . . n) is a formation component of the layer, and is a plate such as a metal or plastics. Each filter plate 16i can individually advance into or retreat from the space where the X-ray passes. In this figure, the advancing state into the X-ray passing space is shown by black-painted section while the retreating state from the X-ray passing space is shown by white sections. The filter plates 161, 162 . . . 16n are formed such that the thickness of each filter plate is successively doubled with the thinnest filter plate 161 defined as a reference. Specifically, supposing that the thickness of the filter plate 161 is defined as AT, each of the filter plates 161, 162 . . . 16n has a thickness of AT, AT*2 . . . AT*2n−1. Since the filter plates 161, 162 . . . 16n has such thickness, selecting the combination of the filter plates relating to the formation of the layer can increase or decrease the sum of the thickness of the filter plates in the layer from 0 to AT*2n−1 at intervals of AT with 2n step. FIG. 3 shows the maximum plate thickness, step number of increasing or decreasing thickness and total sum of the plate thickness corresponding to the number n of the filter plate. From this figure, it is understood that, when the filter plate number is 4, for example, the maximum plate thickness is AT*8, step number of increasing or decreasing thickness is 16 and the total sum of the plate thickness is AT*15. FIG. 4 is a view showing this state in detail. In this figure, AT is 1 mm, for example. In the same figure, “1” represents the advancing state of the filter plate into the X-ray passing space, while “0” represents the retreating state. As shown in this figure, the sum of the thickness of the filter plate can be increased or decreased by 1 mm over 16 steps from 0 mm to 15 mm. This can be said to be substantially continuous thickness adjustment. Specifically, a filter can be obtained that can provide a fine spectrum adjustment over a wide range. Further, the filter layer is formed by the combination of plural filter plates, thereby facilitating miniaturization. In the actual filter, four filter plates advance or retreat alternately from both sides of the X-ray passing space as shown in FIG. 5. Specifically, the filter plates 161 and 162 advance or retreat from the left side of the X-ray passing space, while the filter plates 163 and 164 advance or retreat from the right side of the X-ray passing space, for example. With the state where the layer is formed, the filter plates 161 and 162 at the left side and the filter plates 163 and 164 at the right side are interleaved. It should be noted that the interleave is not essential. The filter plates 163 and 164 at the right side may be on the filter plates 161 and 162 at the left side, or vice versa. FIGS. 6 to 8 show one example of the construction of the filter 16. FIG. 6 is a whole constructional view, and FIGS. 7 and 8 are partial constructional views. As shown in these figures, four filter plates 161, 162, 163 and 164 are supported by a pair of rails 172 and 174 in the filter 16. The rails 172 and 174 are parallel to each other. The rails 172 and 174 each has four parallel grooves corresponding to four filter plates 161, 162, 163 and 164. The filter plates 161, 162, 163 and 164 are supported in such a manner that both end sections of each filter plate are inserted into each of four grooves. The filter plates 161, 162, 163 and 164 are respectively connected to one end of links 261, 262, 263 and 264. The other end of each of the links 261 and 262 is mounted to an axis 272 so as to be rotatable about the axis 272. The other end of the links 263 and 264 is mounted to an axis 274 so as to be rotatable about the axis 274. Plate cams 461, 462, 463 and 464 for driving the links 261, 262, 263 and 264 are respectively mounted corresponding to these links. The plate cams 461, 462, 463 and 464 are cams utilizing the outer peripheral shape. The cam of this type is referred to as a peripheral cam hereinbelow. Each of the links 261, 262, 263 and 264 has each of pins 361, 362, 363 and 364 that come in contact with the outer periphery of each of the plate cams 461, 462, 463 and 464. These pins are always pressed toward the outer periphery of each of the plate cams 461, 462, 463 and 464 by springs 561 and 562 that pull the links 261 and 262 in the leftward direction and springs 563 and 564 that pull the links 263 and 264 in the rightward direction. The plate cams 461 and 462 are mounted to the same rotating axis 602. This rotating axis 602 is driven by a motor 802 via a decelerator 702. The plate cams 463 and 464 are similarly mounted to the same rotating axis. This rotating axis is driven by a motor 804 via a decelerator 704. As described above, the plate cams present at the same side with respect to the X-ray passing space have a common rotation axis, thereby being capable of reducing the number of components. The links 261 and 262 driven by the plate cams 461 and 462 displace reciprocatingly the filter plates 161 and 162 respectively along the rails 172 and 174. The filter plates 161 and 162 are in the advancing state into the X-ray passing space when they are present at the center of the rails 172 and 174, while they are in the retreating state when they are present at the left end section. The links 263 and 264 driven by the plate cams 463 and 464 displace reciprocatingly the filter plates 163 and 164 respectively along the rails 172 and 174. The filter plates 163 and 164 are in the advancing state into the X-ray passing space when they are present at the center of the rails 172 and 174, while they are in the retreating state when they are present at the right end section. The section composed of the links 261 to 264, plate cams 461 to 464, decelerators 702 and 704 and motors 802 and 804 is one example of the adjusting means in the present invention. The section composed of the links 261 and 262, plate cams 461 and 462, decelerator 702 and motor 802 and the section composed of the links 263 and 264, plate cams 463 and 464, decelerator 704 and motor 804 are one example of a pair of moving in/out mechanism in the present invention. The moving in/out mechanism has a construction for moving the filter plates in and out by the reciprocating movement of the links based upon the rotation of the plate cams, thereby being capable of simplifying the mechanism. FIGS. 10 to 12 show one example of another construction of the filter 16. FIG. 10 is a whole constructional view, and FIGS. 11 and 12 are partial constructional views. As shown in these figures, four filter plates 161, 162, 163 and 164 are supported by a pair of rails 172 and 174 in the filter 16. The rails 172 and 174 are parallel to each other. The rails 172 and 174 each has four parallel grooves corresponding to four filter plates 161, 162, 163 and 164. The filter plates 161, 162, 163 and 164 are supported in such a manner that both end sections of each filter plate are inserted into each of four grooves. The filter plates 161, 162, 163 and 164 are respectively connected to one end of links 261, 262, 263 and 264. The other end of each of the links 261 and 262 is mounted to an axis 272 so as to be rotatable about the axis 272. The other end of the links 263 and 264 is mounted to an axis 274 so as to be rotatable about the axis 274. Plate cams 471, 472, 473 and 474 for driving the links 261, 262, 263 and 264 are respectively mounted corresponding to these links. Each of the links 261, 262, 263 and 264 has each of pins 361, 362, 363 and 364 that is engaged with each of the plate cams 471, 472, 473 and 474. The plate cams 471, 472, 473 and 474 are cams utilizing a loop shape drawn by a groove. The cam of this type is referred to as a grooved cam hereinbelow. The plate cams 471 and 472 are grooves formed on both faces of a gear as shown in FIG. 13. The plate cams 471 and 472 are driven by a motor 802 via a decelerator 702. The plate cams 473 and 474 are similarly driven by a motor 804 via a decelerator 704. As described above, the plate cams present at the same side with respect to the X-ray passing space have a common gear, thereby being capable of reducing the number of components. The links 261 and 262 driven by the plate cams 471 and 472 displace reciprocatingly the filter plates 161 and 162 respectively along the rails 172 and 174. The filter plates 161 and 162 are in the advancing state into the X-ray passing space when they are present at the center of the rails 172 and 174, while the filter plates 161 and 162 are in the retreating state when they are present at the left end section. The links 263 and 264 driven by the plate cams 473 and 474 displace reciprocatingly the filter plates 163 and 164 respectively along the rails 172 and 174. The filter plates 163 and 164 are in the advancing state into the X-ray passing space when they are present at the center of the rails 172 and 174, while the filter plates 163 and 164 are in the retreating state when they are present at the right end section. The section composed of the links 261 to 264, plate cams 471 to 474, decelerators 702 and 704 and motors 802 and 804 is one example of the adjusting means in the present invention. The section composed of the links 261 and 262, plate cams 471 and 472, decelerator 702 and motor 802 and the section composed of the links 263 and 264, plate cams 473 and 474, decelerator 704 and motor 804 are one example of a pair of moving in/out mechanism in the present invention. The moving in/out mechanism has a construction for moving the filter plates in and out by the reciprocating movement of the links based upon the rotation of the plate cams, thereby being capable of simplifying the mechanism. The cam will be explained. As described above, the cam has a function for binarily switching the position of the filter plate by its rotation, between the advancing position into the X-ray passing space and the retreating position therefrom. FIG. 14 shows the number of state that can be switched for one rotation and a rotational angle step per one switch corresponding to the number of cam per one axis. As shown in this figure, when the number of cam per one axis is n, the number of state that can be switched for one rotation is 2n and the rotational angle step per one switch is 360°/2n. Accordingly, when the number of cam per one axis is 1, the number of state that can be switched is 2 and the rotational angle step per one switch is 180°. When the number of cam per one axis is 2, the number of state that can be switched is 4 and the rotational angle step per one switch is 90°. When the number of cam per one axis is 3, the number of state that can be switched is 8 and the rotational angle step per one switch is 45°. FIG. 15 shows a shape of the cam when the number of cam per one axis is 1. FIG. 15(a) shows the shape of the peripheral cam, while FIG. 15(b) shows the shape of the grooved cam. As shown in the same figure, the section at 0° becomes a short diameter “0” and the section at 180° becomes a long diameter of “1” in this cam. Accordingly, one rotation of this cam can switch the position of one filter plate between two stages, i.e., between “0” that is the retreating state and “1” that is the advancing state. FIG. 16 shows the position of the filter plate and the rotational angle of the cam corresponding to each switching step. The cam of this shape can be used for the case where two filter plates are separated into one each, each of which is arranged at both sides of the X-ray passing space from which each of filter plates advances or retreats. The shapes of the cams at both sides are the same. It should be noted that the ratio of the rotation of cams at both sides is defined as 1:0.5, wherein the cam at the other side (cam 1′) rotates at 180° every time the cam at one side (cam 1) makes one rotation. Further, the cam 1switches the position of the filter plate having the thickness of 1 and the cam 1′ switches the position of the filter plate having the thickness of 2. It should be noted that the thickness here means a thickness normalized by the minimum thickness. The same is applied to the following description. FIG. 17 shows the position of the filter plate and the rotational angle of the cam corresponding to each switching step with this state. In the same figure, “0” represents the retreating position of the filter plate, while “1” represents the advancing position of the filter plate. As shown in the same figure, a two-layer filter can be obtained in which the thickness is changed from 0 to 3 by 1 in four steps. FIG. 18 shows a shape of the cam when the number of cam per one axis is 2. FIGS. 18(a) and 18(b) respectively show the shape of the peripheral cam and the shape of the grooved cam with respect to one (cam 1) of two cams. In this cam, the sections of 0°, 90°, 180° and 270° become short diameter “0”, long diameter “1”, short diameter “0” and long diameter “1” respectively, as shown in the same figure. Accordingly, one rotation of this cam can switch the position of the filter plate in a four-step manner of 0, 1, 0 and 1. An advancing position or retreating position of a filter plate and a rotational angle of a cam for every step is shown in FIG. 19. FIGS. 18(c) and 18(d) respectively show the shape of the peripheral cam and the shape of the grooved cam with respect to the other (cam 2) of two cams. In this cam, the sections of 0°, 90°, 180° and 270° become short diameter “0”, short diameter “0”, long diameter “1” and long diameter “1” respectively. Accordingly, one rotation of this cam can switch the position of the filter plate in a two-step manner of 0, 0, 1 and 1. The cam of this shape can be used for the case where four filter plates are separated into two filter plates each, and each is respectively arranged at both sides of the X-ray passing space from which each advances or retreats. The shapes of the cams at both sides are the same. It should be noted that the ratio of the rotation of cams at both sides is defined as 1:0.25, wherein the cam at the other side (cam 1′ 2′) rotates at 90° every time the cam at one side (cam 1, 2) makes one rotation. Further, the cam 1 or 2 switches the position of the filter plate having the thickness of 1 or 2 and the cam 1′ or 2′ switches the position of the filter plate having the thickness of 4 or 8. FIG. 20 shows the position of the filter plate and the rotational angle of the cam corresponding to each switching step with this state. In the same figure, “0” represents the retreating position of the filter plate, while “1” represents the advancing position of the filter plate. As shown in the same figure, a four-layer filter can be obtained in which the thickness is changed from 0 to 15 by 1 in sixteen steps. FIG. 21 shows a shape of the cam when the number of cam per one axis is 3. FIGS. 21(a) and 21(b) respectively show the shape of the peripheral cam and the shape of the grooved cam with respect to the first cam (cam 1) of three cams. As shown in the same figure, the sections of 0°, 45°, 90°, 135°, 180°, 225°, 270° and 315° become short diameter “0”, long diameter “1”, short diameter “0”, long diameter “1”, short diameter “0”, long diameter “1”, short diameter “0” and long diameter “1” respectively in this cam. Accordingly, one rotation of this cam can switch the position of the filter plate in eight-step manner of 0, 1, 0, 1, 0, 1, 0 and 1. FIGS. 21(c) and 21(d) respectively show the shape of the peripheral cam and the shape of the grooved cam with respect to the second cam (cam 2) of three cams. As shown in the same figure, the sections of 0°, 45°, 90°, 135°, 180°, 225°, 270° and 315° become short diameter “0”, short diameter “0”, long diameter “1”, long diameter “1”, short diameter “0”, short diameter “0”, long diameter “1” and long diameter “1” respectively in this cam. Accordingly, one rotation of this cam can switch the position of the filter plate in four-step manner of 0, 0, 1, 1, 0, 0, 1 and 1. FIGS. 21(e) and 21(f) respectively show the shape of the peripheral cam and the shape of the grooved cam with respect to the third cam (cam 3) of three cams. As shown in the same figure, the sections of 0°, 45°, 90°, 135°, 180°, 225°, 270° and 315° become short diameter “0”, short diameter “0”, short diameter “0”, short diameter “0”, long diameter “1”, long diameter “1”, long diameter “1” and long diameter “1” respectively in this cam. Accordingly, one rotation of this cam can switch the position of the filter plate in two-step manner of 0, 0, 0, 0, 1, 1, 1 and 1. FIG. 22 shows the position of the filter plate and the rotational angle of the cam corresponding to each switching step. The cam of this shape can be used for the case where six filter plates are separated into three filter plates each, and each is respectively arranged at both sides of the X-ray passing space from which each advances or retreats. The shapes of the cams at both sides are the same. It should be noted that the ratio of the rotation of cams at both sides is defined as 1:0.125, wherein the cam at the other side (cam 1′ 2′, 3′) rotates at 45°every time the cam at one side (cam 1, 2, 3) makes one rotation. Further, the cam 1, 2 or 3 switches the position of three filter plates each having the thickness of 1, 2 or 4 and the cam 1′, 2′ or 3′ switches the position of three filter plates each having the thickness of 8, 16 or 32. This makes it possible to obtain a six-layer filter wherein the thickness is changed from 0 to 63 by 1 in sixty-four steps. An example of main parts of a six layers filter is shown in FIG. 23 of a perspective. As shown in FIG. 23, the six layers filter has six filter plates 161, 162, 163, 164, 165, 166. These filter plates are supported by a pair of rails 172, 174 and can move along the rails without mutual interference. In FIG. 23, the four filter plates 161, 162, 164, 165 out of the six plates advance in X-ray passing space and the other two filter plates 163, 166 retreat from the X-ray passing space. The filter plates 161, 162, 163 come in and out from the left side in the FIG. 23 and the filter plates 164, 165, 166 come in and out from the right side in the FIG. 23. The filter plates 161, 162, 163 come in and out by means of links 261, 262, 263 respectively. The links 261, 262, 263 are driven by means of plate cams 461, 462, 463 respectively and turn round on a common axis 272. The plate cams 461, 462, 463 are fixed on a common rotation axis 602. The plate cams 461, 462, 463 are connected with the links 261, 262, 263 by means of pins 361, 362, 363 respectively. The plate cams 461, 462, 463 are cams shaped like a disk, and the pins 361, 362, 363 are forced into edges of the cams by springs 561, 562, 563. Incidentally, if the plate cams 461, 462, 463 are cams having a groove instead of the cams shaped like a disk, the springs 561, 562, 563 are needless. The filter plates 164, 165, 166 come in and out by means of links 264, 265, 266 respectively. The links 264, 265, 266 are driven by means of plate cams 464, 465, 466 respectively and turn round on a common axis 274. The plate cams 464, 465, 466 are fixed on a common rotation axis 604. The plate cams 464, 465, 466 are connected with the links 264, 265, 266 by means of pins 364, 365, 366 respectively. The plate cams 464, 465, 466 are cams shaped like a disk, and the pins 364, 365, 366 are forced into edges of the cams by springs 564, 565, 566. Incidentally, if the plate cams 464, 465, 466 are cams having a groove instead of the cams shaped like a disk, the springs 564, 565, 566 are needless. FIG. 24 shows another example of the construction of the filter 16. In this example, filter plates 161, 162, 163 and 164 are driven by motors 811, 812, 813 and 814 via decelerators 711, 712, 713 and 714. The filter plates 161, 162, 163 and 164 are mounted to the output section of each decelerator 711, 712, 713 and 714 via arms 911, 912, 913 and 914. Each of the filter plates 161 and 162 moves between the retreating position at the right end section and the advancing position at the left end section by the swing movement of each of the arms 911 and 912 with the rotation of the motors 811 and 812. Each of the filter plates 163 and 164 moves between the retreating position at the left end section and the advancing position at the right end section by the swing movement of each of the arms 913 and 914 with the rotation of the motors 813 and 814. The section composed of the arms 911 to 914, decelerators 711 to 714 and motors 811 to 814 is one example of the adjusting means in the present invention. The section composed of the arms 911 and 912, decelerators 711 and 712 and motors 811 and 812 and the section composed of the arms 913 and 914, decelerators 713 and 714 and motors 813 and 814 are one example of a pair of moving in/out mechanism in the present invention. The moving in/out mechanism has a construction for moving the filter plates in and out by the swing movement of the arms driven by the motors, thereby being capable of facilitating a miniaturization of the mechanism. FIG. 25 shows another example of the construction of the filter 16. In this example, filter plates 161, 162, and 163 are driven by motors 811, 812, and 813 via decelerators 711′, 712′, and 713′, respectively. Filter plate 164 is driven by motor 814 via a fourth decelerator (not shown in FIG. 25). The filter plates 161, 162, 163 and 164 are mounted to the output section of each decelerator 711′, 712′, 713′ and the fourth decelerator via arms 911, 912, 913 and 914. The decelerators 711′, 712′, 713′ and the fourth decelerator are configured to change the direction of the rotating axis by 90° on the way. Therefore, the direction of each rotating axis of each of the motors 811, 812, 813 and 814 is horizontal. Each of the filter plates 161 and 162 moves between the retreating position at the right end section and the advancing position at the left end section by the swing movement of each of the arms 911 and 912 with the rotation of the motors 811 and 812. Each of the filter plates 163 and 164 moves between the retreating position at the left end section and the advancing position at the right end section by the swing movement of each of the arms 913 and 914 with the rotation of the motors 813 and 814. FIG. 26 schematically shows the construction of another example of the filter 16. As shown in the same figure, the filter plate 160 is driven by an actuator 800 via a link 260 so as to move between the retreating position at the left side and the advancing position at the right side. Such mechanism is provided by the number of the filter plates. An electromagnetic solenoid having a movable section that makes a reciprocating movement is used as the actuator 800. It should be noted that an air cylinder or hydraulic cylinder may be used instead of the electromagnetic solenoid. The section composed of the link 260 and the actuator 800 is one example of the adjusting means or moving in/out mechanism in the present invention. The moving in/out mechanism moves the filter plate in or out by utilizing the reciprocating movement of the movable section of the electromagnetic solenoid, air cylinder or hydraulic cylinder, thereby facilitating a linear movement. Many widely different embodiments of the invention may be configured without departing from the spirit and the scope of the present invention. It should be understood that the present invention is not limited to the specific embodiments described in the specification, except as defined in the appended claims. |
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abstract | A method of determining and quantifying the presence and concentration of regulated radionuclides present in filter material used to remove radionuclide contaminants from the cooling water of a nuclear reactor. Multiple samples of the reactor cooling water are taken and the presence and concentration of directly measurable fission and activation produced radionuclides are determined through gamma spectroscopy. The release rate of radioactivity from the reactor as a function of the removal rate of the filter material is determined at equilibrium. The presence and the concentration of the indirectly measured fission regulated radionuclides are determined as a function of release rates of the directly measurable fission produced isotopes. |
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048470090 | description | FIG. 1a depicts schematically a first method for loading and sealing a double container system 2, consisting of a removable inner container 4 of steel and an outer shielding container 6 in six steps, designated A, B, C, D, E, and F. The inner container 4 has a screw-in inner cover 8 and a weld-on outer cover 10 and the shielding container 6 has a screw-on shielding cover 12. For loading and sealing, in a first step A, the empty double container system 2 is injected into a shielded chamber 14, for example, a so-called hot cell. In the second step B, in the shielded chamber, the open inner container 4 is loaded through the top opening of the shielded container 6 with radioactive material 16 which is enclosed in and is to be stored in a sheath (box, metal mould) 16'. In the third step C, the inner container while still in the hot cell is sealed with the screw-in inner cover 8, and the seal of the screw-in cover is tested. In the fourth step D, the ejection from chamber 14 of the double container system which is loaded and sealed with the inner cover 8 takes place. In step E, outside of the shielded chamber, the outer cover 10 is welded to the inner container 4, and after and welding is complete, the weld is tested. Finally, in the last step F, the shielding cover 12 is screwed onto the shielding container 6. FIG. 1b depicts schematically a second method for loading and sealing a double container system 2, consisting of a removable inner container 4 of steel and an outer shielding container 6 in six steps, A, B, C, D, E, and F. The inner container 4 has a screw-in inner cover 8 and a weld-on outer cover 10 and the shielding container 6 has a screw-on shielding cover 12. For loading and sealing in a first step A, the empty and open double container system 2 is locked or gripped from below the hot cell (shielded chamber) 14', specifically within an injection aperture 17 located in the floor of the cell. The seal of the injection aperture is not depicted, but is understood that suitable, known transport and lifting devices are used and that the docked double container system 2 is arranged absolutely sealed and shielded in the injection aperture 17, as is indicated by means of the seal/shield 19. In the second step B, the inner container 4 is loaded from the hot cell 14' with the radioactive material 16 which is to be stored and which is enclosed in a sheath 16'. In the third step C, the inner container 4 is sealed with the screw in cover 8 while the double container system 2 is still locked in the injection aperture 17, and the screw-in cover seal is tested. In the fourth step D, the sealing of the injection aperture 17 and the loosening and removal from the hot cell 14 of the loaded double container system 2, now sealed with the inner cover 8, takes place. After this, in step E, outside of the shielded region, the outer cover 10 is welded to the inner container 4, and after the welding is complete, the weld is tested. Finally, in the last step F, the shielding cover 12 is screwed to the shielding container 6. The inner container 4 which is depicted in greater detail in FIG. 2 consists of a cylindrical jacket 18, a floor 20, and a seal 22. The seal 22 consists of the inner cover 8 which is designed as a sealing plug, and can be screwed into the jacket 18 against bottom and side seals 24, and the outer cover 10, which is designed as a sealing plug with a handle 26, which outer cover is welded to the jacket 18 of the inner container 4. A welding gap 28 is left between the outer cover 10 and the jacket 18 for the application of a weld 29 between the cover and the jacket by means of narrow-gap welding. The sealed container is further provided with a welded-on plasma hot wire cladding layer 30 for corrosion protection. FIG. 3 shows the weld in more detail. The welding gap 28 between the outer cover 10 and the container jacket 18 widens slightly toward the top and is limited on the bottom by means of two surrounding welding flanges 32 and 34 which lie opposite one another, of which one is located on the jacket and the other on the outer cover 10. The welding flanges 32 and 34 are canted to attain a clean weld root 36. The cant 38 amounts to about 45.degree. and is preferably provided both on the top and on the bottom sides of the welding flanges. The canting has the advantage that heat dissipation from the weld point is improved. Also, the canting facilitates the introduction and positioning of the outer cover 10. The root welding is performed preferably with the help of an inert gasshielded arc welding device with tungsten electrodes ("WIG" welding device), with which a very precise weld can be performed. Onto the weld root 36, further weld layers 40 are welded with the inert gas-shielded arc welding device with tungsten electrodes, the purpose being to securely prevent burn-through during the following welding of the remaining weld layers 42 with the help of a submerged arc welding system ("UP" welding system), with which large quantities of welding metal can be applied. The welding of the root thereby proceeds preferably with the help of at least two "WIG" welding heads, which lie opposite one another and operate simultaneously. This prevents distortion of the cover and thus a disruption of the uniformity of the welding gap during the welding process, and finally the danger of forming an uneven welded root and possible fissures. The final welding performed as narrow-gap submerged arc welding has the advantage, since it welds thicker cross-sections than has presently been the case, that it leads to lower production of heat and leads to a more uniform build-up of the weld layers and thus of the weld itself. The weld material of the edge layers molds to the edges of the welding gap between the cover and the container jacket, whereby the coarse grain is almost completely converted to fine grain by the following layer. Thus the necessary condition is created for eliminating a following voltage warming and cooling, and it is assured that the material values of the weld lie within the framework of the material values of the base material. The production of the seal takes place with the help of a device depicted in FIGS. 4 through 6. The shielding container 6, into which the inner container 4 is inserted, is set on a horizontal rotary table 44, which is anchored to the floor 46 of a foundation pit 48. Where applicable, the inner container 4 can be set on the rotary table alone. The rotary table is equipped with a spherical turning connection to absorb horizontal and axial forces. The rotary table is driven by means of a motor within a low rotational speed range. The top of the table has a mechanical stage which is adjustable with a motor for the precision positioning of the container under the welding and testing movable bridge 50. Since the foundation pit 48 should be able to be used for containers of various sizes, spacing pieces 52 (drawn in dot-and-dash lines) are placed on the rotary table for equalization of length so that the sealing weld will always take place at the same height above the floor. A pit cover 54 (FIG. 5) is provided which is put on during loading of the rotary table with the help of a crane. It is moved over the concrete pit and can be walked on during the loading process. All welding, testing and other devices are mounted on the bridge 50. The movable bridge 50 has several interior tracks (not depicted) which serve for guiding transport carriages 56 and the welding and testing devices 58, 60, 62, 64 which are fastened on them. The bridge 50 is equipped with a traveling gear 66 (FIG. 4) with flanged wheels and a direct current drive by which the bridge can be driven at both positioning speed and rapid traverse speed along raised tracks 68 which are arranged on both sides of the foundation pit. The drives of the transport carriages are supplied with power separately, and each transport carriage has a height adjustment with which the complete welding or testing device can be moved into operating or waiting position. The welding and testing devices encompass an inert gas-shielded arc welding device with tungsten electrodes (WIG) 58, submerged arc welding device ("UP") 60, a plasma hot-wire welding device ("PH") 62 and a testing device 64. FIG. 6 shows a part of the welding and testing bridge 50 in greater detail with a testing device 64 in the operating position. The weld 29 and the application of the corrosion protection layer 30 have already taken place. The operating position of the welding devices 58 or 60 are quite analogous in appearance. It can be easily seen from FIG. 6 that the welding and testing of the outer cover 10 is to be performed with the inner container 2 inserted in the shielding container 6. Reference number 70 indicates a moderator. The complete welding and testing process preferably occurs automatically, for which purpose a control device 72 (FIG. 5) is provided. Otherwise, control panels are provided, from which the welding and testing devices can be controlled. If desired, the welding and testing devices can be mounted on bracket arms (not depicted), rather than on bridge 50. To guarantee optimum testing of the weld 29, the shielding container has a step-shaped, annular mouth 74 extending above the weld so that a sufficient free space arises between the shielding container and the inner container to insert and move the testing device 64 (FIG. 6). Furthermore, the outer cover 10 of the inner container 4 has provided on its bottom side below the welding flange 34 an annular, step-shaped recess 78, by means of which an annular chamber 80 is formed between the outer cover 10 and the inner cover 8, which continues in an annular groove 82 formed below the welding flange 32 of the container jacket in the inner wall of this jacket. |
claims | 1. A nuclear reactor modeling system comprising: a modeling interface configured to:determine standardized modeling data of an abstract nuclear reactor model representing a nuclear reactor core, the standardized nuclear reactor modeling data includes a nuclear reactor data structure including a plurality of assembly structures, each assembly structure representative of a physical component that is present in the nuclear reactor model, two or more of the assembly structures representing fuel assemblies of the nuclear reactor, each fuel assembly structure including a plurality of block structures representative of axially distributed blocks of the fuel assembly of the nuclear reactor core, at least one block structure of the plurality of block structures of the fuel assembly structure includes material data and location, the material and location data representative of fuel pin material and location within the fuel assembly of the nuclear reactor core, at least one fuel pin material of a first fuel assembly structure being different from a second fuel pin material of a second fuel assembly structure, one or more of the assembly structures representative of control rods of the nuclear reactor, and one or more of the assembly structures representative of a structure of a fuel assembly receptacle of the nuclear reactor core:converting the standardized modeling data of the abstract nuclear reactor model to nuclear reactor modeling data defining a nuclear reactor model modelling the nuclear reactor core, the nuclear reactor modeling data includinga plurality of cell data for a plurality of cells of the nuclear reactor model, each cell defined by one or both of bounding of surfaces and regions of space, each cell data including a physical location, a material identifier, and a geometry of the associated cell of the nuclear reactor model,nuclear reactor performance data including fuel cell swelling and fuel depletion, andnuclear material data including cycle load data;a simulator including a plurality of simulator modules including a neutronics simulator module, a fuel burn simulator module, a thermal hydraulics simulator module, and a material performance simulator module, the simulator coupled to the modeling interface and configured to generate simulation data for the modeling interface, the neutronics simulator module interacting with the fuel burn simulator module, the thermal hydraulics simulator module and the material performance simulator module to iteratively produce time dependent nuclear reactor simulation of at least a portion of the nuclear reactor simulation data for the modeling interface;the modeling interface configured to:selectively and iteratively send the nuclear reactor modeling data to one or more selected simulator modules of the simulator to generate nuclear reactor simulation data;receive the nuclear reactor simulation data includingneutronics data, fuel burn data, thermal hydraulics data, material performance data;analyze and update the nuclear reactor modeling data and the simulation data to maintain updated nuclear reactor modeling data that represents a new state of the nuclear reactor; andstandardize the updated nuclear reactor modeling data to create updated standardized modeling data of the abstract nuclear reactor model to represent an updated state of the nuclear reactor core for export to a database, the updated standardized modeling data further includes one or more variables including at least one of the group comprising density, flux, power, and flow, wherein standardizing includes defining structural and behavioral patterns in an object oriented environment that are sufficient to describe a state of the nuclear reactor model at an identified moment in time;a database, coupled to the modeling interface, and configured to receive the standardized data, whereby the standardized data is maintained for subsequent analysis. 2. The system of claim 1, wherein the plurality of assembly structures includes one or more assemblies representing shielding of the nuclear reactor core. 3. The system of claim 1, wherein the standardized modeling data is received from a file. 4. The system of claim 1, wherein the standardized modeling data is received via a graphical user interface. 5. The system of claim 1, wherein the nuclear reactor performance data relates to fission product removal. 6. The system of claim 5, wherein the nuclear reactor performance data further relates to coolant expulsion. 7. The system of claim 5, wherein the nuclear reactor performance data relates to fission gas removal. 8. The system of claim 1, wherein the neutronics simulator module includes a stochastic simulation tool. 9. The system of claim 8, wherein the stochastic simulation tool is based on a Monte Carlo N-Particle transport code (MCNP) simulation tool. 10. The system of claim 1, wherein the neutronics simulator module includes a deterministic simulation tool. 11. The system of claim 10, wherein the deterministic simulation tool is a REBUS simulation tool. 12. The system of claim 1, wherein the modeling interface is configured to determine and maintain the standardized modeling data of the abstract nuclear reactor model in an object oriented programming environment. 13. The system of claim 12, wherein the standardizing step further comprises defining creational patterns in the object oriented environment. 14. The system of claim 1 further comprising a data viewer coupled to the database that enables multi-dimensional visualization of the standardized data representing the nuclear reactor core. 15. The system of claim 1 further comprising a graphical user interface (GUI) coupled to the modeling interface, the GUI receives the standardized modeling data;and the modeling interface, in determining the standardized modeling data, receives the standardized modeling data from the GUI. 16. The system of claim 15, wherein the GUI includes a graphics menu wherein a user can select models that represent a nuclear reactor fuel assembly structure, or a block representing an axial portion of the fuel assembly structure. 17. The system of claim 15 further comprising a graphical user interface (GUI) wherein the GUI is used to review a current state of the abstract nuclear reactor. 18. The system of claim 17, wherein the GUI is configured to interact with the abstract nuclear reactor model by modifying or supplementing the standardized modeling data. 19. The system of claim 1 further comprising:a communications interface coupled to the database and a reactor control system and configured to provide the standardized data to the reactor control system. 20. The system of claim 1, wherein the material identifier of the cell data includes material characteristics and amount of material of the at least one material of the associated cell of the nuclear reactor model, the cell data being heterogeneous for the plurality of cells of the nuclear reactor model. 21. The system of claim 20, wherein the simulator determines an average rate of change of the amount of material of the cell. 22. The system of claim 21, wherein the simulator updates the cell data based at least in part on the average rate of change of the amount of material of the cell. 23. The system of claim 22, wherein the simulator includes a deflagration wave builder module that receives an equilibrium case of neutronic activity and builds an enrichment distribution over a plurality of fuel assemblies to achieve a desired deflagration wave and determines at least one move quantity of the amount of material of the cell, and updates the cell data based at least in part on the move quantity. 24. The system of claim 1, wherein the simulator includes a control rod module that simulates insertion and removal of control rods and produces control rod worth curves. 25. The system of claim 1, wherein the simulator includes a deflagration wave builder module that receives an equilibrium case of neutronic activity and builds an enrichment distribution over a plurality of fuel assemblies to achieve a desired deflagration wave. 26. The system of claim 1, wherein the thermal hydraulic module determines flow orificing and temperature distributions across a plurality of cells. 27. The system of claim 19, wherein the reactor control system controls neutron absorption assemblies of the nuclear reactor core in response to the standardized data. 28. The system of claim 19, wherein the reactor control system controls flows of fluids through the nuclear reactor core in response to the standardized data. 29. The system of claim 1, wherein the at least one fuel pin material includes a first material for a first block of the first fuel assembly structure and a third and different fuel pin material for a second block of the first fuel assembly structure. |
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063046296 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention This present invention relates to the field of relatively compact scanner apparatus and methods and more particularly to apparatus and methods for scanning objects which are transported by a conveyor belt through a temporarily sealed tunnel, such as in contraband detection systems. 2. Description of the Prior Art Scanners, particularly "compact" scanners, are used for detecting contraband at schools, correctional mail screening, courthouse security, airport hand parcels, and industrial processing applications. These scanners employ tunnel housing, usually leaded and wired in part on the outside, an isolating device, a conveyor device, a bed assembly housing in which the conveyor device is substantially located, and framing and thinner painted covers to hide the unsightly framing and the lead. The tunnel housing typically has a top portion, and side portions which together with a top portion from the bed assembly housing, form a substantially enclosed area. The tunnel housing is also provided with entrance and exit openings to the substantially enclosed area. The isolating device substantially covers the entrance and exit openings and is typically in the form of two separate lead curtains. One lead preferably fabric curtain is bolted to flat framing at the entrance opening, and the other lead preferably fabric curtain is bolted to flat framing located at the exit opening. Isolating devices permit the passage of conveyed objects into the substantially enclosed area formed by the tunnel housing and the top portion of the bed assembly housing, which is typically x-ray scatter lead shielded on the outside and may also substantially exclude light, noise, heat, cold, moisture, dryness, electrostatic or electromagnetic fields, dust gasses or chemical vapors while the conveyed objects are being analyzed. Scanners analyze objects which are brought into the enclosed area formed by the tunnel housing and the top portion of the bed assembly housing by the conveyor device. The conveyor devices are typically comprised of relatively short lengthened conveyor belts. Short lengthened conveyor belts, particularly those with a relatively low length to width ratio, such as of less than twelve to one, 12 to 1, often mistrack causing damage to the conveyor belts, objects being scanned, and other parts of the system. Currently, expensive and elaborate tracking mechanisms such as precise construction of components, toothed or perforated belting to mesh with drive gears or belt grooves raised, profile rails, servo-drive tracking adjustment mechanisms, and reliance on a human attendant are used for tracking conveyor belts. The framing provided to structurally connect the tunnel housing, the bed assembly housing, the conveyor device and the isolating device is often elaborate, wasteful, and space consuming, and requires a plurality of cover panels to hide the framing, the lead, the detector assembly and wiring. Scanners are needed which are more compact in overall width and length without sacrificing the width of the enclosed area inside the tunnel housing, and which are simpler and less costly to manufacture. SUMMARY OF THE INVENTION A compact and reliable scanning apparatus and method is provided. The scanner in one embodiment comprises a conveyor device, a tunnel housing, a bed assembly housing, an isolating device, and one or more analysis devices. The tunnel housing is comprised of top and side portions which together with a top portion from the bed assembly housing form a substantially enclosed area for analyzing objects. The bed assembly housing encloses most of the components of the conveyor device. The conveyor device is typically comprised of a conveyor belt, rollers and a conveyor tracking device. The conveyor tracking device is preferably comprised of first and second channels formed in first and second rails, as taught in U.S. patent application Ser. No. 08/584,469, the disclosure of which is specifically incorporated herein by reference thereto. The conveyor belt, preferably includes a first edge and a second edge, and an inner surface and an outer surface. The first and second edges of the conveyor belt typically pass through the first and second channels formed in the first and second rails, respectively. The first and second rails inhibit the conveyor belt from misaligning. The conveyor belt preferably traverses a forward path and a return path and the first and second rails are preferably provided in the return path. The first and second rails are preferably opposite one another. The isolating device is preferably comprised of separate first and second curtains which extend outward from separate first and second brackets, respectively. The first and second brackets and curtains are preferably adaptable for insertion into first and second slots, respectively, located in the top portion of the tunnel housing, near the entrance and exit openings, respectively. The slots can also be called slits. The brackets, the curtains, and the slots of the housing, are typically adaptable so that the curtains can be inserted into and through the appropriate slot but the brackets cannot be inserted through the appropriate slot. After the brackets and curtains have been inserted into their respective slot each curtain should entirely the cover either the exit or the entrance opening. The isolating device preferably temporarily seals off the substantially enclosed area bounded by the tunnel housing and the top portion of the bed assembly housing so that no X-rays will leak out. In the preferred embodiment of the present invention the tunnel housing, the bed assembly housing, the conveyor device, the one or more analysis devices, and the isolating device are constructed in a manner which provides a largely frameless scanner apparatus. The bed assembly housing preferably comprises a top portion which is used as with the inverted U-shaped sheet metal housing to form a substantially enclosed area. The bed assembly housing further preferably comprises first and second side portions which are preferably fixed to the first and second side portions of the housing providing structural support and reducing the need for framing. One of the analysis devices may be comprised of detector housing metal members which are attached to the inside side and top of the frameless housing. In this way the long slits on two sides, which weaken the inverted U, are eliminated from conventional housing constructions. The slits appear instead in the thinner, substantially metal, detector housing members which form a side and the top of the tunnel. The analysis devices may be various types including X-ray or electromagnetic generators and detectors or vapour detectors. The efficient construction of the conveyor device, the isolating device, the tunnel housing, the bed assembly housing, and the one or more analysis devices of the scanner maximizes the cross sectional area, and the width and height for the enclosed area formed by tunnel housing and the top portion of the bed assembly housing of the scanner for a given overall cross sectional width and height. In addition, the construction significantly simplifies and reduces the number of metal parts needed to frame, cover the framing, or otherwise make up the scanning device, decreasing its construction and operating costs. |
053655664 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention is directed to a radiation diaphragm of the type having a diaphragm lamella engaged by an adjustment system, so that the lamella is adjustable in the ray beam of a radiation transmitter. 2. Description of the Prior Art European Application 0 373 285 discloses a gating device for an x-ray diagnostics installation which includes a rigid element that is adjustable around a longitudinal axis in a plane perpendicular to the useful ray beam, and which has an arcuate edge. As a result of the variable arrangement of the arcuate edge, the useful ray beam can be individually matched to (gated onto) an arcuately shaped examination subject. German OS 35 00 812 discloses a gating mechanism for x-radiation which includes rigid diaphragm lamellae adjustable into the ray beam, these lamellae lying against one another parallel to both sides of a slot, and being individually adjustable in the longitudinal direction by an adjustment system dependent on the size of a subject. In angiographic examinations, for example of the legs of a patient, x-radiation is incident on regions of an image intensifier, used as the radiation receiver, which are not shadowed by the legs directly. Those regions may comprise a relatively large-area. The excessively high radiation intensity in these regions leads to a degradation of the image quality due to reduction of contrast and glare; moreover, the stray radiation is unnecessarily increased. In order to counter these negative effects, diaphragms, for example of sheet lead, are employed, which are laterally moved into the beam path so that they are matched as well as possible to the contours of the examination subject. This gating is suitable in order to blank the ray beam at the sides of the legs. Strip-shaped diaphragms having a fixed shape that are arranged in the region between the legs are employed for blanking radiation that is incident on the image intensifier between the legs. It is also possible to provide plastically deformable and absorbent volumes, for example sacks of rice flour, in this region. This technique, however, is difficult to manipulate and increases the stray radiation. SUMMARY OF THE INVENTION It is an object of the present invention to implement a radiation diaphragm of the type having a lamella adjustable by an adjustment system, such that the radiation that is incident directly on the radiation receiver, and is thus unattenuated, is blanked as well as possible. This object is achieved in accordance with the principles of the present invention in a radiation diaphragm wherein the diaphragm lamella is flexible, preferably composed of elastic material, and wherein means are provided for flexing the diaphragm lamella so that radiation absorption can be well-matched to examination subjects that differ in shape. Preferably, elements of the adjustment system engage the diaphragm lamella such that the diaphragm lamella is adjustable around an axis aligned perpendicularly relative to the central ray of the ray beam, since the effective width of the diaphragm lamella can thus be set in a continuum within a broad range. For use within an examination subject not having any straight limiting lines but instead having only contours curved, the adjustment system preferably engages end faces of the diaphragm lamella lying opposite one another, and the end faces are adjustable along the aforementioned axis and toward one another. The diaphragm lamella can thus be well-matched to arcuate examination subjects. An improved adaptability of the diaphragm lamella to different examination subjects is achieved in an embodiment wherein the adjustment system includes first and second holders for the diaphragm lamella which are adjustable independently of one another. The adjustment possibilities of the diaphragm lamella are thus enhanced. Preferably the diaphragm lamella is composed of rubber containing lead oxide since it is thus extremely flexible and has a high radiation absorption. The diaphragm lamella may be composed of a plurality of sub-lamellae, or a plurality of elastic diaphragm lamellae may be provided that are adjustable relative to one another, thereby enhancing the adaptability of the diaphragm lamella to examination subjects having different shapes. |
description | This application is a continuation-in-part application from U.S. application Ser. No. 13/065,437 filed on Mar. 22, 2011 and claims priority from the foregoing application. The United States of America may have certain rights to this invention under Management and Operating Contract No. DE-AC05-84ER 40150 from the Department of Energy. The present invention relates to a panel for shielding thermal neutrons through the use of lightweight panels which incorporate a high percentage of the element Boron, and a method of making such a panel. Neutron radiation may be generated as a result of a variety of nuclear reactions or interactions. More specifically, devices such as particle accelerators and nuclear reactors may emit neutrons during operation. A portion of such neutron emissions may subsequently classify as thermal neutrons. Neutrons, including thermal neutrons, have a deleterious effect on both living matter and inanimate objects. Thermal neutrons may also participate in neutron activation, thereby inducing radioactivity in environmental materials, equipment, and structures. It is of vital importance, therefore, to provide adequate shielding from any sources of neutron radiation. Various methods and devices are known to be capable of providing shielding from such radiation. It is known that elemental Boron has beneficial properties when used as a component of shielding devices. The highest density Boron possible is desirable in order to maximize the effectiveness of the shielding. As a result, shielding arrangements such as dry-packed Boron Carbide in metal boxes, Boron-loaded polyethylene plastic sheets, and Boron-loaded drywall have been disclosed in the art. Unfortunately, none of the foregoing technologies or systems are able to achieve a high Boron density. Further, all such technologies are traditionally quite expensive to deploy. It is therefore preferable to have a cost-effective method of shielding that is able to take advantage of the characteristics of the element Boron so as to provide an adequate amount of shielding from thermal neutrons. It is further desirable that such shielding should be lightweight and easily transported and installed. Accordingly, a need exists for boron-enriched panels which can be easily deployed to shield discrete rooms or locations. It is an object of the invention to provide a boron shielding panel and a method of making same which can be used as an effective but low-cost thermal neutron shield, and, further, possess sufficient rigidity, be easily maintainable, cleanable, and customizable. A boron shielding panel which can be used as a thermal neutron shield. Boron, in the form of Boron Carbide of varying grit sizes, is added during panel manufacture. In a preferred embodiment of the invention, the total Boron Carbide content of the mixture includes 50% coarse Boron Carbide particles and 50% fine Boron Carbide particles. The panel provides an efficient and inexpensive shield for thermal neutrons that is easily deployed and customizable for the required application. It is recognized in the art that the element Boron may be used in various fashions in order to provide radiation shielding. Boron is particularly suitable for neutron shielding applications as it has one of the highest neutron absorption cross-sections of all elements. The ability of Boron to effectively capture neutrons makes it ideal for applications involving thermal neutron shielding. A cost-effective method of shielding thermal neutrons can therefore be realized by making composite panels with a high percentage of Boron. It is observed that the compound Boron Carbide (B4C) contains as much as seventy-six percent (76%) Boron by weight and is the highest Boron-containing compound known. Boron Carbide is commonly used as an abrasive, in anti-ballistic materials, and in industrial applications. It is a hard, granular material which can be obtained in various grit or particle sizes. In the preferred embodiment of the invention disclosed herein, the shielding panels are composed of a resin base and Boron Carbide particles. Specifically, the three principal components of the panel are as follows: (1) resin base or glue, (2) hardener, and (3) Boron Carbide. In the preferred embodiment, the glue consists of an unsaturated polyester resin in a styrene monomer (C6H5CH=CH2), such as the commercially available product POLYLITE™ 32132-18. It will be noted that the glue can consist of any resin or resin mixture with similar properties. A hardener (cure initiator), such as the commercially available NOROX™ MEKP-9 in liquid form, would be used in the mixture. Nuclear-grade Boron Carbide of two particle sizes, coarse and fine, are further included in the mixture. In order to prepare the panels, the glue and hardener are mixed together by weight. The Boron Carbide powder is then progressively introduced into the mixture. The final mixture consists of essentially sixty percent (60%) Boron Carbide and forty percent (40%) glue or resin mixture. The mixture is then poured onto a mold and permitted to dry, and, commensurately, harden. If necessary, the mixture may be agitated after pouring so as to facilitate the removal of air from the mix. It will be noted that three-dimensional molds can be used to prepare various customized three-dimensional forms and shapes for these panels. As an example, a mixture could be as follows: (1) 37 pounds of Boron Carbide, (2) 24 pounds of resin, and (3) 106 cubic centimeters of catalyst, for a total wet mixture weight of 61 pounds. As a further example, a panel of 6″×6″×⅜″ with a total weight of 13.4 ounces would consist of sixty percent (60%) Boron Carbide with a weight of 8.04 ounces and forty percent (40%) resin with a weight of 5.36 ounces. The use of varying Boron Carbide grit sizes is critical in order to achieve a high density of Boron Carbide content in the final panel product. In a preferred embodiment of the invention, two particle sizes, coarse and fine, are used. The fine grade consists of particles of an average size of 16.4 microns and a maximum size of 50 microns. The coarse grade consists of particles with an average size of 105 microns and a maximum size of 140 microns. The percentage of any one particular grade can vary between 30% to 70%, with the second grade being of a commensurate percentage. In the preferred embodiment, the total Boron Carbide content of the mixture includes 50% coarse Boron Carbide particles and 50% fine Boron Carbide particles. The final panel would be at least forty-six percent (46%) Boron by weight. The panels possess sufficient strength and rigidity to be utilized and mounted in a variety of shielding applications. The panels possess a hard surface but can be drilled, sawed, glued, or bolted with appropriate tools. The panels can also be prepared with a variety of surface colors so as to insure that they are aesthetically pleasing. Further, the panels are easily cleanable and maintainable. Potential industrial applications would include new nuclear reactor power plants, nuclear detection or fabrication facilities, buildings or rooms containing nuclear medical devices, particle beam facilities, high-density shielding for nuclear propulsion systems, and any other application where the reduction of thermal nuclear radiation must be accomplished. While the invention has been described in reference to certain preferred embodiments, it will be readily apparent to one of ordinary skill in the art that certain modifications or variations may be made to the composition and method without departing from the scope of invention described in the foregoing specification. |
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claims | 1. A method of collecting 3He from a nuclear reactor, the method comprising:a. providing heavy water at least part of which is exposed to a neutron flux of the reactor;b. providing a cover gas in fluid communication with the heavy water;c. operating the nuclear reactor whereby thermal neutron activation of deuterium in the heavy water produces tritium (3H) and at least some of the tritium produces 3He gas by β−decay and at least a portion of the 3He gas escapes from the heavy water and mixes with the cover gas;d. extracting an outlet gas stream, the outlet gas stream comprising a mixture of the cover gas and the 3He gas; ande. separating the 3He gas from the outlet gas stream using at least one of a thermal diffusion process, a fractional diffusion process, a heat flush process, a superleak process and a differential absorption process. 2. The method of claim 1, further comprising outputting a 3He gas stream for further processing. 3. The method of claim 2, further comprising treating the outlet gas stream to provide a treated cover gas stream. 4. The method of claim 3, further comprising mixing at least a portion of the treated cover gas stream into the cover gas in fluid communication with the heavy water. 5. The method of claim 1, wherein the step of extracting the outlet gas stream is performed while nuclear reactor is operating. 6. The method of claim 5, wherein the outlet gas stream is extracted as a generally continuous stream while nuclear reactor is operating. 7. The method of claim 1, wherein the step of separating the 3He gas from the outlet gas stream is an on-line process that is performed while the nuclear reactor is operating. 8. The method of claim 1, wherein when the cover gas contacts the heavy water at a free surface interface. 9. A method of collecting 3He from a nuclear reactor, the method comprising:a. providing heavy water at least part of which is exposed to a neutron flux of the reactor;b. operating the nuclear reactor whereby thermal neutron activation of deuterium in the heavy water produces tritium (3H) and at least some of the tritium produces 3He gas by β−decay and at least a portion of the 3He gas escapes from the heavy water;c. extracting an outlet gas stream including the 3He gas; andd, separating the 3He gas from any other gas in the outlet gas stream using at least one of a thermal diffusion process, a fractional diffusion process, a heat flush process, a superleak process and a differential absorption process. |
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047568526 | abstract | Disclosed is a method of installing a reversibly porous, air-diffusible, water-restrictive, polymer plug in a port that extends through the wall of a nuclear waste storage container. The plug is inserted a predetermined distance, for example, with the aid of a screwdriver applied to a slot in the plug's outer face. When inserted, the plug prevents the loss of nuclear waste through the port while the air-diffusible nature of the material allows gases to pass through the material. The resultant venting action of the plug prevents the creation of pressure differences between the interior of the container and the environment. Thus, the likelihood of the container becoming overpressurized and leaking is minimized. In addition, the water-restrictive nature of the plug material restricts the ingress and egress of water from the container, reducing the likelihood of groundwater contamination during storage. After insertion, a portion of the plug left projecting from the container's surface is removed, protecting the plug from external forces and tampering. |
040654001 | abstract | High level liquid waste solidification is achieved on a continuous basis by atomizing the liquid waste and introducing the atomized liquid waste into a reaction chamber including a fluidized, heated inert bed to effect calcination of the atomized waste and removal of the calcined waste by overflow removal and by attrition and elutriation from the reaction chamber, and feeding additional inert bed particles to the fluidized bed to maintain the inert bed composition. |
047327308 | abstract | A stuck fuel rod capping sleeve to be used during derodding of spent fuel assemblies if a fuel rod becomes stuck in a partially withdrawn position and, thus, has to be severed. The capping sleeve has an inner sleeve made of a lower work hardening highly ductile material (e.g., Inconel 600) and an outer sleeve made of a moderately ductile material (e.g., 304 stainless steel). The inner sleeve may be made of an epoxy filler. The capping sleeve is placed on a fuel rod which is then severed by using a bolt cutter device. Upon cutting, the capping sleeve deforms in such a manner as to prevent the gross release of radioactive fuel material |
046997555 | claims | 1. Ultrafiltration circuit for the primary cooling fluid of a pressurized-water nuclear reactor incorporating, inside a containment shell (1), a primary circuit (2) which communicates with the inner volume of the reactor vessel (3) containing a core consisting of fuel assemblies and in which the pressurized water constituting the primary fluid circulates, and at least one auxiliary circuit (8) taken off from the primary circuit (2) and comprising a discharge branch (10), on which are arranged means of cooling (12) and depressurizing (13) the primary fluid extracted by means of the auxiliary circuit (8) and which passes through the wall (1) of the containment shell, and a charge branch (11) for returning the fluid into the primary circuit (2) and likewise passing through the wall (1) of the containment, and, outside the containment, means (17, 18, 19, 20) of purifying and treating the cooled and depressurized fluid, wherein the containment shell contains (a) a first loop (27) taken off from the discharge branch (10) of the auxiliary circuit (8) and incorporating a pipe (29) for extracting and conveying fluid at its operating pressure and temperature into a first ultrafiltration device (30) located in the first loop (27), a pipe (31) for discharging filtrate at the outlet of the ultrafiltration device (30) and for returning this filtrate into the discharge branch (10) downstream of the point (33) where fluid is extracted by means of the extraction pipe (29), a valve (32) being inserted in the discharge branch (10) between the two pipes (29, 31), and a concentrate discharge pipe (39) which passes through the wall (1) of the containment and on which are arranged, inside the containment, means (40, 41) of cooling and depressurizing the concentrate before it is introduced into the discharge branch (10) of the auxiliary circuit (8) downstream of the cooling and depressurizing means (12, 13) arranged on this discharge branch (10); and (b) a second loop (28) taken off from the charge branch (11) of the auxiliary circuit (8) and incorporating a pipe (49) for extracting and conveying fluid at its operating temperature and pressure into a second ultrafiltration device (50) located in the second loop (28), a pipe (51) for discharging the filtrate at the outlet of the ultrafiltration device (50) and for returning this filtrate into the charge branch (11) downstream of the extraction point (54), a valve (52) being inserted between these two pipes (49, 51), and a concentrate discharge pipe (58) connected to the discharge branch (10) of the auxiliary circuit (8), inside the containment, upstream of the cooling and depressurizing means (12, 13) arranged on this branch (10). 2. Ultrafiltration circuit according to claim 1, wherein the discharge branch (10) of the auxiliary circuit (8) incorporates a part of the primary circuit (2) which includes a primary pump (7), and wherein the first loop (27) of the ultrafiltration circuit is taken off from the primary pump (7). 3. Ultrafiltration circuit according to claim 1, comprising a concentrate purification filter (60) arranged on the discharge pipe (38) outside the containment shell (1). 4. Ultrafiltration circuit according to claim 3, wherein the filter (60) is a filter with mixed-bed ion-exchanger resins. 5. Ultrafiltration circuit according to claim 1, wherein the concentrate retained by the ultrafiltration device (30, 50) is made to circulate by means of a circulating pump (36, 56) in a circuit (35, 55) incorporating part of the extraction and supply pipe (29, 49). 6. Ultrafiltration circuit according to claim 5, wherein the filtrate discharge pipe (38, 58) forms a link to the concentrate circulation circuit (35, 55), and a regulating device (37, 57) makes it possible to extract some of the concentrate circulating in the circuit (35, 55) via the discharge pipe (38, 58). 7. Ultrafiltration circuit according to claim 5, wherein the filtrate discharge pipe (38, 58) forms a link to the concentrate circulation circuit (35, 55), and an adjustable three-way valve makes it possible to extract some of the concentrate circulating in the circuit (35, 55) via the discharge pipe (38, 58). |
045308144 | description | Referring to FIG. 1, there is illustrated a portion of a nuclear power plant 10 wherein steam is supplied from a conventional nuclear steam generator 11 to a high pressure turbine illustrated at 12 (hereinafter referred to as "HP turbine"). After expansion of this main steam through the HP turbine 12 to perform work, it is exhausted to an apparatus for superheating steam such as the moisture separator-reheater illustrated at 14 for removal of moisture therefrom and for reheating the main steam prior to its discharge to a lower pressure steam turbine illustrated at 16 (hereinafter referred to as "LP turbine"). After expanding through the LP turbine 16 while doing work, the main steam is exhausted to a condenser 17 for condensing thereof, and the condensate is then returned via feed pump 19 and other conventional apparatus such as feed water heaters (not shown) to the nuclear steam generator 11 whereby the steam cycle is repeated with the addition of heat to produce the main steam and the return of the main steam to the HP turbine 12. The HP and LP turbines 12 and 16 respectively provide power output such as through means of an electrical generator illustrated at 18. Although only two steam turbines are illustrated in FIG. 1, the power plant may have more than two such turbines and an apparatus for superheating steam may be provided in the flow path of steam between any two successive turbines in accordance with this invention. For example, the power plant may also be provided with an intermediate pressure steam turbine. The output of a nuclear plant may be increased where excess turbine-generator capacity exists if reactor supplied heat which is normally used to reheat the main steam is instead utilized for increasing the number of degrees of superheat of the main steam to perform additional work in the HP turbine and heat from another source is provided to reheat the main steam before its delivery to the LP turbine. Since the expense of a nuclear reactor is such that it is desirable to make maximum use of it, in order to provide such increased output in accordance with one aspect of this invention, the moisture separator-reheater 14 is supplied with vapor such as steam from a separately fired vapor generator such as the steam generator illustrated at 20 for transferring heat to the main steam as it passes through the heat exchange apparatus 14. This steam generator 20 is separately fired in order not to utilize any of the heat supplied by the nuclear reactor whereby the steam generator 20 may function as a peak load or power upgrade device with maximum use being made of the reactor supplied heat to increase the power output of the plant when excess turbine-generator capacity exists. Inotherwords, in cases where the HP turbine has sufficient capacity to handle all of the main steam which can be provided by the nuclear steam generator, then additional heat may be added to the main steam by a separately fired steam generator 20 before its passage through the LP turbine 16 for even greater power output than could otherwise be provided by the nuclear reactor itself. In order to reduce the possiblity of radiation contamination in reheater 14 and the separately fired vapor generator 20, which may otherwise be caused by contamination of nuclear generated steam passing through reheater 14 by fuel clad leaks or nuclear steam generator tube leaks, the steam generator 20 for providing reheat vapor to the reheater 14 is fossil fuel-fired. For the purposes of this specification and the claims, a "fossil fuel-fired vapor generator" is defined as a non-nuclear vapor generator and is meant to include any of the various non-nuclear vapor generators which burn various fossil fuels such as, for example, oil, gas, coal, and coal-water mixtures, and is also meant to include vapor generators supplied with heat from such non-nuclear energy sources as solar and geothermal. After the vapor has given up heat to the main steam in the moisture separator-reheater 14 and has condensed as it passes through the tubes thereof, it is returned to the fossil fuel-fired steam generator 20 such as by means of the feed pump illustrated at 22 so that additional heat may be added to it for its return to the moisture separator-reheater 14, and the cycle is repeated. The steam supplied by many fossil fuel-fired vapor generators 20 can be provided at a higher pressure than the pressure of steam which may otherwise be provided by a conventional nuclear steam generator for reheating the main steam to thereby provide a higher reheat temperature. For example, a fossil fuel-fired vapor generator providing saturated steam at 650.degree. F. (343.degree. C.) should be able to reheat main steam from a high pressure turbine from around 380.degree. F. (193.degree. C.) to more than 600.degree. F. (316.degree. C.) However, many conventional nuclear steam generators are not designed to provide reheat saturated steam having a temperature above about 510.degree. F. (266.degree. C.). Referring to FIGS. 2 and 3 there is shown an apparatus for superheating steam which is indicated generally at 30. This apparatus 30 is a preferred embodiment of the heat exchange apparatus which is illustrated at 14 in FIG. 1. Apparatus 30 is provided with an elongate, horizontally disposed, generally cylindrical shell 32 supported by members 33. A main steam inlet 34 opens into the shell 32 to supply steam to be superheated, such as steam exhausted from HP turbine 12 in FIG. 1 which is to be reheated before its delivery to LP turbine 16. In accordance with one aspect of this invention, a diffuser separator 36 is provided at the main steam inlet 34 to use the velocity head of the incoming steam to remove a major portion of the entrained moisture which is then drained from the apparatus 30 through line 37. For example, a moisture content of 10 to 12 percent may be reduced to a level in the neighborhood of 1 to 2 percent utilizing the energy that would otherwise be lost at the inlet 34 to the apparatus 30. In addition, it is believed that the diffuser separator 36 may actually recover a portion of the velocity head normally lost to yield a pressure rise of perhaps 1/2 lb. per sq. in. (0.04 kg. per sq. cm.) to thereby reduce the power loss which would otherwise result as the main steam passes through the apparatus 30. The moisture which is not separated from the main steam requires additional steam to evaporate it thus resulting in added power loss, and its impingement on reheater tubes is a cause of tube temperature oscillation. In addition to carrying solids into the lower pressure turbine, such moisture contains solids which tend to foul the reheater tubes thereby further reducing heat transfer, increasing pressure drop, and requiring more down-time for cleaning. Therefore, in order to separate substantially all of the remaining moisture from the steam, a group of secondary separators 38 using corrugated scrubber plates is also provided in the path of the main steam through the apparatus 30. This separated moisture is then drained from the apparatus 30 through line 39. It should be readily apparent from viewing the arrangement of components within the shell in FIG. 3 that these separators 36 and 38 take up only a small percentage of the available space within the shell 32 thus leaving space therein to provide increased heat transfer surface. Apparatus 30 is provided with at least one bundle of tubes such as high pressure bundle 40 of tubes which may be supplied with heating vapor from such sources as turbine throttle steam or from a fossil fuel-fired steam generator. A low pressure bundle 42 of tubes is also preferably provided which may use extraction steam so that the amount of throttle steam required for the high pressure bundle may be reduced for improvement in power output, or it also may use steam from a fossil fuel-fired steam generator. When the low pressure tube bundle of a conventional reheater which has flow control devices is supplied with extraction steam, its drains may, due to pressure drop of the reheat steam in its passage through the reheater tubes, have to be routed to a lower pressure feed water heater than the feed water heater to which the drains could have otherwise been directly routed. However, since there should be very little pressure drop of the reheat steam in its passage through the reheater tubes when extraction steam is supplied to the low pressure tube bundle 42 of the present invention, its drains may be routed, for increased efficiency of the plant, to the same feed water heater as the extraction steam could have been routed otherwise. Each of the tube bundles 40 and 42 is provided with an outlet header 44 from which extends drain line 65. Each drain line 65 is routed at illustrated in FIG. 2 to provide some flexibility of movement to allow for expansions and contractions of the respective tube bundles. Each outlet header 44 extends generally longitudinally of heat exchange apparatus 30 and is located preferably about midway between the sides thereof and generally below mid-height of the shell. The tubes illustrated schematically at 46 of the tube bundles 40 and 42 open into and extend in an upwardly direction from the respective outlet header 44 and are inclined. Preferably, these tubes 46 are disposed to lie substantially parallel to a plane which is perpendicular to the longitudinal axis of the shell. FIG. 3 is a view taken is such a plane. In order to utilize the available space within the apparatus 30 for providing greater heat transfer surface than if each tube bundle 40 and 42 consisted of only one bank of tubes and thereby reduce shell-side pressure drops and terminal temperature difference to provide higher main steam temperature at the outlet from the apparatus 30 in accordance with the present invention, each tube bundle 40 and 42 is comprised of two banks 48 and 50 of tubes extending from the respective outlet header 44 in generally the configuration of a "V" as viewed in a cross-section of the apparatus 30 taken in a plane perpendicular to the longitudinal axis thereof as shown in FIG. 3. Inotherwords, a first bank 48 of tubes extends from the respective outlet header 44 in a direction upwardly and outwardly toward one side 52 of the shell 32 on one side of the vertical longitudinal centerplane 60 of the shell, and a second bank 50 of tubes extends upwardly and outwardly from the respective outlet header 44 to the other side 54 of the shell. In a view taken in a plane perpendicular to the longitudinal axis of the shell, as shown in FIG. 3, the angle illustrated at 56 at which each of the tubes 46 of the first bank 48 extends relative to the vertical longitudinal centerplane 60 of the apparatus 30 is opposed to the angle illustrated at 58 at which each of the tubes of the second bank extends relative to the centerplane 60. The particular angle at which each of these tubes 46 extends is not critical to this invention. However, an angle in the range of around 30 to 60 degrees is believed to effectively utilize the space available within the shell and provide adequate inclination of the tubes. The tubes 46 of the first blank 48 terminate at and open into a first inlet header 62 which is preferably adjacent the shell 32 on one side of the centerplane 60 thereof, and the tubes 46 of the second bank 50 terminate at and open into a second inlet header 64 which is preferably adjacent the shell 32 on the other side of the centerplane 60. As shown in FIG. 2, each of the headers 62 and 64 extends in a direction substantially parallel to the longitudinal axis of the shell 32. The outlet header 44 of the high pressure bundle 40 of tubes is disposed above the outlet header 44 of the low pressure bundle 42 of tubes, and each of the inlet headers 62 and 64 thereof is disposed above respective inlet headers 62 and 64 of the low pressure bundle 42 of tubes. The inlet headers 62 and 64 are connected through lines 66 to a source or sources of vapor such as steam from a fossil fuel-fired steam generator, turbine extraction steam, or throttle steam to provide heating steam to the inclined tubes 46 of the tube bundles. These tubes 46 are short (they generally do not extend in a longitudinal direction of the shell 32) and inclined, to quickly drain condensate over a short distance by gravity to thus provide a means for condensing the steam in the tubes without sub-cooling it and to thereby provide more efficient heat transfer and greater power output. By a "short tube" is meant a tube the length of which is less than the shell diameter of the heat exchanger in which the tube is located. The shell diameter is measured in a plane perpendicular to the longitudinal axis of the heat exchanger. However, it is recognized that a negligible amount of sub-cooling may occur. Condensate collecting in each of the outlet headers 44 may then be removed by draining it to respective vented drain tanks or by other suitable means through drain lines 65 after which it may be returned to the main steam cycle or to the fossil fuel-fired steam generator for reheating is previously described. In order to increase resistance to damage from violent temperature transients, each of the tube bundles 40 and 42 is supported at each respective upper inlet header 62 and 64 by a member 67 attached thereto which is supported by a member 68 attached to the shell 32. A portion of each member 67 is slideably supported by a portion of the respective member 68 to allow for expansions and contractions of the inlet headers 62 and 64. Each of the lower outlet headers 44 is suspended from the respective upper headers 62 and 64 by the respective tube banks 48 and 50 to allow for expansions and contractions of the tubes 46 and outlet headers 44. The secondary moisture separators 38 are disposed in the steam path between the primary separator 36 and the low pressure bundle 42 of tubes and are slidably supported by members 63 to allow for expansions and contractions. After passing through the primary separator 36 at the steam inlet 34, the main steam passes through the secondary separators 38 on one side or the other of the outlet headers 44 after which it passes upwardly and between the outlet headers 44 and respective inlet headers 62 and 64 and over the inclined tubes 46 whereby heat is exchanged from the heating steam in the inclined tubes 46 to the main steam after which the resulting dried and superheated main steam continues to pass upwardly and is discharged at an outlet illustrated at 69 which is disposed at the top of the apparatus 30. The main steam is then routed to a steam turbine such as LP turbine 16 in FIG. 1 for use therein. Baffles 61 may also be provided at suitable locations for regulation of steam flow through the reheater 30. The headers 44, 62, and 64 extend through apertures in the baffles 61 which apertures provide sufficient clearance to allow for differential motion between the baffles and headers during expansions and contractions. However, such clearances which are perhaps one-sixteenth to one-eighth inch (about 2 to 3 mm.), allow steam flow through the apertures which results in reduction of efficiency of the reheater 30. In order to reduce such steam flow, plates (not shown) which are anchored to the respective headers and which overlap the respective apertures are provided adjacent and parallel to the respective baffles. Referring to FIGS. 4 and 5, there is shown an alternative embodiment of apparatus for superheating steam indicated generally at 70. This heat exchange apparatus 70 may also be used as the reheater illustrated at 14 in FIG. 1. Apparatus 70 is provided with a horizontally disposed generally cylindrical shell 72 having a main steam inlet 74, a main steam outlet 76, a primary separator 78, a primary separator drain 79, secondary separators 80 and secondary separator drain 81 similar to those provided in the apparatus 30 of FIGS. 2 and 3. Two bundles 82 and 84 of tubes 86, each bundle provided with two banks 88 and 90 of tubes 86 extending in a "V" configuration from a respective lower header 92, are also provided. Each lower header 92 is positioned approximately midway between the sides 94 and 96 of the shell 72 and the tubes 86 extend upwardly and outwardly therefrom to locations on respective sides of the vertical longitudinal centerplane 98 of the shell which are adjacent the shell 72. However, in this embodiment, each lower header 92 serves as both a heating steam inlet header and a condensate outlet header. Rather than extending upwardly and opening into another header, the tubes 86 of each bank 88 and 90 extend upwardly and terminate at closed ends in order to eliminate the expense of providing additional headers and of providing joints between the tubes and headers that would otherwise be required, to increase available room for corrugated scrubbers, and to permit the use of longer tubes in each tube bank 88 and 90 whereby the number of tube-to-header joints may be even further reduced since a lesser number of tubes would thus be required. Such a construction is also provided to simplify fabrication, inspection, and maintenance of the apparatus 70. The upper closed ends 100 of the tubes 86 are inserted in apertures 102 in support plates 104. Support plates 104 slide in and are supported by guides 106 which guides are in turn anchored to the shell 72. Referring to FIG. 6, one of the tubes 86 is shown opening into a header 92 and extending between the header tube sheet 108 and a support plate 104. In order to provide for expansions and contractions of the tubes and to simplify construction, it is preferred that only a few such as 100 tubes out of 5,000 be anchored to a support plate 104 so that the remainder of the tubes are free to expand and contract. Each of the tubes 86 is closed by means such as plugging with plug member 110, it being preferable that they be plugged over a distance illustrated at 112 from the support plate 104 in a direction toward the respective header 92 of at least about 2 inches beyond the edge of the support plate 104 in order to prevent uneven heating of the support plate 104 and thereby reduce stresses from expansions and contractions. The tubes 86 are provided with suitable fins 113 along their length for improved heat transfer in accordance with conventional practice. However, in order to provide increased strength to the header tube sheets 108, the portions of the tubes 86 which are inserted and expanded into the tube sheets 108 are not provided with fins so that reduced diameter apertures 114 may be provided in the tube sheets 108 for insertion of tubes 86. In accordance with this alternative embodiment, heating steam is provided to the inlet headers 92 through lines 116. This heating steam flows upwardly from the respective inlet header 92 into the tubes 86 which are in communication therewith and gives up heat in heat exchange relation to the main steam which is simultaneously flowing past the tubes 86. The heating steam in the tubes 86 is expected to continue giving up heat to the main steam until it condenses after which the condensate thus formed is expected to flow by gravity down along the walls of the tubes 86 and back to the respective inlet header 92, and the condensing of the steam is expected to result in lower pressures for drawing of more steam into the tubes. Thus, the steam is expected to become condensed, but not sub-cooled, for improved heat transfer efficiency. The condensate is then discharged from the inlet header 92 to a conventional vented surge tank (not shown) or other suitable condensate receiving apparatus through condensate removal lines 118. It is commonly understood that at low temperatures and pressures such as a pressure below 100 lbs. per sq. in. (7 kg. per sq. cm.) absolute in the 100.degree. to 400.degree. F. (38.degree. to 204.degree. C.) temperature range the condensing of steam in such closed end inclined tubes as shown in FIGS. 4, 5, and 6 is not feasible. Referring to FIGS. 7, 8, and 9, in which the temperature may be converted from the Farenheit scale to the Centigrade scale by subtracting 32.degree. and multiplying the difference by five-ninths, in such a range of pressure and temperature, the concentrations of non-condensible gases such as oxygen, hydrogen, and nitrogen commonly found in the condensate are not sufficiently soluble in the condensate to prevent accumulation of undissolved gases in the tubes. Such an accumulation of gases will seriously impede heat transfer. For example, at a temperature of 400.degree. F. (204.degree. C.) and a pressure of 100 lbs. per sq. in. (7 kg. per sq. cm.) absolute, the solubility of oxygen in water is only 0.2 ml. per gr. However, it is believed that this difficulty will not be present in a superheater according to the present invention since the amount of dissolved gases in the reheat steam will usually be low and the high temperatures at which condensation will take place provide increased gas solubility in the condensate. Referring to FIGS. 7, 8, and 9, at the higher pressures and temperatures at which the apparatus 70 shown in FIGS. 4 and 5 will normally operate, the solubility of oxygen, hydrogen, and nitrogen in water is greater than about 0.7 ml. per gr. This is believed to be sufficient to prevent undesirable accumulation of non-condensable gases in the tubes 86. Since the tubes 86 are relatively short (the length of each tube is less than the shell diameter which may typically be about 12 ft.), the rate of condensate flow is expected to be not more than about 25 lbs. (11 kg.) per hour from each tube 86. Such a flow rate would not require a significant portion of tube cross-section for a typical tube diameter of about one in. (2.5 cm.). Therefore, it is believed that "flooding" of the tubes 86 and the resulting loss of heat transfer efficiency will not normally occur. In order to achieve a reduced pressure drop of the main steam between exhaust from the HP turbine 12 and the inlet to the LP turbine 16, the reheater 14 preferably is positioned adjacent the turbines 12 and 16, as illustrated in FIG. 1, to thereby reduce the pressure losses which would otherwise result from lengthy pipe runs. This position is distinguished from the locations of the nuclear steam generator and the fossil fuel-fired steam generator 20 which may be positioned remote from the turbines 12 and 16. The heating vapor to be supplied to the tubes of an apparatus 14 in accordance with this invention is not limited to steam but may include other satisfactory vapors such as, for example, the vapors of organic fluids such as diphenyl oxide and silicone fluid. Neither is either of the apparatus 30 and 70 required to have two bundles of tubes. For example, this invention is meant to include a heat exchange apparatus having a single bundle of tubes. In addition to reheating steam from a HP turbine such as an intermediate pressure turbine for delivery to a lower pressure turbine, either of the apparatus 30 and 70 may be utilized in accordance with this invention to superheat steam for other purposes such as superheating steam from a steam generator for delivery to a HP turbine. Certain features of this invention may sometimes be used to advantage without a corresponding use of the other features. It is also to be understood that the invention is by no means limited to the specific embodiments which have been illustrated and described herein, and that various modifications may indeed be made within the scope of the present invention as defined by the claims. |
047755105 | description | DETAILED DESCRIPTION OF THE PREFERRED EMBODIMENT The numeral 10 generally designates the hollow flow deflector of the invention. It is illustrated in FIG. 1 as being mounted on a cylindrical element of a fuel assembly 12 which may typically be a thimble or guide tube of anuclear fuel assembly but which also may be one or more of the cylindrical fuel rods, poison rods or elements 14 making up the square matrix schematically shown in FIG. 2. The cylindrical fuel assembly element or thimble 12 and the fuel rods or poison rods 14 are supported in well-known manner intermediate their ends by a fuel support grid made up of orthogonal strips 16 and 18 as schematically shown in FIG. 1. Ideally, the hollow flow deflector is mounted on the cylindrical fuel assembly element between grids, as shown in FIG. 1, in order to provide the desired subchannel mixing. The hollow flow deflector 10 includes an elongated hollow body with a central cylindrical opening 20 for engaging a cylindrical fuel assembly element 12 in tight fitting relationship therewith. As previously pointed out, while the hollow flow deflectors 10 are capable of attachment to any or all members of the fuel assembly, they are most conveniently mounted on each of these non-fuel bearing members or control element guide tubes 12. The deflectors can be attached to the hollow tubes by means of welding, bulging or other mechanical means. The hollow flow deflector 10 has an upstream end portion 22 of a first diameter or overall transverse outer dimension. The term "upstream" originates from the fact that flow of coolant through the fuel assembly is from bottom to top, in a direction parallel to the axis of the cylindrical fuel assembly elements 12 and 14. The body of the hollow flow deflector 10 also includes a downstream end portion 24 of a second and greater overall transverse dimension and an intermediate transition portion 26 joining the upstream end portion 22 with the downstream end portion 24. The flow deflector 10 includes a plurality of flow channels or grooves 30 regularly spaced about the periphery of its body. As shown in FIG. 2, the flow channels are conveniently made concave in cross-section and extend at least through the transition portion and the downstream end portion. The cylindrical central opening 12, of course, extends through and from the upstream end portion through the transition portion into and through the downstream end portion to permit its mounting on the cylindrical fuel assembly element, whether it is a thimble 12 or a fuel rod 14. The flow deflector 10 is conveniently machined from a length of zircalloy tubing such that the flow channel 30 may be easily machined and the varying transverse outer dimensions of the upstream end portion 22, the transition portion 26, and the downstream end portion 24 can be provided without the expense of complexly shaped dies. The cutting tool marks from the machining operation are visible from a close inspection of the hollow flow deflector 10. Thus, it will be seen that a hollow coolant flow deflector 10 is provided for diverting the flow of coolant fluid from one subchannel to another in order to promote mixing of the fluid. The deflectors are to be positioned at selected locations such that the vane projections formed by the flow channels 30 operate in regions of coolant flow of the fuel assembly where thermal hydraulic performance can be improved by local turbulence production, local coolant flow deflection or fuel assembly coolant flow deflection. The hollow deflectors 10 can be employed with the support grid structures to create a more effective mixing of the coolant without creating an objectionable degree of pressure-loss as the coolant flows through the reactor core, thereby resulting in reduced pumping requirements and concomitant plant operating costs. Flow visualization tests and laser velocity measurements were performed with one hollow flow deflector design in a 7.35:1 scale, 6.times.6 matrix flow model in an air test facility. During the flow visualization tests, it was observed that the flow was deflected into the adjacent subchannel at the downstream end portion of the deflector. Based upon the laser velocity measurements, it was observed that at a distance of one hydraulic diameter after the deflector, the turbulence level in the guide tube subchannel was increased by up to 50% and the flow was slanted towards the guide tube in the wake of the deflector, filling the void adjacent to the guide tube. At a distance of five hydraulic diameters after the deflector, the velocity profile and turbulence levels returned to the normal valves without the deflector. Thus, the flow deflection and increased turbulence in the vicinity of the hollow flow deflector clearly will improve the heat transfer between the fuel adjacent to the guide tube and the moderating coolant medium in that region. |
047175281 | abstract | A control system for a nuclear reactor is disclosed. A control rod strategy computer provides for dynamic control of core power distribution in both radial and axial directions and forms the basis for a partial trip capability. Several microprocessor-based computation centers are combined together in a data-sharing network which processors determine local power density, determine the instantaneous differential and integral reactivity worth of each group of control rods, determine and effectuate partial trip for immediate power reduction, determine and provide for uniform core burnup, and effectuate core reactivity changes by directing the movement of groups of four control rods from zero to one hundred percent of travel while minimizing power distribution factors throughout the core. The power control circuitry to move the groups of control rods are bus arranged such that the power circuitry is shared among the groups of control rods. In this manner, the power to the control rod drive mechanisms is controlled and bus arranged for distribution to these drive mechanisms. This arrangement allows separate housings or cabinets for each holding circuit for each group of control rods and one housing or cabinet for the moving circuit for all of the groups of control rods. |
051035049 | summary | From DE-A-29 23 286, a textile fabric is known whose orthogonally crossing warp threads and weft threads are made of spun mixed yarn of steel fibres of stainless steel and of textile fibres. The steel fibres may be extremely thin, less than 25 micrometers in diameter, for example. According to this document, the mesh is proposed to measure at least 0.5 cm. This textile fabric is designed for carpet floors or working garments, for example, in order to obtain an antistatic network and thus prevent electrostatic charging. However, such a fabric does not provide effective protection against microwaves and other electromagnetic radiation, to which in particular the hospital personnel, for example, when operating electromedical equipment such as X-ray, ECG and EEG apparatus and the like, and the personnel for operating radar installations, for example, are exposed. In the course of electronization, also heart pacemakers are miniaturized and refined, but at the same time rendered susceptible to electromagnetic radiation interference such as occurs in everyday life due to, for example, broadcast and TV stations and various electrical apparatus, electric motors, electrical and electronic ignition control devices in motor vehicles, shavers, electrical household appliances, and electronic computer installations and the like. Modern heart pacemakers are designed so as to take over the pacemaker function only when the normal heart rhythm is disturbed, but not to interfere when the heart rhythm is normal. The occurrence of electromagnetic radiation, however, may result in a disturbance of the heart pacemaker control such that incorrect information on the current heart activity may be received and the control may fail in that case. While it is true that clothing which shields against electromagnetic radiation exist, such as the vest known from US-A-4 196 355. However, this vest, due to its construction and weight, is much too heavy and uncomfortable to be suitable for normal applications. According to the invention, a textile fabric and clothing made thereof are provided, which are effectively suitable for shielding against electromagnetic radiation, in particular in the microwave range, and, at the same time, do not restrict the wearing comfort of usual clothing. The invention is based on the perception that total shielding against electromagnetic radiation is not necessary, since radiation below certain intensities can be accepted without health impairment, and that this provides the possibility of achieving sufficient shielding by means of clothing even without restriction on the wearing comfort. For shielding against electromagnetic radiation, the invention provides a textile fabric whose orthogonally crossing warp threads and weft threads are made of spun mixed yarn of steel fibers of stainless steel and of textile fibers. To this end, the textile fabric has the quality of fabrics for usual clothing, wherein the textile fibers comprise cotton fibers and are twined with the steel fibers which measure 6 to 10 micrometers in diameter and constitute a content of 10 to 15% per weight of the mixed yarn, the distribution of the warp threads and the weft threads in the fabric and the composition of the warp threads and the weft threads being substantially the same, the number of mixed yarn threads in warp direction and in weft direction each is 18 to 20 threads per cm, and the yarn fineness of the textile fabric is in the range of 30 to 50 tex (g per km), especially of 38 to 40 tex, such that a shielding by 20 to 40 dB against electromagnetic radiation at a frequency of 10 GHz is established by the fabric. Owing to the content and fineness of steel fibers suggested in accordance with the invention, a textile fabric suitable for effective shielding against excessive electromagnetic microwave radiation can be obtained in a clothing fabric quality, which textile fabric is not too stiff. Hence, clothing can be made which is comparable to clothing without steel fibers. For the shielding effect, it is essential that the steel fibers and the textile fibers are spun and twined with each other in a manner such that a substantial part of the steel fibers is exposed on the exterior surface of the mixed yarn and sufficient mutual electrical contact of the fibers is achieved in the warp and weft threads at the crossings of the fabric to form a Faraday cage. The average number of the steelfibers in the yarn cross-section is preferably 10 to 15. Preferably, the thickness of the steel fibers measures 8 micrometers, and the content of the steel fibers in the mixed yarn is preferably 13.5% per weight. The length of the steel fibers is preferably in the range of the length of the cotton fibers and, hence, measures 3 to 10 centimeters. Although it is further possible to provide a mixture of cotton fibers and polyamide fibers, it is preferred to provide textile fibers exclusively of cotton. Cotton is capable of absorbing moisture and improves the electrical conductibility with increasing moisture absorption. Although a twined one-thread yarn, i.e., a yarn only single-twined, may be used, it is preferred to use a mixed yarn made of double-twined mixed yarn threads, each of which is made of textile fibers and steel fibers in a twined manner, wherein the single threads have a yarn fineness of 16 to 20 tex and a degree of turns of 550 to 650 turns, especially Z-turns, per m, and wherein the degree of turns of the double-threads is 400 to 480 turns, especially Z-turns, per m. The textile fabric according to the invention is light (the weight of the textile fabric is preferable in the region of 160 g per m.sup.2) and permeable to air and washable like other cloths without impairment of the shielding effect against electromagnetic radiation as would be the case if there were no discrete steel fibers but, instead, for example, a metal coating of the textile fibers. Moreover, the textile fabric according to the invention may be dyed such that pleasing and fashionable articles of clothing can be made of it. A textile fabric according to the invention having been proved and being excellently effective to shield heart pacemaker against microwave radiation was made of pure cotton with spun with steel fibers. The warp threads and weft threads were made of double-twined yarn of a fineness of R 38 to 40 tex made of single-threads of a fineness of R 16 to 18 tex. The degree of turns of the single threads was 590 to 640 Z per m and that of the Yarn was 420 to 440 Z per m. The content of steel fibers spun into the single-threads was about 13.5% per weight of the mixed yarn. The thread density of the textile fabric was 18 to 20 threads per cm, each, in warp direction and in weft direction and the average weight of the textile fabric was 160 g per m.sup.2. The textile fabric had an elongation of 9 to 14% at a breaking force of 638 to 672 N. The clothing according to the invention is completely or partly made of a textile fabric according to the invention. This textile fabric may completely form the clothing or may be provided as its interior lining. It is also possible to line the textile fabric itself with a lining of different textile material. For the manufacture of this textile fabric, usual pieces of fabric are cut out which are sewed together along joint seams. According to the invention, it is essential that the textile fabric is made of the above-presented textile/steel-fiber fabric according to the invention and covers at least the upper part of the body and the hip area of the person wearing the clothing as well as at least the person's upper arms approximately to the elbows. To avoid interruption of the shielding effect in the area of the joint seams, the joint seams should be turned up into each other and sewed together by at least two seams with a sewing yarn that is also a textile-fiber/steel-fiber mixed yarn of the kind according to the invention. However, the content of steel fibers in the sewing yarn may possibly be greater than the content in the threads of the textile fabric. All fasteners of the clothing, such as zip and button fasteners, should be free of metal and be underlayed with an interior border band or flap made of the textile-fiber/steel-fiber fabric according to the invention and provide an overlap breadth of at least 5 and preferably 7 centimeters. If pockets are provided in the textile fabric, these pockets should not be inserted into but put on the fabric, in order to avoid interruptions of the shielding effect. In a preferred embodiment of a clothing according to the invention, the clothing is in the form of a shirt or T-shirt which has at least elbow-length sleeves, covers the hip area of the person wearing the shirt and is provided with a closable neckline that, even when not closed, is covered by an interior flap made of the textile-fiber/steel-fiber fabric according to the invention. As proven by numerous experiments, such a clothing is most suitable for wearers of heart pacemakers to attain an effective shielding of the heart pacemaker against electromagnetic radiation disturbances and, at the same time, achieving the wearing comfort of other shirts or T-shirts. Owing to the invention, shirts or T-shirts of this kind may be manufactured in such a way that they look like other shirts or T-shirts and, hence, the fact that a person needs a heart pacemaker is not discernible from this kind of clothing. In another preferred embodiment, clothing according to the invention suitable for the personnel of radar installations takes the shape of an overall, in which the textile-fiber/steel-fiber fabric according to the invention covers the upper part of the body as well as at least the upper arms and the hip area, at least down to the knees. As protective clothing for hospital staff against the influence of electromagnetic radiation emitted by electromedical apparatus, two-piece clothing consisting of a jacket or blouse and a pair of trousers is suggested, wherein the pair of trousers is overlapped by the jacket or blouse by at least 10 centimeters and the jacket or blouse in its chest area has overlapping parts of textile-fiber/steel-fiber fabric held together by Velcro-type fasteners. Since clothing according to the invention is made of comfortable textile material, it can be worn completely closed, while the Velcro-type fasteners prevent the jacket or blouse from being worn open. In this arrangement, one of the overlapping parts of textile fabric preferably extends to ne shoulder of the wearing person, and the other part of textile fabric extends at least to the middle of the person's body. The waist measurement of the trousers of this clothing is preferably also adjustable by means of Velcro-type fasteners. In such clothing according to the invention, an interior lining may be formed of the textile-fiber/steel-fiber fabric according to the invention, while the exterior side of the clothing may, for example, made of light cotton cloth of a quality usual for hospital clothing. When the clothing according to the invention is embodied by an overall or two-piece suit, a stand-up collar containing textile fabric material according to the invention is preferably additionally provided. |
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