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053274717 | description | Referring now to the figures of the drawing in detail and first, particularly, to FIG. 1 thereof, there is seen a completely assembled fuel assembly having a longitudinal axis being indicated by reference symbol AX, in a view which is not to scale. The fuel assembly is surrounded laterally by a fuel assembly case or jacket WC, which is open at the top and bottom. A fuel assembly head or cap HD and a foot or base part FT are located there. The foot part FT is positioned in the core of a reactor on a base grid by means of a bracket FTa. A transition piece 2 forms a flow channel that leads from an inlet opening 3a to coolant inlets 3 in a bottom plate 1 that covers the foot part or the lower end of the fuel assembly. A lower edge WCa of the fuel assembly case WC is supported against and largely sealed off from the foot part FT and its bottom plate 1 by means of a sealing spring 4. A coolant tube or "water tube" WR extends axially through the interior and preferably the center of the case WC. The coolant tube WC has one respective end piece WRa, WRb and respective openings 5a, 5b for the passage of coolant (water) at its lower end upper ends. The fuel assembly head HD and the bottom plate 1 are supports for the coolant tube WR. Spacers SP are mounted on the coolant tube at predetermined axial positions between retaining means in the form of stop bodies 90, 91. The spacers SP are at right angles to the tube WR and include support ribs. These support ribs form a grid with meshes or holes, which may be formed, for instance, from sheets that are welded together or from lengthwise and crosswise ribs that meet one another at right angles. Supported on these ribs are a number of fuel rods FR. Each of the fuel rods FR is parallel to the case, passes through the meshes of a plurality of spacers and carries closure caps FRa, FRb at the bottom and top. The fuel assembly head HD has a cover plate 6, which covers the fuel assembly case WC on the top and has coolant outlets 7. A handle 9 is disposed on the upper surface of the cover plate, and the cover plate 6, the handle 9 and the upper end piece WRb of the coolant tube are held together by a connecting part 8 that is constructed as a stop body. The fuel rods FR are firmly clamped substantially within the meshes of the spacers. The bottom plate 1 and the cover plate 6 serve only as stops that prevent major axial motions of the rods. The closure caps of the fuel rods therefore have no thread with which they would be screwed to the plates. Instead, the fuel rods stand on the bottom plate with their lower closure caps, and are also only loosely guided by the upper closure caps in appropriate receiving positions on the cover plate. The nuclear reactor fuel assembly of FIG. 2 has a fuel assembly head with a cover plate 6 and a fuel assembly foot part FT, each having a rectangular (in this case square) cross section, an elongated coolant tube WR with a cross-shaped, round or rectangular (in this case square) cross section, and an elongated fuel assembly case WC, which likewise has a rectangular (square) cross section. The coolant tube WR is disposed centrally in gridlike spacers SP, which are located inside the fuel assembly case and are spaced apart from one another, as seen in the longitudinal direction of the coolant tube. One fuel rod FR containing nuclear fuel reaches through each mesh or space of these spacers SP. These fuel rods loosely engage leadthroughs in the fuel assembly head and the fuel assembly foot with bolts that are formed on both ends by closure caps FRa, FRb. As can be seen in FIG. 4, all of the fuel rods are fixed in helical springs 95 which are biased for pressure. These springs are seated on the bolt on the lower surface or underside of the fuel assembly head and are supported on that lower surface and on the applicable fuel rod FR. While the fuel assembly head and foot are formed of stainless steel, the elongated fuel assembly case WC is manufactured from a zirconium alloy. The fuel assembly case WC is open on both ends and is fitted at one end over the cover plate 6 and at the other end over the fuel assembly foot FT. The fuel assembly case accordingly engages the fuel assembly head and foot laterally. On the non-illustrated upper end, the fuel assembly case has sheet-metal strips pointing inward at the corners, which are firmly screwed to stay bolts 97 on the top of the cover plate 6 in the head. The coolant tube WR having the rectangular or square cross section is likewise formed of a zirconium alloy. The sides of the cross section of the coolant tube WR are parallel to the respective adjacent side of the cross section of the fuel assembly case. The ribs of the spacers intersect one another at right angles and are each likewise parallel to two mutually parallel sides of the cross section of the coolant tube WR. All of the sides of the coolant tube cross section are spaced apart from the respective adjacent side of the cross section of the fuel assembly case WC by the same distance. The coolant tube is closed on the upper end with a first end piece WRb and on the lower end with a second end piece WRa. Both end pieces are likewise made of a zirconium alloy and are firmly welded to the coolant tube. The upper end piece WRb has two flow openings 5b for the coolant extending longitudinally of the coolant tube, which are best seen in FIG. 4, while the second end piece on the lower end is provided with a flow opening for the coolant which extends longitudinally of the coolant tube but cannot be seen in FIG. 2. The first end piece WRb on the upper end of the coolant tube has two stay bolts 120 and 121, which have different outside diameters and are each provided with a thread. The threads engage the cover plate 6 and the handle 9 on the outside of the fuel assembly head along a diagonal of the fuel assembly head. The handle 9 has a central part 9a to be screwed to the coolant tube and two lateral cantilever arms 9b, 9c, which are supported on the top of the fuel assembly head and are firmly screwed to two diagonally opposed stay bolts 97 between these stay bolts and the aforementioned sheet-metal strips in the corner of the fuel assembly case WC. Each of the handle and the fuel assembly head are firmly screwed onto the threaded bolts 120 and 121 on the top of the fuel assembly head with a union nut 122, until they meet shoulders 123 and 124 on the stay bolts 120 and 121, as shown in FIG. 5. Due to the two stay bolts, the head part is firmly screwed to the coolant tube in such a manner as to be secure against rotation. Since the stay bolts 120 and 121 have different outside diameters, it is assured that the fuel assembly head will always be firmly screwed to the water tube WR in the same angular position with respect to the longitudinal axis of the water tube. As is seen in FIG. 8, the second end piece WRa on the lower end of the coolant tube WR, has a threaded bolt 130 on the top of the fuel assembly foot, which reaches through a leadthrough of the gridlike bottom plate 1, where it is firmly screwed with a nut until it meets. In a boiling water nuclear reactor, the nuclear reactor fuel assembly of FIG. 2 is vertically disposed, and a coolant that is formed of a two-phase mixture of water and steam flows through it in the fuel assembly case. Water enters the fuel assembly case 5 through the gridlike grate of the fuel assembly foot part having the bottom plate 1, into the fuel assembly case 5, and wet steam exits from the fuel assembly case through a gridlike grate of the fuel assembly head having the cover plate 6. Only water in the liquid phase is located in the coolant tube and effects increased reactivity in the reactor core of the boiling water reactor. In a further feature of the nuclear reactor fuel assembly, as is shown in FIGS. 4 and 6, the lengths of the stay bolts 120 and 121 protruding past the shoulders 123 and 124 on the upper end piece WRb of the coolant tube, are selected to be so large that the fuel assembly head and handle are not retained in contact with the shoulders 123, 124, but instead the fuel assembly head is displaceable counter to the spring force of the compression springs 95 seated on the fuel rods with respect to this end piece WRb. The compression springs 95, which are constructed as helical springs, have one end supported on the cover plate 6 of the fuel assembly head and another end supported on the fuel rods and therefore on the bottom plate 1 of the fuel assembly foot. The compression springs can thus resiliently intercept any displacement of the handle and of the fuel assembly head, for example from forces of gravity in the direction of the shoulders 123 and 124. The shoulders 123 and 124 are advantageously located between the fuel assembly head and the engagement point of the compression springs 95 on the fuel rods FR, so that these compression springs are not subject to blocking, if the fuel assembly head strikes the shoulders 123 and 124 of the stay bolts 120 and 121 acting as stop bodies. The union nuts 122 on the stay bolts 120 and 121 that are located on the upper surface of the fuel assembly head are also stop bodies on the upper end piece. These union nuts define an initial position of the fuel assembly head or cover plate 6. They are each guided in a respective sheath 131 located on the handle 9. The union nuts 122 each have a tang 133 acting as a locking device on the bearing surfaces thereof that are disposed on the handle 9. As is seen in FIG. 7, the tangs each lock into place in an annular groove 134 serving as a counterpart locking device, on the bearing surface of the handle 9 on the union nut 122. As a result, the locking of the union nuts 122 to the handle 9 can be released only if a weight, for instance, acts upon the handle (or the fuel assembly head) in the direction of the fuel assembly foot, and a separating motion between the union nuts 122 and the handle takes place in the longitudinal direction of the stay bolt and counter to the spring force of the compression springs 95. Advantageously, the stay bolts 120 and 121 are firmly screwed to the upper end piece of the coolant tube WR, so that they can be replaced without further action being required, for instance if their threads on the outside of the fuel assembly head 2 are damaged. Non-illustrated compression springs, which may be helical springs, for instance, and are supported on the fuel assembly head and on the shoulders 120 and 121, may also be mounted on the stay bolts 120 and 121. As FIG. 8 shows, the threaded bolt 130 of the end piece WRa on the lower end of the coolant tube may reach through the bottom plate 1 of the fuel assembly foot part and may be displaceable counter to the spring force of a compression spring 96, that is constructed as a helical spring, with respect to the fuel assembly foot. This compression spring 96 is mounted on the threaded bolt 130 of the lower end piece WRa and has one end supported on the end piece and therefore on the fuel assembly head and another end supported on the fuel assembly foot. The fuel assembly head and the handle 9 can then be firmly screwed until they meet the respective shoulders 123 and 124 of the stay bolts 120 and 121. The compression spring 96, which is subject to blocking and is disposed on the inside of the fuel assembly foot, serves on one end as a stop body for the end piece on the fuel assembly foot, and on the other end, a nut 135 which is seated on the stay bolt 130 and is screwed onto the fuel assembly foot from below serves as a stop body and thus defines an initial position for the fuel assembly foot. As FIG. 8 shows, the fuel assembly case WC that is firmly screwed to the fuel assembly head may be displaced on a jacket surface 136 of the fuel assembly foot, along with the fuel assembly head, in the longitudinal direction of the fuel assembly case. In the cover plate 6 shown in FIG. 9, the positions of the fuel rods can be shown as center points of meshes in a grid with square meshes. A central region of 3.times.3 meshes is intended to receive a square coolant tube ("water tube") and is covered by a connecting part 8 for a load-bearing or supporting connection between the cover plate and the water tube. Since it is advantageous for only some of the fuel rods to extend practically over the entire length between the bottom plate and the cover plate, while the other rods are shorter and carry an upper closure cap on the upper end, are already located in the vicinity of a spacer and are spaced apart from the cover plate, the water outlet in the cover plate is formed by outlet openings 7 that are located between the positions of long fuel rods, and by enlarged outlet openings 63 which are disposed in the projection or imaginary extension of the shorter fuel rods. Since screw connections between the cover plate and the fuel rods are to be avoided, the upper closure caps FRb of the longer fuel rods end in the form of smooth, unthreaded bolts, which are only loosely disposed in sleeves 61 at the corresponding positions on the cover plate. Accordingly, the cover plate is not strained by the weight of the fuel rods. Instead, it merely prevents an axial displacement of the fuel rods exceeding a predetermined play. The sleeves 61 that are disposed at the appropriate receiving positions of long fuel rods, are joined together by ribs 62. The cover plate can therefore be made so thin and permeable that the coolant can pass through it virtually unhindered. By suitable dimensioning of the inlet openings 3 in the bottom plate, a uniform flow through the fuel assembly can be achieved, and the pressure loss at the cover plate can be negligible as compared to the pressure loss of the coolant as it flows through the bottom plate and the fuel assembly. At least most of the enlarged outlet openings, that is most positions of the shorter fuel rods, are advantageously each located on the diagonal of the case cross section. Fuel rods that are adjacent to the case wall preferably belong to the group of the long fuel rods. It is particularly with a fuel assembly in which the support ribs of the spacers form 11 rows and 11 columns, that the shorter fuel rods are each disposed in the third row or the third column (counting from the case wall inward), as can be seen in FIG. 9. The cover plate is supported on the case wall and on its rounded corners by lateral bearing lugs 65 and reinforcements 66 of corner sleeves. The handle 9 may have markings 900, 901, which serve to identify the individual fuel assembly and to define right and left, in particular if the fuel fillings in the various fuel rods are different for individual fuel assemblies and their parts. A load-bearing or supporting connection between the handle 9 and the cover plate 6 is advantageously achieved if these two parts are constructed integrally, for instance as a cast workpiece. The head HD formed by the handle and the cover plate need merely be screwed through the connecting part 8 to the upper end piece WRb of the water tube, in order to achieve a load-bearing or supporting mechanical connection with the water tube. A part in the form of a collar 67 shown in FIG. 10 is formed onto the handle, and the upper end piece WRb of the water tube is constructed as a socket pin 92 that is passed through the cover plate and welded to the water tube. This socket pin carries a male thread 93 on its upper end that is passed through the collar, and a union nut 94 is screwed onto this thread from above. The cover plate is supported against the bottom plate by the helical spring 95 that is subject to compression and that advantageously is disposed on the upper end piece or pin 92 between the water tube and the cover plate. If the cover plate is therefore held down counter to the compressive force of the helical spring 95, then the union nut 94 can be screwed onto the male thread 93 of the socket pin 92 up to a desired end position. When the helical spring 95 is relieved, they then form the stop for the collar 67 and the cover plate 6. A securing cap which is also disposed between this collar 67 and the union nut 94, has an upper end that fits partway around the union nut 94 and secures it against falling out. If the helical spring 95 is relieved in the end position, then corresponding profiles of the collar 67 and therefore of the nut 94 mesh with one another and prevent the union nut 94 from being able to rotate relative to the socket pin 92. By using the fuel rod FR1 as an example, it is seen that its upper closure cap FRb is only loosely guided with a certain lateral play in the corresponding sleeve 61. In the illustrated embodiment of these closure caps, the fuel rods (together with the water tube) can also be moved longitudinally counter to the force of the helical spring 95, in order to compensate for an expansion of material such as can occur from radiation and heating during reactor operation. The union nut 94 then serves as a stop body on the upper end piece WRb of the coolant tube and fixes the maximal elongation of the spring 95 and therefore the maximum spacing between the fuel assembly head with the cover plate 6 and the foot part with the bottom part 1. In the longitudinal section of FIG. 10, a fuel rod FR2 is seen directly next to the long fuel rod FR1. The fuel rod FR2 is disposed in a row behind the fuel rod FR1, because the position located directly next to the rod FR1 is occupied by a short rod, that is not visible in FIG. 10. The upper closure caps FRb do not carry any screw thread. This is particularly true for the lower closure caps FRa as well, which are visible in FIG. 11 and which stand unscrewed on the bottom plate 1, as is shown in FIG. 11. This bottom plate is joined rigidly and in a load-bearing or supporting manner to the lower end piece WRa of the water tube, so that the bottom plate and the weight of the fuel rods standing on it is supported practically completely by the end piece of the water tube and is transmitted to the cover plate and the handle 9 through the releasable and force-locking connection of the upper water tube end piece WCb. The lower end piece WRa, with an end pin that is constructed as a tube piece 110, engages the bottom plate 1 and the foot part FT of the fuel assembly. It is screwed there and secured against rotation by a securing bolt 106, which engages a recess of the end piece WRa. The lower closure cap FRa of a long fuel rod FR1 can stand on the bottom plate or it can be loosely guided in appropriate receiving openings. At least for the short fuel rods, a plug connection is provided, for example a base 310 that is constructed as a bayonet mount, having an inside which is engaged by an appropriate adaptor on the lower closure cap of the short fuel rod and which can be locked by a quarter turn, like a plug connection. With the exception of the screw connection between the water tube and the bottom plate, there are accordingly no further screw connections on the foot part. The foot part itself includes the transition piece 2, which has a lower end with star-shaped brackets FTa and an opening 3a and forms a flow channel that is covered on its upper end by the bottom plate 1 with the water inlets 3, through which the coolant enters the interior of the fuel assembly. The lower edge WCa is supported and sealed off from the lateral upper edge of the foot part FT by a sealing spring 4. In the left-hand part of FIG. 11, which shows a side view of the foot part, a lower long side 402 of the sealing spring 4 can be seen. The lower long side 402 is feathered by slits 405, while another long side 401 is bent around an edge between the bottom plate and the transition piece. Since this sealing spring largely prevents an uncontrolled flow of coolant between the foot part and the fuel assembly case, the coolant stream can be split in a defined manner into partial streams flowing inside and outside the case WC, by means of suitable dimensioning of the inlet openings 3 and lateral outlet channels 211. In the present construction, the fuel assemblies are held in a skeleton having a load-bearing or supporting spine which is the water tube. The fuel rods are not part of the skeleton, and the bottom plate does not include any threaded bores for receiving the fuel assembly closure caps. The bottom plates there need merely form suitable stops, on which the closure caps are seated, but can otherwise be constructed freely. In particular, the inlet openings 3 may, for instance, be constructed as narrow slits or in some other way, so that they are not permeable or passable to relatively large foreign bodies. In that event, the bottom plate acts as a sieve at which broken-off pieces or other foreign bodies of certain dimensions are retained, so that they cannot travel into the interior of the fuel assembly and impede the coolant stream there. The water tube WR in FIG. 11 also has stop bodies 90, 91, each of which defines the axial position of one spacer SP. In order to mount these spacers, one stop body is secured to the water tube at a time, for instance by spot welding, beginning at one end of the water tube. Next, from the other, free end, the appropriate spacer is shoved up to the level of the already secured stop body and is retained there by then securing the other stop body to the water tube on the opposite side of the spacer. In this way, all of the spacers can be successively secured to the water tube, as long as at least the upper or lower end of the water tube is still accessible. Before the load-bearing or supporting skeleton is mounted in final form by screw connections of the water tube and the head part, the fuel rods are inserted into the meshes of the various spacers. The final installation of the load-bearing or supporting skeleton takes place then, by screwing the head part onto the water tube, with the fuel rods being introduced into their appropriate receiving positions on the cover plate from below. Then, the fuel assembly case need merely be slipped over this, the rods and the skeleton. As a result, the installation and dismantling of the fuel assembly and replacement of the fuel rods are reduced to a few simple manipulations. In particular, the fuel rods need not be screwed into the bottom plate and unscrewed from it. In FIG. 12, the bottom plate 1 is shown with the closure caps FRa of the long fuel rods standing on them and with the base 310 secured to the bottom plate. The profile of the base is engaged by a profiled extension of the lower closure cap of a shorter fuel rod, serving as an adaptor 305. The base 310 is retained in a bore 308 in the bottom plate. The shorter fuel rods also preferably extend over at least the lower third and, for instance, the lower half of the fuel assembly, so that all of the fuel rods extend through the meshes of the lower spacers. In FIG. 12, one of the spacers SP is shown in the upper part of the fuel assembly. The upper closure cap of a shorter fuel rod ends into its vicinity, while the longer fuel rod extend extends as far as the fuel assembly head. The fuel assembly case preferably has a square cross section and the thickness of its side walls is reduced, at least in some axial regions of the fuel assembly, as is shown in FIG. 13. It may be composed of a plurality of parts, and the side walls have weld seams 30, 31, which serve partly as reinforcements and partly to connect profiled parts. FIG. 14 shows the uppermost spacer, which is located under the cover plate of such a fuel assembly having shorter and longer fuel rods. The support ribs for supporting the fuel rods are constructed as annular sheaths 501, 502, 503, which are secured to one another and carry support knobs 504 and support springs 505 for supporting the fuel rods. Outer ribs 507 are supported by peripheral knobs 508 on the fuel assembly case and carry lugs 509 that face into the interior of the fuel assembly. In this way, a fuel assembly that is very stable mechanically is created. It is assembled from only a few parts in a simple manner and can be easily inspected and dismantled. |
description | The present application claims priority from Japanese Patent Application No. 2011-052917 filed Mar. 10, 2011, and Japanese Patent Application No. 2011-271331 filed Dec. 12, 2011. 1. Technical Field This disclosure relates to a system and a method for generating extreme ultraviolet (EUV) light. 2. Related Art In recent years, semiconductor production processes have become capable of producing semiconductor devices with increasingly fine feature sizes, as photolithography has been making rapid progress toward finer fabrication. In the next generation of semiconductor production processes, microfabrication with feature sizes of 60 nm to 45 nm, and microfabrication with feature sizes of 32 nm or less, will be required. In order to meet the demand for microfabrication with feature sizes of 32 nm or less, for example, an exposure apparatus is needed in which a system for generating EUV light at a wavelength of approximately 13 nm is combined with a reduced projection reflective optical system. Three kinds of systems for generating EUV light are known in general, which include a LPP (Laser Produced Plasma) type system in which plasma is generated by irradiating a target material with a laser beam, a DPP (Discharge Produced Plasma) type system in which plasma is generated by electric discharge, and a SR (Synchrotron Radiation) type system in which orbital radiation is used. An extreme ultraviolet light generation system according to one aspect of this disclosure may include: a laser apparatus configured to output a laser beam; a chamber provided with a window, through which the laser beam from the laser apparatus enters the chamber; a target supply unit configured to output a target toward a predetermined position inside the chamber; a laser beam focusing optical system positioned to reflect the laser beam toward a predetermined position inside the chamber; a detector for detecting an image of the laser beam at the predetermined position; a target position adjusting mechanism for adjusting a direction into which the target is to be outputted; a laser beam focus position adjusting mechanism for adjusting a focus position of the laser beam; and a controller for controlling the target position adjusting mechanism and the laser beam focus position adjusting mechanism based on the image detected by the detector. An extreme ultraviolet light generation system according to another aspect of this disclosure may include: a first laser apparatus configured to output a first laser beam; a second laser apparatus configured to output a second laser beam; a chamber provided with a window, through which the first and second laser beams respectively from the first and second laser apparatuses enter the chamber; a target supply unit for outputting a target toward a predetermined position inside the chamber; a laser beam focusing optical system positioned to reflect the first and second laser beams toward a predetermined position; a detector for detecting an image of the second laser beam at the predetermined position; a target position adjusting mechanism for adjusting a direction into which the target is to be outputted; a laser beam focus position adjusting mechanism for adjusting a focus position of at least one of the first and second laser beams; and a controller for controlling the target position adjusting mechanism and the laser beam focus position adjusting mechanism based on the image detected by the detector. A method according to yet another aspect of this disclosure for generating extreme ultraviolet light in a system including a laser apparatus, a chamber, a target supply unit, a laser beam focusing optical system, a detector, a target position adjusting mechanism, a laser beam focus position adjusting mechanism, and a controller may include: detecting an image of a laser beam reflected by the laser beam focusing optical system at a predetermined position; and controlling the target position adjusting mechanism and the laser beam focus position adjusting mechanism based on the detected image. A method according to still another aspect of this disclosure for generating extreme ultraviolet light in a system including first and second laser apparatuses, a chamber, a target supply unit, a laser beam focusing optical system, a detector, a target position adjusting mechanism, a laser beam focus position adjusting mechanism, and a controller may include: outputting first and second laser beams respectively from the first and second laser apparatuses; detecting an image of the second laser beam reflected by the laser beam focusing optical system at a predetermined position; and controlling the target position adjusting mechanism and the laser beam focus position adjusting mechanism based on the detected image. Hereinafter, selected embodiments of this disclosure will be described in detail with reference to the accompanying drawings. The embodiments to be described below are merely illustrative in nature and do not limit the scope of this disclosure. Further, the configuration(s) and operation(s) described in each embodiment are not all essential in implementing this disclosure. Note that like elements are referenced by like reference numerals and characters, and duplicate descriptions thereof will be omitted herein. This disclosure will be illustrated following the table of contents below. Contents 1. Summary 2. Terms 3. Overview of EUV Light Generation System 3.1 Configuration 3.2 Operation 4. EUV Light Generation System Including Laser Beam Irradiation Image Detector 4.1 Configuration 4.2 Operation 4.3 Effect 4.4 Image when Target is Irradiated by Laser Beam 4.5 Control Flow 4.5.1 Main Flow 4.5.2 Parameter Initialization Subroutine 4.5.3 EUV Light Generation Position Setting Subroutine 4.5.4 EUV Light Generation Subroutine 4.5.5 Laser Beam Irradiation Image Detection Subroutine 4.5.6 Position Determination Subroutine 4.5.7 Target Position Control Subroutine 4.5.7.1 Modification of Target Position Control Subroutine 4.5.8 Laser Beam Focus Position Control Subroutine 5. EUV Light Generation System Including Image Detector for Detecting Images when Target is Irradiated by Pre-Pulse and Main Pulse Laser Beams 5.1 Configuration 5.2 Operation 5.3 Effect 5.4 Image when Target is Irradiated by Main Pulse Laser Beam 5.5 Control Flow 5.5.1 Parameter Initialization Subroutine 5.5.2 EUV Light Generation Subroutine 6. EUV Light Generation System in which Beam Delivery System Includes Actuator for Adjusting Focus of Laser Beam 6.1 Configuration 6.2 Operation 6.3 Effect 7. Supplementary Descriptions 7.1 Two-Axis Tilt Stage 7.2 Focus Position Adjusting Mechanism 7.3 Modification of Focus Position Adjusting Mechanism 7.4 Top-Hat Mechanism 7.5 First Modification of Top-Hat Mechanism 7.6 Second Modification of Top-Hat Mechanism1. Summary An overview of the embodiments is as follows. In the selected embodiments to be described below, an EUV light generation apparatus used with a laser apparatus may be configured to detect an image of a laser beam by which a target has been irradiated. The EUV light generation apparatus may also be configured to control the position at which a laser beam is to be focused and the position of a target, based on the aforementioned detection result. 2. Terms Terms used in this application may be interpreted as follows. The term “droplet” may refer to one or more liquid droplet(s) of a molten target material. Accordingly, the shape of a droplet may be substantially spherical due to its surface tension. The term “plasma generation region” may refer to a three-dimensional space in which plasma is to be generated. In a beam path of a laser beam, a direction or side closer to the laser apparatus is referred to as “upstream,” and a direction or side closer to the plasma generation region is referred to as “downstream.” The “predetermined repetition rate” does not have to be a constant repetition rate but may, in some examples, be a substantially constant repetition rate. The term “diffused target” refers to a target material in a state where at least one of pre-plasma and fragments of the target material is included. The term “pre-plasma” refers to a target material in a plasma state or in a state where plasma is mixed with its atoms or molecules. The term “fragments” may include fine particles such as clusters and microdroplets transformed from a target material as the target material is irradiated by the laser beam, or a mixture of such fine particles. The term “obscuration region” refers to a three-dimensional space defined by the specifications of an external apparatus, such as the exposure apparatus. Typically, the EUV light that passes through the obscuration region is not used for exposure in the exposure apparatus. 3. Overview of EUV Light Generation System 3.1 Configuration FIG. 1 schematically illustrates the configuration of an exemplary LPP type EUV light generation system. An EUV light generation apparatus 1 may be used with at least one laser apparatus 3. In this application, a system including the EUV light generation apparatus 1 and the laser apparatus 3 may be referred to as an EUV light generation system 11. As illustrated in FIG. 1 and described in detail below, the EUV light generation apparatus 1 may include a chamber 2, a target supply unit (droplet generator 26, for example), and so forth. The chamber 2 may be airtightly sealed. The target supply unit may be mounted to the chamber 2 so as to penetrate the wall of the chamber 2, for example. A target material to be supplied by the target supply unit may include, but is not limited to, tin, terbium, gadolinium, lithium, xenon, or any combination, alloy, or mixture thereof. The chamber 2 may have at least one through-hole formed in the wall thereof. The through-hole may be covered with a window 21, and a pulsed laser beam 31 may travel through the window 21 into the chamber 2. An EUV collector mirror 23 having a spheroidal surface may be disposed inside the chamber 2, for example. The EUV collector mirror 23 may have a multi-layered reflective film formed on the spheroidal surface, and the reflective film may include molybdenum and silicon that are laminated in alternate layers, for example. The EUV collector mirror 23 may have first and second foci. The EUV collector mirror 23 may preferably be disposed such that the first focus thereof lies in a plasma generation region 25 and the second focus thereof lies in an intermediate focus (IF) region 292 defined by the specification of an exposure apparatus 6. The EUV collector mirror 23 may have a through-hole 24 formed at the center thereof, and a pulsed laser beam 33 may travel through the through-hole 24. Referring again to FIG. 1, the EUV light generation system 11 may include an EUV light generation controller 5. Further, the EUV light generation apparatus 1 may include a target sensor 4. The target sensor 4 may be equipped with an imaging function and may detect at least one of the presence, trajectory, and position of a target. Further, the EUV light generation apparatus 1 may include a connection part 29 for allowing the interior of the chamber 2 and the interior of the exposure apparatus 6 to be in communication with each other. A wall 291 having an aperture may be disposed inside the connection part 29. The wall 291 may be disposed such that the second focus of the EUV collector mirror 23 lies in the aperture formed in the wall 291. Further, the EUV light generation system 1 may include a laser beam direction control unit 34, a laser beam focusing mirror 22, and a target collection unit 28 for collecting a target 27. The laser beam direction control unit 34 may include an optical element for defining the direction in which the laser beam travels and an actuator for adjusting the position and the orientation (or posture) of the optical element. 3.2 Operation With reference to FIG. 1, the pulsed laser beam 31 outputted from the laser apparatus 3 may pass through the laser beam direction control unit 34, and may be outputted from the laser beam direction control unit 34 after having its direction optionally adjusted. The pulsed laser beam 31 may travel through the window 21 and enter the chamber 2. The pulsed laser beam 31 may travel inside the chamber 2 along at least one beam path from the laser apparatus 3, be reflected by the laser beam focusing mirror 22, and strike at least one target 27, as the pulsed laser beam 33. The droplet generator 26 may output the targets 27 toward the plasma generation region 25 inside the chamber 2. The target 27 may be irradiated by at least one pulse of the pulsed laser beam 33. The target 27, which has been irradiated by the pulsed laser beam 33, may be turned into plasma, and rays of light, including EUV light 251, may be emitted from the plasma. The EUV light 251 may be reflected selectively by the EUV collector mirror 23. EUV light 252 reflected by the EUV collector mirror 23 may travel through the intermediate focus region 292 and be outputted to the exposure apparatus 6. The target 27 may be irradiated by multiple pulses included in the pulsed laser beam 33. The EUV light generation controller 5 may integrally control the EUV light generation system 11. The EUV light generation controller 5 may process image data of the droplet 27 captured by the target sensor 4. Further, the EUV light generation controller 5 may control at least one of the timing at which the target 27 is outputted and the direction into which the target 27 is outputted (e.g., the timing with which and/or direction in which the target is outputted from the droplet generator 26), for example. Furthermore, the EUV light generation controller 5 may control at least one of the timing with which the laser apparatus 3 oscillates (e.g., by controlling laser apparatus 3), the direction in which the pulsed laser beam 31 travels (e.g., by controlling laser beam direction control unit 34), and the position at which the pulsed laser beam 33 is focused (e.g., by controlling laser apparatus 3, laser beam direction control unit 34, or the like), for example. The various controls mentioned above are merely examples, and other controls may be added as necessary. 4. EUV Light Generation System Including Laser Beam Irradiation Image Detector Subsequently, an EUV light generation apparatus including a laser beam irradiation image detector for detecting an image of the laser beam passing around the target will be described with reference to the drawings. FIG. 2 schematically illustrates the configuration of an EUV light generation system 11A including a laser beam irradiation image detector 100. 4.1 Configuration As illustrated in FIG. 2, the EUV light generation system 11A may include the EUV light generation controller 5, the laser apparatus 3, the laser beam direction control unit (hereinafter, also referred to as a beam delivery unit) 34, and the chamber 2. The chamber 2 may include a main chamber 2a, into which the targets 27 are to be supplied, and a sub-chamber 2b, in which a laser beam focusing optical system 220 is disposed. The main chamber 2a and the sub-chamber 2b may be divided by a partition plate 201 having a through-hole formed at the center thereof, through which the pulsed laser beam 33 may pass. Alternatively, the main chamber 2a and the sub-chamber 2b may be separate chambers which may be integrated. However, this embodiment is not limited thereto, and the main chamber 2a and the sub-chamber 2b may be formed by dividing a single chamber into two with the partition plate 201. The laser beam focusing optical system 220 disposed inside the sub-chamber 2b may include an off-axis paraboloidal concave mirror 222 and a high-reflection mirror 223, for example. The off-axis paraboloidal concave mirror 222 may be attached to a base plate 221 through a mirror holder 222a, for example. The high-reflection mirror 223 may be attached to the base plate 221 through a two-axis tilt stage 223a (this may correspond to a laser beam focus position adjusting mechanism), for example. The base plate 221 may be movable in the Z-direction through a single-axis stage 221a (this may correspond to a laser beam focus position adjusting mechanism), for example. The high-reflection mirror 223 may have its tilt angles θx and θy adjusted through the two-axis tilt stage 223a. Here, the tilt angle θx may be a pitch angle and the tile angle θy may be a yaw angle with respect to an angle formed by a normal line at the center of the reflective surface of the high-reflection mirror 223 and the installation surface of the two-axis tilt stage 223a on the base plate 221. The pulsed laser beam 31 may be reflected by high-reflection mirrors 341 and 342 of the beam delivery unit 34 and may enter the sub-chamber 2b via the window 21. The pulsed laser beam 31 that has entered the sub-chamber 2b may be reflected by the off-axis paraboloidal concave mirror 222. With this, the pulsed laser beam 31 may be transformed into a converging pulsed laser beam 33. Thereafter, the pulsed laser beam 33 may be reflected by the high-reflection mirror 223, and may enter the main chamber 2a via a through-hole 201a. The main chamber 2a may include the EUV collector mirror 23, a target supply unit 260, the target sensor 4, and a laser beam irradiation image detector 100. The EUV collector mirror 23 may be attached to the partition plate 201 through an EUV collector mirror holder 231, for example. The through-hole 24 in the EUV collector mirror 23 and the through-hole 201a in the partition plate 201 may each be sized not to block the pulsed laser beam 33 when the pulsed laser beam 33 passes through the respective through-holes. The target supply unit 260 may include the droplet generator 26 and a two-axis stage 261 (this may correspond to a target position adjusting mechanism). The droplet generator 26 may be attached to the main chamber 2a through the two-axis stage 261. The two-axis stage 261 may be configured to move the droplet generator 26 in the Y-direction and the Z-direction, whereby the position at which the target 27 passes through the plasma generation region 25 may be adjusted. The laser beam irradiation image detector 100 may include an off-axis paraboloidal mirror 101, a beam splitter 102, an imaging lens 103, an image sensor 104, and a beam dump 105. The off-axis paraboloidal mirror 101 may be attached to the inner wall of the main chamber 2a through a support 101a, for example. The support 101a may be disposed in the obscuration region of the EUV light 252. The beam splitter 102, the imaging lens 103, the image sensor 104, and the beam dump 105 may be disposed inside a detector chamber 110, which is in communication with the main chamber 2a through a connection hole 110a, for example. The pulsed laser beam 33 that has passed through the plasma generation region 25 may be reflected by the off-axis paraboloidal mirror 101. A pulsed laser beam 253, which includes the pulsed laser beam 33 reflected by the off-axis paraboloidal mirror 101, may enter the detector chamber 110 through the connection hole 110a. Then, the pulsed laser beam 253 may pass through the beam splitter 102, and thereafter may be imaged on the photosensitive surface of the image sensor 104 through the imaging lens 103. At this point, the image sensor 104 may be in a capture mode. For example, when the image sensor 104 is provided with a shutter or the like, the shutter may be operated such that the shutter remains open for a predetermined time in synchronization with the pulsed laser beam 253 being incident on the image sensor 104. In this way, the image sensor 104 may be arranged so as to detect an image of the pulsed laser beam 253 (that is, the pulsed laser beam 33 that has passed through the plasma generation region 25). The beam splitter 102 may transmit a part of the pulsed laser beam 253 and reflect the remaining part. The transmissivity of the beam splitter 102 may be adjusted so that the amount of light incident on the image sensor 104 is retained at or below the saturation amount of light. The pulsed laser beam reflected by the beam splitter 102 may be absorbed by the beam dump 105. The EUV light generation controller 5 may include a reference clock generator 51a, an EUV light generation point controller 51, a laser beam focus control driver 52, a target controller 53, and a target supply driver 54. The EUV light generation controller 5 may integrally control the operation of the EUV light generation system 11a. Specifically, the reference clock generator 51a may generate a reference clock that may serve as a reference for various operations. The EUV light generation point controller 51 may input various signals to the laser beam focus control driver 52, the target controller 53, and the laser apparatus 3, to thereby actuate them. The laser beam focus control driver 52 may actuate the single-axis stage 221a and the two-axis tilt stage 223a of the laser beam focusing optical system 220, based on control signals from the EUV light generation point controller 51. The target controller 53 may input a control signal to the target supply driver 54, based on the control signal inputted from the EUV light generation point controller 51 and the image data inputted from the target sensor 4. The target supply driver 54 may send an output signal to the droplet generator 26 to cause the droplet generator 26 to output the targets 27, based on the control signal inputted from the target controller 53. Further, the target supply driver 54 may actuate the two-axis stage 261, based on the control signal inputted from the target controller 53. The EUV light generation point controller 51 may send an output trigger for the pulsed laser beam 31 to the laser apparatus 3. 4.2 Operation Subsequently, the operation of the EUV light generation system 11A shown in FIG. 2 will be described. The operation of the EUV light generation system 11A may be controlled by the EUV light generation controller 5. Accordingly, the operation of the EUV light generation controller 5 will be described below. The EUV light generation controller 5 may receive an EUV light generation request signal and an EUV light generation position specification signal from the exposure apparatus 6. The EUV light generation request signal may be a signal for requesting the EUV light to start being generated. The EUV light generation position specification signal may include information specifying the position inside the chamber 2 at which the EUV light is to be generated. The EUV light generation controller 5, which has received these signals, may output the output signal for the target 27 to the target supply unit 260. Then, the EUV light generation controller 5 may send the output trigger of the pulsed laser beam 31 (laser output timing) to the laser apparatus so that the target 27 is irradiated by the pulsed laser beam 33 when the target 27 arrives in the plasma generation region 25. The pulsed laser beam 31 outputted from the laser apparatus 3 may travel, as the substantially collimated pulsed laser beam 31, through the beam delivery unit 34 that includes the high-reflection mirrors 341 and 342, and may enter the chamber 2 through the window 21. The pulsed laser beam 31 may be transformed into the pulsed laser beam 33 that is to be focused in the plasma generation region 25 by the laser beam focusing optical system 220 that includes the off-axis paraboloidal concave mirror 222 and the high-reflection mirror 223. The pulsed laser beam 33 may be focused in the plasma generation region 25 in synchronization with the timing at which the target 27 passes through the plasma generation region 25. When the target 27 is irradiated by the pulsed laser beam 33, the target 27 may be turned into plasma, and the EUV light 251, including the EUV light 252, may be emitted from the plasma. Of the emitted EUV light 251, the EUV light 252 may be reflected selectively by the EUV collector mirror 23 so as to be focused in the intermediate focus (IF) region 292. The EUV light 252 that has passed through the intermediate focus region 292 may then enter the exposure apparatus 6. The pulsed laser beam 33 that has passed through the plasma generation region 25 may be reflected by the off-axis paraboloidal mirror 101. Here the off-axis paraboloidal mirror 101 may be positioned such that the pulsed laser beam 33 is incident thereon at 45 degrees. The off-axis paraboloidal mirror 101 may transform the pulsed laser beam 33 into the collimated pulsed laser beam 253. The pulsed laser beam 253 may travel through the connection hole 110a and be incident on the beam splitter 102 disposed inside the detector chamber 110. The beam splitter 102 may transmit a part of the pulsed laser beam 253 incident thereon, and reflect the remaining part. The remaining pulsed laser beam 253 reflected by the beam splitter 102 may be absorbed by the beam dump 105. The pulsed laser beam 253 that has been transmitted through the beam splitter 102 may be focused on the photosensitive surface of the image sensor 104 through the imaging lens 103. With this, the pulsed laser beam 253 (that is, the pulsed laser beam 33 that has passed through the plasma generation region 25) may be imaged on the image sensor 104. In the case where the pulsed laser beam 33 has struck the target 27, the image of the pulsed laser beam 253 may include a shadow of the target 27. The image data captured by the image sensor 104 may be sent to the EUV light generation point controller 51 of the EUV light generation controller 5. The EUV light generation point controller 51 may send control signals to the laser beam focus control driver 52 and the target supply driver 54 based on the image data. The control signal may be inputted to the target supply driver 54 through the target controller 53. With this, the laser beam focusing optical system 220 and the target supply unit 260 may be adjusted so that the pulsed laser beam 33 and the target 27 arrive at the EUV light generation position specified in the EUV light generation position specification signal. Specifically, the laser beam focus control driver 52 may send actuation signals to the two-axis tilt stage 223a for the high-reflection mirror 223 and to the single-axis stage 221a. With this, the laser beam focusing optical system 220 may be controlled so that the pulsed laser beam 33 passes through the EUV light generation position. Further, the target supply driver 54 may send an actuation signal to the two-axis stage 261. With this, the orientation of the target supply unit 260 may be controlled so that the target 27 passes through the EUV light generation position. The EUV light generation point controller 51 may send the output signal to the droplet generator 26 to cause the droplet generator 26 to output the target 27, based on the image data captured by the image sensor 104. The output signal may be inputted to the droplet generator 26 through the target controller 53 and the target supply driver 54. The EUV light generation point controller 51 may send the output trigger to the laser apparatus 3 to cause the laser apparatus 3 to output the pulsed laser beam 31, based on the image data. This may make it possible for the pulsed laser beam 33 to arrive at the EUV light generation position at substantially the same timing as the timing at which the target 27 arrives at the EUV light generation position. With the above operation being repeated, each of the targets 27 passing through the EUV light generation position may be irradiated by the pulsed laser beam 33. As a result, the EUV light generation system 11A may be controlled such that the EUV light is generated at the specified EUV light generation position. Here, the EUV light generation position may be specified by an exposure apparatus controller 61 or may be specified by another external apparatus. Alternatively, the EUV light generation position may be a fixed position determined in advance. 4.3 Effect As has been described so far, the image of the pulsed laser beam 33 that has passed through the plasma generation region 25 may be detected, the image including the shadow of the target 27. With this, both of the positional relationship between the target 27 and the pulsed laser beam when the target 27 is irradiated by the pulsed laser beam 33 and the position at which the pulsed laser beam 33 is focused can be detected directly. Further, based on this detection result, the position at which the pulsed laser beam 33 is focused and the position at which the target 27 passes through the plasma generation region 25 may be controlled. Accordingly, the EUV light generation position may be controlled with high precision. 4.4 Image when Target is Irradiated by Laser Beam FIG. 3 illustrates a positional relationship between the target 27 and the pulsed laser beam 33 when the target 27 is irradiated by the pulsed laser beam 33. FIG. 4 illustrates an image of the pulsed laser beam 253 detected by the image sensor 104 of the laser beam irradiation image detector 100. In FIG. 3, an axis Ab is the beam axis of the pulsed laser beam 33, and an axis Ao is the axis passing through the reference point O. The axis Ao may extend in the Z-direction. In FIG. 4, a center E (Xt, Yt) indicates the EUV light generation position, a center L (Xb, Yb) indicates the center (corresponding to the beam axis Ab) of an image G33 of the pulsed laser beam 253, and a center T (Xd, Yd) indicates the center of an image (shadow) G27 of the target 27. As illustrated in FIG. 3, when the target 27 is irradiated by the pulsed laser beam 33, pre-plasma 271 may be generated toward a side of the target 27 which has been irradiated by the pulsed laser beam 33, and the target material may scatter toward the opposite side, resulting in fragments 272. Further, as illustrated in FIG. 4, the image sensor 104 may capture the image G33 of the pulsed laser beam 253 and the image G27 of the target 27. The image G27 of the target 27 may include the shadow of the target 27 by the pulsed laser beam 33. Here, the posture of each of the stages for the target supply unit 260 and the laser beam focusing optical system 220 and the timing at which the target 27 is outputted may be adjusted so that the center L (Xb, Yb) of the image G33 and the center T (Xd, Yd) of the image G27 approach the EUV light generation position (the center E (Xt, Yt)), respectively. 4.5 Control Flow Subsequently, the operation of the EUV light generation system 11A shown in FIG. 2 will be described in detail with reference to the flowcharts. The operation below may be executed based on the reference clock given by the reference clock generator 51a shown in FIG. 2. In the description to follow, in order to simplify the description, the frequency of the reference clock is assumed to be substantially the same as the repetition rate of the output triggers when the timing is not adjusted. 4.5.1 Main Flow FIG. 5 shows a main flow of the operation carried out by the EUV light generation controller 5. As illustrated in FIG. 5, the EUV light generation controller 5 may first execute a parameter initialization subroutine for setting an initial value in each parameter (Step S101). Then, the EUV light generation controller 5 may execute an EUV light generation position setting subroutine for setting the EUV light generation position specified by the exposure apparatus controller 61, for example (Step S102). Subsequently, the EUV light generation controller 5 may stand by until an EUV light generation request signal for requesting the generation of the EUV light is received from the exposure apparatus 6 (more specifically, the exposure apparatus controller 61) (Step S103; NO). Upon receiving the EUV light generation request signal (Step S103; YES), the EUV light generation controller 5 may sequentially execute an EUV light generation subroutine for generating the EUV light (Step S104), a laser beam irradiation image detection subroutine for detecting an image of the pulsed laser beam 33 passing around the target 27 (Step S105), and a position determination subroutine for determining whether or not the actual EUV light generation position falls within a permissible range (Step S106). Thereafter, the EUV light generation controller 5 may determine, through the position determination subroutine (Step S106), whether or not the actual EUV light generation position falls within the permissible range, which may be either set in advance or inputted from an external apparatus such as the exposure apparatus 6 (Step S107). When the actual EUV light generation position falls within the permissible range (Step S107; YES), the EUV light generation controller 5 may send, to the exposure apparatus 6, an EUV light generation position normal signal indicating that the EUV light generation position falls within the permissible range (Step S108); and thereafter, the EUV light generation controller 5 may proceed to Step S112. In the mean time, when the actual EUV light generation position falls outside the permissible range (Step S107; NO), the EUV light generation controller 5 may send, to the exposure apparatus 6, an EUV light generation position abnormal signal indicating that the EUV light generation position does not fall within the permissible range (Step S109); and thereafter, the EUV light generation controller may proceed to Step S110. In Step S110, the EUV light generation controller 5 may execute a target position control subroutine for controlling the position and the timing at which the target 27 passes through the plasma generation region 25. Subsequently, the EUV light generation controller 5 may execute a laser beam focus position control subroutine for controlling the position and the timing at which the pulsed laser beam 33 is focused (Step S111). Through these two subroutines (Steps S110 and S111), the EUV light generation system 11A may be controlled so that the target 27 is irradiated by the pulsed laser beam 33 at the specified EUV light generation position. Thereafter, the EUV light generation controller 5 may determine whether or not this operation for controlling the EUV light generation position is to be terminated (Step S112). When the operation is to be terminated (Step S112; YES), the EUV light generation controller 5 may terminate this operation. On the other hand, when the operation is not to be terminated (Step S112; NO), the EUV light generation controller 5 may return to Step S102 and repeat the subsequent steps. 4.5.2 Parameter Initialization Subroutine The parameter initialization subroutine shown in Step S101 of FIG. 5 will be described below with reference to FIG. 6. As shown in FIG. 6, in the parameter initialization subroutine, the EUV light generation controller 5 may load an initial value E (Xt0, Yt0) for the EUV light generation position (Step S121). The initial value E (Xt0, Yt0) may be stored in a memory (not shown) or the like, for example. Subsequently, the EUV light generation controller 5 may set an initial value Dd0 in a delay time Dd of an output signal to be inputted to the droplet generator 26 with reference to the reference clock (Step S122). The initial value Dd0 may be stored in a memory (not shown) or the like, for example. Further, the EUV light generation controller 5 may set an initial value Ld0 in a delay time Ld for an output trigger for the pulsed laser beam 31 with respect to the timing at which the target 27 passes through a predetermined position (Step S123). The initial value Ld0 may be stored in a memory (not shown) or the like, for example. Here, the delay time Ld may be in an amount required for the target 27 to be irradiated by the pulsed laser beam 33 at the EUV light generation position, that is, a duration from an output of a passing signal of the target 27 from the target sensor 4 until the output of the output trigger, for example. Subsequently, the EUV light generation controller 5 may load a proportionality constant k, which may serve as a parameter when actuating various actuators for the two-axis stage 261 of the target supply unit 260, the single-axis stage 221a of the laser beam focusing optical system 220, and so forth (Step S124). The proportionality constant k may be stored in a memory (not shown) or the like, or may be given from an external apparatus, such as the exposure apparatus 6, for example. Thereafter, the EUV light generation controller 5 may load permissible ranges for the actual EUV light generation position (Step S125). Subsequently, the EUV light generation controller 5 may return to the operation shown in FIG. 5. Here, the permissible ranges may include a permissible range Ltr for the beam axis of the pulsed laser beam 33 and a permissible range Lbr for the passing position of the target 27. 4.5.3 EUV Light Generation Position Setting Subroutine The EUV light generation position setting subroutine shown in Step S102 of FIG. 5 will be described below with reference to FIG. 7. As shown in FIG. 7, in the EUV light generation position setting subroutine, the EUV light generation controller 5 may determine whether or not a resetting data ΔEs for a target EUV light generation position E has been received from the exposure apparatus 6 (Step S131). The resetting data ΔEs may be sent from the exposure apparatus controller 61 to the EUV light generation controller 5 when the EUV light generation position E requested for the EUV light generation system 11A is changed in the exposure apparatus 6, for example. Further, in this embodiment, the resetting data ΔEs is assumed to be a deviation amount (ΔXs, ΔYs) from the currently requested EUV light generation position E, but this embodiment is not limited thereto. The resetting data ΔEs may be a new EUV light generation position (coordinates). Based on the determination result in Step S131, when the resetting data ΔEs has not been received (Step S131; NO), the EUV light generation controller 5 may return to the operation shown in FIG. 5. On the other hand, when the resetting data ΔEs has been received (Step S131; YES), the EUV light generation controller 5 may load the resetting data ΔEs (ΔXs, ΔYs) (Step S132). Subsequently, the EUV light generation controller 5 may calculate a new EUV light generation position E (Xt, Yt) by adding the resetting data ΔEs (ΔXs, ΔYs) to the current EUV light generation position E (Xt, Yt) (Step S133). With this, the target EUV light generation position E may be updated. Thereafter, the EUV light generation controller 5 may return to the operation shown in FIG. 5. 4.5.4 EUV Light Generation Subroutine The EUV light generation subroutine shown in Step S104 of FIG. 5 will be described in detail with reference to FIG. 8 below. As shown in FIG. 8, in the EUV light generation subroutine, the EUV light generation controller 5 may stand by until it receives the reference clock (Step S141; NO). Upon receiving the reference clock (Step S141; YES), the EUV light generation controller 5 may reset a timer T (not shown) (Step S142). Then, the EUV light generation controller 5 may stand by until a count value T in the timer T is at or exceeds the delay time Dd (Step S143; NO). When the count value T is at or exceeds the delay time Dd (Step S143; YES), the EUV light generation controller 5 may send the output signal to the target supply unit 260 to cause the target supply unit 260 to output the target 27 (Step S144). Thereafter, the EUV light generation controller 5 may stand by until a passing signal indicating that the target 27 has passed through a predetermined position is received from the target sensor 4 (Step S145; NO). Upon receiving the passing signal (Step S145; YES), the EUV light generation controller 5 may reset the timer T (Step S146). Then, the EUV light generation controller 5 may stand by until the count value T in the timer T is at or exceeds the delay time Ld (Step S147; NO). When the count value T is at or exceeds the delay time Ld (Step S147; YES), the EUV light generation controller 5 may send an output trigger for a single pulse to the laser apparatus 3 (Step S148). Thereafter, the EUV light generation controller 5 may return to the operation shown in FIG. 5. With this, the EUV light generation system 11A may be controlled such that the pulsed laser beam 33 is focused at the EUV light generation position in synchronization with the timing at which the target 27 passes through the EUV light generation position. 4.5.5 Laser Beam Irradiation Image Detection Subroutine The laser beam irradiation image detection subroutine shown in Step S105 of FIG. 5 will now be described in detail with reference to FIG. 9. As shown in FIG. 9, in the laser beam irradiation image detection subroutine, the EUV light generation controller 5 may acquire an image data of the pulsed laser beam 253 (that is, the pulsed laser beam 33 having passed through the EUV light generation position) from the image sensor 104 of the laser beam irradiation image detector 100 (Step S151). Then, the EUV light generation controller 5 may detect the image (shadow) G27 of the target 27 and the image G33 of the pulsed laser beam 253 contained in the acquired image data (Step S152). Subsequently, the EUV light generation controller 5 may detect the center T (Xd, Yd) of the detected image (shadow) G27 and the center L (Xb, Yb) of the detected image G33, respectively (Step S153). Thereafter, the EUV light generation controller 5 may return to the operation shown in FIG. 5. 4.5.6 Position Determination Subroutine The position determination subroutine shown in Step S106 of FIG. 5 will now be described in detail with reference to FIG. 10. As shown in FIG. 10, in the position determination subroutine, the EUV light generation controller 5 may first calculate a distance Lt between the EUV light generation position E and the position (the center T (Xd, Yd), for example) of the target 27 (Step S161). The distance Lt may be obtained by calculating a difference in coordinates ΔT (ΔXd, ΔYd) of the target 27 with respect to the EUV light generation position E. The difference in coordinates ΔT (ΔXd, ΔYd) may, for example, be obtained from the target EUV light generation position E (Xt, Yt) and the position (the center T (Xd, Yd), for example) of the target 27. The calculated difference in coordinates ΔT and the calculated distance Lt may be stored in a memory (not shown) or the like, for example. Here, the deviation in the Z-direction is not taken into consideration. However, when the deviation in the Z-direction is to be taken into consideration, the size of the image G27 of the target 27 in the image data may be used. Further, the EUV light generation controller 5 may calculate a distance Lb between the EUV light generation position E and the position (the center L (Xb, Yb), for example) of the pulsed laser beam 33 (Step S162). The distance Lb may be obtained by calculating a difference in coordinates ΔL (ΔXb, ΔYb) of the pulsed laser beam 33 with respect to the EUV light generation position E. The difference in coordinates ΔL (ΔXb, ΔYb) may, for example, be obtained from the target EUV light generation position E (Xt, Yt) and the position (the center L (Xb, Yb), for example) of the pulsed laser beam 33. The calculated difference in coordinates ΔL and the calculated distance Lb may be stored in a memory (not shown) or the like, for example. Here, the deviation of the focus position in the Z-direction is not taken into consideration. However, when the deviation in the Z-direction is to be taken into consideration, the size of the image G33 of the pulsed laser beam 253 in the image data may be used. Subsequently, the EUV light generation controller 5 may determine whether or not the distances Lt and Lb fall within the permissible ranges Ltr and Lbr, respectively (Step S163). When the distances Lt and Lb fall within the permissible ranges Ltr and Lbr, respectively (Step S163; YES), the EUV light generation controller 5 may set “true” in a position normal flag provided in a memory (not shown), for example (Step S164). Thereafter, the EUV light generation controller 5 may return to the operation shown in FIG. 5. On the other hand, when the distances Lt and Lb fall outside the permissible ranges Ltr and Lbr, respectively (Step S163; NO), the EUV light generation controller 5 may set “false” in the position normal flag (Step S165). Thereafter, the EUV light generation controller 5 may return to the operation shown in FIG. 5. In Step S107 of FIG. 5, the determination may be carried out by using this position normal flag. 4.5.7 Target Position Control Subroutine The target position control subroutine shown in Step S110 of FIG. 5 will now be described in detail with reference to FIG. 11. As shown in FIG. 11, in the target position control subroutine, the EUV light generation controller 5 may load the difference in coordinates ΔT (ΔXd, ΔYd) obtained in Step S161 of FIG. 10 (Step S171). Subsequently, the EUV light generation controller 5 may adjust the delay time Dd for the output signal to cause the target supply unit 260 to output the target 27 by k·ΔXd (Dd=Dd+k·ΔXd), based on the difference in coordinates ΔT (Step S172). Then, the EUV light generation controller 5 may actuate the two-axis stage 261 of the target supply unit 260 so as to move the target supply unit 260 in the Y-direction by a Y adjustment amount ΔYd (Step S173). With this, the EUV light generation system 11A may be controlled such that the target 27 and the pulsed laser beam 33 reach the target EUV light generation position E at a predetermined timing. Thereafter, the EUV light generation controller 5 may return to the operation shown in FIG. 5. 4.5.7.1 Modification of Target Position Control Subroutine The target position control subroutine shown in Step S110 of FIG. 5 may be modified as shown in FIG. 12 as well. As shown in FIG. 12, in the modification of the target position control subroutine, the EUV light generation controller 5 may load the difference in coordinates ΔT (ΔXd, ΔYd) obtained in Step S161 of FIG. 10 (Step S175). Subsequently, the EUV light generation controller 5 may adjust the delay time Ld for the output signal to cause the laser apparatus 3 to output the pulsed laser beam 31 by k·ΔXd (Ld=Ld+k·ΔXd), based on the difference in coordinates ΔT (Step S176). Then, the EUV light generation controller 5 may actuate the two-axis stage 261 of the target supply unit 260 so as to move the target supply unit 260 in the Y-direction by the Y adjustment amount ΔYd (Step S177). In this way, controlling the output timing of the pulsed laser beam 31 so as to shift the predetermined timing may also make it possible to control the EUV light generation system 11A such that the target 27 and the pulsed laser beam 33 reach the target EUV light generation position E at a predetermined timing. Thereafter, the EUV light generation controller 5 may return to the operation shown in FIG. 5. 4.5.8 Laser Beam Focus Position Control Subroutine The laser beam focus position control subroutine shown in Step S111 of FIG. 5 will now be described in detail with reference to FIG. 13. As shown in FIG. 13, in the laser beam focus position control subroutine, the EUV light generation controller 5 may load the difference in coordinates ΔL (ΔXb, ΔYb) obtained in Step S162 of FIG. 10 (Step S181). Subsequently, the EUV light generation controller 5 may calculate angle modification amounts Δθx and Δθy of the high-reflection mirror 223 of the laser beam focusing optical system 220 in the X-direction and the Y-direction, respectively (Δθx=f(ΔXb), Δθy=f(ΔYb)), based on the difference in coordinates ΔL (Step S182). Then, the EUV light generation controller 5 may send a control signal for moving the two-axis tilt stage 223a holding the high-reflection mirror 223 by Δθx and Δθy (Step S183). With this, the EUV light generation system 11A may be controlled such that the pulsed laser beam 33 passes through the target EUV light generation position E at a predetermined timing. Thereafter, the EUV light generation controller 5 may return to the operation shown in FIG. 5. Here, when the focus position of the pulsed laser beam 33 is to be controlled, the single-axis stage 221a for the laser beam focusing optical system 220 may be moved. As has been described so far, the EUV light generation position may be controlled with high precision by controlling the focus position of the pulsed laser beam 33 and the passing position of the target 27 based on the detection result of the image of the pulsed laser beam 253 passing though the EUV light generation position. 5. EUV Light Generation System Including Image Detector for Detecting Images when Target is Irradiated by Pre-Pulse and Main Pulse Laser Beams Subsequently, an EUV light generation system 11B configured such that a target is irradiated by laser beams in multiple stages will be described in detail with reference to the drawings. FIG. 14 schematically illustrates the configuration of the EUV light generation system 11B of a multi-stage laser irradiation type. Here, the configuration similar to that of the EUV light generation system 11A shown in FIG. 2 will be referenced by similar reference characters, and duplicate description thereof will be omitted. 5.1 Configuration The EUV light generation system 11B shown in FIG. 14 may be similar in configuration to the EUV light generation system 11A shown in FIG. 2. However, the EUV light generation system 11B may differ from the EUV light generation system 11A in the following. In the EUV light generation system 11B, the laser apparatus 3 may be replaced by a laser apparatus 3B, and the beam delivery unit 34 may be replaced by a beam delivery unit 34B. The laser apparatus 3B may include a main pulse laser apparatus ML configured to output a pulsed laser beam (hereinafter, this will be referred to as a main pulse laser beam) 31 and a pre-pulse laser apparatus PL configured to output a pre-pulse laser beam 41. The beam delivery unit 34B may include a beam combiner 341B and high-reflection mirrors 342 and 343. The EUV light generation point controller 51 may be connected to each of the main pulse laser apparatus ML and the pre-pulse laser apparatus PL. The reflective surface of the high-reflection mirror 343 may be coated with a film configured to reflect the pre-pulse laser beam 41 with high reflectivity. The beam combiner 341B may be coated with a film configured to transmit the pre-pulse laser beam 41 with high transmissivity on one surface thereof on which the main pulse laser beam 31 enters the beam combiner 341B. The beam combiner 341B may also be coated with a film configured to transmit the pre-pulse laser beam 41 with high transmissivity and reflect the main pulse laser beam 31 with high reflectivity on the other surface thereof. The pre-pulse laser beam 41 outputted from the pre-pulse laser apparatus PL may be reflected by the high-reflection mirror 343. The reflected pre-pulse laser beam 41 may enter the beam combiner 341B. The main pulse laser beam 31 outputted from the main pulse laser apparatus ML may enter the beam combiner 341B through the surface opposite to the surface through which the pre-pulse laser beam 41 enters the beam combiner 341B. The beam combiner 341B may be embodied by a dichroic mirror, for example. The beam combiner 341B may be configured to reflect the main pulse laser beam 31 with high reflectivity and transmit the pre-pulse laser beam 41 with high transmissivity. The beam combiner 341B may be positioned such that the beam path of the reflected main pulse laser beam 31 coincides with the beam path of the transmitted pre-pulse laser beam 41. In this way, the beam combiner 341B may function as a beam path adjusting unit for making the beam path of the main pulse laser beam 31 coincides with the beam path of the pre-pulse laser beam 41. The pre-pulse laser beam 41 transmitted through the beam combiner 341B may then be reflected by the laser beam focusing optical system 220, to thereby be focused in the EUV light generation position as a pre-pulse laser beam 43. 5.2 Operation Subsequently, the operation of the EUV light generation system 11B shown in FIG. 14 will be described. Here, the operation of the EUV light generation system 11B may be controlled by the EUV light generation controller 5. Thus, the operation of the EUV light generation controller 5 will be described below. Upon receiving the EUV light generation request signal and the EUV light generation position specification signal from the exposure apparatus 6, the EUV light generation controller 5 may output an output signal for the target 27 to the target supply unit 260. Then, the EUV light generation controller 5 may send an output trigger for the pre-pulse laser beam 41 (laser output timing) to the pre-pulse laser apparatus PL so that the target 27 is irradiated by the pre-pulse laser beam 43 when the target 27 arrives in the plasma generation region 25. Subsequently, the EUV light generation controller 5 may send an output trigger to the main pulse laser apparatus ML (laser output timing) such that, after the target 27 is irradiated by the pre-pulsed laser beam 43 and is diffused to a certain degree, the diffused target is irradiated by the main pulse laser beam 33. Whether the target 27 is diffused to a certain degree may be determined based on whether a predetermined delay time has elapsed since the timing at which the output trigger is sent to the pre-pulse laser apparatus PL. The pre-pulse laser beam 41 may travel through the beam delivery unit 34B. Specifically, the pre-pulse laser beam 41 may be reflected by the high-reflection mirror 343 of the beam delivery unit 34B, be transmitted through the beam combiner 341B, and be reflected by the high-reflection mirror 342. Thereafter, the pre-pulse laser beam 41 may enter the chamber 2 through the window 21. The pre-pulse laser beam 41 may be transformed into the pulsed laser beam 43 that may be focused in the plasma generation region 25 by the laser beam focusing optical system 220 that includes the off-axis paraboloidal concave mirror 222 and the high-reflection mirror 223. The target 27 may be supplied to the plasma generation region 25 in synchronization with the timing at which the pre-pulse laser beam 43 passes through the plasma generation region 25. When the target 27 is irradiated by the pre-pulse laser beam 43, the target 27 may be diffused, resulting in the diffused target. The diffused target may be irradiated by the main pulse laser beam 33, whereby the target material may be turned into plasma with high efficiency. With this, an energy conversion efficiency (CE) into the EUV light may be improved. The main pulse laser beam 33 may strike the diffused target in the same direction as the pre-pulse laser beam 43, for example. The diffused target may include fine particles or the like of the target material. Thus, apart of the main pulse laser beam 33 may pass through the diffused target without striking any of the fine particles. The part of the main pulse laser beam 33 which has passed through the diffused target may be reflected by the off-axis paraboloidal mirror 101. Here, the off-axis paraboloidal mirror 101 may be disposed such that the main pulsed laser beam 33 is incident thereon at 45 degrees. At this point, the main pulse laser beam 33 may be transformed into the collimated main pulse laser beam 253. The laser beam irradiation image detector 100 may detect the image of the main pulse laser beam 253 (that is, the main pulse laser beam 33 that has passed through the diffused target). In the case where the diffused target has been irradiated by the main pulse laser beam 33, the image of the main pulse laser beam 253 may include a shadow of the diffused target. Here, the beam path of the main pulse laser beam 33 may be set to a beam path that is offset from the beam path of the pre-pulse laser beam 43, in consideration of the position at which the diffused target is generated, the distance along which the diffused target drifts after the target 27 is irradiated by the pre-pulse laser beam 43 until the diffused target is irradiated by the main pulse laser beam 33, and so forth. The EUV light generation point controller 51 may send control signals to the laser beam focus control driver 52 and the target supply driver 54, respectively. With this, the target supply unit 260 and the laser beam focusing optical system 220 may be controlled so that the diffused target is irradiated by the main pulse laser beam 33 in the EUV light generation position specified in the EUV light generation position specification signal received from the exposure apparatus controller 61. Other configuration and operation may be similar to those of the EUV light generation system 11A shown in FIG. 2. Thus, detailed description thereof will be omitted here. 5.3 Effect As has been described so far, detecting the image of the main pulse laser beam 253 (that is, the main pulse laser beam 33 that has passed through the diffused target) may make it possible to detect directly both the position at which the diffused target is irradiated by the main pulse laser beam 33 and the position at which the main pulse laser beam 33 is focused. Further, based on this detection result, the positions at which the pre-pulse laser beam 43 and the main pulse laser beam 33 are focused and the position at which the target 27 passes through the plasma generation region 25 may be controlled. Accordingly, the EUV light generation position may be controlled with high precision. 5.4 Image when Target is Irradiated by Main Pulse Laser Beam As an example of the case where the diffused target is irradiated by the main pulse laser beam, the case where fragments are irradiated by the main pulse laser beam will be described. FIG. 15 illustrates a positional relationship between the main pulse laser beam 33 and fragments 372 resulting from the target 27 being irradiated by the pre-pulse laser beam 43. FIG. 16 illustrates the image of the main pulse laser beam 253 detected by the image sensor 104 of the laser beam irradiation image detector 100. In FIG. 15, a broken line 431 indicates a plane with substantially uniform beam intensity distribution in the beam profile of the pre-pulse laser beam 43. As can be seen from the broken line 431, the pre-pulse laser beam 43 used in this embodiment may have a so-called top-hat type beam intensity distribution. Hereinafter, the pre-pulse laser beam with such beam intensity distribution will be referred to as a top-hat pre-pulse laser beam 43T. As illustrated in FIG. 15, when the target 27 is irradiated by the top-hat pre-pulse laser beam 43T, the target 27 may scatter. As a result, the fragments 372 may be generated toward the side of the target 27 opposite to the side irradiated with the top-hat pre-pulse laser beam 43T. As illustrated in FIG. 16, the fragments 372 may be formed generally in a disc-shape. When the beam intensity distribution along a beam profile is substantially uniform within a given region, as in the top-hat pre-pulse laser beam 43T, a center T (Xs, Ys) of the disc-shaped fragments 372 may substantially coincide with the center T (Xd, Yd) of the target 27 in the image detected by the image sensor 104. The rationale for this will be discussed with reference to FIGS. 17 through 19. In FIGS. 17 and 19, the case where the center line Ad that passes through the center of the target 27 and that is parallel to the beam axis Ab of the top-hat pre-pulse laser beam 43T is deviated from the beam axis Ab. Further, the target 27 is assumed to be contained in its entirety in the rays of the top-hat pre-pulse laser beam 43T. In this case, as long as the target 27 is contained in its entirety in the rays of the top-hat pre-pulse laser beam 43T, a heat input region on the surface of the target 27 may have substantially uniform heat input distribution. When the heat input condition on the surface of the target 27 is constant, the direction into which the fragments 372 scatter may be substantially parallel to the direction in which the top-hat pre-pulse laser beam 43T strikes the target 27. As a result, the center line that passes through the center of the fragments 372 and that is parallel to the beam axis Ab may substantially coincide with the center line Ad of the target 27. FIG. 17 illustrates a case where the target 27 is shifted by ΔX in the +X direction with respect to the beam axis Ab of the top-hat pre-pulse laser beam 43T. FIG. 18 illustrates a case where the beam axis Ab of the top-hat pre-pulse laser beam 43T passes through the center of the target 27. FIG. 19 illustrates a case where the target 27 is shifted by ΔX in the −X direction with respect to the beam axis Ab of the top-hat pre-pulse laser beam 43T. As illustrated in FIGS. 17 through 19, when observed in the direction of the beam axis Ab of the top-hat pre-pulse laser beam 43T, the center T (Xs, Ys) of the fragments 372 and the center T (Xd, Yd) of the target 27 may be detected to substantially coincide with each other. By analyzing the image detected by the image sensor 104, the difference in coordinates ΔL between the target EUV light generation position E (Xt, Yt) and a center Lm (Xb, Yb) of the main pulse laser beam 33 may be obtained. The center of the top-hat pre-pulse laser beam 43T (and of the main pulse laser beam 33) may be controlled based on the obtained result. Alternatively, the difference in coordinates ΔT between the target EUV light generation position E (Xt, Yt) and the center T (Xs, Ys) of the fragments 372 may be obtained, and the position of the target 27 may be controlled based on the obtained result. On the other hand, as shown by a broken line 432 in FIGS. 20 through 22, when the beam intensity distribution of a pre-pulse laser beam 43G is Gaussian, the center T (Xs, Ys) of the generated fragments 372 may change depending on the relationship between the center T (Xd, Yd) of the target 27 and a center Lp (Xb, Yb) of the pre-pulse laser beam 43G when the target 27 is irradiated by the pre-pulse laser beam 43G (the center Lp (Xb, Yb) is an intersection of the beam axis Ab and the vertical dashed line) (see FIGS. 20-22). That is, the fragments 372 may be generated in the direction into which the center T (Xd, Yd) of the target 27 is shifted with respect to the center Lp (Xb, Yb) of the pre-pulse laser beam 43G. This direction may not be parallel to the direction in which the pre-pulse laser beam 43 G strikes the target 27. FIG. 20 illustrates a case where the target 27 is shifted by ΔX in the +X direction with respect to the beam axis Ab of the pre-pulse laser beam 43G. FIG. 21 illustrates a case where the beam axis Ab of the pre-pulse laser beam 43G passes through the center of the target 27. FIG. 22 illustrates a case where the target 27 is shifted by ΔX in the −X direction with respect to the beam axis Ab of the pre-pulse laser beam 43G. When the beam intensity distribution of the pre-pulse laser beam 43G is Gaussian, the above shift amounts may preferably be considered for the difference in coordinates ΔL between the target EUV light generation position E (Xt, Yt) and the center Lm (Xb, Yb) of the main pulse laser beam 33. The position of the pre-pulse laser beam 43G (and of the main pulse laser beam 33) or the position of the target 27 may preferably be controlled based on the difference in coordinates where the shift amount is taken into consideration. 5.5 Control Flow The operation of the EUV light generation system 11B shown in FIG. 14 will now be described in detail with reference to the drawings. The operation of the EUV light generation system 11B may be similar to the operation of the EUV light generation system 11A as shown in FIGS. 5 through 13. However, the parameter initialization subroutine shown in FIG. 6 (Step S101 of FIG. 5) may be replaced by a parameter initialization subroutine shown in FIG. 23. Further, the EUV light generation subroutine shown in FIG. 8 (Step S104 of FIG. 5) may be replaced by an EUV light generation subroutine shown in FIG. 24. 5.5.1 Parameter Initialization Subroutine As shown in FIG. 23, in the parameter initialization subroutine of this embodiment, the EUV light generation controller 5 may load an initial value E (Xt0, Yt0) of the EUV light generation position (Step S221). The initial value E (Xt0, Yt0) may be stored in a memory (not shown) or the like, for example. Then, the EUV light generation controller 5 may set an initial value Dd0 in a delay time Dd of an output signal inputted to the droplet generator 26 with respect to the reference clock (Step S222). The initial value Dd0 may be stored in a memory (not shown) or the like, for example. Further, the EUV light generation controller 5 may set an initial value Ldp0 in a delay time Ldp of an output trigger for the pre-pulse laser beam 41 with respect to the timing at which the target 27 passes through a predetermined position (Step S223). Further, the EUV light generation controller 5 may set an initial value Ldm0 in a delay time Ldm of the output trigger for the main pulse laser beam 31 with respect to the timing at which the target 27 passes through the predetermined position (Step S224). These initial values Ldp0 and Ldm0 may be stored in a memory (not shown) or the like, for example. Here, the delay time Ldp may be a delay time required for the target 27 to be irradiated by the pre-pulse laser beam 43 at the EUV light generation position, the delay time being a duration from the output of the signal for detecting that the target 27 has passed a predetermined position from the target sensor 4 until the target 27 is irradiated by the pre-pulse laser beam 43, for example. Further, the delay time Ldm may be a delay time of an irradiation timing of the main pulse laser beam 33 with respect to the pre-pulse laser beam 43. Subsequently, the EUV light generation controller 5 may load a proportionality constant k serving as a parameter when actuating various actuators for the two-axis stage 261 of the target supply unit 260, the single-axis stage 221a of the laser beam focusing optical system 220, and so forth (Step S225). The proportionality constant k may be stored in a memory (not shown) or the like, or may be given from an external apparatus, such as the exposure apparatus 6, for example. Thereafter, the EUV light generation controller 5 may load the permissible ranges for the actual EUV light generation position (Step S226). Then, the EUV light generation controller 5 may return to the operation shown in FIG. 5. Here, the permissible ranges may include the permissible range Ltr for the beam axes of the main pulse laser beam 33 and of the pre-pulse laser beam 43 and a permissible range Lbr for the passing position of the target 27. 5.5.2 EUV Light Generation Subroutine As shown in FIG. 24, in the EUV light generation subroutine of this embodiment, the EUV light generation controller 5 may stand by until it receives the reference clock (Step S241; NO). Upon receiving the reference clock (Step S241; YES), the EUV light generation controller 5 may reset a timer T (not shown) (Step S242). Then, the EUV light generation controller 5 may stand by until a count value T in the timer T is at or exceeds the delay time Dd (Step S243; NO). When the count value T is at or exceeds the delay time Dd (Step S243; YES), the EUV light generation controller 5 may send the output signal to the target supply unit 260 to cause the target supply unit 260 to output the target 27 (Step S244). Thereafter, the EUV light generation controller 5 may stand by until the passing signal indicating that the target 27 has passed through a predetermined position is received from the target sensor 4 (Step S245; NO). Upon receiving the passing signal (Step S245; YES), the EUV light generation controller 5 may reset the timer T (Step S246). Then, the EUV light generation controller 5 may stand by until the count value T in the timer T is at or exceeds the delay time Ldp of the pre-pulse laser beam 41 (Step S247; NO). When the count value T is at or exceeds the delay time Ldp (Step S247; YES), the EUV light generation controller 5 may send an output trigger for a single pulse to the pre-pulse laser apparatus PL (Step S248). Then, the EUV light generation controller 5 may stand by until the count value T in the timer T is at or exceeds the delay time Ldm of the main pulse laser beam 31 (Step S249; NO). When the count value T is at or exceeds the delay time Ldm (Step S249; YES), the EUV light generation controller 5 may send an output trigger for a single pulse to the main pulse laser apparatus ML (Step S250). Thereafter, the EUV light generation controller 5 may return to the operation shown in FIG. 5. With this, the pre-pulse laser apparatus PL and the main laser apparatus ML may be controlled so that the pre-pulse laser beam 43 and the main pulse laser beam 33 are outputted sequentially in synchronization with timing at which the target 27 passes through the EUV light generation position. As has been described so far, the positions at which the pre-pulse laser beam 43 and the main pulse laser beam 33 are focused and the position at which the target 27 passes through the plasma generation region 25 may be controlled, based on the detection result of the image of the main pulse laser beam 253 passing through the EUV light generation position. Accordingly, the EUV light generation position may be controlled with high precision. 6. EUV Light Generation System in which Beam Delivery System Includes Actuator for Adjusting Focus of Laser Beam Subsequently, an EUV light generation system 11C will be described in detail with reference to the drawings. In the EUV light generation system 11C, a beam delivery unit 34C may be provided with a Z-direction laser beam focus adjusting unit 345 for controlling the focus position of the main pulse laser beam 33 and/or the pre-pulse laser beam 43. FIG. 25 schematically illustrates the configuration of the EUV light generation system 11C including the beam delivery unit 34C. In the description to follow, the configuration similar to that of the EUV light generation system 11A or 11B shown in FIG. 2 or 14 will be referenced by similar reference characters, and duplicate description thereof will be omitted. 6.1 Configuration The EUV light generation system 11C shown in FIG. 25 may be similar in configuration to the EUV light generation system 11B shown in FIG. 14. However, the EUV light generation system 11C may differ from the EUV light generation system 11B in the following. In the EUV light generation system 11C, the beam delivery unit 34B may be replaced by the beam delivery unit 34C, and the laser beam focusing optical system 220 may be replaced by a laser beam focusing optical system 220C. The beam delivery unit 34C may be similar in configuration to the beam delivery unit 34B. However, in the beam delivery unit 34C, the high-reflection mirror 342 may be held by a two-axis tilt stage 342a. Here, the high-reflection mirror 342 and the two-axis tilt stage 342a may be disposed inside the chamber 2. Further, in the beam delivery unit 34C, a top-hat mechanism 344 may be provided between the high-reflection mirror 343 and the beam combiner 341B. Alternatively, the top-hat mechanism 344 may be provided between the pre-pulse laser apparatus PL and the high-reflection mirror 343. Here, when the pre-pulse laser apparatus PL is configured to output the pre-pulse laser beam 41 having top-hat type beam intensity distribution, the top-hat mechanism 344 may be omitted. Further, in the beam delivery unit 34C, a Z-direction laser beam focus adjusting unit 345 may be provided between the beam combiner 341B and the high-reflection mirror 342. 6.2 Operation The two-axis tilt stage 342a for holding the high-reflection mirror 342 may be actuated under the control of the laser beam focus control driver 52. With this, the two-axis tilt stage 342a may function similarly to the two-axis tilt stage 223a holding the high-reflection mirror 223 in the laser beam focusing optical system 220 shown in FIG. 14. In that case, in the example shown in FIG. 25, the laser beam focusing optical system 220C may be attached to the sub-chamber 2b or to the partition plate 201. The top-hat mechanism 344 may be configured to transform the beam intensity distribution of the pre-pulse laser beam 41 into a top-hat type beam intensity distribution. The Z-direction laser beam focus point adjusting unit 345 may be configured to adjust the divergence of the main pulse laser beam 31 and of the pre-pulse laser beam 41, whereby the focus points of the main pulse laser beam 33 and of the pre-pulse laser beam 43 may be moved along the Z-direction. The laser beam focusing optical system 220C may include an off-axis paraboloidal convex mirror 224 and an off-axis paraboloidal concave mirror 225. The off-axis paraboloidal convex mirror 224 may expand the pre-pulse laser beam 41 and the main pulse laser beam 31 incident thereon in diameter. The off-axis paraboloidal concave mirror 225 may focus the pre-pulse laser beam 41 and the main pulse laser beam 31, which have been expanded in diameter by the off-axis paraboloidal convex mirror 224, at the EUV light generation position as the pre-pulse laser beam 43 and the main pulse laser beam 33, respectively. The off-axis paraboloidal convex mirror 224 and the off-axis paraboloidal concave mirror 225 may be attached onto the base plate 221 such that a laser beam is incident on the respective mirrors at approximately 45 degrees. The base plate 221 may be attached to the sub-chamber 2b or to the partition plate 201. 6.3 Effect In the EUV light generation system 11C shown in FIG. 25, the image of the main pulse laser beam 253 (that is, the main pulse laser beam 33 that has passed through the fragments 372) may be detected, whereby the position at which the fragments 372 are irradiated by the main pulse laser beam 33 and the position at which the main pulse laser beam 33 is focused may be detected directly. Further, based on this detection result, the position at which the main pulse laser beam 33 is focused and the position at which the target 27 passes through the plasma generation region 25 may be controlled. Accordingly, the EUV light generation position may be controlled with high precision. Further, the mechanisms (the two-axis tilt stage 342a and the Z-direction laser beam focus adjusting unit 345) for controlling the focus points of the main pulse laser beam 33 and of the pre-pulse laser beam 43 may be provided in the beam delivery unit 34C. This may allow the configuration of the laser beam focusing optical system 220C disposed inside the chamber 2 to be simplified. 7. Supplementary Descriptions 7.1 Two-Axis Tilt Stage Now, an example of the aforementioned two-axis tilt stages 223a and 342a will be described with reference to the drawings. FIG. 26 is a perspective view illustrating an example of the two-axis tilt stages 223a and 342a. As illustrated in FIG. 26, the two-axis tilt stage 223a or 342a may include a holder 2231 to which the high-reflection mirror 223 or 342 is attached and two automatic micrometers 2233 and 2234, for example. Mounting the holder 2231 through the automatic micrometers 2233 and 2234 may allow the tilt angle θx in the X-direction and the tilt angle θy in the Y-direction of the high-reflection mirror 223 or 342 attached to the holder 2231 to be adjusted. Here, when the Z-direction is defined as a line normal to the reflective surface of the high-reflection mirror 223 or 342, the tilt angle θx is the pitch angle that rotates about the X-axis, and the tilt angle θy is the yaw angle that rotates about the Y-axis. A commercially available product may be used for such mirror holder 2231 provided with the two-axis tilt stage. Such commercially available products include AG-M100NV6 manufactured by Newport Corporation, for example. 7.2 Focus Position Adjusting Mechanism Subsequently, an example of the aforementioned Z-direction laser beam focus adjusting unit 345 will be described with reference to FIG. 27. As illustrated in FIG. 27, the Z-direction laser beam focus adjusting unit 345 may include high-reflection mirrors 3451 and 3453 and off-axis paraboloidal concave mirrors 3454 and 3455. The high-reflection mirror 3453 and the off-axis paraboloidal concave mirror 3454 may be attached onto a stage 3452, which is movable with respect to the high-reflection mirror 3451 and the off-axis paraboloidal concave mirror 3455. Moving the stage 3452 may allow the distance between the off-axis paraboloidal mirrors 3454 and 3455 to be adjusted. With this, the wavefront of the main pulse laser beam 31 and the pre-pulse laser beam 41 incident thereon may be adjusted to a target wavefront, respectively. As a result, the divergence of the main pulse laser beam 31 and the pre-pulse laser beam 41 may be adjusted. 7.3 Modification of Focus Position Adjusting Mechanism The Z-direction laser beam focus adjusting unit 345 may be modified as shown in FIGS. 28 through 30 as well. FIGS. 28 through 30 illustrate a modification of the Z-direction laser beam focus adjusting unit 345. As illustrated in FIGS. 28 through 30, a Z-direction laser beam focus adjusting unit 345A may include a deformable mirror 3456 having a reflective surface with a curvature that may be modified, for example. The deformable mirror 3456 may reflect the collimated pulsed laser beam 31 incident thereon as a collimated pulsed laser beam, when the reflective surface thereof is adjusted to be flat, as illustrated in FIG. 28. The deformable mirror 3456, when the curvature of the reflective surface thereof is adjusted to be concave, may reflect the collimated pulsed laser beam 31 incident thereon such that the pulsed laser beam 31 is focused at a predetermined focus F12 distanced therefrom by a focal distance +F, as illustrated in FIG. 29. The deformable mirror 3456, when the curvature of the reflective surface thereof is adjusted to be convex, may reflect the collimated pulsed laser beam 31 incident thereon as a convex laser beam such that the pulsed laser beam 31 is focused at a virtual focus F13 distanced therefrom by a focal distance −F, as illustrated in FIG. 30. As has been described so far, using the deformable mirror 3456 having a reflective surface with a curvature that may be modified may make it possible to adjust the wavefront of the reflected laser beam to a predetermined wavefront in accordance with the wavefront of the incident laser beam. As a result, the divergence of the main pulse laser beam 31 and the pre-pulse laser beam 41 may be adjusted. 7.4 Top-Hat Mechanism Subsequently, the aforementioned top-hat mechanism 344 will be described in detail with reference to the drawings. FIG. 31 schematically illustrates the configuration of a top-hat mechanism 344A serving as an example of the top-hat mechanism 344. As illustrated in FIG. 31, the top-hat mechanism 344A may include a high-precision diffractive optical element (DOE) 344a. The DOE 344a may be provided with a high-precision diffraction grating either on a surface on which the pre-pulse laser beam 41 is incident or on a surface through which the pre-pulse laser beam 41 is to be outputted. The pre-pulse laser beam 41 to be outputted from the DOE 344a may be diffracted three-dimensionally. As a result, diffracted rays of the pre-pulse laser beam 41 may be combined. The combined diffracted rays may be the top-hat pre-pulse laser beam 41T having the top-hat type beam intensity distribution. The outputted top-hat pre-pulse laser beam 41T may be converted into the top-hat pre-pulse laser beam 43T through the laser beam focusing optical system 220. The top-hat pre-pulse laser beam 43T may be focused at the EUV light generation position inside the chamber 2 such that the beam intensity distribution thereof is substantially uniform at the position at which the target 27 is irradiated by the top-hat pre-pulse laser beam 43T. Here, a transmissive DOE is illustrated in FIG. 31. However, this disclosure is not limited thereto, and a reflective DOE may be used as well. 7.5 First Modification of Top-Hat Mechanism FIG. 32 schematically illustrates of the configuration of a top-hat mechanism 344B according to a first modification. As illustrated in FIG. 32, the top-hat mechanism 344B may include a phase optical element 344b. The phase optical element 344b may have a wavy surface on which the pre-pulse laser beam 41 is incident or through which the pre-pulse laser beam 41 is outputted. Accordingly, the pre-pulse laser beam 41 that has passed through the phase optical element 344b may be subjected to a phase shift in accordance with the position at which the pre-pulse laser beam 41 passes through the phase optical element 344b. Rays of the pre-pulse laser beam 41 subjected to a phase shift that may differ depending on a section of the phase shift element 344b through which the rays have passed may be converted into the top-hat pre-pulse laser beam 41T having the top-hat type beam intensity distribution. Thereafter, the top-hat pre-pulse laser beam 41T may be converted into the top-hat pre-pulse laser beam 43T through the laser beam focusing optical system 220. Here, a transmissive phase optical element is illustrated in FIG. 32. However, this disclosure is not limited thereto, and a reflective phase optical element may be used as well. 7.6 Second Modification of Top-Hat Mechanism FIG. 33 schematically illustrates of the configuration of a top-hat mechanism 344C according to a second modification. As illustrated in FIG. 33, the top-hat mechanism 344C may include a mask 344c and a collimate lens 344d. The mask 344c may be disposed such that a region of the pre-pulse laser beam 41 in which the beam intensity distribution is relatively uniform passes through the mask 344c. The collimate lens 344d may collimate the pre-pulse laser beam 41 that has been diverged after passing through the mask 344c. With such top-hat mechanism 344C, an image of the pre-pulse laser beam 41 at the mask 344c may be imaged at the EUV light generation position by the collimate lens 344d and the laser beam focusing optical system 220. The above-described embodiments and the modifications thereof are merely examples for implementing this disclosure, and this disclosure is not limited thereto. Making various modifications according to the specifications or the like is within the scope of this disclosure, and other various embodiments are possible within the scope of this disclosure. For example, the modifications illustrated for particular ones of the embodiments can be applied to other embodiments as well (including the other embodiments described herein). The terms used in this specification and the appended claims should be interpreted as “non-limiting.” For example, the terms “include” and “be included” should be interpreted as “including the stated elements but not being limited to the stated elements.” The term “have” should be interpreted as “having the stated elements but not being limited to the stated elements.” Further, the modifier “one (a/an)” should be interpreted as at least one or “one or more.” |
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claims | 1. An electron beam exposure apparatus, comprising:a first shaping aperture having a plurality of rectangular openings, each having sizes different from each other and shaping a beam shape of an electron beam;a rectangular opening selection deflector which controls a path of the electron beam to irradiate the electron beam on one of the plurality of rectangular openings;a second shaping aperture having a plurality of character openings, each having sizes different from each other and shaping a beam shape of the electron beam passing through the first shaping aperture; anda character beam deflector which controls the path of the electron beam to irradiate the electron beam on character openings corresponding to the rectangular openings in the first shaping aperture. 2. The electron beam exposure apparatus according to claim 1,wherein the second shaping aperture includes a character opening of a first size which is suitable for drawing of repetitive structure, a character opening of a second size smaller than the first size which is suitable for drawing of random structure, and a character opening for shaping a variable shaped beam (VSB). 3. The electron beam exposure apparatus according to claim 2,wherein the character beam deflector uses the character opening of the second size when a combination logic circuit and a sequential circuit are drawn, and uses the character opening of the first size when a memory cell circuit is drawn. 4. The electron beam exposure apparatus according to claim 1, further comprising a deflection controller which controls the rectangular opening selection deflector and the character beam deflector in accordance with a circuit type to be drawn. 5. The electron beam exposure apparatus according to claim 1, further comprising:an aperture information storage which stores drawing information showing a correspondence relationship between first information indicative of sizes and types of the rectangular openings in the first shaping aperture and second information relating to positions and types of the character openings in the second shaping aperture; anda drawing controller which controls the rectangular opening selection deflector and the character beam deflector based on drawing information stored in the aperture information storage. 6. The electron beam exposure apparatus according to claim 5,wherein the aperture information storage stores the drawing information including character ID information indicative of character shape to be drawn; andthe drawing controller reads out the drawing information corresponding to the character ID information and controls the rectangular opening selection deflector and the character beam deflector. 7. The electron beam exposure apparatus according to claim 5,wherein the aperture information storage has a character table indicative of a correspondence relationship of the character ID information, the first information and the second information. 8. An electron beam exposure method, comprising:irradiating an electron beam on a first shaping aperture having a plurality of rectangular openings, each having sizes different from each other and shaping a beam shape of the electron beam;controlling a path of the electron beam to irradiate the electron beam on one of the plurality of rectangular openings;irradiating the electron beam passing through the first shaping aperture on a second shaping aperture having a plurality of character openings, each having sizes different from each other and shaping a beam shape of the electron beam; andcontrolling the path of the electron beam by a character deflector to irradiate the electron beam on character openings corresponding to the rectangular openings in the first shaping aperture. 9. The electron beam exposure method according to claim 8,wherein the second shaping aperture includes a character opening of a first size which is suitable for drawing of repetitive structure, a character opening of a second size smaller than the first size which is suitable for drawing of random structure, and a character opening for shaping a variable shaped beam (VSB). 10. The electron beam exposure method according to claim 9,wherein the character beam deflector uses the character opening of the second size when a combination logic circuit and a sequential circuit are drawn, and uses the character opening of the first size when a memory cell circuit is drawn. 11. The electron beam exposure method according to claim 8,wherein the rectangular opening selection deflector and the character beam deflector are controlled in accordance with the types of the circuit to be drawn. 12. The electron beam exposure method according to claim 8, further comprising:controlling the rectangular opening selection deflector and the character beam deflector based on drawing information stored in an aperture information storage which stores the drawing information indicative of a correspondence relationship between first information indicative of sizes or types of the rectangular openings in the first shaping aperture and second information relating to positions and types of the character openings in the second shaping aperture. 13. The electron beam exposure method according to claim 12,wherein the aperture information storage stores the drawing information including character ID information indicative of character shapes to be drawn; andwhen controlling the character beam deflector, the drawing information corresponding to the character ID information is read out from the aperture information storage and controls the rectangular opening selection deflector and the character beam deflector. 14. The electron beam exposure method according to claim 12,wherein the aperture information storage has a character table indicative of a correspondence relationship of the character ID information, the first information and the second information. 15. A method of manufacturing semiconductor device, comprising:irradiating an electron beam on a first shaping aperture having a plurality of rectangular openings, each having sizes different from each other and shaping a beam shape of the electron beam;controlling a path of the electron beam to irradiate the electron beam on one of the plurality of rectangular openings;irradiating the electron beam passing through the first shaping aperture on a second shaping aperture having a plurality of character openings, each having sizes different from each other and shaping a beam shape of the electron beam;controlling the path of the electron beam by a character deflector to irradiate the electron beam on character openings corresponding to the rectangular openings in the first shaping aperture; andirradiating the electron beam passing through the character opening on a wafer to fabricate a semiconductor device. 16. The method of manufacturing semiconductor device according to claim 15,wherein the second shaping aperture includes a character opening of a first size which is suitable for drawing of repetitive structure, a character opening of a second size smaller than the first size which is suitable for drawing of random structure, and a character opening for shaping a variable shaped beam (VSB). 17. The method of manufacturing semiconductor device according to claim 16,wherein the character beam deflector uses the character opening of the second size when a combination logic circuit and a sequential circuit are drawn, and uses the character opening of the first size when a memory cell circuit is drawn. 18. The method of manufacturing semiconductor device according to claim 15,wherein the rectangular opening selection deflector and the character beam deflector are controlled in accordance with the types of the circuit to be drawn. 19. The method of manufacturing semiconductor device according to claim 15, further comprising:controlling the rectangular opening selection deflector and the character beam deflector based on drawing information stored in an aperture information storage which stores the drawing information indicative of a correspondence relationship between first information indicative of sizes or types of the rectangular openings in the first shaping aperture and second information relating to positions and types of the character openings in the second shaping aperture. 20. The method of manufacturing semiconductor device according to claim 19,wherein the aperture information storage stores the drawing information including character ID information indicative of character shapes to be drawn; andwhen controlling the character beam deflector, the drawing information corresponding to the character ID information is read out from the aperture information storage and controls the rectangular opening selection deflector and the character beam deflector. |
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description | This application is a continuation of application Ser. No. 10/743,502 filed Dec. 22, 2003 now U.S. Pat. No. 7,171,257 which claims the benefit of U.S. Provisional Application No. 60/477,573, filed Jun. 11, 2003, and U.S. Provisional Application No. 60,477,551, filed Jun. 11, 2003 which are hereby incorporated by reference. The present invention relates to creation of lesions whose positions are significant during the course of treatment, such as lesions located on the heart, or on organs close to the heart. More particularly, the invention relates to a method and system for treating cardiac-related diseases, and for creating lesions on anatomical regions that undergo motion, such as motion due to pulsating arteries. A number of medical conditions involve creating lesions whose positions are significant during the course of treatment, such as lesions that are located on the heart or on other organs close to the heart. In many cases, it is necessary to create lesions on anatomical regions that undergo rapid motion, for example motion due to pulsating arteries. Traditionally, the creation of such lesions or moving anatomical regions has required invasive surgery, such as open heart surgery for cardiac-related treatments. As one example, atrial fibrillation is a medical condition characterized by an abnormally rapid and irregular heart rhythm, because of uncoordinated contractions of the atria (i.e. the upper chambers of the heart.) A normal, steady heart rhythm typically beats 60-80 times a minute. In cases of atrial fibrillation, the rate of atrial impulses can range from 300-600 beats per minute (bpm), and the resulting ventricular heartbeat is often as high as 150 bpm or above. A curative surgical treatment for atrial fibrillation that is known in the art is the so called “maze procedure,” which is an open heart procedure involving incisions and ablations of tiny areas of the atria. The surgeon makes a plurality of incisions or lesions in the atria, so as to block the re-entry pathways that cause atrial fibrillation. Upon healing, the lesions form scar tissue, which electrically separate portions of the atria, and interrupt the conduction of the abnormal impulses. While this procedure can be effective, with a high cure rate, the procedure is long and difficult to perform. In general, possible complications of an invasive surgery are significant, and include stroke, bleeding, infection, and death. One technique for avoiding the complications of invasive surgery is radiosurgery, which is recognized as being an effective tool for noninvasive surgery. Radiosurgery involves directing radiosurgical beams onto target regions, in order to create lesions to necrotize tumorous tissue. The goal is to apply a lethal or other desired amount of radiation to one or more tumors, or to other desired anatomical regions, without damaging the surrounding healthy tissue. Radiosurgery therefore calls for an ability to accurately direct the beams upon a desired target, so as to deliver high doses of radiation in such a way as to cause only the target to receive the desired dose, while avoiding critical structures. The advantages of radiosurgery over open surgery include significantly lower cost, less pain, fewer complications, no infection risk, no general anesthesia, and shorter hospital stays, most radiosurgical treatments being outpatient procedures. In order to avoid the disadvantages of invasive surgery, such as the open heart surgical procedure described above, it is desirable to provide a method and system for using radiosurgery to treat diseases that require the creation of lesions in specifically targeted anatomical regions. These anatomical regions may be located on a beating heart wall of a patient, or on organs near the heart. Alternatively, these anatomical regions may be located in other places within the patient's anatomy that undergo motion, e.g. due to pulsating arteries. For these reasons, is desirable to provide a method and system in radiosurgery for precisely applying radiosurgical beams onto these moving anatomical regions of a patient. The present invention is directed to the radiosurgical creation of lesions whose positions are significant during the course of treatment, and to the radiosurgical treatment of anatomical regions that undergo motion. For example, these lesions and/or anatomical regions may be located on beating heart walls, or on organs near the heart, or on pulsating arteries. In accordance with one embodiment of the invention, a method is presented for treating a moving target in a patient by applying to the target one or more radiosurgical beams generated from a radiosurgical beam source. The method includes generating a pre-operative 3D scan of the target and of a region surrounding the target, the 3D scan showing the position of the target relative to the surrounding region. Based on the pre-operative 3D scan, a treatment plan is generated, which defines a plurality of radiosurgical beams appropriate for creating at least one radiosurgical lesion on one or more targets within the patient. In a preferred embodiment of the invention, the target undergoes motion. For example, the motion may be caused by heart beat and/or respiration. The movement of the target is detected and monitored. In near real time, the position of the moving target at a current time is determined, and the difference between the position of the target at the current time, as compared to the position of the target as indicated in the 3D scan, is determined. In near real time, the relative position of the radiosurgical beam source and the target is adjusted, in order to accommodate for such a difference in position. This process is repeated continuously throughout the treatment period. In one embodiment of the present invention, a composite motion (caused by respiration and heartbeat, by way of example) of the target is tracked, and one or more signals are generated that are representative of the motion of the target. For example, a breathing sensor and a heart beat monitor may be used to detect the respiration and cardiac pumping of the patient. Information from the breathing sensor and the heartbeat monitor is then combined, in order to enable the surgical x-ray source to track the position of the target as it moves due to respiration and cardiac pumping, and to generate signals representative of the position of the moving target. The signal that represents the composite motion of the target is then processed to generate two separate signals, each signal being characterized by the frequency of the individual motions that make up the composite motion. In an embodiment of the invention in which the composite motion is due to respiration combined with heart beat, the first signal is substantially characterized by the frequency (F1) of the respiratory cycle of the patient, and the second signal is substantially characterized by the frequency (F2) of the heartbeat cycle of the patient. A correction factor is then computed for each signal separately. The correction factor for the first signal is effective to compensate for the movement of the target due to respiration of the patient. The patient's respiratory motion is characterized by a respiratory cycle. The correction factor for the second signal is effective to compensate for the movement of the target due to the cardiac pumping motion in the patient. The cardiac motion of the patient is characterized by a heartbeat cycle. Both correction factors are applied to a controller that controls the position of the radiosurgical beam source, to modify the relative position of the beam source and the target, in order to account for the displacement of the target due to its composite motion. The surgical x-ray beams are applied from the modified position of the beam source in accordance with the treatment plan, so that the lesions are formed at the desired locations in the patient's anatomy. The processes of tracking the motion of the target, computing the resulting difference in target position, and adjusting the relative position of the beam source and the target accordingly, are repeated continuously throughout the treatment. In use, an observer would see the x-ray source move seemingly in synchronization with the chest wall (i.e. with the respiration), but also including short pulsating motion corresponding to the heart beat cycle. The x-ray source tracks the movement caused by both respiration and heartbeat, while delivering x-rays to the target in accordance with the treatment plan. In one form of the invention, using techniques similar to those disclosed in U.S. Pat. No. 6,501,981 (the “'981 patent”)(owned by the assignee of the present application and hereby incorporated by reference in its entirety), the motion of tissue at or near the target is determined. For example, a look-up table of positional data may be established for a succession of points along the each of the respiratory cycle and the heartbeat cycle. Motion points for the respiratory cycle include position information obtained in response to both respiration and heartbeat of the patient. Positional information for the heartbeat cycle can be obtained through imaging of the tissue while the patient is holding his breath. A table (“table 2”) containing this positional information can provide the basis for first signal. The second signal, on the other hand, can be obtained by subtracting data from the table for the heartbeat cycle (obtained by having the patient hold his breath) from the data from the composite motion (formed of both respiration and heartbeat), since the resulting table (“table 1”) corresponds to motion caused substantially only by respiration. Positional changes for the x-ray source can be applied based on superposition of data from table 1 and table 2. In the present invention, the techniques of radiosurgery are used to treat target tissue by creating radiosurgical lesions. These lesions are created in anatomical target regions located in places that undergo constant motion, such as the heart walls of a beating heart. The motion of the target, due to respiration and heart beat, is continuously tracked during treatment, so that the radiosurgical beams remain properly directed onto the desired target regions in the patient's anatomy. FIG. 1 illustrates a radiosurgical treatment system, known in the art. The radiosurgery system 100 shown in FIG. 1 may, for example, represent the CyberKnife® system developed by Accuray, Inc. In overview, the conventional radiosurgery system 100 includes a radiosurgical beaming apparatus 102; a positioning system 104; imaging means 106; and a controller 108. The system 100 may also include an operator control console and display 140. The radiosurgical beaming apparatus 102 generates, when activated, a collimated radiosurgical beam (consisting of x-rays, for example). The cumulative effect of the radiosurgical beam, when directed to the target, is to necrotize or to create a lesion in a target 118 within the patient's anatomy. By way of example, the positioning system 104 is an industrial robot, which moves in response to command signals from the controller 108. The beaming apparatus 102 may be a small x-ray linac mounted to an arm 112 of the industrial robot 104. The imaging means 106 may be an x-ray imaging system, having a pair of x-ray sources 124 and 125 for generating diagnostic imaging beams 126 and 127, and x-ray image detectors 136 and 137. In the prior art system 100, the imaging means 106 generates real-time radiographic images of the anatomical region containing the target, by transmitting one or more imaging beams through the target. The controller 108 determines the real-time location of the target, by comparing the real-time radiographic image with pre-operative CT (or MRI) scans of the target that have been stored within the computer. The positioning system 104 manipulates the position of the radiosurgical beam, in response to control commands from the controller 108, so as to keep the radiosurgical beam properly directed to the target. In order to account for the motion of a moving target, for example due to respiration of the patient, patients have typically been advised to hold their breath while being scanned by the CT scanner, prior to treatment. In this way, the moving patient is fixed, and therefore the scan does not have any motion artifacts. More recently, new radiosurgical devices, such as the CyberKnife® system, have been employing new technologies for treating moving targets. For example, Accuray recently revealed a new product, Synchrony®, which is Accuray's new system for delivering dynamic radiosurgery to tumors that move with respiration. The Synchrony® system is described in U.S. Pat. No. 6,501,981 (the “'981 patent”), entitled “Apparatus And Method For Compensating For Respiratory And Patient Motions During Treatment,” which issued on Dec. 31, 2002 to A. Schweikard and John R. Adler. The '981 patent is owned by the assignee of the present application, and is hereby incorporated by reference in its entirety. The Synchrony® system precisely tracks tumors in or near the lungs as they move, enabling highly directed beams of radiation to destroy the tumors with minimal damage to adjacent normal tissue. In particular, the Synchrony® system records the breathing movements of a patient's chest, and combines that information with sequential x-ray pictures of tiny markers inserted inside or near the tumor. In this way, the Synchrony® system enables precise delivery of radiation during any point in the respiratory cycle. In a preferred embodiment of the present invention, the motion of the target (located e.g. on the atrial walls of a beating heart) is a composite motion caused by at least two factors: a) respiratory movement of the patient; and b) rapid pulsation or pumping motion of the heart of the patient. FIGS. 2A, 2B, and 2C depict the frequency patterns of the motion of a target within the patient, caused by respiratory motion (in FIG. 2A), cardiac pumping motion (in FIG. 2B) and by a composite motion due to the combination of the respiratory motion and the cardiac pumping motion (in FIG. 2C). The target may be located on a heart wall, or on other moving regions of the patient's anatomy. The respiratory motion is characterized by a respiratory cycle, whose frequency (F1) is about an order of magnitude lower, compared to the frequency (F2) of the cardiac pumping motion. The composite or resultant motion of the target, as illustrated in FIG. 2C, is simply a superposition of the respiratory motion (shown in FIG. 2A) and the cardiac pumping motion (shown in FIG. 2B). FIG. 3A provides a schematic block diagram of a radiosurgical system 200, constructed in accordance with one embodiment of the present invention, for treating a patient by creating radiosurgical lesions in moving anatomical regions. The description of FIG. 3A will focus on cardiac-related treatments, although the scope of the present invention is not limited to cardiac-related treatments, but rather encompasses the treatment of any anatomical region that undergoes motion, for example motion due to pulsating arteries. The system 200 includes a CT scanner 212 for generating CT scan data representative of a pre-operative 3-D diagnostic image of the anatomical target, and surrounding tissue. The target is located in those areas in which the formation of lesions would be therapeutic. For example, in the case of atrial fibrillation, the target is located in those areas in which the formation of lesions would cure atrial fibrillation by properly directing electrical impulses to the AV node and onto the ventricles. Because the target undergoes motion due to respiration and cardiac pumping, a plurality of fiducials may be implanted in the atria near the target, so that the pre-operative diagnostic image shows the position of the target in reference to the fiducials. The pre-operative image is a static image, or “snapshot”, of the target and surrounding tissue. The system 200 includes a radiosurgical beam source 202 for generating one or more radiosurgical beams, preferably x-rays. The cumulative effect of applying the radiosurgical x-ray beams during the treatment period is to create at least one lesion in the target, so that the desired clinical purpose can be accomplished. In the illustrated embodiment, as in the prior art, the radiosurgical beam source 202 is a small x-ray linac. The system 200 also includes a surgical beam positioning system 204. As in the prior art, the positioning system 204 in the illustrated embodiment is an industrial robot, which moves in response to command signals from a central controller 208. The x-ray linac 202 is mounted to an arm of the industrial robot 204. It should be noted that other types of beam source 202 and positioning system 204 known in the art may be used, without departing from the scope of the present invention. The central controller 208 is preferably a multi-processor computer. The controller 208 may also include a storage unit 218 (for example, for storing the pre-operative CT scan data), and an operator console and display 240. The controller 208 preferably has a plurality of processing or controller units, including, inter alia: 1) treatment software 210 for generating, based on the CT scan data generated by the CT scanner 212, a treatment plan that defines a plurality of x-ray beams appropriate for creating one or more lesions in an anatomical target region in the heart; and 2) a controller unit 300 for sending command signals to the positioning system 204 (i.e. the robot), so as to adjust the relative position of the beam source 202 and the target. The treatment plan contains information regarding the number, intensities, positions, and directions of the x-ray beams that are effective to create at least one radiosurgical lesion. The system 200 further includes imaging means 206 for generating x-ray radiographs of the target. The imaging means 206 typically includes a pair of x-ray sources for generating x-ray imaging beams, and an x-ray imaging system. The x-ray imaging system generally includes a pair of x-ray detectors (corresponding to the pair of x-ray sources) for detecting x-rays that have passed through the target, and an image processor for generating an image of the target using the detected x-rays. In the illustrated embodiment, the system 200 further includes means for sensing the respiration of the patient and the pumping motion of the heart, and for generating a signal representative of the motion of the target due to respiration and heart beat of the patient. In the illustrated embodiment, the means for sensing the respiration is a breathing sensor 214, and the means for sensing the heart beat is a heart beat monitor 216. In other embodiments of the invention, the system 200 may include means for sensing other types of motion of the patient, for example motion due to pulsating arteries. The breathing sensor 214 may be coupled to an external body part of the patient that moves in synchronization with the respiration of the patient, and a sensor reader (not illustrated) may be provided that takes a reading from the breathing sensor periodically. A number of commercially available sensors may be used as the breathing sensor 214, including infrared tracking systems, force sensors, air flow meters, strain gauges, laser range sensors, and a variety of sensors based on physical principles such as haptic, acoustic/ultrasound, magnetic, mechanical or optical principles. In the illustrated embodiment, the system 200 includes a signal processor 220 for processing the signal representative of the composite motion (due to both breathing and heartbeat) of the target region, to generate therefrom a first signal substantially characterized by a frequency F1 representative of the respiratory motion of the patient, and a second signal substantially characterized by a frequency F2 representative of the cardiac pumping motion. Appropriate processing units in the controller 208, together with the imaging means 206, are used to generate a first correction factor from the first signal, and a second correction factor from the second signal. The first correction factor, when applied to the controller subunit 300, is effective to move the robot (and hence the x-ray source) to adjust the relative position of the x-ray source and the target, in a way that accounts for movement of the target due to respiration of the patient. The second correction factor, when applied to the controller subunit 300, is effective to correct the relative position of the x-ray source and the target, to account for movement of the target due to cardiac pumping. FIG. 3B schematically illustrates the splitting of the signal representing the composite motion of the target, into first (F1) and second (F2) signals, and the generation of the two correction factors. For example, the original signal (representing the composite motion) can be split into two signals, which are processed separately so as to eliminate a different one of the two components (F1 and F2). The processing could be done by filtering, by way of example. In a preferred embodiment, the original composite signal is treated as a signal plus out-of-band noise. The signal processor 220 may include noise canceling software for eliminating one or more undesired frequency components, i.e. out-of-band noise. For example, one or more conventional noise-canceling algorithms known in the art may be used to cancel the undesired component(s). By way of example, the noise canceling algorithms may be effective to extract the undesired component(s), and invert the extracted components. The algorithm may then generate one or more signals that cancel out the undesired frequency component(s). The first and second correction factors are recombined, and superposed, resulting in a combined correction factor. The correction factor, when applied to the controller subunit 300, accounts for the composite motion of the target due to both breathing and heart beat. FIG. 4 provides a schematic flow chart of a method in accordance with the present invention. In operation, CT scan data are generated in step 310. These data are representative of a pre-operative 3-D diagnostic image of the target. Because the target is a moving target, the diagnostic image may show the position of the target with respect to a plurality of fiducials. In the next step 320, a treatment plan is generated, based on the CT scan data generated in step 310. The treatment plan determines a succession of desired beam paths, each having an associated dose rate and duration, at each of a fixed set of locations. In step 330, the position of the moving target is determined, in near real time. Next, in step 340, the relative position of the beam source 202 and the target is adjusted to accommodate for the change in the position of the target, i.e. the difference in the position of the target (e.g. determined relative to the fiducials) at the current time, compared to the position of the target in the pre-operative CT scan. Finally, in step 350 surgical x-rays are applied to the target in accordance with the treatment plan, thereby creating one or more lesions in the desired locations. Because the target is always moving, the step of determining (in near real time) the position of the target includes the step of tracking the motion of the target. In one embodiment, the step of tracking the motion of the target includes generating at least one signal representative of the motion. In one embodiment, in order to track the motion of the target, the following steps may be taken: 1) the breathing sensor and the heart beat monitor are used to detect the respiration and the cardiac pumping of the patient, and record information relating thereto; 2) a plurality of x-ray images of the target and the implanted fiducials are generated in near real time; 3) the recorded information from the breathing sensor and the heart beat monitor are combined with the plurality of real-times x-ray images, thereby tracking the movement of the target (relative to the fiducials), as the patient breathes and the patient's heart beats. In one embodiment, the signal representative of the composite motion of the target is split into two signals, and the two signals may be separately processed through a signal processor, in order to remove undesired frequency components from each signal. A first signal, substantially characterized by a frequency F1 (representing the respiratory cycle of the patient), and a second signal substantially characterized by a frequency F2 (representing the heart beat), are generated. A first correction factor is generated from the first signal (F1), and a second correction factor is generated from the second signal (F2). In one embodiment of the invention, a look-up table of positional data may be established for a succession of points along each of the respiratory cycle and the heartbeat cycle, using techniques similar to those disclosed in the '981 patent. Motion points for the moving target include position information obtained in response to both respiration and heartbeat of the patient. Positional information for the heartbeat cycle can be obtained through imaging of the tissue while the patient is holding his breath. A table (“table 2”) containing this positional information can provide the basis for the second signal. The second signal, on the other hand, can be obtained by subtracting data from the table for the heartbeat cycle (which was obtained by having the patient hold his breath) from the data from the composite motion (formed from both respiration and heartbeat), since the resulting table (“table 1”) corresponds to motion caused substantially only by respiration. Positional changes for the x-ray source can be applied based on superposition of data from table 1 and table 2. As explained earlier, the first correction factor accounts for the breathing motion, and the second correction factor accounts for the cardiac pumping motion. As mentioned earlier, the first and second correction factors are superposed, to generate a combined correction factor that can be applied to the controller subunit 300, so that the composite motion due to both respiration and heart beat can be accounted for. In another embodiment, the step of generating the first and second correction factors may include the step of digitally comparing the plurality of near real-time x-ray images with the pre-operative CT diagnostic image. The digital comparison may be done, for example, by: 1) generating one or more DRRs (digitally reconstructed radiographs), using the pre-operative CT scan information together with the known imaging-beam positions, angles, and intensities; and 2) computing (using one or more processing units in the controller 208) the amount the target must be moved (translationally and rotationally) in order to bring the DRRs into registration with the real-time x-ray images. DRRs are artificial two-dimensional images, which show how an intermediate three-dimensional image would appear, if a hypothetical camera location and angle, as well as a hypothetical imaging beam intensity, were used. In other words, DRRs are synthetically constructed two-dimensional radiographs that are expected to result, if one or more imaging beams having a known intensity were directed to the target from certain known locations and angles. Algorithms known in the art, for example ray-tracing algorithms, are typically used to synthetically reconstruct the DRRs. In one embodiment, the step of generating the requisite corrections (for adjusting the relative position of the x-ray source and the target, in near real time) to the command signals from the subunit 300 may include: 1) extrapolating the detected motion of the target into a complete cycle; and 2) synchronizing the command signals with the extrapolated motion of the target region, so as to modify the relative positions of the beam source and the target based on the extrapolated motion information. The changes in position of the target is constantly tracked over time, throughout the treatment period. The resulting modifications in the relative positions of the beam source 202 and the target are communicated to the beam source 202 and the positioning system 204 by the controller 208. As a result, the position, direction, and intensity of the radiosurgical beams are continuously adjusted, so that an accurate radiation dose can be applied to the appropriate regions of the patient's anatomy in accordance with the treatment plan, throughout the radiosurgical treatment. The plurality of radiosurgical beams remain directed to the target, in accordance with the treatment plan, throughout the duration of the treatment, and the radiosurgical x-ray beam source tracks the movement of the target. As an improvement, instead of tracking the changes constantly over time, the system 200 can, for one component (for example, the lower frequency component F1 derived from the breathing motion), have a relatively static correction appropriate for just the “peak” of the respiratory cycle, in another embodiment of the present invention. In this embodiment, treatment by creating radiosurgical lesions may be performed only at the peaks of the respiratory cycle using the command signals modified by only the static correction factor (from breathing), and a dynamic (constantly monitored and changing) high-frequency correction factor, derived from heartbeat. While the invention has been particularly shown and described with reference to specific preferred embodiments, it should be understood by those skilled in the art that various changes in form and detail may be made therein without departing from the spirit and scope of the invention as defined by the appended claims. |
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description | This application claims the benefit of the filing of U.S. Provisional Patent Application Ser. No. 60/375,746, entitled “Preventive Technologies to End Shipping Cask Weeping”, filed on Apr. 26, 2002, and the specification thereof is incorporated herein by reference. The Government has rights to this invention pursuant to Contract No. DE-AC04-94AL85000 awarded by the U.S. Department of Energy. 1. Field of the Invention (Technical Field) The present invention relates to prevention or mitigation of spent nuclear fuel shipping cask weeping. 2. Background Art Note that the following discussion refers to a number of publications by author(s) and year of publication, and that due to recent publication dates certain publications are not to be considered as prior art vis-à-vis the present invention. Discussion of such publications herein is given for more complete background and is not to be construed as an admission that such publications are prior art for patentability determination purposes. The phenomenon termed “weeping” is characterized by the occurrence of non-fixed, removable radioactive contamination on the surface of a radioactive material package, at a level exceeding regulatory limits and after prior demonstration that such removable contamination was within allowable limits. This is a persistent problem in the transport of spent fuel from commercial reactors. This problem can be traced primarily to two radioisotopes, 137Cs and 60Co, common in spent fuel storage pools. The following lists prior art methods for the prevention or mitigation of cask surface contamination, with variable or unsubstantiated levels of success: Spent-fuel pool cleanliness (contaminant level); pool chemistry (including lower pool pH) Minimize cask immersion time (in spent-fuel pool) Operational/administrative procedures Standardization of methods for contamination measurement and instrumentation. Minimize time interval between removal from the pool and start of decontamination. Decontamination/cleaning detergents and agents (including chemical foams) Blocking agents Low pH cleaning solutions Pressurized water decontamination Cask surface finish Electro-polish cask surface Stainless steel cask (or other material) Protective coatings/barrier on cask surface “Skirt” (mechanical barrier) around cask Paint cask surface Strippable paints Cover cask during transport Cask design for ease of decontamination (surface finish and material; minimize protuberances) Minimize stress to cask body and surface/air temperature gradients 137Cs and 60Co are common radionuclides observed as non-fixed surface contamination when weeping incidents occur. A number of attempts have been made to provide a mechanistic explanation for the seemingly capricious occurrence of cask weeping incidents. Work has been performed at, or sponsored by, Sandia National Laboratories to determine the mechanism(s) of non-fixed surface contamination on package surfaces. This work led to the conclusions that the contamination is 1) an adsorption/desorption phenomenon; or is 2) a physical-chemical process in which radionuclides are incorporated (physically trapped and/or ionically bonded) onto the package surface. The reevaluation of the literature and the Sandia work by the present inventors suggested that while the phenomenological mechanisms previously conjectured were reasonable, a more likely mechanism for 137Cs weeping is ion exchange involving clay particles affixed to the exterior surfaces of shipping casks. D. Jawarani, et al., “Critical discussion of relevant physical issues surrounding the weeping of nuclear-waste casks”, J. of Nuclear Materials, v. 206, p. 57-67 (1993), focused on the role that may be played by the passivating oxide layer on a stainless steel canister, in comparison with Co and Cs retention on painted surfaces. A number of important observations were made: 1. Metal oxides retain radionuclides by a “physical adsorption” mechanism while ion exchange plays a role in the case of painted surfaces. Once the radionuclide is on the surface sorption into the oxide surface coating occurs by diffusion. 2. On painted surfaces the titanium dioxide pigment may play a role in retaining Cs. 3. There is incubation time for the nuclides to reappear on the cask surface. 4. A potential role for road grime is noted but no mechanistic function is assigned to its presence other than to suggest that it initially scavenges radionuclides that are later somehow “locked” into the metal oxide surface coating. 5. Weeping is found to be strongly associated with surface oxide coatings that have developed to the extent that well-defined grain boundaries are present. Blistering, and presumably spalling, of these oxide layers (due to accumulated stresses) may account for weeping incidents. In fact, the oxide breakdown-repassivation cycle may occur numerous times. 6. The possibility of inhibiting the uptake of radionuclides by blocking the sorption sites using Ba2+ was postulated and then verified in a preliminary manner.Although quite general in application, the Jawarani model is somewhat vague regarding the kinds of sites that may be occupied by the various radionuclides in the oxide layers. It is unlikely that surface metal-hydroxyl sites are occupied by cobalt, and particularly not by cesium, since neither element is strongly sorbed by iron oxide. One tentative explanation offered is that radionuclides may be fixed at defects in the lattices where a free electron reduces the ion to a neutral metal atom. After that, it is hypothesized, the neutral atom diffuses with relative ease deeper into the metal oxide layer. In the case of silver, and to a lesser extent cobalt, this is plausible. Cesium, however, is so readily oxidized that this mechanism seems unlikely. Jawarani et al. also make reference to the importance of various environmental factors: 1. For painted canisters, exposure to humid conditions, or rain, increases the likelihood of weeping. 2. Weathering and exposure to abrasion, dirt, sun, and air all increase the chances of a weeping incident. 3. Low temperatures and low differential temperatures are associated with increased weeping but there is no association between weeping and precipitation or humidity. Weeping incidents usually are associated with releases of 60Co and 137Cs, though on occasion other radionuclides may be involved. Generally these two radioisotopes behave differently and may be influenced by different environmental factors. Studies to date have focused on the roles played by the surface coatings, paint and metal oxide passivating coatings, but have ignored the impact of road grime. As will become apparent, this component can significantly influence the retention and later release of 137Cs. In the course of the following discussion it should be kept in mind that the various mechanisms discussed in the literature—and the mechanism emphasized herein—are not mutually exclusive. Thus, depending on local history and environment all may operate in concert and it may not be possible to isolate a single cause for a particular weeping incident. The present invention is of methods and kits for diminishing the incidents of radioactive 137CS contamination arising from the weeping phenomenon. A combination of pre-treatment (prior to placing a shipping cask into a spent fuel storage pool) and post-treatment (after it is removed) decreases residual 137Cs on the cask by a factor of 100 while also achieving smaller but still significant decreases in 60Co retention. Both pre- and post-treatments involve soaking the shipping cask surface in a solution of monovalent ions such as Cs+ (non-radioactive), K+, and NH4+. Pre-treatment works by blocking most of the sorption sites where radionuclide retention may occur. Post-treatment establishes conditions favorable to the displacement of sorbed radionuclides and thus facilitates removal of those radioisotopes that do become affixed to the cask surface as the spent fuel is being loaded. The invention is an improvement over the currently employed washing technology (often a rinse with soapy water) in that it targets specific chemical mechanisms (sorption, surface complexation, and ion exchange) responsible for retaining the chief radionuclides of concern during weeping incidents. The present invention is of a method of treating a radioactive material containing package to reduce subsequent nuclide desorption, comprising: prior to loading radioactive material into the package, contacting the package with a cation-containing solution, the cation being one or more of Cs+, Rb+, Ag+, Tl+, K+, and NH4+; and after loading radioactive material into the package, contacting the package with a cation-containing solution, the cation being one or more of Cs+, Rb+, Ag+, Tl+, K+, and NH4 +. The first contacting step preferably comprises contacting the package with a substantially non-radioactive cation-containing solution comprising Cs+, most preferably a cesium salt solution and most preferably with Cs+ in concentration greater than approximately 0.1 molar. The second contacting step preferably comprises contacting the package with a cation-containing solution, the cation being one or both of K+ and NH4+, most preferably in concentration greater than approximately 0.1 molar. The second contacting solution preferably additionally comprises one or more complexing agents. Either or both of the contacting steps may employ an absorbent wrap containing the cation-containing solution. The invention is also of a method of treating a surface to reduce subsequent nuclide desorption, the method comprising the steps of: contacting the surface with a first cation-containing solution, the cation being one or more of Cs+, Rb+, Ag+, Tl+, K+, and NH4+; and contacting the surface with a second cation-containing solution, the cation being one or more of Cs+, Rb+, Ag+, Tl+, K+, and NH4+; thereby reducing amounts of radioactive cesium embedded in clays found on the surface. The first contacting step preferably comprises contacting the surface with a substantially non-radioactive cation-containing solution comprising Cs+, most preferably a cesium salt solution and most preferably with Cs+ in concentration greater than approximately 0.1 molar. The second contacting step preferably comprises contacting the surface with a cation-containing solution, the cation being one or both of K+ and NH4+, most preferably in concentration greater than approximately 0.1 molar. The second contacting solution preferably additionally comprises one or more complexing agents. Either or both of the contacting steps may employ an absorbent wrap containing the cation-containing solution. The present invention is additionally of a method of treating a radioactive material containing package to reduce subsequent nuclide desorption, comprising: preparing a substantially non-radioactive cation-containing solution, the cation being Cs+; and contacting the package with the cation-containing solution. The contacting step may be performed prior to and/or after loading radioactive material into the package. The solution may additionally comprise one or more complexing agents. The invention is further of a kit for treating a radioactive material containing package to reduce subsequent nuclide desorption, comprising a container of a substance comprising one or more cesium salts and a container of a substance comprising one or more compounds selected from potassium compounds and ammonium compounds, which containers may be the same container or two different containers. The kit may also comprise one or more complexing agents. In each case, the one or more complexing agents are preferably one or more of ammonium fluorosilicate, oxalic acid, disodium chromotropic acid, glutamic acid, and sodium salicylate. Objects, advantages and novel features, and further scope of applicability of the present invention will be set forth in part in the detailed description to follow, taken in conjunction with the accompanying drawings, and in part will become apparent to those skilled in the art upon examination of the following, or may be learned by practice of the invention. The objects and advantages of the invention may be realized and attained by means of the instrumentalities and combinations particularly pointed out in the appended claims. A “cask weeping” incident occurs when a spent fuel shipping cask (or other radioactive material containing package) that swiped “clean” for non-fixed radionuclide contamination at the point of origin is later found to have removable contamination on its surface or to have contaminated the adjacent environment. The present inventors have discovered that particles of clay that are firmly fixed to the shipping cask surface absorb 137Cs when shipping casks are submerged in the spent fuel storage pool during spent fuel loading. Then, during shipping, ion exchange processes occur contaminating moisture on the cask surface and potentially the surrounding environment, i.e., “weeping”. Accordingly, the present invention is of pre- and post-treatment cask loading procedures and kits that significantly diminish the potential for 137Cs weeping incidents. The invention has been verified experimentally with various combinations of pre- and post-treatments lowering the level of 137Cs retained on the surfaces of metal coupons by as much as a factor of 100. Mechanisms responsible for 60Co retention are different and traditionally are ascribed to the mechanical adhesion of minute Fe—Ni rich particles arising from corrosion in reactor cooling systems. 60Co is a trace constituent in these particles and can be released to produce a weeping incident when the particles dissolve or are mechanically abraded from the cask surface. Alternatively, the stainless steel surface oxidation may have sorption sites with the potential for scavenging dissolved 60Co as well as some 137Cs from cooling pond waters. As expected, the chemical processes that reduced 137Cs retention were not as effective at mitigating 60Co retention. The phenomenon discussed herein is described by many descriptive terms. Colloquially, it is described as cask or flask “weeping” (particularly in Japan and North America), or “sweating” or “sweat-out effect” (in Europe). Regulators use phrases such as “removable external radioactive contamination” (U.S. Nuclear Regulatory Commission) or “non-fixed removable radioactive contamination” (U.S. Department of Transportation). The IAEA uses the expression “non-fixed [surface] contamination”. Phrases for the phenomenon based upon permutations of keywords including “radioactive”, “radionuclide”, “surface”, “external”, “contamination”, “excessive”, “maximum accepted”, “non-fixed”, “removable”, “reversible”, “adsorbed”, “ionic”, “ion exchange”, and even “above-norm”, et cetera are found, interchangeably, in the literature. As the mechanism for the phenomenon has become better understood, an appropriate technical term for the phenomenon describing release of previously “fixed” radionuclides from the surface of a transport package to a “non-fixed, removable” condition may be [radio]nuclide adsorption/desorption or simply nuclide desorption. Nuclide adsorption/desorption, although not consistently used herein, is suggested as a more technically rigorous expression of the phenomenon than those previously and widely used in the literature. 137Cs exists as a positive ion dissolved in the cooling pond water and most shipping cask surfaces will not be completely free of dirt when lowered into the pond. The literature on Cs sorption clearly indicates that clays are far more effective scavengers of cesium than are the metal oxides found on stainless steel surfaces. Compare Dzombak D. A. and Morel F. M. M. (1990) Surface Complexation Modeling Hydrous Ferric Oxide. John Wiley and Sons. with T. Tamura, “Sorption phenomena significant in radioactive waste disposal”, in Underground Waste Management and Environmental Implications, Vol. Memoir # 18, pp. 318-330, American Association of Petroleum Geologists, Tulsa, Okla. (1972). When a clay site acquires a positive cesium ion the process produces a local neutral electrical charge balance. Once the negatively charged layers of the clay mineral no longer repel each other they are able to come together and capture the cesium ion in a rigid cage made up of the clay mineral lattice. Thus, the release of the Cs from clays is always much slower than its uptake. However, it eventually must take place if a new fluid comes in contact with the clay that differs from the fluid that was the initial Cs source (e.g., the new fluid has less Cs and/or more of other ions, Na+, Mg2+, NH4+, etc., that can also occupy the exchange sites). Once Cs+ is released to surrounding fluids it is no longer “fixed contamination” and resulting in the phenomenon know as “weeping”. Although data is presented regarding the behavior of 60Co, the present invention primarily mitigates 137Cs releases. In the spent fuel storage ponds it is thought that 60Co occurs largely as a trace constituent in minute particles of Fe—Ni oxides (“CRUD”) derived from the corrosion of stainless steels in the reactor cooling system. Removing 60Co is a matter of freeing the particles from the cask surface or creating local reducing or acidic conditions that dissolve them. However, Jawarani, et al., supra, suggested an alternate mechanism for 60Co retention relating to ion exchange sites that exist in the oxide passivating layers formed on the surface of the stainless steel cask itself. These same mechanisms are also suggested as a means whereby Cs can be retained by the surfaces of clean stainless steel shipping casks. However, an important mechanism has been missed by prior research, namely the retention of Cs in ubiquitous clays, as shown diagramatically in FIG. 1. Clays are any of a group of hydrous aluminum silicates with layered, sheet-like structure and a very small particle size. The term “clay” is generally applied to 1) a natural material with plastic properties, 2) particles of very fine size, customarily those defined as particles smaller than two micrometers, and 3) very fine mineral fragments or particles composed mostly of hydrous-layer silicates and aluminosilicates. Kaolinite clay is illustrated in FIG. 2. Clay minerals are composed essentially of silica and alumina with variable amounts of magnesia and iron accompanying the alumina. In addition, for some clays (not kaolinite) a variety of alkaline or alkaline earth cations are typically found between the layers to maintain overall charge balance along with their various waters of hydration. This chemistry reflects the normal relative abundance of common elements on the earth's surface. Under laboratory (or industrial) conditions a host of rarer elements (such as cesium) can also be placed artificially at various sites in clay mineral lattices. The essential features of hydrous-layer silicates are continuous two-dimensional tetrahedral sheets of composition Si2O5, with SiO4 tetrahedrons linked by the sharing of three corners of each tetrahedron to form a hexagonal mesh pattern. The apical oxygen at the fourth corner of the tetrahedrons, which is usually directed normal to the sheet, forms part of an adjacent octahedral sheet in which (metal containing) octahedrons are linked by sharing edges. The junction plane between tetrahedral and octahedral sheets consists of the shared apical oxygen atoms of the tetrahedrons and unshared hydroxyls that lie at the center of each hexagonal ring of tetrahedrons and at the same level as the shared apical oxygen atoms. TABLE 1Ions Found in ClaysConstituents found inOccasionalinterlayer spacesCommon constituentsconstituentsion orradiusradiusmol-radiusion(nm)rc/roion(nm)rc/roecule(nm)rc/roO2−0.135—Ni2+0.0740.55Na+0.1010.75Si4+0.0400.30Ti4+0.0600.44K+0.1341.00Al3+0.0550.41Zn2+0.0570.42Cs+0.1631.24Fe2+0.0800.59Mn2+0.0830.61Ca2+0.1050.78Fe3+0.0670.54Mn3+0.0720.53Ba2+0.1401.03Mg2+0.0780.58Mn4+0.0520.39Sr2+0.1180.87Li+0.0760.56H2O0.145Cr3+0.0650.48NH4+0.143Cu+0.0950.70 There are two major types for the structural “backbones” of clay minerals called silicate layers. The unit silicate layer formed by aligning one octahedral sheet to one tetrahedral sheet is referred to as a 1:1 silicate layer, and the exposed surface of the octahedral sheet consists of hydroxyls (FIG. 2). In another type, the unit silicate layer consists of one octahedral sheet sandwiched by two tetrahedral sheets that are oriented in opposite directions and is termed a 2:1 silicate layer. Hydrogen bonding such as is illustrated in FIG. 1 does not exist between the layers of a 2:1 because the octahedral sheet is obscured on both sides by the (silica-rich) tetrahedral sheets. However, in 2:1 clays it is common for the layers, themselves, to be deficient in positive (metal) ions and acquire a net negative charge. To achieve charge neutrality positive ions (typically Na+, K+, Ca2+ and Mg2+) are located between the layers. Unlike metal ions in the octahedral sheets these ions are relatively accessible to the environment and may be exchanged by diffusion out of the open spaces between the layers. This accounts for the relatively high exchange capacity of 2:1 clays relative to kaolinite—which at least formally has no net negative layer charge. Referring to the kaolinite structure of FIG. 3, in descending size, the ions represented are O2, Al3+, Si4+, and H+. The structure illustrated is 0.7 nm thick from the bottom oxygen to the third oxygen layer and extends 10 nm and more in the other two directions. This three-dimensional structure is a clay “micelle”. The kaolinite mineral is made up of many micelles piled one atop the other. Cesium ions can bond between the micelles of kaolinite or within the structure of other types of clay. This structure is described as a 1:1 clay since one layer is characterized by having a predominance of Si and the other a predominance of metal cations (Al in this case). In 2:1 clay minerals—like montmorillonite and illite—there is still one metal-rich layer (now with Mg and Fe in addition to Al) but it is now sandwiched between a Si-rich layers so that both the top and bottom surfaces of the layer both look like the lower surface of the kaolin structure illustrated above. A final complication with 2:1 clays occurs when the negative charge on the layers becomes great enough that there is a high cation residency in the interlayer spaces. In this instance the attraction between the layers and the interlayer cations may also become large enough that the hydration waters normally accompanying the interlayer cations are excluded. Loss of these waters allows the layers to come closer together, after which only a small fraction of the interlayer cations have free access to the clay particle surface—and the exchange capacity is significantly reduced. An ideal formula for a 1:1 clay is provided by the mineral kaolinite, 2SiO2Al2O3 2H2O, while a typical 2:1 clay such as montmorillonite has the idealized formula: (Na, K, Mg, Ca)(Al, Mg, Fe)6(Si4O10)3(OH)6-nH2O. All types of clay minerals have been reported in soils: kaolinite dominates in regions with high rainfall and a large amount of biologic activity provides organic acids which keep soil water pH values low. In drier climates, or where weathering is less advanced, it is likely that 2:1 clay minerals will predominate. However, many exceptions to these generalities exist. What is certain is that some sort of clay will be found wherever a soil has developed, where sediments have collected, or where rocks are being actively weathered. Because of their ubiquitous distribution, clays will be a component of road grime deposited on a shipping cask at virtually any location. Ion-exchange capacity is a measure of the ability of an insoluble material to undergo displacement of ions previously attached and loosely incorporated into its structure by similarly charged ions present in the surrounding solution. Zeolite minerals used in water softening, for example, have a large capacity to exchange sodium ions (Na+) for calcium ions (Ca2+) of hard water. High cation-exchange capacities are characteristic of some clay minerals as well as numerous other natural and synthetic substances. Recognition of ion-exchange processes antedates Arrhenius, who formulated the ionic theory. In 1850, nine years before Arrhenius was born, agriculturist Sir H. S. M. Thompson and chemist J. T. Way described the phenomenon of ion exchange as it occurs in soils. Way addressed the question of how soluble fertilizers like potassium chloride were retained by soils even after heavy rains. A box with a hole in the bottom, was filled with soil, a solution of potassium chloride was poured over the soil, the liquid that flowed out of the bottom collected. The soil was then washed with rainwater and the water collected analyzed, from both the solution and the rainwater. The water turned out to contain all of the chloride that had been originally added but none of the potassium; the potassium had been replaced by chemically equivalent amounts of magnesium and calcium. The potassium could be regained by washing the soil with a solution containing a high concentration of calcium chloride (which pushed the equilibrium in the opposite direction). The process has become universally known as “ion exchange”. Soil is able to bind positive ions (like K+ and Ca2+) because in large part it contains clay. The three basic kinds of sites where metal ions may bind are discussed in roughly the order of increasing importance. The quasi-infinite array of exposed oxygen atoms on the surfaces of the clay sheets also constitutes an infinite array of negative charges with the compensating positive charges displaced to the interior of the clay mineral layers. The exposed negatively charged nodes have a slight tendency to attract hydrogen ions (or to a far lesser extent other metals) and thus the surface acquires a net charge depending on the fraction of these sites that are associated with hydrogen ions. That is, the surface charge changes with solution pH, a slight positive charge exists at low pH and a positive charge at high pH. Where the surface has a slight negative charge there is, obviously, a slight tendency for other cations to be attracted to the surface (FIG. 4). Two more numerous types of exchange sites for metal ions arise from imperfections in the clay crystal lattice. Clearly, the aluminosilicate sheets must eventually end at which point the exposed oxygens are no longer surrounded by a full complement of metal ions (FIG. 4, right K+ ion). This leaves oxygens with unsatisfied bonding capacity (in contrast to just comprising local negative charge centers as described above). This unsatisfied bonding capacity is satisfied by binding to ions from adjacent solutions. Hydrogen ion is preferred but at high pH this can be removed and other metal ions can be sorbed. In this regard it is similar to the first type of site just described. Finally, it is relevant to point out that the face and edge sites on clays are similar to those encountered on the surfaces of the oxide coatings that make up the passivating layers on stainless steel. Where the two types of sites just described predominate ion exchange processes are tied to the abundance of surface hydroxyls. An excess of surface hydroxyls will occur at low pH where there is an excess of hydrogen ions in solution. In this setting a net positive surface charge exists and anion sorption is favored. The opposite is true at high pH values where cation sorption is favored. The “point of zero charge” (PZC) is a useful concept that describes when the transition from anion sorber to cation sorber exists for different compounds. For stainless steel oxide coatings the PZC is about pH 8. So, at pH<8 (˜water in a spent fuel pool; rain [pH˜5]), the stainless steel oxide surface of a package is positive and therefore attracts anions (−), not Cs− or Co2+. Thus, if the classical description of pH 5.6 pool water (Jawarani et al., 1993) is universally correct this provides another reason why sorption should not be the mechanism responsible for loading shipping cask surfaces with either of the main radionuclides implicated in weeping. Paints, such as those that contain TiO2 have a PZC at pH˜4-5, and clays such as kaolinite have a PZC at pH˜2-4, implying that cations might be attracted to the edge sites of either a painted or clay-covered surface. Finally, it must be re-emphasized that that they would also still be attracted to the (usually more numerous) sites formed from substituting a lower-valent cation for one with a higher charge. It is also common in clays for sites normally occupied by higher charged ions to be replaced with ions of similar size (Table 1) but having a lower charge. In 2:1 clays this phenomena is prevalent and accounts for almost all of the overall ion exchange capacity of the material. Often 2:1 clays are abundant enough in soils that this type of site completely overshadows all the other kinds of ion exchange sites that are present. In addition, the phenomena are not restricted to 2:1 clays. In the example below (typical of kaolinite) a silicon atom is replaced by an aluminum atom and a charge deficiency develops that must be compensated for by including an extra cation (K+) near by (see top K+ ion, below, see also FIG. 5). Unlike the first two type's sites the third type of site is not sensitive to pH changes. If a charge deficiency arises from substituting a lower valence cation for a higher valence cation deep inside the crystal structure no single oxygen atom is greatly affected. The electrostatic field on the surface of the clay layer is, thus, relatively diffuse and the bounding of the counter-cation is only marginally selective. In particular, there is no special preference for sorbing hydrogen ions over metal ions since there is insufficient local charge balance to actually form a discrete hydroxyl unit. In this setting sorption of a particular metal depends only weakly on solution pH. This accounts for the lack of a pH dependence for sorption of metal ions on sites of this type. One final challenge is to explain why 2:1 clays are so much more important in determining the fate of 137CS than for 60Co. Partly this arises from the fact that most of the cobalt is tied up in CRUD and not in a dissolved form where it can participate in ion exchange reactions. However, there is a second factor that may be equally important since thermodynamics constraints dictate that there must always be at least some dissolved Co. The reason was alluded to earlier and is related to relatively sizes of the two ions. Hydrated cobalt ions are about the same size as the other common interlayer cations and more or less have to compete for sorption sites based on relative abundance. Cs+ ions, on the other hand, are not strongly hydrated and just happen to fit perfectly into the hexagonal cavities formed in the silica sheets facing outward on either side of the 2:1 layer surfaces. Other ions such as K+, Rb+ and NH4+ also satisfy these criteria, though not quite as well as Cs+. This makes cesium sorption an energetically favorable process. Once it is sorbed the two clay layers may approach each other and trap the Cs in the cavity (FIG. 1). Thus, while cobalt is not strongly sorbed, and can be readily washed out, the opposite is true of cesium. The present inventors believe that an ion exchange mechanism involving clays is responsible in large measure for the cesium contamination that is mobilized during transit of spent fuel packages: it is hypothesized that cesium from spent fuel storage pool water is initially sorbed by clays that are strongly attached to the package surface because of the platy nature of the clay mineral crystals. This bond should be strong enough that most of the clay, and cesium, would not be removed by normal cleaning or swiping operations; it is “fixed contamination”. Subsequently, during transport or long-term storage of the package, cations present in some aqueous solutions coming in contact with the package surface undergo ion exchange reactions with cesium ions in the clays affixed to the cask surface. This mobilizes the cesium and it becomes “non-fixed contamination” and can lead to weeping incidents. Ultimately, it is equilibrium thermodynamics which describes the end state of this process in various environments: Suppose one has a clay (X) that is in equilibrium with a fluid. The fluid contains both dissolved cesium (Cs+) and sodium (Na+). As a consequence, both ions will be sorbed onto the clay: Na<X and Cs<X. The partitioning of the two ions onto the solid can be described by the reaction:Cs++Na<XNa++Cs<X (1) For this chemical reaction, it is possible to write an equilibrium constant, where the [ ] notation represents the concentration of the various components: for example, grams of Cs per gram of clay, or grams of Cs per ml of fluid. [ Na + ] × [ Cs < X ] [ Na < X ] × [ Cs + ] = K x ( 2 ) If this mathematical identity is not satisfied when the fluid and clay initially come in contact, exchange reactions, per equation (1), will take place until the mathematical equality is satisfied. For example, assume rain falls onto a cask on which cesium is bound to clay X. [ Na + ] × [ Cs < X ] [ Na < X ] × [ 0 ] = K x ( 3 ) where [Cs+]=0 in equation (2) (no cesium in rainwater). So, to satisfy the chemical equilibrium, [Cs<X] exchanges (“weeps”) off the clay into solution ([Cs+]) with the rainwater. (There may also be an exchange of sodium.) Also, assume that a second clay (Y) is introduced into the system and is also in contact with the same solution. Although clay Y may contain no Cs initially (i.e., [Cs<Y]=0), it will ionically exchange enough Cs so that the following equality is satisfied: [ Na + ] × [ Cs < Y ] [ Na < Y ] × [ Cs + ] = K y ( 4 ) Dividing one equation by the other it is possible to eliminate the solution concentrations from consideration: [ Na < Y ] × [ Cs < X ] [ Na < X ] × [ Cs < Y ] = K x K y ( 5 ) If this second clay (or other sorbing material) is not firmly fixed to the cask surfaces, then partitioning of the Cs onto this material from both clay X and the rainwater will produce a weeping incident. The same types of equilibria would be achieved for each of the cation pairs in the system; the net effect is thus complex and difficult to predict in detail. However, the literature contains general values for the various equilibrium constants, so basic modeling of these processes, at least for montmorillonite-type clays, can be initiated (P. Fletcher, et al., “The chemical modeling of clay/electrolyte interactions for montmorillonite”, Clay Mineral 24, pp. 375-391 (1989)). To summarize, a coating of clay inherently exists on spent-fuel package surfaces. Cesium has a high affinity to clay; Cs+ is ionically bonded between layers of clay. Fixed Cs+ in the clay layers manifests as non-fixed contamination (“weeps”) via an ion exchange process when exposed to solutions containing preferential cations. Cs more readily adsorbs to cask surfaces than Co, although Co activity in pool is generally greater than Cs activity, i.e.:[Cs]cask/[Cs]pool>>[Co]cask/[Co]pool Co can effectively be removed by “mechanical” washing techniques (albeit, most “weeping” incidents are Co related). Some Cs will always be retained on cask surface, even after thorough washing of cask. Thus, there is always some potential for a Cs-related “weeping” incident. An experimental plan was developed to: 1) confirm the physical/chemical process of cesium ion exchange as it occurs on spent fuel packages—this was confirmed; and 2) select an optimum “wash” for mitigation of Cs weeping. The ion-exchange mechanism was demonstrated for Cs adsorption/removal. There was no pH effect on Cs uptake, which supports the ion-exchange hypothesis. Soiling the surface resulted in a marked increase in 137Cs retention (compared to a freshly washed—“as received” surface), but produced only a small increment in the amount of 60Co retained. A factor of >100 reduction in absorbed radioactive Cs was achieved. Optimal Cs mitigation was achieved via a Cs2SO4 (cesium sulfate) pre-wash followed with multiple water or cation post-washes. An aqueous solution is preferred to enhance ion exchange, but other solutions may be feasible. Other cations of similar size and valence that may work approximately as well as cesium include cations of rubidium, silver, and thalium, but these have significant disadvantages in terms of cost, toxicity, and/or reactivity. Additionally, other salts of cesium may be employed, such as cesium chloride, cesium nitrate, and cesium acetate. Note that although the present invention is particularly useful for casks used for transporting spent fuel, it applies equally to any radioactive material containing package as to which removal of radioactive Cs from the surface is desired and, for that matter, from any surface contaminated with radioactive Cs. Furthermore, although the invention is particularly useful as to the outside of casks, it can also be employed to remove radioactive Cs from a cask interior after transport of radioactive substances. Pre-treatment with substantially non-radioactive Cs is the most effective process for minimizing the uptake of radioactive Cs on metal surfaces and markedly decreases retention of Cs. The uptake of Cs is correlated with the presence of road grime. None of the pre-treatments had an impact on Co uptake. Co uptake involves a different mechanism and is unrelated to the presence of road grime. Post-washing removes some Co, but the amount is not related to the chemistry of the wash fluids or the presence of road grime. Preferred concentrations of Cs, K, and ammonia pre-wash solutions are greater than approximately 0.1 molar up to the saturation limit of the salt(s) being used as the source of the cation(s). Tap water, K+, and NH4+ post-wash solutions work well for Cs removal for all pre-treatment conditions. In contrast to pre-treatment, post-washing with Cs is least effective apparently due to a reduced driving force for ion exchange. This trend holds for coupons pre-treated with tap water, K+, NH4 + or with Cs+. Preferred concentrations of non-radioactive Cs post-wash solutions are as low as can be economically achieved and preferably less than approximately 0.001 molar. Preferred concentrations of NH4+ and K+ post-wash solutions are equal to or greater than approximately 0.1 molar. Tap water was the only pre-treatment that was not effective at eliminating retained 137Cs for samples that were not exposed to simulated road grime, but it was the most effective agent in diminishing retention of Co. For soiled samples, the beneficial effect of a tap-water pre-wash on Co retention seems to disappear. For soiled samples, there is a beneficial effect of a Cs pre-wash. A NH4F pre-treatment to attack clay on coupon surface was ineffective. The process for mitigating cesium retention on metal coupons described below has direct, beneficial application to the loading of spent fuel casks: the process 1) will be effective in mitigating the retention of Cs on cask surfaces, and 2) will reduce the number of weeping incidents related to Cs contamination. The following kit components are preferred to accomplish the process of the invention: 1) pre-mixed solutions of K+, NH4+ and/or non-radioactive Cs+; 2) Cs, K and ammonium sulfate salts with instruction on preparing appropriate concentrated washes; 3) a pre-made absorbent wrap containing salts or solutions of any/all of the above salts with or without the option of providing additional water or solutions that perform the function of either a pre or post rinse of the canister surface or provide the moisture needed to wet the pad and activate the components contained therein; (4) absorbent wipes that perform similarly to the wraps described above; and (5) removable absorbent gels, paints, “peels”, or other media for surface application containing the above mentioned components. Kit instructions may stipulate several steps detailing both the pretreatment of the shipping casks (or other equipment that may potentially become contaminated due to immersion in contaminated waters) prior to submersion as well as their post-treatment after immersion. Kit components employed at these two stages may be quite similar or radically different in composition and design. The invention is further illustrated by the following non-limiting examples. The objective of the testing described below was to: 1) Confirm the cesium ion exchange weeping hypothesis, namely, Cs+ fixed on a spent fuel package surface becomes non-fixed when exposed to water with ions which ionically exchange with Cs+ in clay layers; and 2) devise and demonstrate a procedure (a “wash”, pre-pool and/or post-pool) to mitigate or prevent the occurrence of unfixed Cs+ contamination on spent-fuel package surfaces. The following overall plan was followed: 1) Immerse (˜15-20) stainless steel coupons (the surrogate package surface) into water containing the radioactive tracer isotope 137Cs (the surrogate spent fuel pool). Soak as f(time) and f(pH) [f(pH) tests to demonstrate that Cs bonding is not an adsorption mechanism]). Coupons shall be of a type of stainless steel used in packages and shall have a surface finish(es) typical of packages. (The amount of Cs “bonded” to stainless steel surfaces after exposure to solution will be measured by the difference between the Cs in the original solution minus the amount of cesium remaining in the solution after exposure to the coupons.) 2) Remove stainless steel coupons and rinse with deionized H2O and “tap” water. These coupons shall be swiped and scintillation counted (both swipe and coupon) for 137Cs. (Ideally, most of the Cs will be fixed to the coupons.) 3) Wash a subset of the coupons with solutions containing candidate cations that may ionically exchange with Cs+ cations (e.g., Na+, Li+, K+, Rb+, NH3+, “rain water”, others). (“Pre-immersion washes” containing ions, possibly including non-radioactive Cs that may preferentially occupy “Cs sites” may also be applied to the coupons prior to step 1). 4) Analyze residue solutions from step 3) after washing stainless steel coupons to determine amount of 137Cs removed. (Repeated washes are expected to be most effective in removing cesium.) 5) Assess the impact of the pH of the post-immersion wash solution on the release of 137Cs and 60Co. 6) Interpret experimental results to: i) confirm the ion exchange mechanism, and ii) to describe the physical/chemical process of ion exchange as it occurs on spent fuel packages. In additional research directed at finding effective post-wash solutions, road grime (in the absence of coupons) was loaded with cesium and the soluble salts were then rinsed out. The Cs-loaded road grime was then exposed to various chemical agents and their effectiveness judged by monitoring the Cs buildup in the coexisting solution. A number of identical coupons were machined to fit conveniently into the automated gamma-ray counter that were 6.3 cm long, 0.75 cm wide and 0.25 cm thick. Prior to being cut to size the surfaces were finished to be representative of what might exist on a shipping canister (3.2 μm). Subsequent to machining all of the coupons were washed with detergent to remove machine oils and then air-dried. Half of the samples were then exposed to road grime in the following manner. The undercarriage of an automobile was lightly abraded with a wire brush and the spalled material was collected. A thick suspension of this was then prepared using deionized water. The coupons were submersed in the slurry and let stand over the weekend while the slurry dried on the coupon surfaces. Next, the coupons were rinsed under the tap while being lightly abraded with a test tube brush until they appeared to be clean. No detergent was used in cleaning the coupons during this washing. These “soiled” coupons were then air dried. Pre-treatment was carried out using 0.1 M solutions of Cs2SO4, (NH4)2SO4 and K2SO4. It was thought desirable to keep the counter-anion the same for the three pretreatment fluids. However, selection of sulfate was arbitrary and merely reflected the inventory of Cs salts on hand at the time the pretreatment was carried out. Four clean coupons and four soiled coupons were then immersed in each of the three salt solutions, and a final four of both designations (“clean” and “soiled”) were left untreated (i.e., merely washed with “tap” water prior to initiating the tests) to act as controls. Pretreatment consisted of immersing the coupon in the appropriate fluid and then thoroughly rinsing the coupon with deionized water and letting it air dry. These activities produced a 2×4 matrix of test coupons (soiled or unsoiled X four pre-treatments) that were then exposed to fluids containing both radioactive Cs and Co. Each batch of four coupons was then placed (together) in a 25 ml test tubes containing tap water spiked with 10 μCi/L 137Cs and 25 μCi/L 60Co and allowed to equilibrate for about a day. Coupons were then removed, individually cleaned of adhering droplets test fluid by a quick rinse with deionized water and dried again prior to counting. Thus, for each pretreatment four replicate measurements were made of the amount of surface contamination retained by the coupons. Average surface contamination levels for each type of pretreatment along with the standard deviation of the four measurements are given in Table 2. TABLE 2Results of initial loading after different pretreatments: Average surfacecontamination on four coupons (10−3 μCi) for each pretreatment groupand standard deviations for each group.Cation & surface conditionPre-TreatmentCs, cleanCs, soiledCo, cleanCo, soiledTap Water2.44 ± 0.2842.9 ± 5.61.77 ± 1.182.61 ± 0.410.2 M Cs+0.40 ± 0.46 4.24 ± 0.372.43 ± 0.352.78 ± 0.300.2 M K+3.21 ± 0.3811.2 ± 1.72.16 ± 0.202.72 ± 0.160.2 M NH4+2.99 ± 0.3614.4 ± 2.82.01 ± 0.182.62 ± 0.14 The next step in the processing was to investigate whether there might be a difference in the ability of various fluids to remove radioactivity that had already become affixed to coupon surfaces. In these tests, for each of the four pretreatment groups, one coupon was placed in each of (radionuclide-free) pretreatment fluids that was used to initially precondition the coupons. Samples were again allowed to equilibrate with the fluid over the weekend and then removed, cleaned, and counted in the manner described earlier. Tables 3-6 give residual surface loadings after this rinse process for respectively: residual 137Cs on unsoiled coupons, residual 137CS on soiled coupons, residual 60Co on unsoiled coupons and residual 60Co on soiled coupons. Results for each coupon are listed as follows, the residual activity prior to post-washing (upper left), the residual activity after post washing (upper right), and the percentage reduction resulting from the post-wash of that particular coupon. Parenthetically, for each row the upper left values in the boxes represents the numbers that were used in calculating the averages and standard deviations presented in Table 2. TABLE 3137Cs Retention and release by unpainted coupons that were notexposed to road grime: results are activity (10−3 μCi) remaining onthe coupon surfaces.Post-TreatmentPre-TreatmentTap Water0.2 M Cs+0.2 M K+0.2 M NH4+Tap Water2.4 to N.D.,2.6 to 1.7,2.1 to N.D.,2.7 to 1.1,~100%35%~100%59%removedremovedremovedremoved0.2 M Cs+N.D.0.72 to 0.20,N.D.0.87 to N.D.,72%~100%removedremoved0.2 M K+2.8 to 0.30,3.7 to 2.4,3.2 to 1.1,3.3 to 1.5,89%35%66%55%removedremovedremovedremoved0.2 M NH4+2.8 to 0.39,2.8 to 2.0,2.8 to 1.3,3.5 to 1.5,87%29%54% removal57% removalremovedremovedN.D. not detectable TABLE 4137Cs Retention and release by unpainted coupons that were exposed toroad grime: results are activity (10−3 μCi) remaining onthe coupon surfaces.Post-TreatmentPre-TreatmentTap Water0.2 M Cs+0.2 M K+0.2 M NH4+Tap Water47 to 17,39 to 32,37 to 18,49 to 25,64% removal18% removal51% removal49% removal0.2 M Cs+4.2 to 0.56,4.0 to 1.3,4.8 to 0.41,4.0 to 0.53,87% removal68% removal91% removal87% removal0.2 M K+12 to 3.2,13 to 8.3,8.9 to 1.9,10 to 5.3,73% removal36% removal79% removal47% removal0.2 M NH4+13 to 3.6,12 to 6.5,16 to 4.8,18 to 2.3,72% removal46% removal70% removal82% removal TABLE 560Co Retention and release by unpainted coupons that were notexposed to road grime: results are activity (10−3 μCi) remaining on thecoupon surfaces.Post-TreatmentPre-TreatmentTap Water0.2 M Cs+0.2 M K+0.2 M NH4+Tap Water2.4 to 1.7,2.5 to 1.5,2.2 to 1.1,N.D.29% removal40% removal50% removal0.2 M Cs+2.9 to 1.8,2.1 to 1.3,2.5 to 1.5,2.3 to 1.5,38% removal38% removal40% removal35% removal0.2 M K+2.3 to 1.5,2.0 to 1.3,2.4 to 1.5,1.9 to 1.3,35% removal35% removal38% removal32% removal0.2 M NH4+2.1 to 1.4,1.8 to 1.3,2.2 to 1.3,2.0 to 1.3,33% removal28% removal41% removal35% removal TABLE 660Co Retention and release by unpainted coupons that were exposedto road grime: results are activity (10−3 μCi) of remaining on thecoupon surfaces.Post-TreatmentPre-TreatmentTap Water0.2 M Cs+0.2 M K+0.2 M NH4+Tap Water2.9 to 1.7,2.8 to 1.4,2.9 to 1.2,2.0 to 1.1,39% removal50% removal59% removal45% removal0.2 M Cs+2.7 to 1.5,3.1 to 1.5,2.8 to 1.2,2.4 to 1.4,44% removal52% removal57% removal42% removal0.2 M K+2.7 to 1.7,3.0 to 1.6,2.7 to 1.1,2.5 to 1.2,37% removal47% removal59% removal52% removal0.2 M NH4+2.6 to 1.6,2.5 to 1.2,2.6 to 1.0,2.8 to 1.4,38% removal52% removal62% removal50% removal Table 2 (results of initial loading experiments) illustrates the importance of road grime as a sink for cesium uptake and that the behaviors of Cs and Co in the presence of road grime are distinctly different. The particular affinity of clays, particularly illitic clays, for cesium has been known for some time (Tamura, 1972) and X-ray diffraction studies reveal the presence of minor amounts of illite, kaolinite and chlorite (no smectite) as well as larger proportions of quartz, calcite and various feldspars. The fact that the cobalt content of soiled coupons are only marginally greater than that of unsoiled coupons is consistent with the fact that the interlayer exchange sites on clays do not have a strong affinity for this element (Fletcher et al., supra). The behavior of Co on both sites is probably ascribable to the interactions with the metal oxide surfaces, as described by Jawarani et al., supra. The loading of metal coupons that were not exposed to road grime but still retained cesium can also be ascribed to such interactions. The other feature that is clearly illustrated by Table 2 is the effectiveness of pre-exposing coupons to ions likely to block Cs sorption sites prior to placing them into the tap water spiked with 137Cs. Where clays are involved (“Cs, soiled”, column 3, Table 2) the effect is particularly noticeable, although a smaller benefit can also be obtained using surrogates for Cs with similar ionic size and charge (K+, NH4+). It is worth noting that Jawarani et al., 1993, reports a similar effect using divalent Ba2+, an ion that is almost the same size as K+. In the case of unsoiled coupons only a pre-treatment with (non-radioactive) cesium seems to be effective at inhibiting the uptake of 137Cs. It is possible that there is something about metal-oxide sites that have a unique affinity for Cs. Alternatively, even “clean” coupons would have had some exposure to dust and may have acquired a small amount of clay contamination. Because just a trace of clay may be present it is possible that only the large effect of resulting from a Cs pretreatment actually produces a detectable difference. Interpreting the impact of different post-immersion washings is complicated by the fact that coupons were loaded with very different amounts of 137Cs and 90Co. Tables 3-6 separately compare the effects of post-washing for the two radionuclides and the two surface preparations. For 137Cs on clean coupons (Table 3) it is apparent that tap water is better than any of the salt solutions, and that solutions laced with non-radioactive Cs are particularly ineffective at removing sorbed 137Cs. In the presence of Cs-spiked removal solutions the samples that were pre-washed with Cs clearly released a higher percentage of their Cs than those that received other pretreatments. This same removal pattern is evident for 137Cs on soiled samples. For 60Co the trends are less obvious but there is a slight trend for the K-spiked solutions to work best and tap water is, marginally, the poorest performer. To summarize, if the metal surface is free from road grime only small amounts of Cs are likely to be sorbed and most of this can be removed with a simple tap water rinse without pretreating with non-radioactive Cs. Soiled surfaces pick up significantly more Cs and require both a pretreatment (ideally with non-radioactive Cs) as well as a post-wash to get the activities down to the range observed for unsoiled samples. Soiled coupons without Cs pretreatment acquired an average 137Cs loading of 42.9×10−3 μCi. If a coupon is pretreated with non-radioactive Cs and post-washed with tap water the residual coupon had 0.56×10−3 μCi (Table 4). Or, if post-washed with the K+ solution, a residual activity of 0.41×10−3 μCi. This suggests a methodology for achieving overall decontamination factors of about 76 to 105 (FIG. 6). Clean coupons that had been pre-treated with Cs had, on average, less 137Cs than was obtained by the complete treatment of soiled coupons. In a number of cases the simple washing of an unsoiled sample resulted in contamination that was below our detection limit. Significantly less cobalt was removed by these approaches with decontamination factors generally between a third and two thirds. FIG. 6 shows reduction in Cs contamination on an unpainted, soiled stainless steel coupon by treatment with a Cs+ pre-wash and a K+ post-wash. The total reduction in the Cs level on the surface of the coupon with both the pre- and post-wash is approximately a factor of 105. As research progressed a question was raised regarding the effect of coupon surface preparation; in particular whether painted surfaces would respond similarly. To address this a number of coupons were painted in the appropriate manner and a smaller experimental matrix was devised for comparison (a number of combinations were deleted based on the lessons already learned on the bare metal surface coupons). In this case “clean” refers to as-received painted coupons and “soiled” refers to samples that were immersed in the same batch of road grime slurry (rewetted) as was used the first time. After pretreatment, the samples were then exposed to radionuclide doped tap water as before, then rinsed, dried and counted (Table 7). Because it was expected that the loading of grime (and hence radionuclides) might not be reproducible a few bare metal coupons were inserted as “controls” in this matrix as well, and are compared with their analogues from the previous test sequence. This proved correct and this suite of samples did not retain nearly as much grime (even on unpainted metal finish samples) as had adhered to first set of coupons. TABLE 7Comparison of initial loadings onto painted and unpainted coupons.60Co Retention137Cs RetentionFinishExposurePre-treatment(10−3 μCi)(10−3 μCi)Metal/FreshTap Water2.512.64Painted0.960.72Metal/Fresh0.1 M Cs2SO43.310.86Painted1.60N.D.Metal/SoiledTap Water2.976.94Painted1.251.68Metal/Soiled0.1 M Cs2SO43.501.94Painted0.89-1.190.53-1.24 TABLE 8Results of pre plus post treatment for 137Cs137Cs137Cs(10−3 μCi)Spl.RetainedPost-%NameFinishHistoryPre-Treatment(10−3 μCi)treatRemovedM1UMetalFreshTap Water2.62Tap0.86, 67% M2UMetalFresh0.1 M Cs2SO40.86TapN.D., ~100%M3GMetalSoiledTap Water6.94Tap1.31, 81%43.0M4GMetalSoiled0.1 M Cs2SO41.94TapN.D. ~100%P1GPaintedSoiled0.1 M K2SO43.76K1.36, 64%P2GPaintedSoiled0.1 M2.83NH40.83, 71%(NH4)2SO4P3GPaintedSoiledTap Water1.68TapN.D.,~100%P4GPaintedSoiled0.1 M Cs2SO40.74Dl0.34, 54%P5GPaintedSoiled0.1 M Cs2SO41.24TapN.D.,~100%P6GPaintedSoiled0.1 M Cs2SO40.53DlN.D.,~100%P7UPaintedFresh0.1 M Cs2SO4N.D.TapN.D.,~100%P8UPaintedFresh0.1 M K2SO4N.D.TapN.D.,~100%P9UPaintedFreshTap Water0.72TapN.D.,~100%P10UPaintedFresh0.1 M Cs2SO4N.D.TapN.D.,~100%Bold Italic results are from the previous trial and are averages of four replicate measurements. TABLE 9Results of pre and post treatment for 60Co60Co60CoSpl.RetentionPost-RetentionNameFinishHistoryPre-Treatment(10−3 μCi)Treat(10−3 μCi)M1UMetalFreshTap Water2.51TapN.D., ~100%M2UMetalFresh0.1 M Cs2SO43.31Tap1.06, 68% M3GMetalSoiledTap Water2.97Tap0.90, 69%2.61M4GMetalSoiled0.1 M Cs2SO43.50Tap1.19, 66% P1GPaintedSoiled0.1 M K2SO41.38KN.D.,~100%P2GPaintedSoiled0.1 M1.53NH4N.D.,(NH4)2SO4~100%P3GPaintedSoiledTap Water1.25TapN.D.,~100%P4GPaintedSoiled0.1 M Cs2SO40.89DlN.D.,~100%P5GPaintedSoiled0.1 M Cs2SO41.19Tap0.55, 54%P6GPaintedSoiled0.1 M Cs2SO41.04DlN.D.,~100%P7UPaintedFresh0.1 M Cs2SO41.60Tap0.41, 74%P8UPaintedFresh0.1 M K2SO41.33TapN.D.,~100%P9UPaintedFreshTap Water0.96TapN.D.,~100%P10UPaintedFresh0.1 M Cs2SO41.38TapN.D.,~100%Bold Italic results are from the previous trial and are averages of four replicate measurements. For the unpainted coupons the trends observed are similar to those noted in previous tests except, in this case the loading of road grime was somewhat less so the difference in 137Cs loading between soiled and unsoiled coupons was smaller. For the painted samples: 1. In all cases painting had a beneficial effect in cutting down on the amounts of both 60Co and 137Cs that were retained on the surfaces during the initial loading process (Table 7). 2. Again, soiling the surface resulted in a significant increase in Cs (but not Co) retention relative to samples that had not been exposed to simulated road grime. 3. Tap water was the only pretreatment that was not effective at eliminating retained 137Cs for samples that were not exposed to simulated road grime. However, it was (marginally) the most effective agent in diminishing retention of 60Co. For soiled samples the beneficial impact of a tap-water wash seems to disappear for preventing the uptake of cobalt. 4. For soiled samples with painted surfaces the beneficial effect of a Cs pre-wash relative to a tap water pre-wash is again visible but not as large as with the previous suite of samples. 5. Treatment with ammonium or potassium may have actually made things worse for painted—soiled samples, possibly because they saturated the exchange sites with something that cesium could readily displace. A very slight tendency toward this trend was apparent for the metal-finish samples in the first suite of tests but not for the samples exposed to road grime. However, in the current testing the amount (and impact) of road grime on the samples is smaller so the effect of the coatings placed on the coupon surfaces may be playing a larger relative role in overall responses. 6. For final Cs removal tap water and deionized water performed better than either ammonium or potassium solutions on both clean and soiled painted surfaces. For cobalt removal only tap water left enough contamination so that a signal was still detectable. In a regional sense it is likely that the pH of the spent fuel storage ponds may vary from slightly basic to slightly acidic, though the best-documented pH is about 5.6 (Jawarani et al., supra). Consequently a small experimental matrix (Tables 10 and 11) was carried out where the coupons were pre and post treated with normal tap water but the pH of the Cs and Co loading solutions was adjusted downward to a pH near 6 and upward to 8 (initial loading experiments, Tables 2-6 had a pH of 7.3). With regard to pre and post-treatment these samples compare to results in the upper left entry of Tables 3-6 and were loaded at the same time that these coupons were (i.e., not with the later suite of samples that were loaded for the painted and unpainted comparison in the preceding section). TABLE 10Cobalt uptake from solutions with a pH of about 6 and about 8 andeffect of subsequent soaking in tap water after being loaded.(All coupons were pre-treated with tap water)Exposure SolutionAs MachinedSoiled by Road GrimepH 63.85 to 2.193.04 to 1.96pH 63.12 to 2.502.23 to 1.73pH 847.3 to 41.425.3 to 17.7 pH 8.42.9 to 36.923.0 to 16.1 TABLE 11Cesium uptake from solutions with a pH of about 6 and about 8 andeffect of subsequent soaking in tap water after being loaded.(All coupons were pre-treated with tap water)Exposure SolutionAs MachinedSoiled by Road GrimepH 62.96 to 0.4040.1 to 13.6pH 62.43 to not detectable34.7 to 12.4pH 83.09 to not detectable36.4 to 12.3 pH 8.2.89 to not detectable31.6 to 10.0 The pH of the 137Cs treating solution had no obvious effect on the uptake of 137Cs but the treatment with road grime clearly did. Soaking the loaded coupons in tap water removed almost all of the Cs that was sorbed by clean metal surfaces and about two thirds of the cesium that was sorbed by surfaces soiled by road grime. The largest factor in governing the uptake of 60Co is the pH of the loading solution. Possible explanations include (a) the (unlikely) formation of particulate Co(OH)2 particles or (b) the fact that the Co sorbing sites are more receptive to cation sorption at high pH values (as would be expected of sorption sites on the surfaces of metal oxides). It also appears that exposure to road grime blocked about 40% of these sites. Soaking the loaded coupons in tap water left a significant fraction of the cobalt that had loaded onto the surfaces without regard to whether the surfaces were clean or soiled or whether the surfaces had been loaded at high or low pH. Earlier studies had shown that washing a shipping cask after immersion in contaminated water was likely to provide a higher level of overall decontamination than was achievable by simply blocking the sorption sites prior to immersion. These studies also suggested that reversing the ion exchange process to displace Cs using either potassium or ammonium were not significantly more effective than just washing the contaminated surface with low ionic strength solutions such as tap water. The underlying cause is, of course, that during sorption the clay layers come together to encapsulate the Cs. This prevents it from having ready access to fluids on the surface of the clay particle. Reversing the ion exchange is, thus, generally a slow and inefficient process. To get around this difficulty a number of chemical washes were selected with the potential for simply dissolving at least the outer layer of the clay particles where the balance of the sorbed Cs should reside. These reagents were selected based on their ability to complex aluminum or silicon, the principal elements making up the backbone of the clay structure. The road grime needed for these tests was loaded with Cs by exposing 8.077 grams of material to 101.7 grams of a solution containing 10 parts per thousand-dissolved (non-radioactive) Cs for a period of 15 days at room temperature. After the loading period the fluid was repeatedly decanted off and replaced with fresh batches of deionized water for a period of six weeks. Finally, the Cs-loaded road grime was allowed to equilibrate with the final wash solution for another six-week period. At the conclusion of the equilibration period the water contained about 20 ppm Cs, 0.02 ppm Al and 6.5 ppm Si, and the bottle contained about 400 ml of fluid. Stock solutions of each of the complexing agents were prepared by dissolving them in deionized water. In some cases not all of the salt dissolved, in which case the concentration in the stock solution (Table 12) is simply listed as “saturated”. TABLE 12Concentrations of complexing agents used in Cs release studiesAgentGrams WaterGrams Agent(NH4)2SiF627 4.96Sodium Salicylate19.65.01Salicylic AcidsaturatedsaturatedGallic AcidsaturatedsaturatedOxalic Acid35.75.00Catechol35.85.038-HydroxyquinolinesaturatedsaturatedDi-Sodium EDTAsaturatedsaturatedDi-Sodium Chromotropic Acid17.91.22Glutamic AcidsaturatedsaturatedSodium Isosaccharinic Acid17.91.00 The bottle containing the Cs-loaded road grime was then shaken vigorously to suspend the solids and 18 ml of suspension was pipetted into each of 11 polyethylene bottles. The process went quickly and the vigorous shaking was maintained until all the bottles were loaded. Consequently it is reasonable to assume that all of the bottles contained about the same mass of road grime as well as the same volume of fluid. Next, one ml of the various complexing agent stock solutions was added, individually, into one of the slurry-filled bottles. The bottles sat at room temperature for 7 days and then an aliquot of fluid was removed, passed through a 0.2-micron filter and analyzed for Al, Si and Cs. In addition, reagent blanks were run to see if the complexing agents themselves could introduce Cs that might be mistaken for Cs that had been leached from the road grime. However, the reagent blanks were all clean so any increment in Cs in the test fluid must reflect Cs released from the road grime. In addition to giving the Al, Si and Cs concentrations for each leach test solution the far right hand column in Table 13 is a metric that highlights the best performers (Cs-f is final Cs concentration and Cs-i is that concentration of the initial Cs in the fluid, 20 ppm). It is evident that a number of agents successfully scavenged Cs from the clays in the road grime. In order of decreasing efficiency they are: (NH4)2SiF6 (ammonium fluorosilicate)>oxalic acid>disodium chromotropic acid>glutamic acid>sodium salicylate. Since these samples had already equilibrated for a considerable period of time with fresh water it is likely that any of these solutions would be superior to the post-wash fluids that were previously evaluated. It is also noteworthy that it should be possible to use ammonium fluorosilicate and oxalic acid together since the former specifically targets the silica in the clay structure while the latter is effective only against metals such as aluminum, iron and magnesium (and potentially cobalt as well). TABLE 13Leaching Test ResultsAgentAl ppmCs ppmSi ppm(Cs − f/Cs − i) − 1(NH4)2SiF66.5734.816720.736Sodium Salicylate1.1122.98.290.141Salicylic Acid0.01720.716.250.031Gallic Acid0.05221.610.30.079Oxalic Acid39.830.873.80.539Catechol1.7921.212.50.0608-Hydroxyquinoline0.0319.97.07−0.009Di-Sodium EDTA3.7917.812.19−0.113Di-Sodium1.7425.99.410.291Chromotropic AcidGlutamic Acid0.01124.36.700.212Sodium0.26620.87.200.039Isosaccharinic AcidStarting Fluid0.01920.06.480 The preceding examples can be repeated with similar success by substituting the generically or specifically described reactants and/or operating conditions of this invention for those used in the preceding examples. A main thrust of the present invention is to block Cs-sorbing sites prior to exposing metal or painted surfaces to contaminated waters. However, an alternate approach attacks the aluminosilicate minerals that scavenge the Cs prior to placing the surface in contact with a source of contamination. Table 13 indicates that several reagents have this ability since at the conclusion of the tests dissolved levels of Al, Si and occasionally both were above those in the starting solutions. Thus, in addition to prewashing with monovalent cations it may be advantageous to also use chemicals such as are tabulated in Table 13 as part of the pre-wash procedure, either separately or mixed together with the monovalent cation containing solutions. Although the invention has been described in detail with particular reference to these preferred embodiments, other embodiments can achieve the same results. Variations and modifications of the present invention will be obvious to those skilled in the art and it is intended to cover in the appended claims all such modifications and equivalents. The entire disclosures of all references, applications, patents, and publications cited above are hereby incorporated by reference. |
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abstract | A device for the dry handling of nuclear fuel assemblies is provided. The device includes a transfer basket which can be connected to a lifter and which includes a gripper for gripping the fuel assembly to be transferred, the gripper being supported by a lift built into the basket; and an indexing table which can be placed on a cask and which comprises a positioner for positioning the basket over a slot in the cask. |
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summary | ||
abstract | A fuel assembly 20 is constituted by combining a plurality of fuel rods 21, holding the fuel rods 21 by a plurality of support grids 22, and arranging a lower nozzle 24 and an upper nozzle 23 at opposite ends of the fuel rods 21. A shock-absorbing device 10 for a fuel assembly is fitted to the lower nozzle 24 and the upper nozzle 23. The shock-absorbing device 10 for a fuel assembly is constituted a nozzle support 12 fitted to a depression 24U of the lower nozzle 24 and a depression 23U of the upper nozzle 23, and a buffer 11 combined with the nozzle support 12, with stiffness thereof in a longitudinal direction of the fuel rods 21 being equal to or lower than that of the nozzle support 12. |
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claims | 1. An industrial device for receiving a maintenance operation by remote control from an operation device and being connected to a first management device, said operation device and a network, the industrial device comprising:communication means for periodically monitoring said industrial device for a maintenance request, transmitting the maintenance request received from said industrial device to said operation device via said first management device and a second management device, said second management device transmitting said maintenance request to said first management device, and receiving maintenance operation message from said operation device via said first and second management devices, wherein said maintenance operation message is composed by a command inputted to said operation device and formed from an identifier of a program for executing in said industrial device and a parameter to be given to said program;processing means for storing programs indicating conditions and procedures of processes and for executing a process according to said maintenance operation message received by said communication means; andoutput means for outputting to an output device a whole or a part of charge information transmitted from said second management device, wherein said second management device generates said charge information indicating a maintenance charge according to a process content executed by said processing means. 2. An industrial device according to claim 1,wherein said communication means holds a user of said operation device and attribute information indicating a degree of art held by said user in connection with each other,and obtains said attribute information related to said user of said operation device transmitting said maintenance operation message, andsaid charge information is generated according to said obtained attribute information. |
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048184764 | abstract | The reactor vessel thread protector is a metal cover that is placed over the exposed stud, nut and washer projecting above the closure head to protect the threads from accidental corrosive leakage impingement emitted from mechanical seals on the top of the reactor vessel head. The stud protector includes a drip pan to collect any such fluid being emitted from the reactor vessel. Additionally, the reactor vessel stud thread protector will provide protection from accidental deformation to a stud due to operator maintenance around adjacent studs. The reactor vessel stud thread protector can be in the form of a cylinder which would cover a single stud bolt or a curved box having a rectangular cross-section that would cover a plurality of stud bolts. |
046684441 | description | The following example explains the process of the invention in more detail. The process can comprise, consist essentially of or consist of the stated steps with the recited materials. Unless otherwise indicated all parts and percentages are by weight. DETAILED DESCRIPTION EXAMPLE Isotropic graphitized soft coal secondary pitch coke powder was treated with phenol formaldehyde in the ratio of 4:1 to 6:1, there was added to the resin between 3 and 10 weight percent of hexamethylenetetramine as hardening agent. Half shells having a green density of about 1.2 g/cm.sup.3 were preliminary pressed from the thus produced molding powder at elevated temperature (about 100.degree. C.). A further portion of the molding powder was mixed with coated particles and preliminarily pressed to spherical nuclei having a green density of likewise about 1.2 g/cm.sup.3 (based on the molding powder components). The preliminarily pressed portions were preheated to about 80.degree. C., assembled in the final mold matrix and final molded, whereby to harden the binder the temperature of the charge was increased up to 170.degree. C. After the final molding the elements were carbonized in known manner and finally calcined in a vacuum at temperatures up to 2000.degree. C. The elements thus produced on the average exhibited higher crush strengths than the conventionally produced elements. In the case of using floating matrices the anistropy of the elements defined as the quotient of the linear thermal coefficient of expansion parallel and perpendicular to the molding direction is about 1.2 and accordingly within the band width of the elements conventionally produced by quasiisostatic molds. It is not necessary to encase the coated particles during the molding process to avoid particle damage with these elements. |
description | This application is a division of U.S. patent application Ser. No. 14/533,055 filed on Nov. 4, 2014, incorporated herein by reference in its entirety, which claims priority to, and the benefit of, U.S. provisional patent application Ser. No. 61/899,867 filed on Nov. 4, 2013, incorporated herein by reference in its entirety. This invention was made with Government support under DE-AC07-051D14517 awarded by the United States Department of Energy (DOE). The Government has certain rights in the invention. Not Applicable A portion of the material in this patent document is subject to copyright protection under the copyright laws of the United States and of other countries. The owner of the copyright rights has no objection to the facsimile reproduction by anyone of the patent document or the patent disclosure, as it appears in the United States Patent and Trademark Office publicly available file or records, but otherwise reserves all copyright rights whatsoever. The copyright owner does not hereby waive any of its rights to have this patent document maintained in secrecy, including without limitation its rights pursuant to 37 C.F.R. § 1.14. This technology pertains generally to high temperature reactor containments, and more particularly to isolation of containments for high temperature reactors using low volatility, liquid reactor coolants. Fluoride-salt-cooled, high-temperature reactors (FHRs), molten salt reactors (MSRs), sodium fast reactors (SFRs), lead fast reactors (LFRs) and other liquid cooled high temperature reactors use high-temperature, low-volatility liquid coolant to transfer heat from the reactor containment structure to a process heat or power conversion heat exchanger. Because the process heat fluid or power conversion fluid (such as water, air, carbon dioxide, or helium) is normally at a higher pressure than the low-volatility liquid coolant, and may react chemically with the coolant, leaks in these process heat and power conversion heat exchangers have the potential to cause excessive pressurization, water-hammer transmission, or shock transmission into the reactor vessel and the reactor containment structure. Because the reactor vessel and containment structure provides key barriers to the release of radioactive material from the reactor during any accident, overpressure and damage to the reactor vessel and piping, as well as the containment structure, must be avoided. The issue of mitigating overpressure has been particularly important for SFRs that use steam-Rankine power cycles, where the power conversion heat exchanger is a steam generator and very high overpressures can potentially occur if a tube leaks in the steam generator. Prior measures to prevent overpressure in SFR containment due to steam generator tube leaks include the use of double-walled steam generator tubes, the integration of a rupture disk into the steam generator, and the integration of one or more compressible gas cavities inside a steam generator. While it is highly unlikely for a reactor vessel to leak or rupture, reactors like FHRs, SFRs and LFRs that use low-pressure liquid coolants are designed to prevent fuel from becoming uncovered and to allow continued removal of decay heat even if this occurs. Prior measures to prevent fuel from being uncovered if the reactor vessel leaks have used a “guard vessel” located outside the reactor vessel to collect the leaked coolant and limit the amount that can leak. An aspect of the present disclosure is a high-temperature containment-isolation system for transferring heat from a nuclear reactor containment to a high-pressure heat exchanger using a high-temperature, low-volatility liquid coolant such as a molten salt or a liquid metal, where the coolant flow path provides liquid free surfaces a short distance from the containment penetrations for the reactor hot-leg and the cold-leg, where these liquid free surfaces have a cover gas maintained at a nearly constant pressure and thus prevent high-pressures from being transmitted into the reactor containment. Another aspect is a containment isolation system where the hot coolant exiting the reactor containment flows through an isolation valve into a hot-well vessel, which has a significant total free liquid surface and contains a centrifugal pump that circulates the hot fluid through a crossover-leg to a high-pressure heat exchanger. In one embodiment of the current aspect, the hot well has sufficient volume to accommodate level changes due to thermal expansion of the coolant in the loop and due to changes in level at other free surfaces in the loop when pump speeds are changed. In another embodiment of the current aspect, the hot well has a gate valve to seal and isolate the hot-leg pipe entering the hot well, to provide isolation of the reactor vessel and containment. In another embodiment of the current aspect, the hot well is configured such that the bottom of the suction of the pump is above the top of the hot leg pipe and isolation valve, so draining the cross-over leg to the high-pressure heat exchanger does not break the seal or drain the hot leg pipe. Another aspect is a containment isolation system where the cold coolant reentering the reactor containment flows upward in a standpipe and then across through a cold leg pipe into the reactor containment, and where the standpipe has a liquid free surface, and this liquid free surface has a cover gas maintained at a nearly constant pressure and thus prevents high-pressures from being transmitted into the reactor containment. In one embodiment of the current aspect, the standpipe has a nearly neutrally buoyant plug that fills a significant fraction of the volume above the horizontal penetration into containment, that reduces the volume of coolant in the standpipe above the penetration and thus the level change in the hot well when pump speed is changed, while still limiting the maximum pressure that can be established at the cold leg penetration. In another embodiment of the current aspect, the standpipe has a gate valve to seal the cold leg pipe entering the reactor containment, to provide isolation of the reactor containment. In another embodiment of the current aspect, the standpipe includes a seal loop, which prevents the cold leg from draining when the standpipe is drained. In another embodiment of the current aspect, the standpipe has a drain tank located below it, with a drain line and a freeze or other type of isolation valve to allow the coolant loop to be drained and cooled for inspection and maintenance, and then refilled. In another embodiment of the current aspect, the standpipe has an overflow line above the normal coolant level to return coolant to the hot well or a drain tank if over-pressure occurs. Another aspect is a containment isolation system with a plurality of coolant loops, hot wells, main coolant pumps, and high pressure heat exchangers. In one embodiment of the current aspect, for a plurality of loops each hot well is connected to adjacent hot wells by conduits above the elevation of the cross-over legs, to allow cross flow between the hot wells to maintain a constant hot-well coolant level. In another embodiment of the current aspect, each coolant loop in the plurality of loops can be drained and cooled independently for inspection and maintenance while other coolant loops remain filled. Another aspect is a high temperature reactor (HTR) cooled by a low-volatility, liquid coolant, comprising a reactor cavity having an inner cavity wall, and a reactor vessel disposed within the reactor cavity, wherein the reactor vessel is suspended within the reactor cavity with a plurality of refractory insulator blocks disposed between the inner cavity wall and the reactor vessel. In one embodiment of the current aspect, the reactor vessel is suspended in the reactor cavity via a conical support ring having a first end coupled to the inner cavity wall and a second end coupled to the reactor vessel. Further aspects of the technology will be brought out in the following portions of the specification, wherein the detailed description is for the purpose of fully disclosing preferred embodiments of the technology without placing limitations thereon. 1. Isolation of High-Temperature Reactor Containments One aspect of the present disclosure is a system to prevent a leak from a high-pressure process heat or power conversion heat exchanger from causing excessive pressurization or shock loading inside the reactor containment. In a conventional configuration to transfer heat from a nuclear reactor containment to a process heat or power conversion heat exchanger outside of the containment using a low-volatility liquid, the hot liquid exits the reactor containment structure through one or more hot-leg penetrations. The coolant flows to the high-pressure heat exchanger, and then returns via one or more cold-leg penetrations back inside the reactor containment to be heated again. The pump that circulates the coolant may be in a variety of locations around the loop, either inside or outside containment. In the systems 10A and 10B shown in FIG. 1A and FIG. 1B, the hot coolant 20 exits the containment through a hot-leg pipe 18 that passes through an insulated penetration in the reactor cavity wall 14 with a bellows seal 24 around the pipe. Immediately outside the containment penetration the hot coolant 20 flows through an isolation valve 36, and then into a hot-well vessel 12 located close to the hot-leg penetration to minimize the length of the hot leg 18. The hot well vessel 12 preferably has a substantial liquid free surface area 40 within its internal chamber 38 and sufficient volume to accommodate level changes due to thermal expansion of the coolant in the loop and due to changes in level at other free surfaces in the loop when pump speeds are changed. FIG. 1A shows a containment configuration 10a with hot leg 18 connecting to hot well vessel 12. The steel and concrete reactor cavity wall 14 has a steel liner 16 that is actively cooled by water or air, and that extends through the cavity wall 14. This penetration has internal insulation 22a, around the hot leg pipe 18 and the containment boundary is formed by a bellow seal 24 between the hot leg 18 and the containment liner 16. After the bellow seal 24, all coolant piping and the hot well 12 preferably use an external insulation system with internal trace heating (not shown). The primary monitoring of coolant inventory for protective functions uses level instrumentation located in the hot well. The hot well 12 is designed to have a sufficiently large salt surface area that level swell due to thermal expansion of the coolant 20 and due to level changes at other free surfaces in the primary system result in acceptable level changes in the hot well. If a plurality of hot well cavities and pumps is used, conduits 34 to connect the cavities may be positioned below the Lnom level to provide cross flow between the cavities and maintain their surface levels equally. FIG. 1A and FIG. 1B also show a maximum coolant level (Lmax) for hot accident condition and minimum salt level (Lmin) for maintenance condition. If a plurality of hot well cavities is used, then the conduits 34 connecting the cavities are located above the minimum coolant level (Lmin) to allow individual loops to be drained for maintenance. The gate valve 36 in the hot well 12 allows the hot leg 18 to be isolated (e.g. when gate valve 36 is extended to the bottom location 26 in the vessel), and thus seals and isolates the hot-leg pipe 18 to provide isolation of the reactor vessel (see 110 in FIG. 3), concrete reactor cavity wall 14 and containment liner 16. The hot well vessel 12 comprises a centrifugal pump 28 configured for discharging outbound coolant 30 to crossover leg 32. The crossover leg 32 transfers coolant to the high pressure heat exchanger. The bottom of the suction of the centrifugal pump 28 is preferably above the top of the hot leg pipe 18 and isolation valve 36, so draining the cross-over leg 32 does not break the seal or drain the hot leg pipe 18. The volume of the internal chamber 38 of the hot well vessel 12 may have a cover gas 42 over the surface of the coolant and liquid free surfaces 40 of the chamber 38. The hot well cover gas system maintains a constant gas pressure over the hot well liquid free surfaces 40 and coolant level L. FIG. 1B shows an alternative embodiment where the bellows 24 connects to a flange 42 integrated into the hot-well vessel 12. In this configuration, a break in the hot-leg pipe 18 would not result in a coolant leak outside of the reactor containment boundary. In a preferred embodiment, the hot well 12 is supported by vertical bearings (not shown) that allow horizontal movement to accommodate thermal expansion of the hot leg 18. Snubbers (not shown) may be used to restrain the hot well 12 from rapid motion during seismic events. After exiting a high-pressure heat exchanger (see a representative embodiment, the coiled tube air heater 210 of FIG. 7), the coolant 30 flows back to a vertical cold-leg stand pipe that is adjacent to the reactor containment cold-leg penetrations of cavity liner 16. FIG. 2A through FIG. 2D illustrate various embodiments of cold-leg stand-pipe systems 50a-50d in accordance with the present description. Cold coolant 70 reentering the reactor containment flows upward in a standpipe 52/58 and then across through a cold leg pipe 68 into the reactor containment liner 16. The standpipe 52/58 preferably has a liquid free surface 72, and this liquid free surface 72 has a cover gas 74 maintained at a nearly constant pressure and thus prevents high-pressures from being transmitted into the reactor containment. FIG. 2A shows a basic stand-pipe configuration 50a, while FIG. 2B shows a stand-pipe configuration 50b with a gate-type isolation valve 64. FIG. 2C shows a stand pipe configuration 50c with anti-syphon seal loop and neutrally buoyant plug 82. FIG. 2D shows a stand pipe configuration 50d similar to the embodiment of FIG. 2C with a connection of the containment boundary bellows 66 to the stand-pipe vessel 88. Referring to FIG. 2A and FIG. 2B, the standpipe 52 extends above the cold leg pipe 68 penetration sufficiently high such that under the highest fluid flow rate, the rise of liquid level 62 in the cold-leg standpipe due to the liquid flow 54 remains below the top 58 of the stand pipe 52. Referring to FIG. 2A through FIG. 2D, the standpipe 52/88 has a gas supply system 56 that is capable of maintaining the gas pressure on the surface of the liquid at nearly atmospheric pressure. The existence of free liquid surfaces at both the hot and cold leg penetrations makes it difficult, if not impossible, to transmit high pressure into the reactor containment. The top 58 of each stand pipe 52/88 extends above the upper horizontal cold leg 68 to accommodate the level swell that occurs under pump operation due to the pressure difference across the reactor vessel. An overflow pipe 60 is provided above the normal maximum free surface elevation 62, which will drain salt back to the hot well 12 or a drain tank (see 222 in FIG. 7) if the pressure drop across the reactor exceeds its normal design value or if the cold leg 68 is over pressurized. The cold leg pipe 68 penetration into the reactor cavity wall 14 includes internal insulation 22b and bellows seal 66 connecting to the containment liner 16 for containing coolant in the event of a break in the cold leg pipe 68. In embodiments 50b through 50d of FIG. 2B through FIG. 2D, an isolation gate valve 64 in the standpipe 52/88 is included to provide additional capability to isolate the flow of coolant 70 into the reactor 110 from cold leg pipe 68. Referring to FIG. 2C and FIG. 2D, the stand pipe vessel 88 has integrated into its geometry an anti-siphon seal loop 86 and a nearly neutrally buoyant float 82. The seal loop created by the pipe extension 86 allows the stand pipe vessel 88 and stand pipe 52 to be drained for maintenance without draining the horizontal cold leg pipe 68. The nearly-neutrally buoyant plug 82 is designed to displace most of the coolant volume (i.e. fill a significant portion of the volume) above the upper horizontal cold leg 68 to minimize the level drop that occurs in the hot well 12 when the pump operates and the coolant level increases in the cold-leg stand pipe. The plug 82 is configured to move freely in the vertical direction, to prevent overpressure of the reactor vessel from occurring during high-pressure heat exchanger tube (see FIG. 7) rupture events. Effectively, the plug 82 reduces the volume of coolant in the standpipe 88 above the penetration and thus the level change in the hot well 12 when pump speed is changed, while still limiting the maximum pressure that can be established at the cold leg 68 penetration. The bottom of the plug 82 may also incorporate a burst diaphragm (not shown), compressible cavity (not shown), in addition to flow diverter 84 to mitigate transmission of hypothetical water hammer shocks into the upper horizontal cold leg. FIG. 2D shows a configuration 50d similar to the configuration 50c of FIG. 2C, yet wherein the bellows 66 extends to a flange 90 integrated into the stand-pipe vessel 88, so that a break in the cold leg pipe 68 would be contained inside the containment boundary. The bellows 66 (in addition to bellows 24 for the hot well system 10a/10b) are configured to be installed in a pre-stressed state, so that at normal operating temperatures the reactive forces imposed by the bellows, and the stresses in the bellows, are minimized. All salt pipe penetrations, both for the reactor containment 16 and for the high-pressure heat exchanger vessels (see 210 in FIG. 7), are oriented so that the displacements caused by thermal expansion are primarily axial along the length of the pipes. Therefore, the bellows seals must accommodate substantial axial displacements, as well as moderate transverse displacements. It is appreciated that the embodiments shown in FIG. 1A to FIG. 2D may also be configured for use with reactors that have multiple coolant loops, hot wells, pumps and high pressure heat exchangers. When the reactor has two or more coolant loops, then the hot wells are connected by conduits 34 that allow cross flow between hot wells to maintain uniform coolant levels in all of the hot wells. The crossover legs 32 for each loop are then located at an elevation above the hot leg 18 and hot-leg isolation valve 36, and below the connecting pipes 34 between hot wells. 2. Reactor Cavity Configuration Referring now to FIG. 3 through FIG. 6, features of a reactor cavity and reactor vessel system are shown that have desirable features for use with high temperature reactors (HTRs) cooled by low-volatility coolants, including fluoride salt cooled high temperature reactors (FHRs), lead fast reactors (LFRs), and sodium fast reactors (SFRs). These reactors generally use a pool-type vessel, which only has pipe penetrations at high elevations so that any break in a pipe cannot drain coolant below the level of fuel in the reactor. The general design practice in FHRs, LFRs, and SFRs is to have a separate guard vessel to maintain the coolant inventory in the primary system if the reactor vessel leaks or ruptures. The system of the present disclosure uses a different design approach that eliminates the need for a guard vessel. As illustrated in the schematic side section view of FIG. 3 and side section view of FIG. 4 of the reactor and containment system 100, the HTR reactor vessel 110 of the present disclosure is instead suspended in a reinforced concrete reactor cavity structure 14 with an actively cooled (water or external air) steel inner liner wall 16. Inside the reactor vessel many designs for cores and reactor vessel internals are possible, with FIG. 3 showing a highly schematic view of a reactor vessel internal core barrel assembly 116 and core 126, and FIG. 4 showing an alternative but similar configuration for core internal structures 116 and core 126. The majority of the volume between the reactor vessel 110 and the cavity liner 16 is filled with refractory insulator blocks (e.g. upper refractory blocks 22c and lower refractory blocks 22d), which are preferably fabricated from a refractory material such as ceramic insulating fire-brick or baked carbon blocks that is chemically compatible with the reactor coolant. Blocks of the same material also line the hot leg pipe penetrations 18 and cold leg penetrations 68. Under normal operation the blocks 22a/22b/22c/22d act as an insulator, to minimize parasitic heat loss. However, upon addition of salt into the reactor cavity due to a reactor vessel leak or from a buffer salt store, natural circulation of the salt in the gaps between blocks in the lower cavity 22d, and thus can remove heat from the reactor vessel to prevent overheating. The volume of coolant in the vessel and the volume of free space in the lower cavity are designed so that if the vessel leaks, the coolant surface level in the vessel 120 will remain above the reactor core 126. The reactor vessel 110 is suspended within the steel/concrete composite wall structure 14 and inner cavity liner 16 via a conical reactor vessel support ring 112 that has an upper mounting surface 144 that mates with an annular protrusion 142 in the metallic core structure 110 to form a joint (see FIG. 6) at roughly the vessel diameter D (see FIG. 4). The lower refractory blocks 22d form a conical configuration that terminates at the mounting surface 144 of support ring 112. An annular thermal expansion gap 114 is formed between the inner wall of the refractory blocks 22d and the outer wall of the between the reactor vessel 110. As seen in FIG. 4, an insulated cavity cover structure 132 is provided to define the upper boundary of the reactor cavity system 100. Also shown in FIG. 4 is the hot leg 18, with penetration formed by insulation 22a and bellow 24. If water is used to cool the cavity liner 16, a liner cooling leak drain sump 130 is provided at the lower end of system 100. Because many different realizations for cores 126 and reactor vessel internals 116 are possible, FIG. 4 shows a different schematic configuration than FIG. 3. The normal reactor coolant level 120 is also shown in FIG. 4. As seen in FIG. 5 and FIG. 6, a plurality of slots 140 are created in the lower refractory blocks 22d to allow for positioning of electric heating rods 127 down the length of the outside of the reactor vessel. System 100 is configured to provide auxiliary heating of the reactor cavity using electrically heated rods 127 inserted into slots 140. The slots 140 and rods 127 are preferably configured with an adiabatic section (not shown) that extends through the thermal shield 132 above the reactor cavity and internal reactor structures 128, where electrical power connections are made, and the rods can be replaced through openings in the refueling deck above the thermal shield. The refractory liner blocks 22a/22b/22c/22d are ideally configured to provide insulation and minimize heat losses from the hot legs, cold legs, and reactor vessel to the reactor cavity liner under normal operation. The free volume around the lower cavity blocks 22d is minimized to reduce the amount of coolant volume in the cavity if the reactor vessel leaks or ruptures, such that the level of the coolant 120 in the vessel remains above fuel in the core. The refractory liner block 22a/22b/22c/22d material is preferably selected to have appropriate chemical compatibility with the reactor coolant, to prevent chemical reactions that would generate gases or excessively corrosive conditions. One such material that is compatible with fluoride salts and with lead is a mixture of ceramic oxides (commonly called “fire brick” and “insulating fire brick”) containing AlO3, MgO, SiO2, and other refractory oxides. Another exemplary material is baked carbon. Because sodium is highly reactive, metallic or metal-lined insulation blocks are required. The lower cavity refractory liner blocks 22d are preferably aligned with gaps designed to facilitate natural circulation heat transfer by the salt coolant, if coolant leaks into the cavity space, to create a frozen zone for coolant near the cavity liner and a circulating zone of coolant near the reactor vessel, which assists to transfer heat from the reactor vessel to the cavity liner under accident conditions. In situations where an accident occurs beyond the design basis, additional coolant may be deliberately injected into the space between the reactor vessel 110 and the cavity liner 16, to establish a frozen coolant zone near the cavity wall and to increase heat transfer from the reactor vessel wall to the cavity wall, to assist in the removal of decay heat. The refractory liner blocks 22a/22b/22c/22d are also preferably configured to provide sufficient space for the reactor vessel 110, hot legs 18, and cold legs 68 to undergo normal thermal expansion, without contact occurring between the blocks and the hot legs, cold legs, or reactor vessel. For the reactor vessel, the gap is made sufficiently small for contact to occur to provide support to the reactor vessel during beyond design basis accidents, if the vessel reaches very high temperatures where it may undergo accelerated creep deformation. The refractory liner blocks 22c/22d may be configured with keys and restraint mechanisms (not shown) to transfer horizontal accelerations into the cavity liner under seismic motion. The refractory liner blocks 22c/22d may be configured with additional gaps or channels for the insertion of inspection instruments such as boroscopes, or cavity electrical heaters may be designed to be removed to allow insertion of inspection instruments. The refractory liner blocks 22c/22d may be configured to include an annular region near the reactor vessel with a higher thermal conductivity, for example with graphite, and with electric heaters, to allow heating of the reactor vessel under start up and during over cooling events. 3. Overall Heat Transport Configuration FIG. 7 shows a high-level schematic view of a heat transport and containment isolation system 200 comprising a plurality of coolant loops, hot wells 12, main coolant pumps, systems for the emergency removal of decay heat in the form of a Direct Reactor Auxiliary Cooling System (DRACS) 202, and high-pressure heat exchangers in the form of coiled tube air heaters (CTAH's) 210. Each hot well cavity 12 is connected to adjacent hot wells by connecting pipes or ducts 34 (see FIG. 1A) above the elevation of the cross-over legs 32 to allow cross flow between the hot wells 12 to maintain a constant hot-well coolant level. Each coolant loop (two are shown in FIG. 7, one for each CTAH 210) can be drained to a tank 222 using a drain line with a freeze valve 224 and the coolant loop can then be cooled independently for inspection and maintenance while other coolant loops remain filled. When a loop is drained for maintenance, the cover gas system 56 supplies the gas that fills the loop as the cross-over leg 32 drains toward the CTAH 210. For CTAH 210 tube rupture events, the cover gas system 56 has over-pressure control that can direct gas, air, and entrained salt to a knock-out drum and filter system (not shown). Each CTAH has a hot air duct 212 and cold air duct 214, all configured with expansion bellows 216 similar to bellows 24 and 68 at the hot leg 18 and cold leg 68 penetrations with the containment wall 16 of the reactor cavity structure 14. The standpipes 52 also comprise at their bottoms cold traps 226, which also serve as low-point drain locations for the main salt drain tanks 222. Cold traps 226 comprise redox control and filter cartridges (not shown) inside the cold trap 226 at the bottom of the stand pipe 52. 4. Exemplary PB-FHR Gas-Gap Isolation System and Refractory Reactor Cavity Liner a. Refractory Reactor Cavity Liner Characteristics A gas-gap containment isolation system and a Refractory Reactor Cavity Liner System (RCCLS) was configured for a Mark 1 Pebble-Bed Fluoride-Salt-Cooled High Temperature Reactor (Mk1 PB-FHR) commercial prototype, the details and components of which may be understood with reference to FIG. 1A through FIG. 7 above. The Mk1 PB-FHR design is a small, nominally 236 megawatt thermal pebble-fueled FHR with a rail transportable reactor vessel. The vessel configuration is similar to the earlier ORNL SmAHTR design, with a relatively tall, skinny reactor vessel. The vessel is 350 cm in diameter and nominally 10 m tall, with a 360-cm diameter skirt, making it rail transportable, and the vessel wall operates at the Mk1 core inlet temperature of 600° C. In the reactor cavity system 100 of the present description, the guard vessel is eliminated. Instead, the volume between the reactor vessel and the reactor cavity liner plate is filled with refractory insulation blocks, a refractory reactor cavity liner system 100 (RRCLS), as shown schematically above in FIG. 3 through FIG. 6. This RRCLS 100 utilizes the high freezing temperature of the low volatility coolant (the salt coolants used in FHRs typically freeze between 450° C. to 510° C., with the freezing temperature being 459° C. for the Mk1 primary salt flibe (Li2BeF4)). The high freezing temperature replaces the coolant containment function provided by a guard vessel with coolant containment provided by the generation of a “cold crucible” of frozen salt near the reactor cavity liner and wall, which remains at temperatures below the freezing temperature of the coolant during beyond design basis accidents. Because the majority of the volume of the cavity between the reactor vessel 110 and the reactor cavity liner 102 is filled with the refractory blocks 22d, the volume of coolant required to fill the remaining volume is reduced, and therefore the drop in level 120 in the reactor vessel 110 is also reduced. Under normal operation the refractory liner blocks 22a/22b/22c/22d provide insulation and limits parasitic heat loss. Conduction through the blocks, and natural circulation of the cavity gas in the spaces between the blocks, creates the primary mechanism for heat loss from the reactor vessel surface, which remains at the nominal coolant core inlet temperature of 600° C. under normal operation. The steel cavity liner plate 16 is actively cooled by water (or alternatively air) flowing in pipes behind the plate to maintain the cavity liner plate at an acceptable temperature for the concrete behind the plate (nominally 30° C.). Under beyond design basis accident conditions where the reactor vessel 110 fails and coolant drains into the cavity, or when additional coolant is injected deliberately into the cavity (from either a frozen storage location located above the cavity, or an emergency coolant injection system), the coolant near the reactor vessel remains molten, because the reactor vessel surface temperature exceeds the salt melting temperature. The coolant freezes at the cavity liner 16, because the temperature is below the coolant freezing temperature. The fact that the Mk1 coolant freezes at a relatively high temperature allows leaks from the reactor cavity to be controlled—even if a crack occurs in the cavity liner 16 or the reactor cavity wall 14, any salt that enters the crack would freeze and plug the leak path. Initially the cavity liner 16 is at its nominal temperature, but if active cooling is not maintained the temperature climbs over time. If water is supplied to the cavity liner cooling system, it will boil and maintain the liner at a maximum temperature near 100° C. Without water supply, the liner temperature will climb further, and heat removal occurs due to transient conduction into and heating of the steel-concrete composite structure of the reactor cavity wall 14 and then into surrounding building structures. An interface forms inside the cavity between frozen and molten coolant, and the molten coolant flows by buoyancy driven natural circulation to transfer heat to this interface. For initial start-up and to control overcooling after shut down, electric heating elements 127 are provided near the reactor vessel 110 surface, to allow heating of the vessel. In the baseline design, these heaters are rods inserted into vertical slots 140 in the cavity liner blocks 22d, adjacent to the reactor vessel 110. A number of options exist for refractory insulation materials 22a/22b/22c/22d for the Mk1 PB-FHR. Candidate materials include baked carbon insulating material, a low thermal conductivity graphite-like material used in modular helium reactors, fire brick, and insulating fire brick. Baked carbon is an insulating material that is formed by vibration molding of an isotropic pitch coke filler and coal tar pitch binder. It is baked at a temperature of 800° C. to 1100° C., pitch impregnated, and baked a second time. Because the carbon is heated sufficiently to cause extensive graphitization, the thermal conductivity remains lower than graphite, but is higher than oxide ceramic insulation materials. Baked carbon will have good compatibility with fluoride salts, but may oxidize in the presence of air. FIG. 8 shows property data for baked carbon insulation material. Insulating fire brick is composed of a mix of different oxides. In general, these oxides will react slowly with fluoride salts. The fire bricks are porous, so may require surface coatings to prevent salt from imbibing into the porosity. One option is firebrick from BNZ Materials. Type C22Z is preferably selected due to its relatively high strength. Its thermal conductivity at 538° C. is 0.26 W/m° C. The thermal conductivities of insulating fire brick (<0.5 W/m° C.) are significantly lower than the conductivities of full density fire brick (1.0 to 2.8 W/m° C., 2,050 to 2,645 kg/m3). These higher-density brick materials all have higher thermal conductivity, and lower porosity than the low-density insulating brick. The Mk1 reactor vessel 110 is preferably fabricated from a high-temperature, high-nickel alloy (e.g. 316 SS, 304 SS, or Alloy N). The reactor vessel dimensions change as it is heated. The percent expansion of 316 SS when heated from room temperature to 600° C. is calculated at:100(18.2×10−6° C.−1)(600° C.−20° C.)=1.06%. The mean coefficient of thermal expansion coefficient of Hastelloy N, an alternative material for the reactor vessel, is 14.0×10−6° C.−1. The thermal expansion coefficient for Alloy N is about 26% lower than for 316 SS. Three primary design considerations were calculated with respect to performance of the refractory reactor cavity liner system 100: 1) the parasitic heat loss from the reactor vessel to the reactor cavity wall, 2) the volume of coolant that would drain into the cavity if the reactor vessel fails, and 3) heat transfer to the reactor cavity wall via natural circulation of coolant if the cavity is flooded with coolant. With respect to heat loss to the reactor cavity, the options for refractory ceramic materials to fill and insulate the reactor cavity have thermal conductivities that range from 0.25 to 8 W/m° C. The nominal Mk1 reactor vessel size, used in this calculation, is 10.0 m high and 3.5 m diameter, with a 0.50-m thick refractory insulation layer between the reactor vessel and the reactor cavity liner. The effective area of the insulated surface, including the sides and bottom, is then:3.14((10.0 m)(3.50 m+2(0.25 m))+(3.50 m)2/4)=135 m2. FIG. 9 is a plot of heat loss from a the Mk1 PB-FHR reactor vessel to reactor cavity wall, for a 0.50-m thick insulation layer, with a total effective surface area of 135 m2, and vessel and cavity wall temperatures of 600° C. and 30° C. respectively, as a function of the insulation thermal conductivity. As shown in FIG. 9, the resulting heat loss depends significantly on the effective thermal conductivity of the liner blocks 22c/22d, and would be below 100 kW for insulating fire brick (0.26 W/m° C.) up to approximately 1.0 MW for baked carbon (7 W/m° C.). This heat loss rate is low compared to the nominal reactor thermal power of 240 MWt. So the selection of the specific refractor material can be made based upon criteria in addition to heat loss, including chemical compatibility with coolant, impermeability, mechanical strength, and cost. With respect to drain volume, it is desired that if the reactor vessel 110 leaks, the maximum drop of level 120 in the vessel remains above the top of pebble fuel located in the defueling chute above the reactor core, called the faulted level here. The top of the system 100 defueling chute is nominally 7.0 m above the bottom of the reactor vessel. The total volume of the reactor cavity, below this elevation, is calculated at:Vcf=(7.0 m)3.14((3.5 m+(0.5 m))2−(3.5 m)2)/4+(0.5 m)3.14(3.5 m+(0.5 m))2/4=52 m3. The free volume in the cavity includes the volume in the gap 114 between the reactor vessel 110 and the blocks 22d, and the volume of the gaps between blocks 22d, other blocks, and the reactor cavity wall 102. The gap between the reactor vessel and the refractory blocks is established to allow the vessel to undergo thermal expansion under “beyond design basis accidents” from its normal operating temperature of 600° C. up to the temperature where the vessel material rate of thermal creep would become large. At this beyond design basis accident temperature, selected here to be 900° C., it is desired that the vessel begin to contact the refractory blocks, so that the blocks provide physical support to limit further expansion and creep of the vessel. In heating from 600 to 900° C., the vessel diameter increases by 1.9 cm, and the bottom of the vessel drops 3.8 cm. The volume change of the vessel, due to this expansion in size, is 1.1 m3. The volume of the gaps between refractory blocks is a design parameter that depends upon the geometry of the blocks and the precision of their manufacture. The nominal blocks 22d are 0.5 m×0.5 m×0.5 m. While this exceeds the maximum dimensions of commercial insulating fire brick, these larger blocks are fabricated by cementing together assemblies of smaller blocks. If the porosity of the blocks is found to be an issue for absorbing Mk1 coolant, the blocks are coated with a glazing material to reduce the permeability of the surface. Gaps between blocks 22d will vary between 0.5 and 1.0 cm, with an average gap width of 0.7 cm. In this case, the fraction of the total lower cavity volume occupied by gaps between the blocks and the cavity wall, and blocks with neighboring blocks, will be (1+0.5(4))(0.7 cm)/(50 cm)=0.042, or 4.2% of the cavity volume. In this case, the total void volume in the cavity will be:Vv=0.042(52 m3)+1.1 m3=3.3 m3. In the reactor vessel of the present description, if salt occupies 75% of the cross-sectional area of the vessel at the elevation above the defueling chute and below the bottoms of the hot and cold legs, then the level drop caused by a leak in the reactor vessel would be:H=(3.3 m3)/0.25(3.1415)(3.5 m)2)=0.34 m. This value is sufficiently small so that high assurance exists that fuel in the reactor will remain covered with coolant. Heat transfer from the reactor vessel 110 to the cavity wall 102 changes greatly under “beyond design basis accident” conditions where the cavity is flooded with coolant. The volumetric heat capacity of the Mk1 coolant (flibe) is nominally 4500 kJ/m3° C., compared to argon and air (approximately 0.4 kJ/m3° C. at 627° C.). b. Primary Loop “Gas-Gap” Isolation System The Mk1 PB-FHR heat transport and containment isolation system 200 (FIG. 7) uses the primary coolant to directly heat air for the nuclear air combined cycle power conversion system. Compared to previous FHR designs, as well as the long-term conventional practice for sodium fast reactors (SFRs), the elimination of an intermediate loop in the Mk1 PB-FHR has major implications for simplification, cost reduction, and overall plant safety. The reason that intermediate loops have been conventional for SFRs has been the requirement to mitigate the potential over pressure, water hammer, and shock loading that would occur if sodium and water were to interact following tube failure in a steam generator. Because fluoride salt coolants are chemically stable, no significant chemical reactions will occur under a tube failure in a coiled tube air heater (CTAH) (or a steam generator). However, it remains possible that a tube failure could cause temporary pressure increase and generate water hammer in the salt piping subsystem. The system 200 of the present disclosure introduces the concept of using “gas gaps” in the hot and cold salt piping system to isolate the reactor vessel and reactor cavity/containment from the external salt piping system and CTAHs. Immediately adjacent to the containment penetration, both the hot and cold salt legs 18/68 have free surfaces communicating with the salt cover gas system, which make it deterministically impossible to transmit high pressure into the reactor vessel and containment. For each hot leg 18, the free surface exists in the hot well 12, which may include a gate-type isolation valve 36 and inside a cantilevered, sump-type main-salt pump's (28) take suction. For each cold leg 68, a free surface is provided in a standpipe 52/88, which may also contain an isolation valve 64. The isolation of the hot and cold legs 18/68 is further enhanced by the gate valve, which has an activation stem that extends up through the liquid free surface. The gate valves 36/64 may use a carbon fiber reinforced seat material on the valve wedge to prevent self-welding when closed. The hot well or wells 12 are located immediately adjacent to the reactor cavity structure 14 to minimize the length of the hot legs 18 and cold legs 68, as shown in FIG. 1A, FIG. 1B, FIG. 2A, FIG. 2B, FIG. 2C, and FIG. 2D. The hot wells 12 are designed to have a sufficiently large coolant surface area that level swell due to thermal expansion of the coolant and due to level changes at other free surfaces in the primary system result in acceptable level changes in the hot well. Under normal power operation, the coolant level in the hot wells is above cross over legs 32 that connect the hot wells to the high pressure heat exchangers. For multiple hot wells, conduits 34 between the wells, above the cross over legs 32, keep the level uniform in the hot wells. Gate valves 36 in each hot well allow each hot leg to be isolated. Because the Mk1 system 200 primary coolant, flibe, has a very high volumetric heat capacity, and because the design temperature drop across the Mk1 core is relatively large (100° C.), the total volumetric flow of primary coolant to transport 236 MWth in a Mk1 PB-FHR is only 0.50 m3/sec. This makes it possible to design the reactor to operate with a relatively small pressure drop between the cold legs 68 and hot 18 legs, with the design goal being a total pressure drop between 2 and 3 meters of head. Each of the two cold legs 68 has a vertical standpipe 52/88 located adjacent to the reactor cavity/containment wall. FIG. 2A through FIG. 2D show various realizations of this standpipe. Because each standpipe communicates with the salt cover gas system, they provide a “gas gap” to prevent overpressure of the reactor vessel and containment. Chilled coolant exiting the bottom of each CTAH 210 flows in a downward sloping pipe over to the stand pipe 52/88, up the stand pipe 52/88, and then horizontally through a containment wall penetration in a cold leg pipe 68 into the reactor vessel 110. FIG. 2A shows the simplest realization of a standpipe 50a, where the top of each standpipe extends approximately 3 m above the upper horizontal cold leg 68, to accommodate the level swell that occurs under pump operation due to the pressure difference across the reactor vessel. The specific height is determined by the detailed design of the flow path in the reactor and the resulting pressure drop. An overflow pipe 60 is provided above the normal maximum free surface elevation 80, which will drain coolant back to the hot well 12 if the pressure drop across the reactor exceeds its normal design value or if the cold leg is over pressurized. From the discussion above it will be appreciated that the technology can be embodied in various ways, including the following: 1. A high temperature reactor (HTR) cooled by a low-volatility, liquid coolant, comprising: a reactor cavity having an inner cavity liner; a reactor vessel disposed within the reactor cavity; and wherein the reactor vessel is suspended within the reactor cavity with a plurality of refractory insulator blocks disposed between the inner cavity liner and the reactor vessel. 2. The HTR of any preceding embodiment: wherein the reactor vessel is suspended in the reactor cavity via a conical support ring having a first end coupled to the inner cavity liner and a second end coupled to the reactor vessel; and wherein the conical support ring accommodates temperature differentials between the reactor vessel and reactor cavity. 3. The HTR of any preceding embodiment: wherein the second end of the conical support ring attaches to the reactor vessel via a conical joint; and wherein the conical joint is configured to allow for differential thermal expansion or horizontal loading within the reactor vessel. 4. The HTR of any preceding embodiment, wherein gaps between the plurality of refractory insulator blocks act to provide flow paths for natural circulation of coolant to thereby remove heat from the reactor vessel in the event of a reactor vessel coolant leak. 5. The HTR of any preceding embodiment, further comprising: an annular thermal expansion gap formed between the plurality of refractory insulator blocks and the reactor vessel. 6. The HTR of any preceding embodiment, wherein the inner cavity liner comprises an actively cooled steel liner plate. 7. The HTR of any preceding embodiment, further comprising: a plurality of slots disposed within an internal surface of the refractory blocks adjacent said annular thermal expansion gap; and a plurality of electric heating rods disposed within the plurality of slots, the electric heating rods running substantially along the length of the reactor vessel. 8. The HTR of any preceding embodiment, the reactor cavity having an outer containment wall, the HTR further comprising: a hot-leg pipe coupled to the reactor vessel at a penetration of the outer containment wall; wherein the hot-leg pipe provides a coolant flow path from the reactor vessel through the outer containment wall to a main-salt loop external to the reactor vessel; wherein the hot-leg penetration is insulated and has a flexible bellows seal. 9. The HTR of any preceding embodiment, further comprising: an isolation valve and a hot-well vessel coupled to the hot leg; wherein hot coolant exiting the reactor vessel flows through the isolation valve into the hot-well vessel. 10. The HTR of any preceding embodiment, wherein the hot-well vessel provides a substantial free liquid surface, the hot-well vessel further comprising: a centrifugal pump submerged in the hot well and configured to circulate the hot coolant into a crossover-leg to a high-pressure heat exchanger; and wherein the coolant free surface has a cover gas maintained at a nearly constant pressure and thus prevents high-pressures from being transmitted into the reactor vessel. 11. The HTR of any preceding embodiment, the reactor cavity having an outer containment wall, the HTR further comprising: a cold-leg pipe coupled to the reactor vessel at a penetration of the outer containment wall; wherein the cold-leg pipe provides a coolant flow path into the reactor vessel through the outer containment wall from a main-salt loop external to the reactor vessel; wherein the cold-leg penetration is insulated and has a flexible bellows seal 12. The HTR of any preceding embodiment, further comprising: a standpipe coupled to the cold-leg pipe; wherein the standpipe provides a free liquid surface; and wherein the liquid free surface has a cover gas maintained at a nearly constant pressure and thus prevents high-pressures from being transmitted into the reactor vessel. 13. A high-temperature containment-isolation system for an HTR, comprising: a reactor vessel disposed inside a reactor cavity having an outer containment wall; a hot-leg pipe coupled to the reactor vessel at a hot-leg penetration of the outer containment wall; wherein the hot-leg pipe provides a coolant flow path having a liquid free surface near the hot-leg penetration; and wherein the liquid free surface has a cover gas maintained at a nearly constant pressure and thus prevents high-pressures from being transmitted into the reactor vessel. 14. The system of any preceding embodiment, further comprising: an isolation valve and a hot-well vessel coupled to the hot-leg coolant flow path; and wherein hot coolant exiting the reactor vessel flows through the isolation valve into the hot-well vessel. 15. The system of any preceding embodiment, wherein the hot-well vessel provides a substantial free liquid surface and a submerged centrifugal pump that circulates the hot coolant through a crossover-leg. 16. The system of any preceding embodiment, wherein the hot-well vessel has sufficient volume to accommodate level changes due to thermal expansion of the coolant and level changes elsewhere in the loop due to flow caused by operation of the pump. 17. The system of any preceding embodiment, wherein the isolation valve comprises a gate valve to seal and isolate the hot-leg pipe and reactor vessel from the hot-well vessel. 18. The system of any preceding embodiment, further comprising: a standpipe to allow for reentry of cold coolant into the reactor vessel through a cold-leg pipe entering through a cold-leg penetration of the outer containment wall; wherein the standpipe has a liquid free surface having a cover gas maintained at a nearly constant pressure to prevent high-pressures from being transmitted into the reactor vessel. 19. The system of any preceding embodiment, wherein the standpipe comprises a substantially neutrally buoyant plug configured to displace a significant portion of the volume of coolant above the cold-leg penetration. 20. The system of any preceding embodiment, wherein the standpipe comprises a gate valve to seal a cold leg pipe entering the reactor vessel. 21. The system of any preceding embodiment, wherein the standpipe comprises a seal loop to prevent the cold leg from draining when the standpipe is drained. 22. The system of any preceding embodiment, further comprising: a second hot-leg pipe coupled to the reactor vessel at a second hot-leg penetration of the outer containment wall; and wherein the second hot-leg pipe provides a coolant flow path with a second hot-well vessel coupled to the second hot-leg, the second hot-well vessel having a liquid free surface near the second hot-leg penetration. 23. The system of any preceding embodiment, wherein the hot-well vessel and second hot-well vessel are connected by a conduit to allow flow between the hot-well vessel and second hot-well vessel and to maintain a nearly equal liquid level between the hot-well vessel and second hot-well vessel. 24. The system of any preceding embodiment, wherein one or more of the hot-well vessel and second hot-well vessel comprises a gate valve to seal and isolate the hot-leg pipe and reactor vessel from either the hot-well vessel or second hot-well vessel. 25. The system of any preceding embodiment, wherein the conduit is located above the elevation of a crossover leg allowing coolant to exit either the hot-well vessel or second hot-well vessel. 26. The system of any preceding embodiment, further comprising: a plurality of standpipes to allow for reentry of cold coolant into the reactor vessel through a plurality of cold-leg pipes entering through a cold-leg penetration; and wherein each standpipe has a liquid free surface having a cover gas maintained at a nearly constant pressure to prevents high-pressures from being transmitted into the reactor vessel. 27. The system of any preceding embodiment, wherein each standpipe comprises a substantially neutrally buoyant plug configured to displace a significant fraction of the volume of coolant above the penetration into containment. 28. The system of any preceding embodiment, wherein each standpipe comprises gate valve to seal each cold leg pipe entering the reactor vessel. 29. The system of any preceding embodiment, wherein each standpipe comprises a seal loop to prevent each cold leg from draining when the standpipe is drained. 30. A high-temperature containment-isolation system for an HTR, comprising: a reactor vessel disposed inside a reactor cavity having an outer containment wall; a standpipe to allow for reentry of cold coolant into the reactor vessel through a cold-leg pipe entering through a cold-leg penetration of the outer containment wall; and wherein the standpipe has a liquid free surface having a cover gas maintained at a nearly constant pressure to prevent high-pressures from being transmitted into the reactor vessel. 31. The system of any preceding embodiment, wherein the standpipe comprises a substantially neutrally buoyant plug configured to displace a significant portion of the volume of coolant above the cold-leg penetration. 32. The system of any preceding embodiment, wherein the standpipe comprises a gate valve to seal a cold leg pipe entering the reactor vessel. 33. The system of any preceding embodiment, wherein the standpipe comprises a seal loop to prevent the cold leg from draining when the standpipe is drained. 34. The system of any preceding embodiment, further comprising: a hot-leg pipe coupled to the reactor vessel at a hot-leg penetration of the outer containment wall; wherein the hot-leg pipe provides a coolant flow path having a liquid free surface near the hot-leg penetration; and wherein the liquid free surface has a cover gas maintained at a nearly constant pressure and thus prevents high-pressures from being transmitted into the reactor vessel. 35. The system of any preceding embodiment, further comprising: an isolation valve and a hot-well vessel coupled to the hot-leg coolant flow path; and wherein hot coolant exiting the reactor vessel flows through the isolation valve into the hot-well vessel. 36. The system of any preceding embodiment, wherein the hot-well vessel provides a substantial free liquid surface and a submerged centrifugal pump that circulates the hot coolant through a crossover-leg. Although the description herein contains many details, these should not be construed as limiting the scope of the disclosure but as merely providing illustrations of some of the presently preferred embodiments. Therefore, it will be appreciated that the scope of the disclosure fully encompasses other embodiments which may become obvious to those skilled in the art. In the claims, reference to an element in the singular is not intended to mean “one and only one” unless explicitly so stated, but rather “one or more.” All structural, chemical, and functional equivalents to the elements of the disclosed embodiments that are known to those of ordinary skill in the art are expressly incorporated herein by reference and are intended to be encompassed by the present claims. Furthermore, no element, component, or method step in the present disclosure is intended to be dedicated to the public regardless of whether the element, component, or method step is explicitly recited in the claims. No claim element herein is to be construed as a “means plus function” element unless the element is expressly recited using the phrase “means for.” No claim element herein is to be construed as a “step plus function” element unless the element is expressly recited using the phrase “step for.” |
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042630967 | claims | 1. In a device having a toroidal plasma containment vessel with walls confining a toroidal region encircling a central opening through the vessel, a toroidal coil system for producing a toroidal magnetic field in the vessel, said toroidal coil system comprising fixed field generating linking coil means comprising at least one linking coil disposed in fixed relation to said vessel with the coil linking the vessel, each linking coil being formed of first and second sections with the first section passing through said central opening through the vessel and the second section completing the linking coil to link the vessel, a plurality of movable unlinked field correcting coils each formed of first and second C-shaped sections joined to each other at their open ends with their bights spaced apart, means for removably mounting the second C-shaped section of each movable coil adjacent said second section of one of said linking coils, with the plasma containment vessel disposed between the open ends of the respective first and second C-shaped sections, with neither C-shaped section passing through said central opening through the plasma containment vessel to link the vessel, and means for passing electric current through said linking and movable coils in opposite sense in the respective adjacent second sections. 2. A toroidal coil system according to claim 1 wherein with said second C-shaped sections mounted adjacent said second section of a linking coil, said first C-shaped sections are substantially evenly distributed about the periphery of said vessel, and the net magnetic field strength at said adjacent second sections approximates the magnetic field strength at each of said first C-shaped sections when said means for passing electric current is energized. 3. A toroidal coil system according to claim 2 wherein said linking coils are disposed at respective azimuthal positions 180.degree. apart and said sections of said unlinked coils are substantially evenly distributed therebetween. 4. A toroidal coil system according to claim 2 wherein said first sections of said linking coils are substantially evenly distributed around said central opening. |
description | The present application is a continuation-in-part of U.S. patent application Ser. No. 14/534,391, filed Nov. 6, 2014, which is a continuation of U.S. patent application Ser. No. 13/208,915, filed Aug. 12, 2011, which in turn claims the benefit of U.S. Provisional Patent Application Ser. No. 61/373,138, filed Aug. 12, 2010. The present application is also a continuation-in-part of U.S. patent application Ser. No. 14/394,233, filed Oct. 13, 2014, which is a United States national stage application under 35 U.S.C. §371 of PCT Application No. PCT/US2013/036592, filed on Apr. 15, 2013, which claims the benefit of U.S. Provisional Patent Application No. 61/624,066 filed Apr. 13, 2012. The present application is also a continuation-in-part of U.S. patent application Ser. No. 14/395,790, filed Oct. 20, 2014, which is a United States national stage application under 35 U.S.C. §371 of PCT Application No. PCT/US2013/037228, filed on Apr. 18, 2013, which claims the benefit of U.S. Provisional Patent Application 61/625,869, filed Apr. 18, 2012. The present application is also a continuation-in-part of U.S. patent application Ser. No. 14/424,201, filed Feb. 26, 2015, which is a United States national stage application under 35 U.S.C. §371 of PCT Application No. PCT/US2013/057855, filed Sep. 31, 2013, which claims priority to U.S. Provisional Application Ser. No. 61/695,837, filed Aug. 31, 2012. The present application is also a continuation-in-part of U.S. patent application Ser. No. 14/655,860, filed Jun. 26, 2015, which is a United States national stage application under 35 U.S.C. §371 of PCT Application No. PCT/US2013/077852 filed Dec. 26, 2013, which claims priority to U.S. Provisional Application Ser. No. 61/746,094 filed Dec. 26, 2012. The present application is also a continuation-in-part of U.S. patent application Ser. No. 14/762,874, filed Jul. 23, 2015, which is a United States national stage application under 35 U.S.C. §371 of PCT Application No. PCT/US2014/013185, filed Jan. 27, 2014, which claims priority to U.S. provisional application No. 61/756,787, filed Jan. 25, 2013, and to U.S. provisional application No. 61/902,559, filed Nov. 11, 2013. The disclosures of the aforementioned priority applications are incorporated herein by reference in their entireties. The storage, handling, and transfer of high level waste, (hereinafter, “HLW”) such as spent nuclear fuel (hereinafter, “SNF”), requires special care and procedural safeguards. For example, 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 spent nuclear fuel 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, spent nuclear fuel is first placed in a canister. The loaded canister is then transported and stored in large cylindrical containers called casks. A transfer cask is used to transport spent nuclear fuel from location to location while a storage cask is used to store spent nuclear fuel 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. Spent nuclear fuel is loaded into the canister while the canister and transfer cask remain submerged in the pool of water. Once fully loaded with spent nuclear fuel, a lid is typically placed atop the canister while in the pool. The transfer cask and canister are then removed from the pool of water, the lid of the canister is welded thereon and a lid is installed on the transfer cask. The canister is then properly dewatered and filled with 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 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 (or other HLW). VVOs stand above ground and are typically cylindrical in shape and extremely heavy, weighing over 150 tons and often having a height greater than 16 feet. VVOs typically have a flat bottom, a cylindrical body having a cavity to receive a canister of spent nuclear fuel, and a removable top lid. In using a VVO to store spent nuclear fuel, a canister loaded with spent nuclear fuel is placed in the cavity of the cylindrical body of the VVO. Because the spent nuclear fuel 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, and exits the VVO at an elevated temperature through top ventilation ducts. The bottom and top ventilation ducts of existing VVOs are located near the bottom and top of the VVO's cylindrical body respectively. 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 spent nuclear fuel 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 overpacks, must place themselves in close vicinity of the ducts for short durations. Thus, a need exists for a VVO system for the storage of high level radioactive waste that has an inlet duct that reduces the likelihood of radiation exposure while providing extreme radiation blockage of both gamma and neutron radiation emanating from the high level radioactive waste. The effect of wind on the thermal performance of a ventilated system can also be a serious drawback that, to some extent, afflicts all systems in use in the industry at the present time. Storage VVO's with only two inlet or outlet ducts are especially vulnerable. While axisymmetric air inlet and outlet ducts behave extremely well in quiescent air, when the wind is blowing, the flow of air entering and leaving the system is skewed, frequently leading to a reduced heat rejection capacity. The thick top lid is one of the most expensive components of a radioactive waste canister. Such canisters may be used to store and transport non-fuel radioactive waste from nuclear generation plants such as activated reactor internals, control components, sundry non-fissile materials, and waste from operations such as resins, and in some applications vitrified nuclear waste fuel (“glass logs”) encased in an outer metal cylinder. On existing canisters, the thick top lid is needed to shield personnel from radiation who are working on the lid (e.g. welding, bolting, fluid operations, etc.). The lid must also be thicker because the lid further performs the main canister lifting connection, and therefore must have the thickness needed for structural reasons to support the weight of the entire canister when hoisted via a crane or similar equipment used to move the canister. For these reasons, the thick top lid of a waste canister adds considerably to the overall weight and expense of the canister. An improved radioactive waste canister is desired. A need also exists periodic leak testing is often required for monitoring the integrity of the inner and outer confinement boundaries on canisters holding radioactive materials. Some present leak testing processes involve removing the cask lid, which is undesirable, as doing so has the potential to increase radiation exposure to workers. Other leak testing processes and systems involve installing a continuous leak testing monitoring system that uses a compressed helium tank and pressure transducers. Such a system, however, requires periodic replacement of the transducers and replenishment of the helium gas stored in the tank. In view of the shortcomings of present leak detection processes and systems, improvements are desirable which reduce the on-site maintenance requirements, improve leak detection capabilities, and reduce potential radiation exposure to workers. A need also exists for the ability to better examine welds formed on containers that are used to store spend nuclear fuel. Finally, a need exists to better enable spent nuclear fuel to be transferred from place to place as necessary. These, and other drawbacks, are remedied by the present invention. In one embodiment, the invention can be a system for storing high level radioactive waste comprising: an overpack body extending along a vertical axis and having a cavity for storing high level radioactive waste, the cavity having an open top end and a floor; an overpack lid positioned atop the overpack body to enclose the open top end of the cavity; an air inlet vent for introducing cool air into the cavity, the air inlet vent extending from an opening in an outer surface of the overpack body to an opening in the floor, the opening in the outer surface of the overpack body extending about an entirety of a circumference of the outer surface of the overpack body; and an air outlet vent in the overpack lid for removing warmed air from the cavity. In another embodiment, the invention can be a system for storing high level radioactive waste comprising: an overpack body extending along a vertical axis and having a cavity for storing high level radioactive waste, the cavity having an open top end and a floor, the overpack body comprising an air inlet vent for introducing cool air into a bottom portion of the cavity; a plurality of plates disposed within a portion of the air inlet vent, each of the plates extending along a reference line that is tangent to a third reference circle having a center point coincident with the vertical axis; and an overpack lid positioned atop the overpack body to enclose the open top end of the cavity, the overpack lid comprising an air outlet vent for removing warmed air from the cavity. In yet another embodiment, the invention can be a system for storing high level radioactive waste comprising: an overpack body extending along a vertical axis and having a cavity for storing high level radioactive waste, the cavity having an open top end and a floor, the overpack body comprising an air inlet vent for introducing cool air into a bottom portion of the cavity; an overpack lid positioned atop the overpack body to enclose the open top end of the cavity, the overpack lid comprising an air outlet vent for removing warmed air from a top portion of the cavity; and the air inlet vent comprising a first section that extends substantially horizontally from an outer surface of the overpack body to a terminal end and a second section extending from the first section of the air inlet vent to an opening in the floor at an oblique angle relative to the vertical axis. In still another embodiment, the invention can be a radioactive waste container system comprising: a canister having an interior chamber for holding radioactive waste and an open top; a lid assembly comprising a confinement lid and a shielded lifting lid, the confinement lid being detachably mounted to the lifting lid; the confinement lid being configured for mounting on the canister and having a first thickness; the lifting lid including a lifting attachment and having a second thickness; wherein the confinement lid is independently mountable on canister from the lifting lid. In still a further embodiment, the invention can be a radioactive waste container system comprising: a canister having an interior chamber for holding radioactive waste and an open top; a lid assembly comprising a lower confinement lid and an upper shielded lifting lid, the confinement lid being detachably bolted to the lifting lid; the lifting lid including a plurality of first bolt holes having a first diameter and a plurality of second bolt holes having a second diameter, the first diameter being larger than the second diameter; the confinement lid including a plurality of third bolt holes having a third diameter, wherein each of the third bolt holes is concentrically aligned with one of the first or second bolt holes of the lifting lid; and a plurality of first mounting bolts inserted through the first bolt holes and threadably attaching the confinement lid to the canister without engaging the lifting lid. In a yet further embodiment, the invention can be a method for storing radioactive waste using a container system, the method comprising: detachably mounting a confinement lid to a shielded lifting lid, the confinement lid and shielded lifting lid collectively forming a lid assembly; placing a canister having an interior chamber for holding radioactive waste into an outer protective overpack; lifting the lid assembly using the lifting lid; placing the lid assembly on an open top of the canister; attaching the confinement lid to the canister using a first set of mounting bolts without threadably engaging the lifting lid with the bolts; detaching the lifting lid from the confinement lid; and removing the lifting lid from the canister. In another embodiment, the invention can be 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, 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 inlet ducts, each inlet duct connecting at least one of the inlet passageways to ambient atmosphere and each comprising an inlet duct cover affixed over a surrounding inlet wall, the inlet wall being peripherally perforated; and a 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. In still another embodiment, the invention can be a system for storing radioactive materials, the system comprising: a canister comprising: a first hermetically sealed vessel having a first cavity; a second hermetically sealed vessel having a second cavity, wherein the first vessel is positioned in the second cavity; an interstitial space between the first and second vessels; and a test port through the second vessel in fluidic communication with the interstitial space; a conduit having a first end fluidically coupled to the test port; and a removable seal operably coupled to a second end of the conduit. In yet another embodiment, the invention can be a method of storing radioactive materials, the method comprising: a) providing a cask having a cask body that forms a cask cavity having an open top end; b) positioning a canister loaded with the radioactive materials in the cask cavity, the canister comprising a first hermetically sealed vessel having a first cavity in which the radioactive materials are disposed and a second hermetically sealed vessel having a second cavity, wherein the first vessel is positioned in the second cavity, such that an interstitial space exists between the first and second vessels, and wherein the second vessel includes a test port that is in fluidic communication with the interstitial space; c) fluidically coupling a first end of a conduit to the test port, the conduit extending from the first end to a second end located outside of the cask; and d) securing a cask lid to the cask body to substantially enclose the open top end of the cask cavity. In another embodiment still, the invention can be a system for leak testing a canister containing radioactive materials, the system comprising: a canister comprising: a first hermetically sealed vessel having a first cavity; a second hermetically sealed vessel having a second cavity, wherein the first vessel is positioned in the second cavity; an interstitial space between the first and second vessels; and a test port through the second vessel in fluidic communication with the interstitial space; a conduit having a first end fluidically coupled to the test port; a removable seal operably coupled to a second end of the conduit; and a leak detector configured to operably couple to the second end of the conduit and to detect whether a leak exists in at least one of the first vessel and the second vessel. In a further embodiment, the invention can be a method of leak testing a storage canister for radioactive materials, the method comprising: a) positioning the canister in a cask cavity of a cask body, the canister comprising a first hermetically sealed vessel having a first cavity in which the radioactive materials are disposed and a second hermetically sealed vessel having a second cavity, the first vessel positioned in the second cavity such that an interstitial space exists between the first and second vessels, and wherein the second vessel includes a test port that is in fluidic communication with the interstitial space; b) coupling a first end of a conduit to the test port, the conduit extending from the first end to a second end located outside of the cask body; c) securing a cask lid to the cask body to substantially enclose the cask cavity; and d) operatively coupling a leak detector to the second end of the conduit to perform a leak test comprising determining whether a leak exists in at least one of the first vessel and the second vessel In a still further embodiment, the invention can be a method of leak testing a canister containing radioactive materials, the method comprising: a) coupling a first end of a conduit to a test port of the canister that is in fluid communication with an interstitial space of the canister, the conduit extending from the first end to a second end; and b) operatively coupling a leak detector to the second end; c) drawing gas from the conduit using the leak detector to establish a vacuum within the conduit and the interstitial space; and d) monitoring the drawn gas for the presence of a first indicator which is representative of a leak in a fluidic containment boundary of the canister that contains the radioactive materials. In another embodiment, the invention can be a canister for storing radioactive materials, the canister comprising: a base plate; a side wall having a bottom sealed to the base plate; and a top plate including a top surface with a top edge having a bevel and with a channel set in from the top edge, wherein a weld is formed between the beveled top edge and a top of the side wall to seal the top plate to the side wall, and wherein the base plate, side wall, and top plate form a sealed vessel. In another embodiment, the invention can be a method of forming a sealed canister, the method comprising: placing a top plate on a top opening of a side wall, a bottom of the side wall being sealed to a base plate, wherein the top plate includes a top surface with a top edge having a bevel and with a channel set in from the top edge; and forming a weld between the beveled top edge and the top opening of the side wall to seal the top plate to the side wall. In another embodiment, still, the invention can be a method of storing radioactive materials, the method comprising: placing radioactive materials in a cavity formed by a side wall having a bottom sealed to a base plate; placing a top plate on a top opening of the side wall, the top plate including a top surface with a top edge having a bevel and with a channel set in from the top edge; forming a weld between the beveled top edge and the top opening of the side wall to seal the top plate to the side wall, so that the cavity is sealed; placing a first probe in the channel and a second probe opposite the first probe and adjacent the side wall, such that the weld is disposed between the two probes; activating the first and second probes to determine an integrity of a volume of the weld between the probes; and moving the first and second probes synchronously around the top plate to determine the integrity of an entire volume of the weld. In another embodiment, the invention can be an apparatus for transferring spent nuclear fuel, the apparatus comprising: a cylindrical inner shell forming a cavity configured to receive a canister containing spent nuclear fuel, the cavity configured so that an annulus is formed between a canister placed in the cavity and an inner wall of the cylindrical inner shell; an intermediate shell disposed concentrically around and spaced apart from the inner shell; an outer shell disposed concentrically around and spaced apart from the intermediate shell; a bottom flange affixed to bottoms of each of the shells; a bottom lid removably affixed to the bottom flange and including at least one first channel fluidically connecting the annulus to an exterior of the bottom lid, wherein the at least one first channel is configured to preclude a direct line of travel from within the cavity to the exterior of the bottom lid; a top flange affixed to tops of each of the shells and including at least one second channel fluidically connecting the first annulus to an exterior of the top flange, wherein the at least one second channel is configured to preclude a direct line of travel from within the cavity to the exterior of the top flange; and a top lid removably affixed to the top flange. In yet another embodiment, the invention can be an apparatus for transferring spent nuclear fuel, the apparatus comprising: a cylindrical inner shell forming a cavity configured to receive a canister containing spent nuclear fuel; an intermediate shell disposed concentrically around and spaced apart from the inner shell; an outer shell disposed concentrically around and spaced apart from the intermediate shell; a bottom flange affixed to bottoms of each of the shells; a bottom lid removably affixed to the bottom flange; a top flange affixed to tops of each of the shells, the top flange including at least two integrally formed trunnions configured to enable hoisting of the apparatus; and a top lid removably affixed to the top flange. In still another embodiment, the invention can be an apparatus for transferring spent nuclear fuel, the apparatus comprising: a cylindrical inner shell forming a cavity configured to receive a canister containing spent nuclear fuel; an intermediate shell disposed concentrically around and spaced apart from the inner shell; an outer shell disposed concentrically around and spaced apart from the intermediate shell; a bottom flange affixed to bottoms of each of the shells; a bottom lid removably affixed to the bottom flange, the bottom lid including an impact zone comprising an impact absorbing structure; a top flange affixed to tops of each of the shells; and a top lid removably affixed to the top flange. In another embodiment, the invention can be a method for transferring spent nuclear fuel from a pool, the method comprising: lifting a transfer cask from a pool, the transfer cask comprising: a cylindrical inner shell forming a cavity configured to receive a canister containing spent nuclear fuel, the cavity configured so that an annulus is formed between a canister placed in the cavity and an inner wall of the cylindrical inner shell; an intermediate shell disposed concentrically around and spaced apart from the inner shell; an outer shell disposed concentrically around and spaced apart from the intermediate shell; a bottom flange affixed to bottoms of each of the shells; a bottom lid removably affixed to the bottom flange and including at least one first channel fluidically connecting the annulus to a channel inlet at an exterior of the bottom lid, wherein the at least one first channel is configured to preclude a direct line of travel from within the cavity to the exterior of the bottom lid; a removable plug sealingly affixed to the channel inlet; a top flange affixed to tops of each of the shells and including at least one second channel fluidically connecting the first annulus to an exterior of the top flange, wherein the at least one second channel is configured to preclude a direct line of travel from within the cavity to the exterior of the top flange; and a top lid removably affixed to the top flange; removing the removable plug from the channel inlet, thereby allowing ambient air to enter the at least one first channel; draining the pool water from the canister; and moving the transfer cask to a staging area. Further areas of applicability of the present invention will become apparent from the detailed description provided hereinafter. It should be understood that the detailed description and specific examples, while indicating the preferred embodiment of the invention, are intended for purposes of illustration only and are not intended to limit the scope of the invention. All drawings are schematic and not necessarily to scale. Parts given a reference numerical designation in one figure may be considered to be the same parts where they appear in other figures without a numerical designation for brevity unless specifically labeled with a different part number and described herein. The following description of the preferred embodiment(s) is merely exemplary in nature and is in no way intended to limit the invention, its application, or uses. The description of illustrative embodiments according to principles of the present invention is intended to be read in connection with the accompanying drawings, which are to be considered part of the entire written description. In the description of embodiments of the invention disclosed herein, any reference to direction or orientation is merely intended for convenience of description and is not intended in any way to limit the scope of the present invention. Relative terms such as “lower,” “upper,” “horizontal,” “vertical,” “above,” “below,” “up,” “down,” “top,” and “bottom” as well as derivatives thereof (e.g., “horizontally,” “downwardly,” “upwardly,” etc.) should be construed to refer to the orientation as then described or as shown in the drawing under discussion. These relative terms are for convenience of description only and do not require that the apparatus be constructed or operated in a particular orientation unless explicitly indicated as such. Terms such as “attached,” “affixed,” “connected,” “coupled,” “interconnected,” and similar refer to a relationship wherein structures are secured or attached to one another either directly or indirectly through intervening structures, as well as both movable or rigid attachments or relationships, unless expressly described otherwise. Moreover, the features and benefits of the invention are illustrated by reference to the exemplified embodiments. Accordingly, the invention expressly should not be limited to such exemplary embodiments illustrating some possible non-limiting combination of features that may exist alone or in other combinations of features; the scope of the invention being defined by the claims appended hereto. Multiple inventive concepts are described herein and are distinguished from one another using headers in the description that follows. Specifically, FIGS. 1-8 are relevant to a first inventive concept, FIGS. 9-26 are relevant to a second inventive concept, FIGS. 27-34 are relevant to a third inventive concept, FIGS. 35-47 are relevant to a fourth inventive concept, FIGS. 48-52B are relevant to a fifth inventive concept, and FIGS. 53-59 are relevant to a sixth inventive concept. The first through sixth inventive concepts should be considered in isolation from one another. It is possible that there may be conflicting language or terms used in the description of the first through sixth inventive concepts. For example, it is possible that in the description of the first inventive concept a particular term may be used to have one meaning or definition and that in the description of the second inventive concept the same term may be used to have a different meaning or definition. In the event of such conflicting language, reference should be made to the disclosure of the relevant inventive concept being discussed. Similarly, the section of the description describing a particular inventive concept being claimed should be used to interpret claim language when necessary. I. Inventive Concept 1 With reference to FIGS. 1-8, a first inventive concept will be described. Referring to FIGS. 1-4 concurrently, a system for storing high level radioactive waste will be described in accordance with an embodiment of the present invention. The system can be considered a VVO 100. The VVO 100 is a vertical, ventilated dry spent fuel storage system that is fully compatible with 100 ton and 125 ton transfer casks for spent fuel canister operations. Of course, the VVO 100 can be modified/designed to be compatible with any size or style transfer cask. The VVO 100 is designed to accept spent fuel canisters for storage. All spent fuel canister types engineered for storage in free-standing and anchored overpack models can be stored in VVO 100. As used herein the term “canister” broadly includes any spent fuel containment apparatus, including, without limitation, multi-purpose canisters and thermally conductive casks. For example, in some areas of the world, spent fuel is transferred and stored in metal casks having a honeycomb grid-work/basket built directly into the metal cask. Such casks and similar containment apparatus qualify as canisters, as that term is used herein, and can be used in conjunction with VVO 100 as discussed below. In certain embodiments, the VVO 100 is a substantially cylindrical containment unit having a vertical axis A-A and a horizontal cross-sectional profile that is substantially circular in shape. Of course, it should be understood that the invention is not limited to cylinders having circular horizontal cross sectional profiles but may also include containers having cross-sectional profiles that are, for example, rectangular, ovoid or other polygon forms. While the VVO 100 is particularly useful for use in conjunction with storing and/or transporting SNF assemblies, the invention is in no way limited by the type of waste to be stored. The VVO cask 100 can be used to transport and/or store almost any type of HLW. However, the VVO 100 is particularly suited for the transport, storage and/or cooling of radioactive materials that have a high residual heat load and that produce neutron and gamma radiation, such as SNF. This is because the VVO 100 is designed to both provide extreme radiation blockage of gamma and neutron radiation and facilitate a convective/no force cooling of any canister contained therein. The VVO 100 of the present invention generally comprises an overpack body 110 for storing high level radioactive waste and a removable overpack lid 120 that is positioned atop the overpack body 110. The overpack body 110 extends along the vertical axis A-A. The overpack lid 120 generally comprises a primary lid 121 and a secondary lid 122. The primary lid 121 is secured to the overpack body 110 by bolts 123 that restrain separation of the primary lid 121 of the overpack lid 120 from the overpack body 110 in case of a tip over situation. Moreover, the secondary lid 122 is secured to the primary lid 121 by bolts 124. The overpack lid 120 is a steel/concrete structure that is equipped with an axisymmetric air outlet vent or passageway 145 for the ventilation/removal of air as will be discussed in more detail below. An annular opening 157 is formed in an outer sidewall surface 178 of the overpack lid 120 that forms a passageway from the air outlet vent 145 to the external environment. More specifically, the annular opening 157 is a 360° opening in the outer sidewall surface 178 of the overpack lid 120. The overpack lid 120 has a quick connect/disconnect joint to minimize human activity for its installation or removal. In certain embodiments, the overpack lid 120 may weigh in excess of 15 tons. The VVO 100 further comprises shock absorber or crush tubes 102 in its top region. The shock absorber tubes 102 are arranged at suitable angular spacings to serve as a sacrificial crush material if, for any reason, the VVO 100 were to tip over. The shock absorber tubes 102 also facilitate guiding and positioning of a canister within a cavity 111 of the VVO 100 in a substantially concentric disposition with respect to the VVO 100. Referring to FIGS. 1, 4 and 6 concurrently, the overpack body 110 comprises a cylindrical wall 112, a bottom enclosure plate 130 and the overpack lid 120 described above. The cylindrical wall 112 has an inner shell 113, an intermediate shell 114 and an outer shell 115. In the exemplified embodiment, each of the inner, intermediate and outer shells 113, 114, 115 are formed of one-inch thick steel. Of course, the invention is not to be so limited and in other embodiments the inner, intermediate and outer shells 113, 114, 115 can be formed of metals other than steel and can be greater or less than one-inch in thickness. The inner shell 113 has an inner surface 116 that defines an internal cavity 111 for containing a hermetically sealed canister that contains high level radioactive waste (FIG. 5). The inner surface 116 of the inner shell 113 also forms the inner wall surface of the overpack body 110. Furthermore, the outer shell 115 has an outer surface 117. The outer surface 117 of the outer shell 115 also forms the outer sidewall surface of the overpack body 110. In the exemplified embodiment, the inner, intermediate and outer shells 113, 114, 115 are concentric shells that are rendered into a monolithic weldment by a plurality of connector plates 105a, 105b. The inner shell 113 is spaced from the intermediate shell 114 by connector plates 105a and the intermediate shell 114 is spaced from the outer shell 115 by connector plates 105b. Of course, in certain other embodiments the connector plates 105a, 105b can be altogether omitted. The space between the inner shell 113 and the intermediate shell 114 is intended for placement of a neutron shielding material. For example, in certain embodiments the neutron radiation shielding material is a hydrogen-rich material, such as, for example, Holtite, water or any other material that is rich in hydrogen and a Boron-10 isotope. In certain embodiments, there is approximately seven inches of Holtite filling the space between the inner and intermediate shells 113, 114. Thus, the space between the inner and intermediate shells 113, 114 serves to prevent neutron radiation from passing through the VVO 100 and into the external environment. An axially intermediate portion of the space between the intermediate shell 114 and the outer shell 115 is filled with a heavy shielding concrete to capture and prevent the escape of both gamma and neutron radiation. The density of the concrete is preferably maximized to increase the radiation absorption characteristics of the VVO 100. In certain embodiments, there is approximately twenty-eight inches of concrete filling the intermediate portion of the space between the intermediate and outer shells 114, 115. In some embodiments, steel plates are placed within the concrete to serve as a supplemental radiation curtain. There are no lateral penetrations in the multi-shell weldment that may provide a streaming path for the radiation issuing from the high level radioactive waste. The top and bottom portions of the space between the intermediate and outer shells 114, 115 (both above and below the concrete) are top and bottom forgings 128, 129 in the form of thick annular rings made of a metal material, such as steel. The top forging 128 comprises machine threaded holes 126 that are sized and configured to receive the bolts 123 of the primary lid 121 therein during attachment of the overpack lid 120 to the overpack body 110. As noted above, the inner surface 116 of the inner shell 113 defines the cavity 111. In the exemplified embodiment, the cavity 111 is cylindrical in shape. However, the cavity 111 is not particularly limited to any specific size, shape, and/or depth, and the cavity 111 can be designed to receive and store almost any shape of canister. In certain embodiments, the cavity 111 is sized and shaped so that it can accommodate a canister of spent nuclear fuel or other HLW. More specifically, the cavity 111 has a horizontal cross-section that can accommodate no more than one canister. Even more specifically, it is desirable that the size and shape of the cavity 111 be designed so that when a spent fuel canister is positioned in the cavity 111 for storage, a small clearance exists between outer side walls of the canister and the inner surface 116 of the inner shell 113, as will be discussed in more detail below with reference to FIG. 5. Referring to FIGS. 4 and 5 concurrently, the present invention will be further described. The cavity 111 comprises a floor 152 and an open top end 151 that is enclosed by the overpack lid 120 as has been described herein above. A plurality of support blocks 153 are disposed on the floor 152 of the cavity 111 to support a canister 200 contained within the cavity 111 above the floor 152. In the exemplified embodiment, four support blocks 153 are illustrated (see FIG. 6). However, more or less than four support blocks 153 can be used in alternate embodiments. Each of the support blocks 153 is a low profile lug that is welded to the inner surface 116 of the inner shell 113 and/or to the floor 152. In the exemplified embodiment, the canister 200 is a hermetically sealed canister for containing the high level radioactive waste. When the canister 200 is positioned within the cavity 111, it rests atop the support blocks 153 so that a space 154 exists between a bottom 202 of the canister 200 and the floor 152. The space 154 is a bottom plenum that serves as the recipient of ventilation air flowing up from an inlet vent as will be described below. Furthermore, when the canister 200 is positioned within the cavity 111, an annular gap 155 exists between the inner surface 116 of the inner shell 113 (i.e., the inner wall surface of the overpack body 110) and an outer surface 201 of the canister 200. The annular gap 155 is an uninterrupted and continuous gap that circumferentially surrounds the canister 200. In other words, the canister 200 is concentrically spaced apart from the inner shell 113, thereby creating the annular gap 155. As described in more detail below, the annular gap 155 forms an annular air flow passageway between an annular air inlet passageway 142 and the air outlet vent 145. The VVO 100 is configured to achieve a cyclical thermosiphon flow of gas (i.e., air) within the cavity 111 when spent nuclear fuel emanating heat (i.e., the canister 200) is contained therein. In other words, the VVO 100 achieves a ventilated flow by virtue of a chimney effect. Such cyclical thermosiphon flow of the gas further enhances the transmission of heat to the environment external to the VVO 100. The thermosiphon flow of gas is achieved as a result of an air inlet vent 140 that introduces cool air into the bottom of the cavity 111 of the overpack body 110 from the external environment and an air outlet vent 145 for removing warmed air from the cavity 111. Thus, as a result of thermosiphon flow, cool external air can enter into the space 154 of the cavity 111 between the bottom 202 of the canister 200 and the floor 152 via the air inlet vent 140, flow upward through the cavity 111 within the annular gap 155 between the canister 200 and the inner surface 116 of the inner shell 113, and flow back out into the external environment as warmed air via the air outlet vent 145. The newly entered air will warm due to proximity to the extremely hot canister 200, which will cause the natural thermosiphon flow process to take place whereby the heated air will continually flow upwardly as fresh cool air continues to enter into the cavity 111 via the air inlet vent 140. Thus, the air inlet vent 140 provides a passageway that facilitates cool air entering the cavity 111 from the external environment and the air outlet vent 145 provides a passageway that facilitates warm air exiting the cavity back to the external environment. In the exemplified embodiment, the air outlet vent 145 is formed into the overpack lid 120. The air outlet vent 145 provides an annular passageway from a top portion of the cavity 111 to the external environment when the overpack lid 120 is positioned atop the overpack body 110 thereby enclosing the top end 151 of the cavity 111. Specifically, the air outlet vent 145 has a vertical section 174 that extends from the cavity 111 upwardly into the overpack lid 120 in the vertical direction (i.e., the direction of the vertical axis A-A) and a horizontal section 175 that extends from the vertical section 174 to the annular opening 157 in the horizontal direction (i.e., the direction transverse to the vertical axis A-A). More specifically, the vertical section 174 of the air outlet vent 145 extends from an annular opening 176 in a bottom surface 177 of the overpack lid 120 and the horizontal section 175 extends from the vertical section 174 to the annular opening 157 in the outer sidewall surface 178 of the overpack lid 120. As described above, the annular opening 157 is a circumferential opening that extends around the entirety of the overpack lid 120 in a continuous and uninterrupted manner and circumferentially surrounds the vertical axis A-A. The overpack body 110 additionally comprises a bottom block 160 disposed within the cylindrical wall 112, and more specifically within the inner shell 113 of the cylindrical wall 112, and a base structure at a bottom end 179 of the cylindrical wall 112. The base structure comprises a base plate 161 and an annular plate 162. The air inlet vent 140 is formed directly into the bottom block 160, which is a thick sandwich of steel and concrete. The bottom block 160 is positioned below the floor 152 of the cavity 111. More specifically, the bottom block 160 extends between the floor 152 of the cavity 111 and the base plate 161, which forms the bottom end of the VVO 100. The bottom block 160 has a columnar portion 163 and a horizontal portion 164. The annular plate 162 is a donut-shaped plate having a central hole 181. The annular plate 162 is axially spaced from the base plate 161, thereby creating a space or gap in between the annular plate 162 and the base plate 161. Moreover, the annular plate 162 extends from the outer surface 117 of the overpack body 110 inwardly towards the vertical axis A-A a radial distance that is less than the radius of the overpack body 110. More specifically, the annular plate 162 extends from the outer surface 117 of the overpack body 110 to the columnar portion 163 of the bottom block 160. Thought of another way, the columnar portion 163 of the bottom block 160 extends through the central hole 181 of the annular plate 162 and rests atop the base plate 161. Referring to FIGS. 1, 4, 6 and 8 concurrently, the air inlet vent 140 will be described in more detail. In the exemplified embodiment, the air inlet vent 140 is formed into the bottom closure plate 130 and extends into the bottom block 160 and comprises an annular air inlet plenum 141 and an annular air inlet passageway 142. The annular air inlet plenum 141 is formed in the space/gap between the annular plate 162 and the base plate 161. Thus, the annular air inlet plenum 141 is substantially horizontal and extends radially inward from the outer surface 117 of the overpack body 110. More specifically, the annular air inlet plenum 141 extends horizontally from the outer surface 117 of the overpack body 110 at an axial height below the floor 152 of the cavity 111. An opening 143 is formed in the outer surface 117 of the overpack body 110 that forms a passageway from the external environment to the annular air inlet plenum 141 to enable cool air to enter into the annular air inlet plenum 141 from the external environment as has been described above. The opening 143 circumferentially surrounds the vertical axis A-A around the entirety of the outer surface 117 of the overpack body 110 in an uninterrupted and continuous manner. In other words, the opening 143 is a substantially 360° opening in the outer surface 117 of the overpack body 110. The annular air inlet passageway 142 extends upward from a top surface 144 of the annular air inlet plenum 141 to the floor 152 of the cavity 111. More specifically, the annular air inlet passageway 142 extends upwardly from an opening 147 in the top surface 144 of the annular air inlet plenum 141 to an opening 146 in the floor 152. The annular air inlet passageway 142 is wholly formed within the bottom block 160. The opening 147 in the top surface 144 of the annular air inlet plenum 141 is proximate an end of the annular air inlet plenum opposite the opening 143 in the outer surface 117 of the overpack body 110. The opening 146 in the floor 152 is an annular opening that extends 360° around the floor 152. The annular air inlet plenum 141 circumferentially surrounds the vertical axis A-A. In the exemplified embodiment, the annular air inlet passageway 142 also circumferentially surrounds the vertical axis A-A and has an inverted truncated cone shape. Thus, the annular air inlet passageway 142 extends upward from the air inlet plenum 141 to the opening 146 in the floor 152 of the cavity 111 at an oblique angle relative to the vertical axis A-A. Thought of another way, the annular inlet passageway 142 extends from the air inlet plenum 141 at a first end 183 to the floor 152 at a second end 184. The first end 183 is located a first radial distance R1 from the vertical axis A-A and the second end 184 is located a second radial distance R2 from the vertical axis A-A. The second radial distance R2 is greater than the first radial distance R1. Of course, the invention is not to be so limited and in certain other embodiments the annular air inlet passageway 142 can take on other shapes as desired. Referring to FIGS. 1, 4, 7 and 8 concurrently, the annular air inlet plenum 141 will be further described. The annular air inlet plenum 141 comprises a plurality of plates 148 therein. Each of the plates 148 extends from a first end 149 to a second end 159. The first ends 149 of the plates 148 are proximate the outer surface 117 of the overpack body 110 and the second ends 159 of the plates 148 are proximate the columnar portion 163 of the bottom block 160. A line connecting the first ends 149 of the plates 148 forms a first reference circle 171 having a diameter D1 and a line connecting the second ends 159 of the plates 148 forms a second reference circle 172 having a diameter D2, wherein the first diameter D1 is greater than the second diameter D2. Each of the plates 148 in the annular air inlet plenum 141 extend along a reference line 169 that is tangent to a third reference circle 170. Although the reference line 169 is only illustrated with regard to two of the plates 148, it should be understood that each of the plates has a reference line that is tangent to the third reference circle 170. The circumference of the third reference circle 170 is formed by an outer surface 165 of the columnar portion 163 of the bottom block 160. The third reference circle 170 has a center point that is coincident with the vertical axis A-A. In the exemplified embodiment, the plates 148 are thin steel plates that facilitate transferring the weight of the VVO 100 to the base plate 161 and also provide a means to scatter and absorb any errant gamma radiation that may attempt to exit the air inlet plenum. Furthermore, in the exemplified embodiment sixty plates 148 are illustrated. However, the invention is not to be so limited and in certain other embodiments more or less than sixty plates 148 may be disposed within the annular air inlet plenum 141. Due to the axisymmetric configuration of the air inlet plenum 141, the annular air inlet vent 140 is configured so that aerodynamic performance of the air inlet vent 140 is independent of an angular direction of a horizontal component of an air-stream applied to the outer surface 117 of the overpack body 101. Similarly, due to the axisymmetric configuration of the air outlet vent 145, the air outlet vent 145 is configured so that the aerodynamic performance of the air outlet vent 145 is independent of an angular direction of a horizontal component of an air-stream applied to the outer surface 117 of the overpack body 110. II. Inventive Concept 2 With reference to FIGS. 9-26, a second inventive concept will be described. The present invention provides a separate, reusable shielded lifting lid for waste canister lid bolting and lifting. Accordingly, the lifting lid is bolted and not welded to the canister. The canister loading is dry in an overpack such as a metal cylindrical jacket holding the radioactive waste inside. Canisters typically have thick (e.g. 10 inch) steel lids on each canister to protect the operator from radiation during canister closure operations. The thick lids are heavy and expensive, and further not reusable as they remain attached to the canister for longer-term storage. Advantageously, the present invention allows use of a significantly thinner main closure confinement lid (e.g. about 3 to 5-inch thick in exemplary embodiments) for radionuclides containment. After radioactive waste contents are placed in the canister, the confinement lid is installed and held in place by gravity alone in some embodiments. The confinement lid thickness, however, has generally poor radiation shielding value. Accordingly, the confinement lid is installed using a thicker and reusable shielded lifting lid which serves as an upper over-lid to the lower confinement lid. The two-part lid system combination of the confinement lid and shielded lifting lid provide the thickness required to shield the operator from the radioactive canister contents during the canister closure bolting operations. In use, the shielded lifting lid in one exemplary and non-limiting embodiment has holes that match the bolt spacing to allow the operator to install the confinement lid bolts in a radiation shielded environment. After the lifting lid bolts are installed, the operator hooks up the lifting rigging to the shielded lifting lid and moves away from the canister to a more distal and remote location. The shielded lifting lid may then be removed from the top of the canister, preferably with the confinement lid remaining in place, and a heavy overpack lid is installed for longer term storage and radiation shielding. Using this method, the waste canister and overpack advantageously are shorter, lighter, better shielded, and less expensive to fabricate. FIGS. 9 and 10 depict a radioactive canister system according to the present disclosure including a waste canister 1100 having a generally cylindrical body defining an interior chamber 1101 and comprised of a top 1102, bottom 1104, and cylindrical sidewall 1106 extending therebetween. Top 1102 is open for insertion of radioactive waste and bottom 1104 is preferably closed in one embodiment. A main closure confinement lid 1200 is shown attached to top 1102 of canister 1100 by a plurality of fasteners such as mounting bolts 1154 which may be circumferentially spaced apart around the top of the canister, as further described herein. In one embodiment, canister 1100 may be a non-fuel radioactive waste canister (NWC). Referring to FIG. 10, canister 1100 has an interior configured to store the size and shape of radioactive waste to be deposited in the canister. In one embodiment, the canister may include a basket insert 1120 configured for holding a plurality of metal waste cylinders 1121 (see, e.g. FIG. 14) each containing radioactive waste materials. Basket insert 1120 includes a pair of vertically spaced apart top and bottom plates 1122, 1124 which are connected via a plurality of tie rods 1126. Top plate 1122 and bottom plate 1124 include a plurality of horizontally spaced apart circular openings 1123 each having a diameter which is configured and dimensioned to receive waste cylinders 1121 therethrough, as shown in FIG. 14. Referring to FIGS. 10 and 11, the top portion of tie rods 1126 may be threaded for attachment to top plate 1122 by a threaded nut 1125. Top plate 1122 may be spaced by a vertical distance below the top 1102 of canister 1100. Bottom plate 1124 may be elevated by a vertical distance above the bottom 1104 of canister 1100 by a plurality of vertical tubular sleeves 1128 having a bottom end resting on bottom 1104 of the canister 1100 and a top end attached to bottom plate 1124 as better shown in FIG. 12. In one embodiment, sleeves have an inside diameter sized to receive the bottom end portion of tie rods 1126 which are slidably received in the sleeves. This provides for vertical adjustment in the height of the basket insert 1120 to accommodate the height of waste cylinders 1121 to be stored inside canister 1100. Bottom plate 1124 remains fixed and stationary in position. The top plate 1122 with attached tie rods 1126, however, is movable upwards and downwards with respect to the canister and bottom plate 1124 to reach a desired position depending on the height of waste cylinders 1121. In some embodiments, the top plate 1122 may be thereafter be fixed in the desired position after vertical adjustments are made by securing the top plate to the interior of the canister sidewall 1106 such as by welding or other suitable means. Accordingly, adjustable basket insert 1120 may accommodate a variety of waste cylinder heights. Basket insert 1120 (i.e. top plate, bottom plate, tie rods, etc.) may be made of any suitable material, including without limitation a corrosion resistant metal such as stainless steel in one embodiment. FIG. 13 shows canister 1100 loaded into an outer overpack 1130 for transport and storage of radioactive waste. The overpack provides protection during transport and storage of the waste by encapsulating the waste canister in an outer protective jacket. Overpack 1130 has an open top 1132, and is configured and dimensioned to completely receive canister 1100 through the top 1102. Overpack 1130 has an open interior defining an interior surface 1133 and an exterior surface 1135 (see also FIG. 17). Overpack 1130 is generally cylindrical in shape further including a cylindrical sidewall 1134 and flat closed bottom 1136 (see FIG. 23) configured for resting on a flat surface such as concrete slab. Preferably, in one embodiment, overpack 1130 has a greater height than canister 1100 so that the canister is recessed below the open top 1132 of the overpack when fully inserted therein. Overpack 1130 may be made of any suitable material or combination of materials (see, e.g. FIG. 17) which may include neutron absorbing materials such as without limitation concrete, lead, or boron. An example of a suitable overpack for use with canister 1100 may be a HI-SAFE™ transport overpack as used in vertical non-fuel waste storage systems available from Holtec International of Marlton, N.J. The sidewalls 1134 forming the spaced apart cylindrical walls that define an annular space between the inner and outer surfaces 1133 and 1135 respectively may be formed of a corrosion resistant metal also selected for strength to protect the inner canister 1100, such as stainless steel as one non-limiting example. The neutron absorbing material may be disposed between the inner and outer surfaces 1133 and 1135. In some embodiments, overpack 1130 may also include Metamic® for radiation shielding which is a discontinuously reinforced aluminum/boron carbide metal matrix composite material also available from Holtec International. Referring to FIGS. 10-11 and 13, the top of the canister 1100 may include a peripheral contamination boundary seal which cooperates with the confinement lid 1200 to prevent leakage of radiation from the canister, particularly at the lid bolting locations. In particular, the boundary seal shields the mounting blocks 1150 to prevent radiation streaming. In one embodiment, the boundary seal may be configured as an annular shielding flange 1140 that extends circumferentially around the upper peripheral edge of the top 1102 of the canister. Confinement lid 1200 rests on the shielding flange when bolted to the canister 1100. Shielding flange 1140 may be horizontally flat and extend inwards in a direction perpendicular to and from sidewall 1106 towards the vertical axial centerline CL of the canister 1100. In one embodiment, shielding flange 1140 is attached to the uppermost top edge of the sidewall 1106 as shown. Shielding flange 1140 may have an at least partially scalloped configuration in top plan view in some embodiments as shown to accommodate insertion of waste cylinders 1121 into the canister. According, the scallops 1142 if provided are preferably concentrically aligned with the circular openings 1123 in basket insert 1120 in top plan view. This minimizes the required diameter of the canister 1100 for holding the waste cylinders 1121. In other possible embodiments, however, shielding flange 1140 may have an uninterrupted shape forming a continuous ring in top plan view. At the lid bolting locations, shielding flange 1140 is configured to cover a with a plurality of mounting blocks 1150 which are circumferentially spaced around the interior of canister 1100 disposed adjacent to sidewall 1106 to provide a radiation-shielded bolting system for attaching confinement lid 1200 and shielded lifting lid 1300 to the canister. Shielding flange 1140 may be formed of any suitable material including metals which are corrosion resistant such as stainless steel. With continuing reference to FIGS. 10-11 and 13, mounting blocks 1150 may have a generally arcuate and curved shape in top plan view which complements the inside radius of curvature of the sidewall 1106 to which mounting blocks 1150 may be attached. Mounting blocks 1150 may be rigidly/fixedly attached to the canister sidewall 1106 by a suitably strong mechanical connection capable of supporting at least the entire dead weight of canister 1100 and basket insert 1120 for lifting and loading the canister into overpack 1130. Accordingly, in one preferred embodiment, mounting blocks 1150 are welded to at least sidewall 1106 of the canister body for strength. In some embodiments, the mounting blocks 1150 may be abutted against but are not fixedly connected to the underside of radiation shielding flange 1140 so that lifting loads are not transferred to the flange directly but rather bypass the flange to the mounting blocks 1150 via the bolting provided. Any suitable number of mounting blocks 1150 may be provided; the number and circumferential spacing being dependent on the magnitude of the structural load imparted to the blocks dependent on whether the canister 1100 will be lifted in an empty condition or in a fully loaded condition with filled waste cylinders 1121 positioned in the canister. It is well within the ambit of those skilled in the art to determine an appropriate number and circumferential spacing of the mounting blocks 1150. In one embodiment, the mounting blocks 1150 are each configured for both lifting canister 1100 and attaching both the lower confinement lid 1200 and upper lifting lid 1300. As best shown in FIGS. 11 and 17, mounting blocks 1150 each include a plurality of threaded mounting sockets 1152 for forming a threaded connection with complementary threaded mounting bolts 1154 and 1156 used for attaching confinement lid 1200 and shielded lifting lid 1300 respectively to the canister 1100. In one non-limiting example, three threaded mounting sockets 1152 may be provided in each mounting block. However, other suitable numbers of mounting sockets may be used. In certain embodiments, the mounting sockets 1152 extend only partially into the mounting blocks 1150 as shown. Radiation shielding flange 1140 includes mating holes 1144 which are each concentrically aligned with the threaded mounting sockets 1152 of the mounting block to provide access for mounting bolts 1154, 1156 to the mounting sockets in the block. Because shielding flange 1140 in some embodiments in not intended to be a load-bearing member relied upon for lifting the canister, holes 1144 may not be threaded so that the weight of the canister is transferred through the flange via the mounting bolts 1156 to the shielded lifting lid 1300. In one embodiment, mounting bolts 1154 and/or 1156 may be threaded bolts having an integral or separate washer disposed adjacent to the head, as best shown in FIG. 19. Mounting bolts 1154 are used for attaching the lower confinement lid 1200 to canister 1100 via mounting blocks 1150. In one embodiment, mounting bolts 1154 are not used for lifting the canister 1100 but rather for lid securement. By contrast, mounting bolts 1156 serve a dual purpose and may be used for both attaching the lower shielded lifting lid 1300 to canister 1100 and supporting the weight of the canister during lifting operations via mounting blocks 1150 engaged by bolts 1156. In one preferred embodiment, mounting bolts 1156 may have a longer shank than mounting bolts 1154 as shown. This arrangement ensures that the depth of threaded engagement between the threaded mounting sockets 1152 of the mounting blocks 1150 and mounting bolt 1156 is sufficient for lifting the canister 1100, as further explained herein. The confinement lid 1200 is generally circular in shape (top plan view) and shown in FIGS. 8, 17, and 19. Confinement lid 1200 includes a plurality of bolt holes 1202 spaced circumferentially around the peripheral side 1204 of the lid as best shown in FIG. 9 (including at locations where mounting bolts 1154 are shown installed). Bolt holes 1202 penetrate top surface 1206 of the confinement lid, and in one embodiment are not threaded. The bolt holes 1202 may be arranged in groups corresponding to the location and arrangement of the mounting blocks 1150 inside the canister 1100. The bolt holes 1202 have a diameter sized to at least pass the shank of mounting bolts 1154 and 1156 through the holes to threadably engage the mounting blocks 1150. Accordingly, some of the bolt holes 1202 are configured to receive the shanks of the confinement lid mounting bolts 1154 and others are configured to receive the shank of shielded lifting lid mounting bolts 1156. In cases where the mounting bolts 1154 and 1156 have shanks of the same diameter, the bolt holes 1202 may all have the same diameter. Where the shanks of bolts 1154 and 1156 are different in diameter, the holes 1202 may have correspondingly different diameters for each bolt. The confinement lid 1200 may have a uniform thickness from peripheral side 1204 to peripheral side 1204 as best shown in FIG. 17 in one embodiment. In other embodiments, the thickness may vary at different locations on the lid 1200. Confinement lid 1200 may be made of any suitable material, preferably an appropriate metal for the application. In an exemplary embodiment, without limitation, the confinement lid 1200 for example may be made of stainless steel for corrosion resistance. The upper shielded lifting lid 1300 is not intended to remain on canister 1100 for longer term waste storage. Instead, in some embodiments, the lifting lid 1300 is configured and structured for transporting and initially lifting the canister 1100 into position in the cylindrical overpack 1130 prior to loading the waste cylinders 1121 after which the lifting lid is removed, and then after the waste cylinders are loaded in the canister, the lifting lid is replaced on the canister to shield the operator for bolting the lower confinement lid 1200 in place after which the lifting lid is removed again. It will be appreciated that this scenario for using the shielded lifting lid 1300 may be varied in other embodiments. Referring to FIGS. 15-20, shielded lifting lid 1300 is generally circular in shape (top plan view) and includes a plurality of bolt holes 1302 spaced circumferentially around the peripheral side 1304 of the lid as best shown in FIG. 9. In one embodiment, holes 1302 are not threaded. The bolt holes 1302 may be arranged in clustered groups or sets corresponding to the location and arrangement of the mounting blocks 1150 inside the canister 1100. The bolt holes 1302 have a diameter sized to at least pass the shank of mounting bolts 1154 and 1156 through the holes to threadably engage the mounting blocks 1150. Accordingly, some of the bolt holes 1302 are configured to receive the shanks of the confinement lid mounting bolts 1154 and others are configured to receive the shank of shielded lifting lid mounting bolts 1156. In cases where the mounting bolts 1154 and 1156 have shanks of the same diameter, the bolt holes 1302 may all have the same diameter. Where the shanks of bolts 1154 and 1156 are different in diameter, the holes 1302 may have correspondingly different diameters for each bolt. According to another aspect of the invention, bolt holes 1302 have different diameters in one embodiment even if the mounting bolts 1154, 1156 are used have the same shank diameter. The confinement lid mounting bolts 1154 need not engage the upper shielded lifting lid because bolts 1154 are only required to secure the lower confinement lid to canister 1100. Accordingly, in the embodiment shown in FIG. 19, the bolt holes 1302 for the confinement lid mounting bolts 1154 may have a larger diameter than the bolt holes 1302 for the lifting lid mounting bolts 1156. In this arrangement, the bolt holes 1302 for the confinement lid mounting bolts 1154 are sized with a diameter large enough to allow the shank and entire head of bolts 1154 to pass through the bolt holes so that the head and integral washer directly engage the top surface 1206 of the confinement lid 1200 (see, e.g. FIG. 9). When completely installed, the heads of the mounting bolts 1154 are recessed below the top surface of the lifting lid 1300 as shown. By contrast, since the mounting bolts 1156 for the lifting lid 1300 also serve a lifting function for the canister 1100, the bolt holes 1302 have a diameter sized so that the heads of bolts 1156 do not pass through the bolt holes and instead engage the top surface 1306 of the lifting lid (thereby projecting above the top surface and remaining exposed as shown in FIG. 19). In this manner, the bolts 1156 transfer the dead load and weight of the canister 1100 from the mounting blocks 1150 directly to the shielded lifting lid 1300 without involvement of the confinement lid 1200. Accordingly, to accommodate the foregoing arrangement, the lifting lid mounting bolts 1156 preferably have a longer shank than the confinement lid mounting bolts 1154 in this embodiment. As shown in FIGS. 17 and 18, several spaced apart clusters comprised of three bolt holes 1302 may be provided in the non-limiting embodiment shown which are spaced circumferentially around and proximate to the peripheral side 1304 of the shielded lifting lid 1300. Each cluster of bolt holes 1302 is spaced apart by an arcuate distance from adjacent clusters of holes 1302. The clusters of bolts holes 1302 are each vertically aligned with a corresponding mounting block 1150 (see also FIG. 11). In this embodiment, the center hole 1302 has a smaller diameter for the lifting lid mounting bolt 1156 than the two adjacent outer holes 1302 have larger diameters for the confinement lid mounting bolts 1154. Other suitable arrangements of holes 1302 may be provided. The bolt holes 1202 in the confinement lid 1200 may also arranged in clusters of three to mate with the bolt holes 1302 of the lifting lid 1300. All three of the bolt holes 1202 in each cluster in the confinement lid, however, may have the same diameter. Advantageously, having two different size bolt holes 1302 for the confinement lid mounting bolts 1154 and the lifting lid mounting bolts 1156 reduces possible installation error and ensures that the operator will not confuse which holes are intended for each. This plays a role in deploying the two-part lid system when the confinement lid 1200 and its respective bolts 1154 are eventually left in place after bolting the confinement lid to the canister 1100 and the lifting lid mounting bolts 1156 are removed by the operator, as further described herein. The shielded lifting lid 1300 may have a non-uniform thickness from peripheral side 1304 to peripheral side 1304 as best shown in FIG. 17. Accordingly, in one possible embodiment as shown, the peripheral portion of lifting lid 1300 may include an outer annular step or shoulder 1308 having a smaller thickness than the inner central portion 1314 of the lid. The shoulder 1308 is configured to complement and abuttingly engage a corresponding top annular rim 1138 of the overpack 1130 such that portions of the lifting lid 1300 adjacent to peripheral side 1304 overlap the top of the rim to prevent radiation streaming as shown. Rim 1138 therefore defines an annulus for receiving shoulder 1308. Accordingly, as shown in FIG. 17, shielded lifting lid 1300 has a larger diameter than confinement lid 1200 to account for the overlap with the annular rim 1138 of the overpack 1130. The central portion 1314 of the lifting lid 1300 preferably has a thickness and a diameter sized to allow at least partial insertion of the central portion into the overpack 1130 such that the outwards facing annular sides of the central portion abuts the interior surface 1133 of the overpack as shown. This arrangement further prevents radiation streaming from the canister 1100 when the lifting lid 1300 is in place on the canister. Because shielded lifting lid 1300 serves a structural purpose for lifting the canister 1100, the lifting lid preferably has a thickness which is greater than the confinement lid 1200. In one embodiment, the lifting lid has a thickness which is at least twice the thickness of the confinement lid. Shielded lifting lid 1300 may be made of any suitable material, preferably an appropriate metal for the application. In exemplary embodiments, without limitation, the lifting lid 300 for example may be made of carbon steel or stainless steel. Referring to FIGS. 15 and 16, the lower confinement lid 1200 is detachably mounted to upper shielded lifting lid 1300 so that the lid assembly 1200/1300 may be lifted and moved as a single unit as shown with the lifting lid supporting the confinement lid when not attached to the canister 1100. When needed during the canister closure operations, the lifting lid 1300 may be uncoupled from the confinement lid 1200. In one embodiment, a plurality of circumferentially spaced fasteners such as threaded assembly bolts 1131 may be provided to attach lifting lid 1300 to confinement lid 1200. Assembly bolts 1131 which are inserted through the lifting lid 1300 and engage complementary threaded sockets 1208 (shown in FIG. 9) formed in the confinement lid (such arrangement and operation being apparent to those skilled in the art without further elaboration). A suitable number of assembly bolts 1131 are provided to support the lower confinement lid 1200 from the upper shielded lifting lid 1300 during hoisting. Accordingly, confinement lid 1200 may be considered to be fully supported by the lifting lid 1300 during lifting of the lid assembly 1200/1300. As shown in FIGS. 15 and 16, shielded lifting lid 1300 includes a lifting attachment such as lifting lugs 1402 and pin 1404 for grappling and hoisting the lid. Other suitable lifting attachments configured for grappling such as for example lifting bails may be used. An exemplary method for storing radioactive waste using the present container system with two-part lid assembly 1200/1300 (confinement lid 1200, lifting lid 1300) according to the present disclosure will now be described. As a preliminary step, the lower confinement lid 1200 is detachably mounted to the upper shielded lifting lid 1300 using assembly bolts 1131 to collectively form the lid assembly 1200/1300, shown in FIG. 15. Referring to FIGS. 9 and 10, the method begins with a canister 1100 first being provided with an empty basket insert 1120 disposed inside the canister as shown. Next, the empty canister 1100 is lifted and placed into the overpack 1130 as shown in FIG. 13. In one embodiment, this step may be performed by bolting the lid assembly 1200/1300 to canister 1100 using the mounting bolts 1156 to threadably engage the mounting blocks 1150, and grappling and attaching a hoist 1400 to the upper lifting lid 1300 using lifting lugs 1402 and pin 1404 as shown in FIG. 15. The hoist 1400 may be part of the lifting equipment such as a crane or other suitable equipment operable to raise and lower the canister. After positioning the basket insert 1120 into the canister 1100, the mounting bolts 1156 may be removed to disconnect the canister from the lid assembly. The lid assembly 1200/1300 may then be lifted by the hoist and removed (see FIG. 13). Next, one or preferably more lid alignment pins 1406 may be threaded into some of the threaded sockets 1152 of the mounting block to eventually help properly align the lid assembly 1200/1300 with the canister (see FIG. 13). In one non-limiting example, three alignment pins 1406 are used spaced apart on the canister. The alignment pins 1406 are preferably installed locally by an operator prior to loading the radioactively “hot” waste cylinders 1121 into the canister. Following installation of the alignment pins 1406, the waste cylinders 1121 are loaded into the canister 1100, and more specifically positioned in their respective locations provided in basket insert 1120 as shown in FIG. 14. Loading of the waste cylinders is performed remotely (i.e. at a distance) by an operator using suitable equipment to protect the operator from radiation. After loading the waste cylinders 1121, the lid assembly 1200/1300 is remotely hoisted by the operator over and vertically positioned above the top 1102 of the canister 1100, as shown in FIG. 15. Using the lid alignment pins 1406, the operator vertically aligns holes 1302 in shielded lifting lid (with holes 1202 in confinement lid being concentrically aligned with holes 1302) with corresponding pins 1406 to properly orient the lid rotationally with respect to the canister. When the pins 1406 and their corresponding holes have been axially aligned, the operator lowers lid assembly 1200/1300 onto the canister 1100 as shown in FIG. 16 (see pins 1406 extending through holes 1302). The operator will now be shielded from radiation emitted from the canister so that the confinement lid 1200 may be bolted in place locally. Next, the lid alignment pins 1406 and assembly bolts 1131 which hold the lower confinement lid 1200 to upper shielded lifting lid 1300 may be removed (see, e.g. FIG. 18). All of the confinement lid mounting bolts 1154 may then be installed to mount the confinement lid 1200 to the canister 1100 using the mounting blocks 1150. The mounting bolts 1154 are threaded through bolt holes 1302 until the heads of the bolts engage the top surface 1206 of the confinement lid 1200 and the bolts are tightened to the required torque (see FIGS. 19 and 20). Prior to removing the shielded lifting lid 1300, a set of overpack lid alignment pins 1408 may next be installed in threaded sockets 1510 of the overpack 1130. With the confinement lid 1200 now fully fastened to canister 1100, the shielded lifting lid 1300 may next be removed via the hoist remotely by an operator as shown in FIG. 23. In the following steps, the overpack lid 1500 is installed on overpack 1130 following closure of canister 1100 described above. FIG. 23 shows the shielded lifting lid 1300 being removed and the overpack lid 1500 staged for installation. FIG. 21 shows overpack lid 1500 in greater detail. Overpack lid 1500 is circular in shape (top plan view) and includes a plurality of mounting holes 1502, top surface 1504, peripheral sides 1506, and a lifting bail 1508 attached towards the center of the lid for engagement by a hoist. Overpack lid 1500 serves a structural role of protecting the canister 1100 disposed inside the overpack 1130, and in some embodiments supporting the weight of the overpack when mounted thereto for transport and lifting. Accordingly, overpack lid 1500 may have a thickness greater than the thickness of the confinement lid 1200. Referring now to FIGS. 23 and 24, the overpack lid 1500 is grappled and lifted via the attached hoist 1400 by crane or other equipment, vertically aligned with overpack 1130 using the alignment pins 1408 in a manner similar to alignment pins 1406, and lowered onto the overpack. Alignment pins 1408 are then removed and mounting bolts 1512 are then installed in the threaded sockets 1510 of the overpack 1130 to complete installation and securement of the overpack lid 1500, as shown in FIG. 25. Optionally, the lifting bail 1508 may be removed. FIG. 26 shows the overpack 1130 with overpack lid 1500 fully installed and canister 1100 disposed inside loaded with waste cylinders 1121. Protective caps 1514 may be installed over mounting bolts 1512. An operator is shown in FIG. 26 to provide perspective on the size of overpack 1130 in one non-limiting embodiment, which may be about 6 or more feet in diameter and about 6 or more feet in height. Any suitable size overpack may be used. As noted herein, the shielded lifting lid 1300 is reusable. Accordingly, in some embodiments, the exemplary method described above may further comprise a step of detachably mounting a second different confinement lid 1200 to the shielded lifting lid 1300; the second confinement lid and shielded lifting lid collectively forming a second lid assembly. It will be appreciated that the two-part lid assembly 1200/1300 may also be used in applications where the confinement lid 1200 is intended to be welded to the canister 1100 for closure rather than by bolting. III. Inventive Concept 3 With reference to FIGS. 27-34, a third inventive concept will be described. FIG. 27A illustrates a high level waste (“HLW”) storage container 2010, encased in surrounding concrete 2011, as it would be in an installation. FIG. 28 illustrates the storage container 2010 in a sectional view, still with the surrounding concrete 2011. While the HLW storage container 2010 will be described in terms of being used to store a canister of spent nuclear fuel, it will be appreciated by those skilled in the art that the systems and methods described herein can be used to store any and all kinds of HLW. The HLW storage container 2010 is designed to be a vertical, ventilated dry system for storing HLW such as spent fuel. The HLW storage container 2010 is fully compatible with 100 ton and 125 ton transfer casks for HLW transfer procedures, such as spent fuel canister transfer operations. All spent fuel canister types engineered for storage in free-standing, below grade, and/or anchored overpack models can be stored in the HLW storage container 2010. As used in this section the term “canister” broadly includes any spent fuel containment apparatus, including, without limitation, multi-purpose canisters and thermally conductive casks. For example, in some areas of the world, spent fuel is transferred and stored in metal casks having a honeycomb grid-work/basket built directly into the metal cask. Such casks and similar containment apparatus qualify as canisters, as that term is used herein, and can be used in conjunction with the HLW storage container 2010 as discussed below. The HLW storage container 2010 can be modified/designed to be compatible with any size or style of transfer cask. The HLW storage container 2010 can also be designed to accept spent fuel canisters for storage at an Independent Spent Fuel Storage Installations (“ISFSI”). ISFSIs employing the HLW storage container 2010 can be designed to accommodate any number of the HLW storage container 2010 and can be expanded to add additional HLW storage containers 2010 as the need arises. In ISFSIs utilizing a plurality of the HLW storage container 2010, each HLW storage container 2010 functions completely independent form any other HLW storage container 2010 at the ISFSI. The HLW storage container 2010 has a body 2020 and a lid 2030. The lid 2030 rests atop and is removable/detachable from the body 2020. Although an HLW storage container can be adapted for use as an above grade storage system, by incorporating design features found in U.S. Pat. No. 7,933,374, this HLW storage container 2010, as shown, is designed for use as a below grade storage system. Referring to FIG. 28, the body 2020 includes an outer shell 2021 and an inner shell 2022. The outer shell 2021 surrounds the inner shell 2022, forming a space 2023 therebetween. The outer shell 2021 and the inner shell 2022 are generally cylindrical in shape and concentric with one another. As a result, the space 2023 is an annular space. While the shape of the inner and outer shells 2022, 2021 is cylindrical in the illustrated embodiment, the shells can take on any shape, including without limitation rectangular, conical, hexagonal, or irregularly shaped. In some embodiments, the inner and outer shells 2022, 2021 will not be concentrically oriented. The space 2023 formed between the inner shell 2022 and the outer shell 2021 acts as a passageway for cool air. The exact width of the space 2023 for any HLW storage container 2010 is determined on a case-by-case design basis, considering such factors as the heat load of the HLW to be stored, the temperature of the cool ambient air, and the desired fluid flow dynamics. In some embodiments, the width of the space 2023 will be in the range of 1 to 6 inches. While the width of space 2023 can vary circumferentially, it may be desirable to design the HLW storage container 2010 so that the width of the space 2023 is generally constant in order to effectuate symmetric cooling of the HLW container and even fluid flow of the incoming air. As discussed in greater detail below, the space 2023 may be divided up into a plurality of passageways. The inner shell 2022 and the outer shell 2021 are secured atop a floor plate 2050. The floor plate 2050 is hermetically sealed to the outer shell 2021, and it may take on any desired shape. A plurality of spacers 2051 are secured atop the floor plate 2050 within the space 2023. The spacers 2051 support a pedestal 2052, which in turn supports a canister. When a canister holding HLW is loaded into the cavity 2024 for storage, the bottom surface of the canister rests atop the pedestal 2052, forming an inlet air plenum between the underside of the pedestal 2052 and the floor of cavity 2024. This inlet air plenum contributes to the fluid flow and proper cooling of the canister. Preferably, the outer shell 2021 is seal joined to the floor plate 2050 at all points of contact, thereby hermetically sealing the HLW storage container 2010 to the ingress of fluids through these junctures. In the case of weldable metals, this seal joining may comprise welding or the use of gaskets. Most preferably, the outer shell 2021 is integrally welded to the floor plate 2050. An upper flange 2077 is provided around the top of the outer shell 2021 to stiffen the outer shell 2021 so that it does not buckle or substantially deform under loading conditions. The upper flange 2077 can be integrally welded to the top of the outer shell 2021. The inner shell 2022 is laterally and rotationally restrained in the horizontal plane at its bottom by support legs 2027 which straddle lower ribs 2053. The lower ribs 2053 are preferably equispaced about the bottom of the cavity 2024. The inner shell 2022 is preferably not welded or otherwise permanently secured to the bottom plate 2050 or outer shell 2021 so as to permit convenient removal for decommissioning, and if required, for maintenance. The inner shell 2022, the outer shell 2021, the floor plate 2050, and the upper flange 2077 are preferably constructed of a metal, such as a thick low carbon steel, but can be made of other materials, such as stainless steel, aluminum, aluminum-alloys, plastics, and the like. Suitable low carbon steels include, without limitation, ASTM A516, Gr. 70, A515 Gr. 70 or equal. The desired thickness of the inner and outer shells 2022, 2021 is matter of design choice and will determined on a case-by-case basis. The inner shell 2022 forms a cavity 2024. The size and shape of the cavity 2024 is also a matter of design choice. However, it is preferred that the inner shell 2022 be designed so that the cavity 2024 is sized and shaped so that it can accommodate a canister of spent nuclear fuel or other HLW. While not necessary, it is preferred that the horizontal cross-sectional size and shape of the cavity 2024 be designed to generally correspond to the horizontal cross-sectional size and shape of the canister-type that is to be used in conjunction with a particular HLW storage container. More specifically, it is desirable that the size and shape of the cavity 2024 be designed so that when a canister containing HLW is positioned in the cavity 2024 for storage (as illustrated in FIG. 30A), a small clearance exists between the outer side walls of the canister and the side walls of the cavity 2024. Designing the cavity 2024 so that a small clearance is formed between the side walls of the stored canister and the side walls of the cavity 2024 limits the degree the canister can move within the cavity during a catastrophic event, thereby minimizing damage to the canister and the cavity walls and prohibiting the canister from tipping over within the cavity. This small clearance also facilitates flow of the heated air during HLW cooling. The exact size of the clearance can be controlled/designed to achieve the desired fluid flow dynamics and heat transfer capabilities for any given situation. In some embodiments, for example, the clearance may be 1 to 3 inches. A small clearance also reduces radiation streaming. The inner shell 2022 is also equipped with multiple sets of equispaced longitudinal ribs 2054, 2055, in addition to the lower ribs 2053 discussed above. One set of ribs 2054 are preferably disposed at an elevation that is near the top of a canister of HLW placed in the cavity 2024. This set of ribs 2054 may be shorter in length in comparison to the height of the cavity 2024 and a canister. Another set of ribs 2055 are set below the first set of ribs 2054. This second set of ribs 2055 is more elongated than the first set of ribs 2054, and these ribs 2055 extend to, or nearly to, the bottom of the cavity 2024. These ribs 2053, 2054, 2055 serve as guides for a canister of HLW is it is lowered down into the cavity 2024, helping to assure that the canister properly rests atop the pedestal 2052. The ribs also serve to limit the canister's lateral movement during an earthquake or other catastrophic event to a fraction of an inch. A plurality of openings 2025 are provided in the inner shell 2022 at or near its bottom between the support legs 2027. Each opening 2025 provides a passageway between the annular space 2023 and the bottom of the cavity 2024. The openings 2025 provide passageways by which fluids, such as air, can pass from the annular space 2023 into the cavity 2024. The openings 2025 are used to facilitate the inlet of cooler ambient air into the cavity 2024 for cooling a stored HLW having a heat load. As illustrated, eight openings 2025 are equispaced about the bottom of the inner shell 2022. However, any number of openings 2025 can be included, and they may have any spacing desired. The exact number and spacing will be determined on a case-by-case basis and will be dictated by such considerations as the heat load of the HLW, desired fluid flow dynamics, etc. Moreover, while the openings 2025 are illustrated as being located in the side wall of the inner shell 2022, the openings can be provided in the floor plate in certain modified embodiments of the HLW storage container. The openings 2025 in the inner shell 2022 are sufficiently tall to ensure that if water enters the cavity 2024, the bottom region of a canister resting on the pedestal 2052 would be submerged for several inches before the water level reaches the top edge of the openings 2025. This design feature helps ensure thermal performance of the system under accidental flooding of the cavity 2024. With reference to FIG. 29, a layer of insulation 2026 is provided around the outside surface of the inner shell 2022 within the annular space 2023. The insulation 2026 is provided to minimize heating of the incoming cooling air in the space 2023 before it enters the cavity 2024. The insulation 2026 helps ensure that the heated air rising around a canister situated in the cavity 2024 causes minimal pre-heating of the downdraft cool air in the annular space 2023. The insulation 2026 is preferably chosen so that it is water and radiation resistant and undegradable by accidental wetting. Suitable forms of insulation include, 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 desired thickness of the layer of insulation 2026 is matter of design and will be dictated by such considerations such as the heat load of the HLW, the thickness of the shells, and the type of insulation used. In some embodiments, the insulation will have a thickness in the range ½ to 6 inches. As shown in FIGS. 28 and 29, inlet ducts 2060 are disposed on the top surface of the upper flange 2077. Each inlet duct 2060 connects to two inlet passageways 2061 which continue from under the upper flange 2077, into the space 2023 between the outer and inner shells 2021, 2022, and then connect to the cavity 2024 by lower openings 2062 in the bottom of the inner shell 2022. Within the space 2023, the inlet passageways 2061 are separated by dividers 2063 to keep cooling air flowing through each inlet passageway 2061 separate from the other inlet passageways 2061 until the cooling air emerges into the cavity 2024. FIGS. 30A and 30B illustrate the configuration of the inlet passageways 2061 and the dividers 2063. Each inlet passageway 2061 connects with the space 2023 by openings 2064 in the top of the outer shell 2021. From the openings 2064, the cooling air continues down the in the space, via the individual inlet passageways 2061 created by the dividers 2064, and into the cavity 2024, where it is used to cool a placed HLW canister. The dividers 2063 are equispaced within the space 2023 to aid in balancing the air pressure entering the space 2023 from each inlet duct and inlet passageway. Also, as shown in the figures, each of the lower ribs 2053 is integrated with one of the dividers 2063, such that the lower ribs form an extension of the dividers, extending into the cavity 2024. Referring back to FIG. 29, each inlet duct 2060 includes a duct cover 2065, to help prevent rain water or other debris from entering and/or blocking the inlet passageways 2061, affixed on top of an inlet wall 2066 that surrounds the inlet passageways 2061 on the top surface of the upper flange 2077. The inlet wall 2066 is peripherally perforated around the entire periphery of the opening of the inlet passageways 2061. At least a portion of the lower part of the inlet ducts are left without perforations, to aid in preventing rain water from entering the HLW storage container. Preferably, the inlet wall 2066 is perforated over 60% or more of its surface, and the perforations can be made in any shape, size, and distribution in accordance with design preferences. When the inlet ducts 2060 are formed with the inlet wall 2066 peripherally perforated, each of the inlet ducts has been found to maintain an intake air pressure independently of each of the other inlet ducts, even in high wind conditions, and each of the inlet ducts has been found to maintain an intake air pressure substantially the same as each of the other inlet ducts, again, even in high wind conditions. The lid 2030 rests atop and is supported by the upper flange 2077 and a shell flange 2078, the latter being disposed on and connected to the tops edge of the inner shell 2022. The lid 2030 encloses the top of the cavity 2024 and provides the necessary radiation shielding so that radiation does not escape from the top of the cavity 2024 when a canister loaded with HLW is stored therein. The lid 2030 is designed to facilitate the release of heated air from the cavity 2024. FIG. 31A illustrates the HLW storage container 2010 with a canister 2013 placed within the cavity 2024. As shown in the FIG. 31B detailed view, the bottom of the canister 2013 sits on the pedestal 2052, and the lower ribs 2053 maintain a space between the bottom of the canister 2013 and the inner shell 2022. Similarly, the FIG. 31C detailed view shows that the upper ribs 2054 maintain a space between the top of the canister 2013 and the inner shell 2022. The FIG. 31D detailed view shows the lid 2030 resting atop the upper flange 2077 and the shell flange 2078. The lid 2030 includes a closure gasket 2031 which forms a seal against the upper flange 2077 when the 20 lid 30 is seated, and a leaf spring gasket 2032 which forms a seal against the shell flange 2078. FIGS. 32 and 33 illustrate the lid 2030 removed from the body of the HLW storage container. Referring first to FIG. 32, the lid 2030 is preferably constructed of a combination of low carbon steel and concrete (or another radiation absorbing material) in order to provide the requisite radiation shielding. The lid 2030 includes an upper lid part 2033 and a lower lid part 2034. The upper lid part 2033 preferable extends at least as high as, if not higher than, the top of each inlet duct 2060. Each lid part 2033, 2034 includes an external shell 2035, 2036 encasing an upper concrete shield 2037 and a lower concrete shield 2038. One or more outlet passageways 2039 are formed within and around the body parts 2033, 2034 to connect the cavity with the outlet duct 2040 formed on the top surface of the lid 2030. The outlet passageways 2039 pass over the lower lid part 2034, between the upper and lower lid parts 2033, 2034, and up through a central aperture within the upper lid part 2034. The outlet duct 2040 covers this central aperture to better control the heated air as it rises up out of the. By being disposed on the top of the lid 2030, the outlet duct 2040 may also be raised up significantly higher than the inlet ducts, using any desired length of extension for the outlet duct. By raising up the outlet duct higher, mixing between the heated air emitted from the outlet duct and cooler air being drawn into the inlet ducts can be significantly reduced, if not eliminated altogether. The outlet duct 2040, which is constructed similar to the inlet ducts, includes a duct cover 2041, to help prevent rain water or other debris from entering and/or blocking the outlet passageways 2039, affixed on top of an outlet wall 2042 that surrounds the outlet passageways 2039 on the top surface of the upper lid part 2033. The outlet wall 2042 is peripherally perforated around the entire periphery of the opening of the outlet passageways 2039. At least a portion of the lower part of the outlet duct is left without perforations, to aid in preventing rain water from entering the HLW storage container. Preferably, the outlet wall 2042 is perforated over 60% or more of its surface, and the perforations can be made in any shape, size, and distribution in accordance with design preferences. The external shell of the lid 2030 may be constructed of a wide variety of materials, including without limitation metals, stainless steel, aluminum, aluminum-alloys, plastics, and the like. The lid may also be constructed of a single piece of material, such as concrete or steel for example, so that it has no separate external shell. When the lid 2030 is positioned atop the body 2020, the outlet passageways 2039 are in spatial cooperation with the cavity 2024. As a result, cool ambient air can enter the HLW storage container 2010 through the inlet ducts 2060, flow into the space 2023, and into the bottom of the cavity 2024 via the openings 2062. When a canister containing HLW having a heat load is supported within the cavity 2024, this cool air is warmed by the HLW canister, rises within the cavity 2024, and exits the cavity 2024 via the outlet ducts 2040. Because the inlet ducts 2060 are placed on different sides of the lid 2030, and the dividers separate the inlet passageways associated with the different inlet ducts, the hydraulic resistance to the incoming air flow, a common limitation in ventilated modules, is minimized. This configuration makes the HLW storage container less apt to build up heat internally under high wind conditions. A plurality of HLW storage containers 2100 can be used at the same ISFSI site and situated in arrays as shown in FIG. 34. Although the HLW storage containers 2100 are closely spaced, the design permits a canister in each HLW storage container 2100 to be independently accessed and retrieved easily. In addition, the design of the individual storage containers 2100, and particularly the design and positioning of the inlet and outlet ducts, enables the inlet ducts of a first of the storage containers to maintain air pressure independently of the inlet ducts of a second of the storage containers. Each storage container therefore will operate independently of each of the other storage containers, such that the failure of one storage container is unlikely to lead directly to the failure of other surrounding storage containers in the array. IV. Inventive Concept 4 With reference to FIGS. 35-47, a fourth inventive concept will be described. Referring to FIG. 35, a dual-walled DSC 3099 according to one embodiment of the present invention is disclosed. The dual-walled DSC 3099 and its components are illustrated and described as an MPC style structure. However, it is to be understood that the concepts and ideas disclosed herein can be applied to other areas of high level radioactive waste storage, transportation and support. Moreover, while the dual-walled DSC 3099 is described as being used in combination with a specially designed fuel basket 3090 (which in of itself constitutes an invention), the dual-walled DSC 3099 can be used with any style of fuel basket, such as the one described in U.S. Pat. No. 5,898,747, issued Apr. 27, 1999. In fact, in some instances it may be possible to use the dual-walled DSC 3099 without a fuel basket, depending on the intended function. Furthermore, the dual-walled DSC 3099 can be used to store and/or transport any type of high level radioactive materials and is not limited to SNF. As will become apparent from the structural description below, the dual-walled DSC 3099 contains two independent containment boundaries about the storage cavity 3030 that operate to contain both fluidic (gas and liquid) and particulate radiological matter within the cavity 3030. As a result, if one containment boundary were to fail, the other containment boundary will remain intact. While theoretically the same, the containment boundaries formed by the dual-walled DSC 3099 about the cavity 3030 can be literalized in many ways, including without limitation a gas-tight containment boundary, a pressure vessel, a hermetic containment boundary, a radiological containment boundary, and a containment boundary for fluidic and particulate matter. These terms are used synonymously throughout this application. In one instance, these terms generally refer to a type of boundary that surrounds a space and prohibits all fluidic and particulate matter from escaping from and/or entering into the space when subjected to the required operating conditions, such as pressures, temperatures, etc. Finally, while the dual-walled DSC 3099 is illustrated and described in a vertical orientation, it is to be understood that the dual-walled DSC 3099 can be used to store and/or transport its load in any desired orientation, including at an angle or horizontally. Thus, use of all relative terms through this specification, including without limitation “top,” “bottom,” “inner” and “outer,” are used for convenience only and are not intended to be limiting of the invention in such a manner. The dual-walled DSC 3099 includes a first shell that acts as an inner shell 3010 and a second shell that acts as an outer shell 3020. The inner and outer shells 3010, 3020 are preferably cylindrical tubes and are constructed of a metal. Of course, other shapes can be used if desired. The inner shell 3010 is a tubular hollow shell that includes an inner surface 3011, an outer surface 3012, a top edge 3013 and a bottom edge 3014. The inner surface 3011 of the inner shell 3010 forms a cavity/space 3030 for receiving and storing SNF. The cavity 3030 is a cylindrical cavity formed about a central axis. The outer shell 3020 is also a tubular hollow shell that includes an inner surface 3021, an outer surface 3022, a top edge 3023 and a bottom edge 3024. The outer shell 3020 circumferentially surrounds the inner shell 3010. The inner shell 3010 and the outer shell 3020 are constructed so that the inner surface 3021 of the outer shell 3020 is in substantially continuous surface contact with the outer surface 3012 of the inner shell 3010. In other words, the interface between the inner shell 3010 and the outer shell 3020 is substantially free of gaps/voids and are in conformal contact. This can be achieved through an explosive joining, a cladding process, a roller bonding process and/or a mechanical compression process that bonds the inner shell 3010 to the outer shell 3020. The continuous surface contact at the interface between the inner shell 3010 and the outer shell 3020 reduces the resistance to the transmission of heat through the inner and outer shells 3010, 3020 to a negligible value. Thus, heat emanating from the SNF loaded within the cavity 3030 can efficiently and effectively be conducted outward through the shells 3010, 3020 where it is removed from the outer surface 3022 of the outer shell via convection. Even though the interface is formed in any of these manners, there still remains an interstitial space 3097 between the inner shell 3010 and the outer shell 3020. Alternatively, the interstitial space may be formed without the inner surface of the outer shell being in substantially continuous surface contact with the outer surface of the inner shell. As is discussed in more detail below, the presence of this interstitial space is used advantageously during a leak testing process. The inner and outer shells 3010, 3020 are preferably both made of a metal. As used herein, the term metal refers to both pure metals and metal alloys. Suitable metals include without limitation austenitic stainless steel and other alloys including Hastelloy™ and Inconel™. Of course, other materials can be utilized. The thickness of each of the inner and outer shells 3010, 3020 is preferably in the range of 5 mm to 25 mm. The outer diameter of the outer shell 3020 is preferably in the range of 1700 mm to 2000 mm. The inner diameter of the inner shell 3010 is preferably in the range of 1700 mm to 1900 mm. The specific size and/or thickness of the shells 3010, 3020, however, is a matter of design choice. In some embodiments, it may be further preferable that the inner shell 3010 be constructed of a metal that has a coefficient of thermal expansion that is equal to or greater than the coefficient of thermal expansion of the metal of which the outer shell 3020 is constructed. Thus, when the SNF that is stored in the cavity 3030 and emits heat, the outer shell 3020 will not expand away from the inner shell 3010. This ensures that the continuous surface contact between the outer surface 3012 of the inner shell 3010 and the outer surface 3021 of the outer shell 3020 will be maintained and a gaps will not form under heat loading conditions. The dual-walled DSC 3099 also includes a first lid that acts as an inner top lid 3060 for the inner shell 3010 and a second lid that acts as an outer top lid 3070 for the second shell 3020. The inner and outer top lids 3060, 3070 are plate-like structures that are preferably constructed of the same materials discussed above with respect to the shells 3010, 3020. Preferably the thickness of the inner top lid 3060 is in the range of 99 mm to 300 mm. The thickness of the outer top lid is preferably in the range of 50 mm to 150 mm. The invention is not, however, limited to any specific dimensions, which will be dictated on a case-by-case basis and the radioactive levels of the SNF to be stored in the cavity 3030. Referring to FIG. 36, the inner top lid 3060 includes a top surface 3061, a bottom surface 3062 and an outer lateral surface/edge 3063. The outer top lid 3070 includes a top surface 3071, a bottom surface 3072 and an outer lateral surface/edge 3073. When fully assembled, the outer lid 3070 is positioned atop the inner lid 3060 so that the bottom surface 3072 of the outer lid 3070 is in substantially continuous surface contact with the top surface 3061 of the inner lid 3060. The outer lid 3070 also includes a test port 3095, to which one end of conduit is coupled (see FIGS. 44 and 45) in fluidic communication therewith. As is discussed below, the other end of the conduit is fitted with both a removable seal, to enable leak testing, and valve, both being included to comply with ASME Code. During an SNF underwater loading procedure, the inner and outer lids 3060, 3070 are removed. Once the cavity 3030 is loaded with the SNF, the inner top lid 3060 is positioned so as to enclose the top end of the cavity 3030 and rests atop the brackets 3015. Once the inner top lid 3060 is in place and seal welded to the inner shell 3010, the cavity 3030 is evacuated/dried via the appropriate method and backfilled with nitrogen, helium or another inert gas. The drying and backfilling process of the cavity 3030 is achieved via the holes 3064 of the inner lid 3060 that form passageways into the cavity 3030. Once the drying and backfilling is complete, the holes 3061 are filled with a metal or otherwise plugged so as to hermetically seal the cavity 3030. Referring now to FIGS. 35 and 37 concurrently, the outer shell 3020 has an axial length L2 that is greater than the axial length L1 of the inner shell 3010. As such, the top edge 3013 of the inner shell 3010 extends beyond the top edge 3023 of the outer shell 3020. Similarly, the bottom edge 3024 of the outer shell 3020 extends beyond the bottom edge 3013 of the inner shell 3010. The offset between the top edges 3013, 3023 of the shells 3010, 3020 allows the top edge 3013 of the inner shell 3010 to act as a ledge for receiving and supporting the outer top lid 3070. When the inner lid 3060 is in place, the inner surface 3011 of the inner shell 3010 extends over the outer lateral edges 3063. When the outer lid 3070 is then positioned atop the inner lid 3060, the inner surface 3021 of the outer shell 3020 extends over the outer lateral edge 3073 of the outer top lid 3070. The top edge 3023 of the outer shell 3020 is substantially flush with the top surface 3071 of the outer top lid 3070. The inner and outer top lids 3060, 3070 are welded to the inner and outer shells 3010, 3020 respectively after the fuel is loaded into the cavity 3030. Conventional edge groove welds can be used. However, it is preferred that all connections between the components of the dual-walled DSC 3099 be through-thickness weld. The dual-walled DSC 3099 also includes a first plate that acts as an inner base plate 3040 and a second plate that acts as an outer base plate 3050. The inner and outer base plates 3040, 3050 are rigid plate-like structures having circular horizontal cross-sections. The invention is not so limited, however, and the shape and size of the base plates 3040, 3050 is dependent upon the shape of the inner and outer shells 3010, 3020. The inner base plate 3040 includes a top surface 3041, a bottom surface 3042 and an outer lateral surface/edge 3043. Similarly, the outer base plate 3050 includes a top surface 3051, a bottom surface 3052 and an outer lateral surface/edge 3053. The top surface 3041 of the inner base plate 3040 forms the floor of the cavity 3030. The inner base plate 3040 rests atop the outer base plate 3050. Similar to the other corresponding components of the dual-walled DSC 3099, the bottom surface 3042 of the inner base plate 3040 is in substantially continuous surface contact with the top surface 3051 of the outer base plate 3050. As a result, the interface between the inner base plate 3040 and the outer base plate 3050 is free of gaseous gaps/voids for thermal conduction optimization. An explosive joining, a cladding process, a roller bonding process and/or a mechanical compression process can be used to effectuate the contact between the base plates 3040, 3050. Preferably, the thickness of the inner base plate 3040 is in the range of 50 mm to 150 mm. The thickness of the outer base plate 3050 is preferably in the range of 99 mm to 200 mm. Preferably, the length from the top surface of the outer top lid 3070 to the bottom surface of the outer base plate 3050 is in the range of 4000 mm to 5000 mm, but the invention is in no way limited to any specific dimensions. The outer base plate 3050 may be equipped on its bottom surface with a grapple ring (not shown) for handling purposes. The thickness of the grapple ring is preferably between 50 mm and 150 mm. The outer diameter of the grapple ring is preferably between 350 mm and 450 mm. Referring now to FIGS. 36 and 38 concurrently, the inner shell 3010 rests atop the inner base plate 3040 in a substantially upright orientation. The bottom edge 3014 of the inner shell 3010 is connected to the top surface 3041 of the inner base plate 3040 by a through-thickness single groove (V or J shape) weld. The outer surface 3012 of the inner shell 3010 is substantially flush with the outer lateral edge 3043 of the inner base plate 3040. The outer shell 3020, which circumferentially surrounds the inner shell 3010, extends over the outer lateral edges 3043, 3053 of the inner and outer base plates 3040, 3050 so that the bottom edge 3024 of the outer shell 3020 is substantially flush with the bottom surface 3052 of the outer base plate 3050. The inner surface 3021 of the outer shell 3020 is also connected to the outer base plate 3050 using a through-thickness edge weld. In an alternative embodiment, the bottom edge 3024 of the outer shell 3020 could rest atop the top surface 3051 of the outer base plate 3050 (rather than extending over the outer later edge of the base plate 3050). In that embodiment, the bottom edge 3024 of the outer shell 3020 could be welded to the top surface 3051 of the outer base plate 3050. When all of the seal welds discussed above are completed, the combination of the inner shell 3010, the inner base plate 3040 and the inner top lid 3060 forms a first hermetically sealed structure surrounding the cavity 3030, thereby creating a first pressure vessel. Similarly, the combination of the outer shell 3020, the outer base plate 3050, and the outer top lid 3070 form a second sealed structure about the first hermetically sealed structure, thereby creating a second pressure vessel about the first pressure vessel and the cavity 3030. With the inclusion of the test port 3095, the seal of the second pressure vessel also effectively includes the conduit, sealed at the end not coupled to the test port. Theoretically, the first pressure vessel is located within the internal cavity of the second pressure vessel. Each pressure vessel is engineered to autonomously meet the stress limits of the ASME Code with significant margins. Unlike the prior art DSC, all of the SNF stored in the cavity 3030 of the dual-walled DSC 3099 share a common confinement space. The common confinement space (i.e., cavity 3030) is protected by two independent gas-tight pressure retention boundaries. Each of these boundaries can withstand both sub-atmospheric supra-atmospheric pressures as needed, even when subjected to the thermal load given off by the SNF within the cavity 3030. In the event of a failure of the first hermetically sealed structure surrounding the cavity 3030, at least some of the backfilled helium will leak into the interstitial space 3097. Because helium is both an inert gas and a small molecule, the testing equipment and processes, described in greater below, are able to draw helium through the interstitial space 3097 for detection and determination of whether the first hermetically sealed structure has failed. A ventilated system 3101 is shown in FIGS. 39A & 39B. The cask lid 3107 of a ventilated cask 3103 is shown in FIG. 39A, and a cross section of the ventilated cask 3103 is shown in FIG. 39B. As can be seen in FIG. 39B, the ventilated cask 3103 includes a cylindrical cask body 3105 and a cask lid 3107. The cylindrical cask body 3105 includes a set of air inlet ducts 3109 near its bottom and a set of air outlet ducts 3111 near its top. A dual-walled DSC 3099 containing decaying spent nuclear fuel stands upright inside the ventilated cask 3103, with a small diametrical clearance, in the form an annular gap 3113, being formed between an inner surface of the cylindrical cask body 3105 of the ventilated cask 3103 and the outer surface 3115 of the DSC 399. The outer surface 3115 of the DSC 3099 becomes heated due to the thermal energy being generated by the spent nuclear fuel sealed in the DSC 3099. The heat of the outer surface 3115 causes the surrounding air column to heat and rise, resulting in a continuous natural convective ventilation action. The cold air entering the air inlet ducts 3111 at the bottom of the cylindrical cask body 3105 is progressively heated as it rises in the annular gap 3113, reaching its maximum value as it exits the cylindrical cask body 3105. Different designs of such casks are known and described in greater detail in the prior art, e.g., U.S. patent publication No. 2003/0147486, published Aug. 7, 2003, and WO 2013/115881, published Aug. 8, 2013, the disclosures of which are incorporated herein by reference in their entirety. An assembled cask 3151 is shown in FIG. 40. The cask lid 3153 includes ventilation ducts 3155, through one of which the conduit 3157 runs to the outside of the cask 3151. The conduit 3157 extends down the side of the cask body 3159, and into an enclosure 3161 which is affixed to the exterior of the cask body 3159. Although not shown, the conduit may be secured to the cask body 3159 by appropriate brackets affixed to the cask body 3159. As an alternative, the conduit may extend away from the cask body entirely, to an enclosure that is affixed to an independent support (such as a nearby pole or other wall). The conduit 3157 is preferably ¼ inch stainless steel conduit, as such conduit can be evacuated without collapsing. Other conduit materials and sizes that exhibit a similar strength and properties as stainless steel conduit may also be used. Also, the conduit 3157 follows a tortuous path from the first end, where it is coupled to the test port, to the second end, to which the seal, valve, and alternately the testing equipment are coupled. The tortuous path is included so that there is no line of sight path for radiation to escape from the DSC to the outside of the cask 3151. Also, by running the conduit to the outside of the cask, the testing described below may be performed while the cask remains in its storage position and the cask lid remains on the cask, thereby minimizing the amount of time needed to perform the test and significantly reducing the amount of radiation to which workers are exposed. FIG. 41 shows a detailed view of the enclosure 3161 with a cover 3163 in place, which serves to protect contents of the internal chamber of the enclosure 3161, and may be used to make the enclosure waterproof, if desired. One sidewall 3165 of the enclosure 3161 and cover 3163 may include features for locking the cover in place—as shown these features are a pair of aligned rings 3167 on the sidewall 3165 and on the cover 3163, which enable a lock or other security feature (e.g., a tag) to be placed on the enclosure 3161. The conduit 3157 passes through sidewall 3169 and into the internal chamber 3171 of the enclosure 3161, as shown in FIG. 42. Within the enclosure 3161, the second end 3173 of the conduit 3157 includes one test apparatus connector 3175 and a secondary connector 3177. The two connectors 3175, 3177 provide a dual failsafe boundary in compliance with ASME Code. When no test is being performed, a removable seal 3179 is coupled to the test apparatus connector 3175. The removable seal 3179 may be of any type suitable for sealing the test apparatus connector 3175 and for use under the operating conditions described herein. The test apparatus connector 3175 is otherwise configured for coupling to the test apparatus to be used, which may be a mass spectrometer leak detector (MSLD) of the kind which are readily available on the market today, and one of ordinary skill in the art would be aware of the types of different MSLDs available. The secondary connector 3177 is regulated by a valve 3181 which is suitable for the operating conditions described herein. During the testing process, once tests are performed by the MSLD, a source of a second inert gas (different from the inert gas which is filled in the canister) may be connected to the secondary connector so that the conduit and at least part of the interstitial space are backfilled with this second inert gas. An alternative for extending the conduit 3157 to the outside of the cask 3151 is shown in FIG. 43. In this embodiment, a groove 3191 is formed in the cask lid 3153, and the conduit 3157 is positioned in the groove 3191, with the cask lid 3153 in place on the cask body 3159 so that the conduit 3157 may extend to the outside of the cask 3151. FIG. 44 shows this same embodiment without the cask lid in place. As shown, the conduit 3157 extends across the top of the cask body 3159 from the test port 3193 formed in the outer top lid 3195 of the second pressure vessel. The conduit 3157 is coupled to the test port 3193 with an appropriate pressure fitting 3199, which may also be constructed from stainless steel. FIGS. 45 and 46 illustrate the test port 3193 in greater detail—in FIG. 46, the cask is not shown for additional clarity. A portion of the interstitial space 3201 exists between the inner top lid 3203 and the outer top lid 3195. As indicated above, although the interstitial space 3201 may be very small, in such a small space, small, inert helium atoms may still move around within such a space. In the event that larger inert atoms are used to fill the cavity of the canister, the choices of how to form the interstitial space may be more limited to take into consideration the presently disclosed system and method of leak detection. The test port 3193 extends through the outer top lid 3195 so that it is in fluidic communication with the interstitial space 3201. Thus, when the vacuum is created in the conduit, if helium molecules are present within the interstitial space, at least some of them will be drawn into the conduit, and from there into the attached MSLD, so that they may be detected. A block diagram showing the leak detection system and illustrating the method for detecting leaks is depicted in FIG. 47. The interstitial space 3251 is formed between the inner pressure vessel 3253 and the outer pressure vessel 3255. The first end 3257 of the conduit 3259 is coupled to the test port 3261, and the second end 3263 of the conduit 3259 is coupled to the leak detector 3265, so that the interstitial space 3251, the test port 3261, the conduit 3259, and the leak detector 3265 are all in fluidic communication. The leak detector 3265 includes a vacuum system 3267, which is used to draw gas from the conduit 3259, and thus also from the interstitial space 3251, into the leak detector 3265 for analysis. The leak detector also includes a gas sensor 3269, which is preferably a mass spectrometer, and a pressure sensor 3271 to monitor the state of the vacuum established in the conduit 3259. The gas sensor 3269 is configured to detect the presence of the inert gas backfilled into the cavity 3273 of the inner pressure vessel 3253. During operation of the leak detector 3265, in one embodiment, the mass spectrometer of an MSLD is used to analyze the gas being drawn from the interstitial space while the vacuum is being established. An analysis is performed to determine if the gas being drawn contains helium atoms, and the number of helium atoms are counted. Depending upon the conditions existing at the time of testing, once the count of helium atoms passes a predetermined number, then a leak in the fluidic containment boundary that is formed by the inner pressure vessel may be said to exist. This predetermined number may vary, depending upon the particular storage container, conditions at the time the storage container was manufactured, or the conditions existing at the storage site. In other words, the presence of a single helium atom is not necessarily indicative of a leak in the inner storage container. However, a count of several helium atoms may be indicative of a leak. Further, because of the ease of the testing procedures, a particular canister might be tested two or more times to confirm the presence of excess helium in the interstitial space before a leak is determined to be positively identified. Also during operation of the leak detector 3265, in one embodiment, the pressure sensor of the MSLD is used to monitor the established vacuum in the conduit and in the interstitial space. In the event that the vacuum decreases over a short period of time from its initially established level, or alternatively if the MSLD needs to perform additional work to maintain the vacuum once established, then a leak in the fluidic containment boundary that is formed by the outer pressure vessel may be said to exist. In one embodiment, an MSLD is able to establish a vacuum in the conduit and in the interstitial space at about 10−8 atms, and if that established vacuum changes by about an order of magnitude, to about 10−7 atms within a time period of about 1 second, then this is an indicator that there is a breach in the containment provided by the outer pressure vessel. Once a test is complete, and whether or not a potential or actual leak is identified, the MSLD is decoupled from the conduit, and the removable seal may be put back in place on the test apparatus connector. Alternatively, before the removable seal is put back in place, the conduit may be backfilled with an inert gas that is different from the inert gas used to backfill the cavity of the inner pressure vessel. The two tests performed by the leak tester are very accurate, and unlike current testing systems, they do not require further investigation to determine if the test resulted in a false positive identification of a leak. The simplicity of the leak testing system and processes described above enables testing of radioactive materials containment on a regular basis, such as monthly, semi-annually, annually, or at any other chosen interval, without requiring dedicated (and costly) test equipment being connected to every individual containment system. Although dedicated equipment permits constant monitoring, it has been found that intermittent testing is sufficient and more cost effective. In addition, testing a single radioactive materials canister may be performed quickly, meaning that a reduction in manpower may be realized by implementing such systems and methods. Finally, the additional equipment that is added to a canister for performing these leak tests is not complex and requires little maintenance, thereby enabling further cost savings to be realized. V. Inventive Concept 5 With reference to FIGS. 48-52B, a fifth inventive concept will be described. The lid 4011 and top portion of a side wall 4013 for an MPC of the prior art are shown in FIG. 48. The top surface 4015 of the lid 4011 includes a beveled edge 4017, and the closure weld 4019 joining the lid 4011 to the side wall 4013 is formed in the space between the half V-shaped space between the beveled edge 4017 and the top portion of the side wall 4013. As shown, the weld is a through-thickness single groove weld V-shaped groove, although the groove could instead be J-shaped. Due the physical configuration of the lid, the sidewall, and the closure weld, this type of closure weld is not susceptible to 100% volumetric examination. A dual-walled MPC 4201 is illustrated in FIG. 49A, and this MPC 4201 is configured so that the closure weld may be subjected to 100% volumetric examination. The dual-walled MPC 4201 may be used with any style of fuel basket, such as the one described in U.S. Pat. No. 5,898,747, issued Apr. 27, 1999. In some instances it may be possible to use the dual-walled MPC 4201 without a fuel basket, depending on the intended function. Furthermore, the dual-walled MPC 4201 may be used to store and/or transport any type of high level radioactive materials and is not limited to spent nuclear fuel. As will become apparent from the structural description below, the dual-walled MPC 4201 creates two independent containment boundaries about the storage cavity 4203 which operate to contain both fluidic (gas and liquid) and particulate radiological matter within the cavity 4203. As a result, if one containment boundary were to fail, the other containment boundary will remain intact. While theoretically the same, the containment boundaries formed by the dual-walled MPC 201 about the cavity 4203 can be literalized in many ways, including without limitation a gas-tight containment boundary, a pressure vessel, a hermetic containment boundary, a radiological containment boundary, and a containment boundary for fluidic and particulate matter. These terms are used synonymously throughout this application. In one instance, these terms generally refer to a type of boundary that surrounds a space and prohibits all fluidic and particulate matter from escaping from and/or entering into the space when subjected to the required operating conditions, such as pressures, temperatures, etc. Finally, while the dual-walled MPC 4201 is illustrated and described in a vertical orientation, it is to be understood that the dual-walled MPC 4201 can be used to store and/or transport its load in any desired orientation, including at an angle or horizontally. Thus, use of all relative terms through this specification, including without limitation “top,” “bottom,” “inner” and “outer,” are used for convenience only and are not intended to be limiting of the invention in such a manner. The dual-walled MPC 4201 includes a first shell that acts as an inner shell 4205 and a second shell that acts as an outer shell 4207. The inner and outer shells 4205, 4207 are preferably cylindrical tubes and are constructed of a metal. Of course, other shapes can be used if desired. The inner shell 4205 is a tubular hollow shell that includes an inner surface 4209, an outer surface 4210, a top edge 4212 and a bottom edge 4215. The inner surface 4209 of the inner shell 4205 forms a cavity/space 4203 for receiving and storing SNF. The cavity 4203 is a cylindrical cavity formed about a central axis. The outer shell 4207 is also a tubular hollow shell that includes an inner surface 4221, an outer surface 4223, a top edge 4225 and a bottom edge 4227. The outer shell 4207 circumferentially surrounds the inner shell 4205. The inner shell 4205 and the outer shell 4207 are constructed so that the inner surface 4221 of the outer shell 4207 is in substantially continuous surface contact with the outer surface 4223 of the inner shell 4205. In other words, the interface between the inner shell 4205 and the outer shell 4207 is substantially free of gaps/voids such that the two shells 4205, 4207 are in conformal contact. This can be achieved through an explosive joining, a cladding process, a roller bonding process and/or a mechanical compression process that bonds the inner shell 4205 to the outer shell 4207. The continuous surface contact at the interface between the inner shell 4205 and the outer shell 4207 reduces the resistance to the transmission of heat through the inner and outer shells 4205, 4207 to a negligible value. Thus, heat emanating from the spent nuclear fuel loaded within the cavity 4203 can efficiently and effectively be conducted outward through the shells 4205, 4207 where it is removed from the outer surface 4223 of the outer shell via convection. The inner and outer shells 4205, 4207 are preferably both made of a metal. As used herein, the term metal refers to both pure metals and metal alloys. Suitable metals include without limitation austenitic stainless steel and other alloys including Hastelloy™ and Inconel™. Of course, other materials can be utilized. The thickness of each of the inner and outer shells 4205, 4207 is preferably in the range of 5 mm to 25 mm. The outer diameter of the outer shell 4207 is preferably in the range of 1700 mm to 2000 mm. The inner diameter of the inner shell 4205 is preferably in the range of 1700 mm to 1900 mm. The specific size and/or thickness of the shells 4205, 4207, however, is a matter of design choice. In some embodiments, it may be further preferable that the inner shell 4205 be constructed of a metal that has a coefficient of thermal expansion that is equal to or greater than the coefficient of thermal expansion of the metal of which the outer shell 4207 is constructed. Thus, when the spent nuclear fuel that is stored in the cavity 4203 emits heat, the outer shell 4207 will not expand away from the inner shell 4205. This ensures that the continuous surface contact between the outer surface 4210 of the inner shell 4205 and the outer surface 4223 of the outer shell 4207 will be maintained and a gaps will not form under heat loading conditions. The dual-walled MPC 4201 also includes a first top plate that acts as an inner top lid 4229 for the inner shell 4205 and a second top plate that acts as an outer top lid 4231 for the outer shell 4207. The inner and outer top lids 4229, 4231 are plate-like structures that are preferably constructed of the same materials discussed above with respect to the shells 4205, 4207. Preferably the thickness of the inner top lid 4229 is in the range of 99 mm to 300 mm. The thickness of the outer top lid 4231 is preferably in the range of 50 mm to 150 mm. The invention is not, however, limited to any specific dimensions, which will be dictated on a case-by-case basis and the radioactive levels of the spent nuclear fuel to be stored in the cavity 4203. The inner top lid 4229 includes a top surface 4233 with a beveled edge 4235, a bottom surface 4237, an outer lateral surface/edge 4239, and a channel 4241 formed in the top surface 4233 and set in from the beveled edge 4235. The outer top lid 4231 includes a top surface 4243 with a beveled edge 4245, a bottom surface 4247, an outer lateral surface/edge 4249, and a channel 4251 formed in the top surface 4243 and set in from the beveled edge 4245. When fully assembled, the outer lid 4231 is positioned atop the inner lid 4229 so that the bottom surface 4247 of the outer lid 4231 is in substantially continuous surface contact with the top surface 4233 of the inner lid 4229. Both the inner top lid 4229 and the outer top lid 4231 also include vent and/or drain ports 4253, 4255. During loading procedure involving spent nuclear fuel, the cavity 4203 is loaded with the spent nuclear fuel, then the inner top lid 4229 is positioned so as to enclose the top end of the cavity 4203 and rests atop brackets (not shown). Once the inner top lid 4229 is in place, a closure weld is formed to seal the inner top lid 4229 to the inner shell 4205. The top lid 4229 may be welded to the inner shell 4205 using any suitable welding technique or combinations of techniques that use a filler material. Examples of suitable welding techniques include resistance seam welding, manual metal arc welding, metal inert gas welding, tungsten inert gas welding, submerged arc welding, plasma arc welding, gas welding, electroslag welding, thermit welding. After the cavity 4203 is sealed by the closure weld, it may then be evacuated/dried via the appropriate method and backfilled with nitrogen, helium or another inert gas using the ports 4249 of the inner lid 4229 that form passageways into the cavity 4203. The ports 4249 may thereafter be filled with a metal or other wise plugged so as to hermetically seal the cavity 4203. The outer shell 4207 has an axial length that is greater than the axial length of the inner shell 4205. As such, the top edge 4225 of the outer shell 4207 extends beyond the top edge 4211 of the inner shell 4205. Similarly, the bottom edge 4227 of the outer shell 4207 extends beyond the bottom edge 4215 of the inner shell 4205. The offset between the top edges 4211, 4225 of the shells 4205, 4207 allows the top edge 4211 of the inner shell 4205 to act as a ledge for receiving and supporting the outer top lid 4231. When the inner top lid 4229 is in place, the inner surface 4209 of the inner shell 4205 extends over the outer lateral edges 4239. When the outer top lid 4231 is then positioned atop the inner lid 4229, the inner surface 4221 of the outer shell 4207 extends over the outer lateral edge 4249 of the outer top lid 4231. The top edge 4225 of the outer shell 4207 is substantially flush with the top surface 4253 of the outer top lid 4231. The inner and outer top lids 4229, 4231 are welded to the inner and outer shells 4205, 4207 respectively after the fuel is loaded into the cavity 4203. Similar to the inner top lid 4229, once the outer top lid 4231 is in place, a closure weld is formed to seal the outer top lid 4231 to the outer shell 4207. The outer top lid 4231 may be welded to the outer shell 4207 using any suitable welding technique or combinations of techniques that use a filler material. Examples of suitable welding techniques include resistance seam welding, manual metal arc welding, metal inert gas welding, tungsten inert gas welding, submerged arc welding, plasma arc welding, gas welding, electroslag welding, thermit welding. The closure welds sealing the inner and outer top lids 4229, 4231 to the inner and outer shells 4205, 4207 may be subjected to 100% volumetric examination once the welds are formed. It is to be understood that the closure weld for the inner top lid 4229 is to undergo volumetric examination before the outer top lid 4231 put in place. The dual-walled MPC 4201 also includes a first plate that acts as an inner base plate 4265 and a second plate that acts as an outer base plate 4267. The inner and outer base plates 4265, 4267 are rigid plate-like structures having circular horizontal cross-sections. The invention is not so limited, however, and the shape and size of the base plates is dependent upon the shape of the inner and outer shells. The inner base plate 4265 includes a top surface 4269, a bottom surface 4271 and an outer lateral surface/edge 4273. Similarly, the outer base plate 4267 includes a top surface 4275, a bottom surface 4277 and an outer lateral surface/edge 4279. The top surface 4269 of the inner base plate 4265 forms the floor of the cavity 4203. The inner base plate 4265 rests atop the outer base plate 4267. Similar to the other corresponding components of the dual-walled MPC 201, the bottom surface 4271 of the inner base plate 4265 is in substantially continuous surface contact with the top surface 4275 of the outer base plate 4267. As a result, the interface between the inner base plate 4265 and the outer base plate 4267 is free of gaseous gaps/voids for thermal conduction optimization. An explosive joining, a cladding process, a roller bonding process and/or a mechanical compression process can be used to effectuate the contact between the base plates 4265, 4267. Preferably, the thickness of the inner base plate 4265 is in the range of 50 mm to 150 mm. The thickness of the outer base plate 4267 is preferably in the range of 99 mm to 200 mm. Preferably, the length from the top surface of the outer top lid 4231 to the bottom surface of the outer base plate 4267 is in the range of 4000 mm to 5000 mm, but the invention is in no way limited to any specific dimensions. The outer base plate 4267 may be equipped on its bottom surface with a grapple ring (not shown) for handling purposes. The thickness of the grapple ring is preferably between 50 mm and 150 mm. The outer diameter of the grapple ring is preferably between 350 mm and 450 mm. The inner shell 4205 rests atop the inner base plate 4265 in a substantially upright orientation. The bottom edge 4215 of the inner shell 4205 is connected to the top surface 4275 of the inner base plate 4265 by a through-thickness single groove (V or J shape) weld. The outer surface 4210 of the inner shell 4205 is substantially flush with the outer lateral edge 4273 of the inner base plate 4265. The outer shell 4207, which circumferentially surrounds the inner shell 4205, extends over the outer lateral edges 4273, 4279 of the inner and outer base plates 4265, 4267 so that the bottom edge 4227 of the outer shell 4207 is substantially flush with the bottom surface 4277 of the outer base plate 4267. The inner surface 4221 of the outer shell 4207 is also connected to the outer base plate 4267 using a through-thickness edge weld. In an alternative embodiment, the bottom edge 4227 of the outer shell 4207 could rest atop the top surface 4275 of the outer base plate 4267 (rather than extending over the outer later edge of the base plate 4267). In such an embodiment, the bottom edge 4227 of the outer shell 4207 could be welded to the top surface 4275 of the outer base plate 4267. When all of the seal and closure welds discussed above are completed, the combination of the inner shell 4205, the inner base plate 4265 and the inner top lid 4229 forms a first hermetically sealed structure surrounding the cavity 4203, thereby creating a first pressure vessel. Similarly, the combination of the outer shell 4207, the outer base plate 4267, and the outer top lid 4231 form a second sealed structure about the first hermetically sealed structure, thereby creating a second pressure vessel about the first pressure vessel and the cavity 4203. Theoretically, the first pressure vessel is located within the internal cavity of the second pressure vessel. Each pressure vessel is engineered to autonomously meet the stress limits of the ASME Code with significant margins. FIG. 49B illustrates a single-walled MPC 4285 which is constructed in a similar manner as each pressure vessel of the double-walled MPC 4201 discussed above. This single-walled MPC 4287 includes a side wall 4289 seal welded to a base plate 4291, and a top plate 4293. The top surface 4295 of the top plate 4293 includes a beveled top edge 4297 and a channel 4299 set in from the top edge 4297. Having the lid configured with the channel 4299 makes it so that the closure weld may be subjected to 100% volumetric examination. All other parts of the single-walled MPC 285 may be constructed in the same manner described above. A detailed view a top plate 4311 and the closure weld 4313 sealing the top plate 4311 to a side wall 4315 of an MPC are illustrated in FIG. 49C. The channel 4317 in the top surface 4319 is set in from the beveled top edge 4321. The channel 4317 extends below the top surface 4319 at least as much as does the bevel of the beveled top edge 4321. In some embodiments, depending upon the configuration of the probe being used, it may be desirable to have the channel 4317 extend deeper below the top surface than the bevel in order to accommodate the probe. The channel 4317 is sufficiently wide so that a probe used for examining the closure weld may be placed within the channel 4317 and moved circumferentially around the top plate 4311 for purposes of achieving 100% volumetric examination of the closure weld. For some types of probes, the channel may be as wide as 2″ to 3″, although these dimensions may vary significantly to accommodate the configuration of the probe used to examine the closure weld. The side wall 4323 of the channel 4317 nearest the beveled top edge 4321 is placed at an angle that is approximately parallel to the angle of the beveled top edge 4321. However, in some embodiments the angle of this channel side wall may vary from the angle of the top beveled edge by 5°-20° or more, depending upon the configuration of probe being used. The side wall 4323, however, may be formed at any angle relative to the beveled top edge 4321. The opposite wall 4325 of the channel 4317 may have any configuration, from a well-defined wall, as is shown, to a curved or flat surface adjoining the bottom 4327 of the channel 4317. One embodiment of a top plate 4331 is shown in FIG. 50 with ports 4333 positioned in the central portion 4335 of the top surface 4337 of the top plate 4331, radially inward from the channel 4339. The ports 4333 may serve any desired purpose for the MPC for which the top plate 4331 is used, and the different ports may be used for different purposes. Examples of purposes for the ports include their use as vent ports, as vacuum ports, as drain ports, as backfill ports, as test ports, among others. Another embodiment of a top plate 4341 is shown in FIG. 51. In this embodiment, the ports 4343 are positioned within the channel 4345. In other embodiments, ports may be positioned both within the channel and in the central portion of the top surface of the top plate. FIGS. 52A and 52B illustrate the process of performing the 100% volumetric examination of the closure weld after it has been formed. With the top plate in place on the top opening of the sidewall, the top plate having a channel as described above, the closure weld may be formed by automated equipment, such as is well known in the art. In order to volumetrically examine the closure weld, probes are mounted on a support arm capable of rotating and positioning the probes to perform the volumetric examination of the closure weld. For example, the probes may be mounted on the same type of weld arm that is used in the automated process for forming the closure weld. The volumetric examination may be carried out once the entire weld is formed. Only the end of the support arm 4371 is illustrated in FIG. 52A to simplify the drawing. It is to be understood that the support arm may have any appropriate configuration that is capable of supporting the probes and moving them around the top plate to perform the volumetric examination, as many different types and configurations of such support arms are well-known in the arts, including combination rotary/articulating robotic arms. Two probes 4373, 4375 are affixed to the end of the support arm 4371, and the support arm is configured for automated or remote positioning of the probes so that the volumetric examination of the closure weld may be performed. The first probe 4373 is positioned on the outside of the top of the side wall 4377, and the second probe 4375 is shown just prior to being positioned within the channel 4379 formed in the top surface 4381 of the top plate 4383. This second probe 4375 is shown positioned within the channel 4379 in FIG. 52B. Once the two probes are in position, the entire volume of a portion of the closure weld is disposed between the two probes, and that entire volume may be volumetrically examined. By activating the two probes and moving them synchronously around the top plate, maintaining their relative position with respect to the closure weld, the entirety of the weld is passed between the two probes in one circumscription of the top plate. It is therefore possible, with the appropriate examination technology, to perform a 100% volumetric examination of the closure weld. Using well-known processes associated with the selected examination technology, the integrity of the entire closure weld may be determined from the examination. In the embodiment of FIG. 52A, the entire closure weld is formed first, followed by the volumetric examination of the closure weld. In the embodiment of FIG. 52B, the weld head 4385 extends from the same support arm (not shown in FIG. 52B) as the probes 4373, 4375. The weld arm then moves the weld head around the top edge of the top plate to form the closure weld, and the probes trail the weld head to perform the volumetric examination. This embodiment may be used to form the weld and substantially concurrently volumetrically examine the weld. For a multi-pass closure weld, having the probes trail the weld head in this manner enables a separate volumetric examination of each pass of the closure weld. Due to the heat generated from the welding process, which may interfere with the examination process, this embodiment may be best suited for use in pools or in the presence of a coolant, such as a flow of demineralized water In certain embodiments, a Linear Scan-Phased Array UT system may be used to examine the closure weld, and for such embodiments the probes are ultrasound transducer probes. Such a UT system is capable of conducting the 100% volumetric examination of the closure weld within a matter of minutes. Beneficially, with the top plate configured as described above and with use of the two probes, no human activity needs to be directly involved for placing the top plate, forming the closure weld, or examining the integrity of the closure weld, so that work crews are not exposed to any significant doses of radiation. In embodiments where a UT system is used outside of a pool of water or other fluid, a coupling agent, such as demineralized water or an appropriate gel, may be introduced between the transducer probes and the top plate and/or side wall to increase the amount of ultrasound energy that passes into the closure weld, thereby improving the volumetric examination. As is well known in the art of UT, only small amounts of the coupling agent are needed to form a thin film, minimizing air gaps, between the transducer probe and the parts of the MPC into which the ultrasound energy is being directed. Therefore, a simple drip system suffices to introduce a coupling agent such as demineralized water to the process of volumetric examination described herein. In embodiments involving a high heat load canister, to ensure that the metal temperature of the weld mass is not too high for an accurate UT reading, it may be necessary to circulate cooling water through the MPC using the vent and drain ports in the lid before performing the volumetric examination. As an alternative, the use of a coupling agent for ultrasound energy, such as demineralized water, between the transducer probes and the MPC helps to insure that the volumetric examination is performed at a uniform temperature, thereby preserving the UT calibration integrity. VI. Inventive Concept 6 With reference to FIGS. 53-59, a sixth inventive concept will be described. FIG. 53 illustrates an apparatus for transferring spent nuclear fuel in the form of a transfer cask 5011. The transfer cask 5011 includes a cylindrical inner shell 5013 which forms a cavity 5015 along with the top lid 5017 and the bottom lid 5019. As shown, a canister 5021 for holding spent nuclear fuel is disposed within the cavity 5015. The inner shell 5013 has a longitudinal axis 5023, and the inner shell 5013 has a slightly larger radius, measured from the longitudinal axis 5023, as compared to the canister 5021, to create an annulus 5025 of space between the inner shell 5013 and the canister 5021 disposed in the cavity 5015. This annulus 5025, as discussed in greater detail below, serves to enable cooling of the canister 5021 by ventilation with atmosphere. The transfer cask further includes an intermediate shell 5027 and an outer shell 5029. Each of the inner shell 5013, the intermediate shell 5027, and the outer shell 5029 are preferably made from carbon steel, with the top of each welded to a top flange 5031, and the bottom of each welded to a bottom flange 5033. The intermediate shell 5027 is disposed concentrically around and spaced apart from the inner shell 5013, thereby forming a second annulus 5035. This second annulus 5035 is capable of holding a gamma absorbing material such as concrete, lead, or steel. Lead is preferred because it most effectively provides gamma shielding for the radioactive spent nuclear fuel once it is placed within cavity 5015. The outer shell 5029 is disposed concentrically around and spaced apart from the intermediate shell 5027, thereby forming a third annulus 5037. This third annulus 5037 is capable of holding a neutron absorbing material such as water or the aforementioned aluminum trihydrate-boron carbide-epoxy mixture. As shown, the third annulus 5037 includes panels of a metal matrix composite. For alternative embodiments in which water is to be used in the third annulus, U.S. Pat. No. 7,330,525 describes a manner in which the outer shell may be formed, in order to contain water, and a process for using water as a neutron absorber in the transfer cask during transfer of a canister containing spent nuclear fuel. The top lid 5017 is securable to the top flange 5031 by extending bolts (not shown) through the top lid 5017 to engage the top flange 5031. The top lid 5017 is typically only secured to the top flange 5031 once the canister 5021 is in place within the cavity 5015 during the transfer process. A central opening 5039 in the top lid 5017 provides access to the canister 5021 for performing certain handling operations with respect to the canister 5021 while the top lid 5017 is secured to top flange 5031. Referring to FIG. 54A, the top flange 5031 is integrally formed through forging and machining so that it does not include any joints, welds, or seams, and so that it does not include parts that are separately formed and then subsequently joined together. The top flange 5031 is machined to include two trunnions 5041 to be used for lifting the transfer cask with a crane. As shown in FIGS. 54A-54C, the trunnions may be of a variety of cross sections such as round trunnions 5041 (FIG. 54A), rectangular trunnions 5041b (FIG. 54B), obround trunnions 5041c (FIG. 54C), oblong trunnions, and the like. The cross-sectional form of the trunnions may be any shape according to design choice, with specific implementations limited only by the equipment used to hoist the transfer cask. More than two trunnions may be machined as part of the top flange, based upon design choices and the lifting system with which the transfer cask is to be used. For purposes of stability during lifting, the trunnions are distributed approximately equidistantly around the top flange. The top flange 5031 also includes a seating groove 5043 for a sealing ring (not shown), which serves as a seal, against the canister and within the annulus, when the canister is placed in the cavity. A plurality of ventilation channels 5045 are included in the top flange 5031, with internal channel inlets 5047 on the interior surface 5049 of the top flange 5031 located below the seating 5043 so that when a canister is placed, air is directed through the ventilation channels 5045. The ventilation channels 5045 open up to the exterior of the top flange 5031, and to the exterior of the transfer cask, at external channel outlets 5051 so that the ventilation channels fluidically connect the annulus 5025 with the exterior of the top flange 5031 and the transfer cask. The ventilation channels 5045 through the top flange 5031 may have a variety of forms or paths, however, because air is being used to ventilate the transfer cask, and unlike water, air is not a good neutron absorber, the one design constraint for the ventilation channels is that the paths of the ventilation channels preclude a direct line of travel from within the cavity to the exterior of the top flange. With this design constraint on the ventilation channels of the top flange, emissions from the canister cannot pass through an all-air pathway from the canister to the exterior of the transfer cask. The integral design of the trunnions 5041 as part of the top flange 5031 serves to eliminate joints between the top flange and the trunnions, thereby significantly improving the fidelity of structural integrity of the overall lifting system (as compared to the prior art, in which the trunnions are joined to the top flange by welding or a threaded joint). The top flange 5031 is also enlarged as compared to top flanges of the prior art, but still keeping within the constraints of the size of the cask pit in the pool and the lifting limit of the cask crane. Even though enlarged, the top flange 5031, inclusive of the integral trunnions 5041, has a smaller outer diameter as compared to the outer shell 5029. To aid in preventing damage that may be caused by protruding trunnions in the event of a transfer cask accidentally tipping into other casks, each trunnion 5041 is disposed within a recess 5053 of the top flange 5031. The larger top flange 5031 also serves to provide increased shielding in the top region of the cask where most human activity (to weld and dry the canister) occurs. Turning back to FIG. 53, the bottom lid 5019 is secured to the bottom flange 5033 by a plurality of bolts (not shown) that extend through holes in the bottom flange 5033 the engage the bottom lid 5019. The bottom lid 5019 includes an impact zone 5061 positioned directly beneath the cavity 5015. The bottom lid 5019 also includes a gamma-absorbing layer 5063, such as lead, below the impact zone 5061. To be most effective in absorbing impacts from accidental falls of the transfer cask, the impact zone 5061 extends substantially under the entirety of the cavity 5015. The impact zone includes an impact absorbing structure 5065 which can serve to cushion the fall of a canister loaded into the transfer cask, thereby providing some damage protection to the fuel in the event of a handling mishap while the transfer cask is being moved around the building or plant site. As shown, the impact absorbing structure 5065 is formed by a plurality of cylindrical tubes 5067 within the bottom lid 5019. These tubes 5067 are distributed throughout the impact zone 5061, with their longitudinal axes aligned with a major dimension (i.e., the diameter) of the bottom lid 5019. The thickness, number of tubes, and the cross-sectional shape of the tubes are a matter of design choice based upon the particular implementation. Factors that may be taken into consideration for these design choices include estimated drop height (based on the operational procedures of the facility), the weight of the canister, and the weight of the loaded transfer cask. Computations have shown that a set of parallel 2-inch tubes distributed throughout the impact zone 5061 can limit the impact load experienced by a 40-ton canister, placed with a transfer cask, falling from 18 inches onto a concrete pad to a g-force of less than 25 (in the absence of the impact limiter, the g-force may shoot up to over 100). A plurality of ventilation channels 5071 are included in the bottom lid 5029, with external channel inlets 5073 on the external surface 5075 of the bottom lid and internal channel outlets 5077 located so that the ventilation channels 5071 can direct an air flow into the annulus 5025. A plurality of ventilation channels configured in this manner are formed approximately equidistantly around the bottom lid to provide cooling ventilation to the canister 5021 outside of the storage pool. At the point of intersection between the channel outlets 5077 and the annulus 5025, the bottom flange 5033 is configured with a chamfered surface 5079 to broaden out the annulus 5025, thereby providing an enlarged space about the base of the canister 5021 into which air may be drawn through the ventilation channels 5071. Each channel inlet 5073 is configured to receive a sealing plug (not shown), which may threadably engage the channel inlet 5073 to provide a seal and turn the ventilation channel and annulus into a “blind” cavity that does not have ingress through the bottom lid. Similar plugs may be placed in the channel outlets of the top flange, thereby rendering the entire annuls cavity into a “blind” cavity. Such plugs may be placed under circumstances where it is desirable to protect the ventilation channel from ingress of contaminated water or other matter, either solely at the bottom of the transfer cask, or at the top and the bottom. A second example of a ventilation channel 5081 is shown in FIG. 57, and a plurality of ventilation channels 5081 configured in this manner are formed approximately equidistantly around the bottom lid to provide cooling ventilation to the canister 5021 outside of the storage pool. Again, at the point of intersection between the channel outlets 5083 and the annulus 5025, an enlarged space 5085 is included about the base of the canister 5021 into which air may be drawn through the ventilation channels 5081. The channel inlets 5087 may also be configured to receive a sealing plug (not shown). The ventilation channels 5071 through the bottom lid 5029 may have a variety of forms or paths, however, because air is being used to ventilate the transfer cask, and unlike water, air is not a good neutron absorber, the one design constraint for the ventilation channels is that the paths of the ventilation channels preclude a direct line of travel from within the cavity to the exterior of the bottom lid. With this design constraint on the ventilation channels of the bottom lid, emissions from the canister cannot pass through an all-air pathway from the canister to the exterior of the transfer cask. FIGS. 56 and 57 illustrate another alternative embodiment of the bottom lid 5091 and an integrated ventilation channel. In this embodiment, the ventilation channel is a toroidal-shaped distribution channel 5093 having a single channel inlet 5095 and a plurality of channel outlets 5097 which are positioned to fluidically connect the annulus, formed between the inner shell of the transfer cask and the canister placed in the cavity, with the exterior of the bottom lid 5091 and the transfer cask. The radial position of the channel inlet 5095 is different than the radial position of the channel outlets 5097 so that the configuration of the ventilation channel 5093 precludes a direct line of travel from within the cavity to the exterior of the bottom lid. A transfer cask which includes the annulus between the inner shell and the canister, the ventilation channels in the top flange, and the ventilation channels in the bottom flange, configured in any of the manners discussed above, when out of a storage pool allows ambient air to ventilate up the annulus to enhance the heat removal efficacy of the cask. Calculations have shown that a mere ¾ inch wide annulus can reduce the fuel cladding temperature by as much as an additional 20° C., in comparison to a blind annulus with stagnant air (which is the state-of-the-art). And, as compared to a water-cooled annulus, a passive ambient air-cooled annulus is much simpler, easier to use, and easier to maintain, thereby resulting in greater operational reliability. Such a transfer cask will remove decay heat from the canister by ventilation action. For low heat canisters (those generating less than about 18 kW), the natural ventilation through the annulus coupled with heat dissipation from the external surfaces of the cask are sufficient to keep the contents of the canister from overheating. In circumstances where additional cooling is needed for higher heat load canisters, beyond the cooling that can be provided by ventilation of ambient air, chilled air can be forced through the annulus. One such system is shown schematically in FIG. 58. And, even a forced air system is simpler and easier to use and maintain than a cooled water system. A forced air system is most easily used when the bottom lid includes an integrated ventilation channel with a single channel inlet, such as is shown in FIG. 56. During use, an air compressor 5111 operates to store compressed air in a compressed air tank 5113, and the air outlet 5115 of the compressed air tank 5113 is fluidically coupled though an appropriate air line 5117 to the channel inlet 5119 of the bottom lid of the transfer cask 5121. The compressed air tank 5113 itself may be cooled by ambient air, or it may be cooled by an active cooling or refrigeration system 5123. As those of skill in the art will recognize, decompression of air naturally decreases the temperature of that air, so that the amount of cooling needed for the compressed air tank 5113 will depend upon the heat dissipation needs of the transfer cask. For example, a refrigeration system may be used to cool the compressed air tank to a temperature as low as 5° C., thereby causing the decompressed air from the compressed air tank to be cooler still when it is directed into the annulus of the transfer cask. The decompressed air is delivered into the ventilation channel of the bottom lid, and then into the annulus, by the positive pressure of expansion upon release from the compressed air tank. The air compressor and compressed air tank are sized to provide the cooled air at a sufficiently high velocity to ensure turbulent flow conditions within the annulus. Calculations have shown that a 50 HP compressor is adequate to cool a canister with as much as 35 kW heat load. The chilled air is heated within the annulus and exits the transfer cask through the ventilation channels in the flange. As an alternative to using a compressed air tank and an air compressor, chilled air may alternatively be forced into the annulus by use of a blower. The advantages of a forced air cooling system include greater simplicity, as compared to a water cooled system, use of single phase cooling medium (air rather than water) and mitigation of the concerns of leakage (no water spillage) at the flanged or screwed joints. The performance of the system is easily monitored by measuring the temperature of the exiting heated air from the cask FIG. 59 is a flowchart showing the process of moving a transfer cask, as described above with ventilation channels, loaded with a canister from a pool for transport or storage of the canister. The process starts 5121 with a fully loaded canister in the cavity of transfer cask without the top lid in place. The process of loading the canister is well-known to those of skill in the art, and so they are not discussed herein. As the transfer cask sits in the pool, one or more plugs may be in place in the bottom lid to seal off the ventilation channels to make the ventilation channels and the annulus a “blind” cavity, thereby protecting from ingress of contaminated water. Without the plugs in place, water fills the annulus and helps to remove heat generated by the spent nuclear fuel in the canister. The hoist of a crane is lowered into the pool and secured to the trunnions of the transfer cask. Once the hoist is secured to the trunnions, the crane lifts 5123 the transfer cask, along with the canister payload, out of the storage pool. The transfer cask is designed so that at this stage in the process, the combined weight of the transfer cask and payload is equal to or less than the rated lifting capacity of the crane. Once lifted out of the storage pool, the crane sets transfer cask down 5125 in a staging area. At this point, the canister contains pool water in addition to the spent nuclear fuel. This pool water acts as a neutron absorber as long as it is in the canister, and it removed from the canister in order to store the spent nuclear fuel in a dry-state. In the event that one or more plugs are in place in the bottom lid, they are removed 5127 to allow ventilated cooling by circulation of atmospheric air through the annulus. As an alternative, at this point, a compressed air tank is fluidically coupled to the channel inlet of the bottom lid using an appropriate hose and coupling. The compressed air tank is coupled to an air compressor so that compressed air is maintained in the tank during use. Compressed air from the tank is decompressed and passed into the channel inlet during the remaining steps of moving the transfer cask while it is loaded with the canister. Once the transfer cask is ventilated, the pool water in the canister is pumped out 5129, and the spent nuclear fuel in the canister is allowed to dry. The canister is then backfilled with an inert gas, such as helium, and sealed. The cask lid is then secured 5131 to transfer cask. The transfer cask is then lifted by the crane and moved to a position above another cask 5133, at which point the bottom lid is removed and the canister is lowered into the other cask 5135. The other cask may be a storage cask, if the spent nuclear fuel is to be stored long-term, or it may be a transport cask suitable for moving spent nuclear fuel over long distances. Once the canister is removed from the transfer cask, the transfer cask may be reused to perform the above described procedure again. To reuse the transfer cask, the one or more plugs are again put in place in the bottom lid to seal off the ventilation channels. As used throughout, ranges are used as shorthand for describing each and every value that is within the range. Any value within the range can be selected as the terminus of the range. In addition, all references cited herein are hereby incorporated by referenced in their entireties. In the event of a conflict in a definition in the present disclosure and that of a cited reference, the present disclosure controls. 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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description | This application claims priority from U.S. Provisional Patent Application No. 60/773,396, filed Feb. 15, 2006, which is hereby incorporated by reference. The present invention relates to charged particle beam systems and, in particular, to applying a sputter coating to a work piece. Integrated circuits and other microscopic or nano-scale devices are often processed using charged particle beams for imaging or altering a work piece. For example, focused or shaped ion beams can be directed toward a work piece to micromachine it or to deposit material by beam-induced deposition. Electron beams can similarly be used to etch, deposit, or form images of the work piece using scanning or transmission electron microscopy. In one application, an ion beam is used to mill a work piece to expose buried features of a work piece to view or modify the buried feature. For example, layers of an integrated circuit may be milled away to expose a buried conductor, which may then be severed or connected to another conductor by ion beam deposition of a new conductive pathway. Ion beams are also used to expose a cross section of a work piece so that the cross section can be viewed using charged particle beam microscopy. A cross section exposed by an ion beam can then be observed using scanning ion microscopy or electron microscopy. For example, U.S. Pat. Pub. No. 20040158409 of Teshima et al. for a “Defect Analyzer” describes methods of cutting cross sections using a focused ion beam. In one technique, referred to as “slice and view,” a cross section is exposed by focused ion beam machining and an image is formed of the cross section. Additional material is then removed from the cross section wall to expose a new wall face offset slightly from the first wall face, and an additional image is formed. By sequentially removing a small amount of material from the cross section wall and forming a series of images, information about a three dimensional structure is obtained. Ion beam machining can be facilitated by using an etch-enhancing gas, such as chlorine, fluorine, or iodine, as described in U.S. Pat. No. 5,188,705 for “Method of Semiconductor Manufacture” to Swanson et al. Ions beams are also used to extract samples from a substrate for viewing in a transmission electron microscope, as described, for example, in U.S. Pat. Pub. 2004/0251412 of Tappel for “Method and Apparatus for Manipulating a Microscopic Sample.” A sample is freed from the substrate by milling around the sample and undercutting the sample. The sample is typically attached to a probe and moved to a TEM sample holder. The sample also may be thinned or shaped by the ion beam. The impact of the ions or electrons in a charged particle beam can damage sensitive work pieces. It is a common practice, therefore, to apply a protective layer over the surface of the work piece before performing charged particle beam processing. Also, non-conductive work pieces tend to accumulate electric charges during charged particle beam processing, both from the charges delivered by the beam and from the ejection of secondary charges caused by the impact of the beam. When a work piece becomes charged, it deflects the charged particles in the primary beam, thereby reducing resolution. Work piece charging can also reduce the number of secondary particles emitted upon impact of the charged particle beam. Because the secondary particles are often used to form images of the work piece, the reduction in the emission of secondary charges degrades the ability of the system to form an image of the work piece. One method of reducing sample charging, described in U.S. Pat. No. 4,639,301 to Doherty for “Ion Beam Processing,” involves the use of an electron flood gun to neutralize accumulated positive charges. Another method of preventing the accumulation of charge is by depositing a conductive layer onto the surface of the work piece to provide a path through which the electrical charge can be discharged. One method of depositing a protective coating, whether conductive or non-conductive, is by charged-particle-beam-induced deposition. In charged-particle-beam-induced deposition, a charged particle beam is scanned over the work piece surface while a precursor gas is introduced in the vicinity of the beam impact area. The precursor gas molecules are chemically decomposed in the presence of the beam to leave a coating on the surface. Volatile decomposition products float away from the surface and are removed by the system vacuum pump. For example, in ion-beam-assisted gas deposition of a conductive coating, a charged particle beam of between about 5 kV and 30 kV is scanned over the surface while a precursor gas, such as a metaloorganic gas, for example, tungsten hexacarbonyl, is directed to the substrate in the vicinity of the beam landing area. Either an ion beam or an electron beam can be used to decompose the precursor gas to deposit a coating, but electron beams cause less surface damage because electrons have significantly less mass than ions. Although electron beams cause less surface damage than ion beams, the electron beams can still damage some sensitive materials, such as the photoresist or low-k or ultra low-k dielectric layers that are used in integrated circuit fabrication. The beam-induced damage can cause changes in the dimensions being measured on the work piece and changes in the profile being observed. Another method of covering a surface with a layer of a material is sputtering. Sputtering, also known as physical vapor deposition, is a physical, as opposed to chemical, process in which molecules or atoms are knocked from a material source by momentum transfer and are then deposited onto a target surface. Sputter coating systems are available commercially and are used in integrated circuit fabrication. Such systems are typically designed to deposit a metal layer over an entire wafer. The metal layer is then patterned using a photolithography process to form a conductive pattern to connect elements of the circuit. Such systems typically use a plasma of ionized argon gas, the argon ions colliding with a target to knock material from the target onto the work piece. Such systems typically use electric fields to provide energy to ionize the gas, and may use magnetic fields to trap electrons to facilitate ionization. The use of these production systems to sputter coat samples for quality control or defect analysis may be prohibitively expensive. Other systems, such as the Gatan Ion Sputter Coater can provide a sputter coating on small samples, but will not accommodate a typical semiconductor wafer, and so the wafer must be broken. In either case, coating requires moving the work piece between the sputter system and the charged particle beam system, which can entail multiple evacuations of vacuum chambers, potential contamination as the work piece is moved between system, and additional time and manpower. An object of the invention is to provide a coating onto a work piece surface for charged particle beam processing while eliminating or minimizing damage to the surface. The invention includes a method and apparatus for providing a sputter coating on a surface to protect the work piece for charged particle beam processing. In one embodiment, the system includes a source material within a charged particle beam system. A charged particle beam is directed toward a source material to sputter the material from the source material onto the work piece surface. The charged particle beam does not need to be directed to the work piece, so damage from the impact of the charged particles onto a sensitive work piece surface can be avoided. In other embodiments, the sputtering may be performed using a conventional sputtering apparatus, either within a separate chamber or within the same vacuum chamber as the charged particle beam system. After the sputter coating is applied, the work piece is processed with a charged particle beam. The foregoing has outlined rather broadly the features and technical advantages of the present invention in order that the detailed description of the invention that follows may be better understood. Additional features and advantages of the invention will be described hereinafter. It should be appreciated by those skilled in the art that the conception and specific embodiment disclosed may be readily utilized as a basis for modifying or designing other structures for carrying out the same purposes of the present invention. It should also be realized by those skilled in the art that such equivalent constructions do not depart from the spirit and scope of the invention as set forth in the appended claims. The invention provides for the sputter application of a protective or conductive layer or coating for charged particle beam processing. Charged particle beam processing can include, for example, micromachining or imaging by an ion beam or by an electron beam. For many applications of charged particle beam processing, it is desirable to protect the surface of the work piece before performing the charged particle beam operation. Charged particle beam deposition typically requires the charged particles in the beam to impact the surface, which can damage the surface. Embodiments of the invention can be used to apply a sputter coat to protect the work piece surface materials, and can also be used to provide a conductive material over an area to reduce charge to achieve better resolution and charge control of the work piece. Sputtering provides for charged particle beam coating without the charged particles landing on the sample, thereby preventing or minimizing damage to the work piece from the charged particle beam. FIG. 1 shows a preferred embodiment of the invention. In step 102, the work piece is loaded into a vacuum chamber for sputtering. In step 104, a protective or conductive coating is sputtered onto the work piece. The coating can cover the entire work piece surface, or just a portion of the surface. In optional step 106, the work piece is moved to a charged particle beam system vacuum chamber. Step 106 is optional, because in some embodiments, the sputtering and the charged particle beam processing occur in the same vacuum chamber. For example, the sputtering may be performed by the charged particle beam, or by a sputtering system that uses a plasma to knock material from a sputter material source within the charged particle beam vacuum chamber. In step 108, the work piece, having a sputtered coating, is processed using a charged particle beam. For example, a cross section may be milled in the work piece, and then imaged, or a sample may be extracted from the work piece using the charged particle beam. Some embodiments of the invention provide for in-situ sputter work piece coating for sample protection and local area charge control. By in-situ is meant that the coating can be applied in the sample vacuum chamber as the charged particle beam system, so the work piece does not need to be moved between systems for coating and charged particle beam processing. Such embodiments can be used with currently available charged particle system with little or no system modifications, and do not require an additional plasma-type sputtering system. These embodiments provide a broad capability and can be implemented in many different ways depending on the requirements of any specific application. In-situ embodiments of the system can provide a new capability for both fab-based and lab-based systems for little or no extra cost depending on the material one wants to sputter onto the work piece. In-situ embodiments allow for controlled local sputter coating on a localized portion of a work piece surface. Almost any material that can be sputter coated using conventional sputtering techniques can also be sputtered using embodiments of the invention. Sputter coatings can be applied onto any work piece in the charged particle beam vacuum chamber. Sputter coating is often a preferred method for sample protection when working with materials such as low k dielectrics, including ultra-low k dielectrics, and photoresists. Embodiments can be used to apply a sputter coating to protect the work piece surface materials, and can also be used to provide a conductive material over an area to reduce charge to achieve better resolution and charge control of the sample. In accordance with an in-situ embodiment, a sputter material source is provided between the ion beam source and the work piece. The charged particle beam is directed toward the sputter source material, and material is sputtered from the sputter source material onto the work piece. The source material can comprise, for example, tungsten, chromium, titanium, copper, aluminum, or any other metallic or non-metallic material having a suitable vapor pressure that allows operation of the charged particle beam system. FIG. 2 shows a typical dual beam FIB/SEM system 200 that can be used to implement an embodiment of the invention. Dual beam systems are available commercially from FEI Company, the assignee of the present invention. Focused ion beam system 200 includes an evacuated envelope 210 having an upper neck portion 212 within which are located an ion source 214 and an ion beam focusing column 216 including extractor electrodes and an electrostatic optical system (not shown). Ion beam 218 passes from ion source 214 through ion beam focusing column 216 and between electrostatic deflectors schematically indicated at 220 toward work piece 222, which comprises, for example, a semiconductor device positioned on movable X-Y stage 224 within lower chamber 226. An ion pump 228 is employed for evacuating neck portion 212. The chamber 226 is evacuated with turbomolecular and mechanical pumping system 230 under the control of vacuum controller 232. The vacuum system provides within chamber 226 a vacuum of between approximately 10−7 Torr (10−7 mbar) and 5×10−4 Torr (1.7×10−4 mbar). If an etch assisting gas, an etch retarding gas, or a deposition precursor gas is used, the chamber background pressure may rise, typically to about 10−5 Torr (10−5 mbar). High voltage power supply 234 is connected to ion source 214 as well as to appropriate electrodes in ion beam focusing column 216 for forming an ion beam 218 and directing the same toward work piece 222. Deflection controller and amplifier 236, operated in accordance with a prescribed pattern provided by pattern generator 238, is coupled to deflection plates 220 whereby ion beam 218 may be controlled to trace out a corresponding pattern on the upper surface of work piece 222, or, in accordance with the invention, on a sputter material source such as a nozzle 240 of a gas injection system 242 described below. Because of its small diameter, nozzle 240 is also referred to sometimes as a needle. In some systems the deflection plates 220 are placed before the final lens, as is well known in the art. The ion source 214 typically provides a metal ion beam of gallium, although other ion sources, such as a multicusp or other plasma ion source, can be used, using for example, oxygen, argon or another noble gas for ions. The ion source 214 typically is capable of being focused into a sub one-tenth micron wide beam at work piece 222 for either modifying the work piece 222 by ion milling, chemically-enhanced etch, material deposition, or for the purpose of imaging the work piece 222. Rather than focusing the beam to a point, the ion column may also provide a shaped beam. Shaped beams can provide larger currents to reduce sputtering time. A secondary electron detector 244, such as a charged particle multiplier used for detecting secondary ion or electron emission for imaging is connected to amplifier 246. The amplified signals are converted into digital signals and subjected to signal processing by the signal processor unit 248. The resulting digital signal is to display an image of work piece 222 on the monitor 250. A scanning electron microscope 260, along with a power supply and control unit 262, is also provided with the FIB system 200. An electron beam 263 is emitted from a cathode 264 by applying voltage between cathode 264 and an anode 268. Electron beam 263 is focused to a fine spot by means of a condensing lens 270 and an objective lens 272. Electron beam 263 is scanned two-dimensionally on the specimen by means of a deflection coil 276. Operation of condensing lens 270, objective lens 272, and deflection coil 276 is controlled by power supply and control unit 262. Some embodiments include a detector for detecting electrons that are transmitted through a thin work piece, for use in a scanning transmission electron microscopy mode. While the presence of an electron beam and an ion beam is useful for most applications, the invention is not limited to a dual beam system. Electron beam 263 can be focused onto work piece 222, which is on movable X-Y stage 224 within lower chamber 226. When the electrons in the electron beam strike work piece 222, secondary electrons are emitted. These secondary electrons are detected by secondary electron detector 244 or by backscattered electron detector 266, which are connected to an amplifier 246. The amplified signals are converted into digital signals and subjected to signal processing by the signal processor unit 248. The resulting digital signal is to display an image of work piece 222 on the monitor 250. Gas injection system 242 extends into lower chamber 226 for introducing and directing a gaseous vapor toward work piece 222. U.S. Pat. No. 5,851,413 to Casella et al. for “Gas Delivery Systems for Particle Beam Processing,” assigned to the assignee of the present invention, describes a suitable gas injection system 242. Another gas delivery system is described in U.S. Pat. No. 5,435,850 to Rasmussen for a “Gas Injection System,” also assigned to the assignee of the present invention. The gas injection system includes nozzle 240, typically made of chromium. As describe below, nozzles made of various material, coated with various materials, or having affixed thereto, volumes of different material, can be used to provide a variety of materials to sputter onto work piece 222. Nozzle 240 can be moved into the path of ion beam 218 so that ions in the beam sputter material from nozzle 240 onto the work piece 222. Nozzle 240 can then be moved out of the beam path to allow processing of work piece 222 by the beam. One or more electric motors 280 are used to move the nozzle along one or more axes. A door 282 is opened for inserting work piece 222 onto stage 224, which may be heated or cooled, and also for servicing an internal gas supply reservoir, if one is used. The door 282 is interlocked so that it cannot be opened if the system is under vacuum. The high voltage power supply provides an appropriate acceleration voltage to electrodes in ion beam focusing column 216 for energizing and focusing ion beam 218. A “load lock” can be used so to avoid having to evacuate the entire sample area to load or unload a work piece. As describe above, in some embodiments, the source material for the sputtering is in the form of the needle of a gas injection system (GIS). Because focused ion beam systems often include one or more of GIS systems, the invention can be readily implemented in existing systems. The GIS system is typically secured to the vacuum chamber wall, and is mounted on a positioned adjustment mechanism that allows the nozzle to be moved to direct the gas to the desired location, or to be withdrawn from the vicinity of the sample when not in use. For a charged particle beam system that includes multiple GIS systems, the GIS nozzles of different GIS systems can be made of different materials, so that sputter coatings of different materials can be applied by using different ones of the GIS nozzles as the sputtering material source. The GIS nozzles can be made from the desired sputter source material, or can be coated with the sputter source material. Alternatively, one or more volumes of source material can be attached to one or more the GIS nozzles, for example, by using FIB deposition, brazing, welding, or by using an adhesive. FIG. 3 shows an embodiment of the invention. An ion column 302 provides an ion beam 310 that is directed to a sputter material source, such as GIS nozzle 304. A work piece 306 is supported on a work piece stage 308 having at least an x-y translation capability. The ion beam 310 sputters atoms 314 or molecules from GIS nozzle 304, atoms 314 being ejected toward the surface of work piece 306. The ion beam typically has a diameter (full width, half maximum) upon impact of less than 10 μm, more preferably less than, 1 μm, and typically less than 0.5 μm While it is convenient to use the GIS nozzle as the sputter material source because many charged particle beam systems already include a GIS, in-situ embodiments are not limited to using a GIS needle. For example, some charged particle beam systems include a micromanipulator for manipulating small items within the vacuum chamber. Micromanipulator systems typically include a probe for attaching to a sample or other item in the vacuum chamber. The probe, or a material attached to the probe could be used as the sputter material source. Alternative, the sputter material source could be on or part of any device introduced into the vacuum chamber that has the ability to be positioned to sputter onto the work piece and to be withdrawn to allow the charged particle beam to process the work process. FIG. 4 is a flow chart showing the steps for a preferred in-situ embodiment of the invention using a GIS needle as the sputter material source. In step 402, the work piece is loaded into the vacuum chamber. In step 404, the stage is moved to position the area of the work piece to be coated near, but slightly offset from, the impact point of the beam. In step 406, a sputter material source, such as a GIS nozzle, is moved into the ion beam path. Skilled persons will be able to determine empirically the optimum position of the area to be coated relative to GIS needle and the charged particle beam. In step 408, the ion beam is directed to the sputter material source and material is sputtered from the source onto the work piece surface. If required, the stage can be moved during the sputtering process to coat a larger area. The position of the coating on the work piece can be adjusted by adjusting the position of beam impact on the GIS nozzle, the position of the stage, or the position of the GIS nozzle. After the coating is applied to the work piece, the sputter material source is removed from the beam path in step 410. In step 412, the ion beam is re-focused onto the work piece if the ion beam is to be used to process the work piece, and in step 414, the work piece is processed by the ion beam or by the electron beam. Some embodiments can use multiple GIS needles. For example, one GIS could dispense gas for beam-assisted gas deposition and one could be used as a source of material for sputtering, or a single needle can be used for both functions sequentially or at the same time. Sputtered particles may be useful for decomposing a precursor to deposit material, in addition to the deposition of the sputtered material itself. The multiple needles could be made or coated with different materials so that sputtered layers of different materials could be applied without removing the work piece from the vacuum chamber. For example, one needle could be coated with one material and a second needle could be made from a different material, or a small target of material could be attached to the needle(s) to allow a selection of various sputter target materials with the use of a single or multiple GISs. Needles could be partially coated, or coated in layers, for example, with later-applied coatings leaving portions of the earlier coatings exposed, so that one could sputter from the underlying material or from one or more of the coatings to provide different materials for coating. Sequential processes can be used, in which multiple coatings are applied; some using sputtering deposition and some using beam-induced gas deposition. The orientation or position of the sputter material source can be varied for precise targeting of the sputtering onto the features of interest on the work piece. That is, the area coated can be controlled by “sputter steering,” for example, by changing the position or orientation of the sputter source, as well as the beam parameters, the distance from the sputter source to the work piece, and other parameters. Various embodiments can be used for relatively small, selected area coverage, or for large area coverage. Multiple materials are available for sputtering, including but not limited to, chromium, titanium, tungsten, copper, and aluminum. In one embodiment, chromium from a GIS needle was deposited on a sputtered area on the order of 10,000 square microns using an ion beam having a landing energy of 30 keV, with a beam current of 20 nA. Larger beam currents result in faster sputter deposition. Larger coverage areas can be achieved, for example, by moving the sample using an X-Y translation stage or by increasing the distance between the sputter source and the work piece. The beam current can be adjusted to speed up or slow down deposition and to control the deposition regions. Beam currents as low as 100 pA or 300 pA may be suitable for coating small areas. The time required to produce a protective coating will vary with the size of the area coated and the thickness of the coat. Sputtering for thirty minutes has been shown to produce a significant amount of deposition. Coating thicknesses from a 30 minute process were estimated to be greater than 150 nm, so coatings in the 10 s of nanometers are easily achievable in less time. It is expected that much shorter sputter times, on the order of five to ten minutes or less, will produce a coat of adequate thickness over a typical area of interest. The ion beam is typically rastered on the GIS needle or other sputter material source in a rectangular pattern. A typical raster size is 40 μm by 5 μm. A relatively large raster size is preferred to avoid removing too much material from the GIS needle in one position. A typically GIS needle has a tip diameter of about 1.5 mm. The beam focus is changed from the work piece to the needle for sputtering, and then the focus can be changed back to the work piece for further processing. The amount of material sputtered is typically relatively small compared to the thickness of the GIS needle, so in most embodiments, wear of the GIS needle is not a problem. The raster can be moved to different positions on the GIS to reduce wear. In most embodiments, the sample position and the GIS position are fixed, while the beam rasters on the GIS needle. The landing position of the primary beam on the sputter material source can be varied to adjust the position of the sputter coating and to coat a larger area. Of course, any or all of the sample, the GIS needle, or the beam can be moved to increase the size or adjust other parameters of the sputter coating on the work piece. The invention provides for producing a coating on the work piece in a charged particle beam system without requiring that the beam be directed toward the work piece. There may be incidental impact of the charged particle beam onto the work piece while the beam is directed to the sputter source, without departing from the scope of the invention. Any incidental impact may be out of focus and can cause less damage than directing a focused or shaped particle beam toward the work piece. The invention can be applied to any work piece, including integrated circuits, thin film read/write heads, microelectromechanical assemblies (MEMS) and other devices. Unlike a prior art plasma sputter system, some embodiments of the present invention use charged particle beam optics, such as an objective lens and optionally a condenser lens, to converge the beam before sputtering. Unlike prior art wafer sputtering systems that produce a coating over an entire wafer, some embodiments of the invention provide the capability for coating a relatively small portion of a wafer or other work piece, thereby allowing other die on a wafer to be unaffected. In one embodiment, a circle of about 200 μm diameter was coated. The size of the coated area varies with the size of the pattern milled by the ion beam on the sputter material source. Smaller areas, such as within the 20 μm to 50 μm meter range should be readily achievable. Besides the in-situ embodiments described above, some embodiments sputter without using a particle beam. For example, material may be sputtered using a diode, triode, or magnetron-type sputtering system. Such systems provide a protective layer onto a work piece surface, without directing a charged particle beam toward the surface, thereby reducing or eliminating damage to the surface. After the protective coating is applied, the work piece can be processed using the charged particle beam, for example, to form an image, to extract a sample, or to expose a cross section or buried layer. The sputtering mechanism can be located in a vacuum chamber different from the one in which the sample is processed using a charged particle beam, or the sputtering mechanism can be located within the same vacuum chamber, so that the work piece does not need to be transported between machines. Although the present invention and its advantages have been described in detail, it should be understood that various changes, substitutions and alterations can be made herein without departing from the spirit and scope of the invention as defined by the appended claims. Moreover, the scope of the present application is not intended to be limited to the particular embodiments of the process, machine, manufacture, composition of matter, means, methods and steps described in the specification. As one of ordinary skill in the art will readily appreciate from the disclosure of the present invention, processes, machines, manufacture, compositions of matter, means, methods, or steps, presently existing or later to be developed that perform substantially the same function or achieve substantially the same result as the corresponding embodiments described herein may be utilized according to the present invention. Accordingly, the appended claims are intended to include within their scope such processes, machines, manufacture, compositions of matter, means, methods, or steps. |
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abstract | A method for carrying out a reactor internal, comprising steps of: forming a first opening portion in a ceiling of a reactor building at a position directly above an equipment pool in said reactor building; cutting a cylindrical reactor internal surrounding a core in a reactor pressure vessel disposed in said reactor building, at one position in an axial direction; surrounding said cut cylindrical reactor internal with a radiation shield; and carrying out said cylindrical reactor internal surrounded by said radiation shield out of said reactor building through said first opening portion. |
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046559973 | abstract | The combination in a nuclear reactor of a concrete containment and thermal insulation for the roof of the containment, the thermal insulation being secured to the roof by hangers each of which has a linkage connector at its upper end by means of which it is secured to the roof, the linkage connector permitting movement of the hanger in two dimensions, and means at the lower end of each hanger serving to support the thermal insulation from beneath and permitting lateral movement of the insulation in two dimensions. |
051401657 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS A system for solidifying radioactive waste will be described with reference to FIG. 1. The radioactive waste is formed into powder by a vertical thin film dryer (not shown). Then, in a pelletizer 24, the powder is formed into pellets of a predetermined size by the addition of a binder. A solidifying vessel or container 30 is placed on a conveyor device 23, and is conveyed to a pellet charge device 22. In the pellet charge device 22, the vessel 30 is lifted, so that a pellet charge pipe 10 extending downward from the pelletizer 24 is introduced into the vessel 30. Then, a predetermined amount of the pellets are charged into the vessel 30 via the pellet charge pipe 10. Then, the vessel 30 is moved downward, so that the pellet charge pipe 10 is separated or removed from the vessel 30. Then, the vessel 30 is conveyed by the conveyor device 23 to a solidifying material-pouring device 26. In the solidifying material-pouring device 26, a solidifying material in the state of a liquid or a slurry is poured into the vessel 30 from a solidifying material tank 25. Thereafter, the vessel 30 is conveyed to a capping device 27 where a metallic top lid or cap 4 with a packing, as well as a clamping band 5, are attached to the vessel 30 (see FIG. 2). Then, the vessel 30 is transferred to a predetermined storage place. A first embodiment of a radioactive waste pellet solidifying vessel of the invention will now be described with reference to FIGS. 2 to 7. For example, a metal drum having a volume of 200 liters is used as the solidifying vessel. A liner 3 of concrete is formed on an inner surface of a body 1 of the vessel, and a concrete lid 2 constituting an inner lid is bonded to the upper end of the liner 3 by an epoxy resin adhesive or the like, the concrete lid 2 being disposed slightly below the upper end of the vessel body 1. The concrete lid 2 has a central opening 31, and a ring-shaped iron plate 6 is bonded by an epoxy resin adhesive or the like to the inner lid 2 in concentric relation to the central opening 31. A plurality of (16 in this embodiment) coil springs 7 are bonded at their one ends to the lower surface of the ring-shaped iron plate 6, and extend in such a manner that the axis of each of the coil springs 7 is displaced an angle of about 7.degree. from the center of the central opening 31. The coil spring 7 has an outer diameter of 10 mm, and is made of a wire having a diameter of 1.0 mm. The space or gap between any two adjacent coil springs 7 is smaller than the size of one radioactive waste pellet. As shown in FIG. 3, four coil springs 7 are longer than the other coil springs 7 in order to prevent the floating of the pellets at the central portion of the vessel. Air bleed holes 32 are formed through the inner lid 2 so as to prevent the air from being trapped in the vessel body 1 when the solidifying material is poured into the vessel as later described. In FIG. 2, reference numeral 4 denotes the metallic top lid (having a packing) for the vessel 30, and reference numeral 5 denotes the clamping band for securing the top lid 4 to the vessel 30. Next, a method of charging the radioactive waste pellets into the vessel of the above construction, as well as a method of pouring the solidifying material into the vessel, will now be described. In this example, the volume of one pellet is about 17 ml, and a paste of cement glass (mixture of cement and water-glass) which is about 0.3 gr/cm.sup.3 higher in specific gravity than the pellet is used as the solidifying material. First, in the pellet charge device 22, the vessel 30 is lifted in such a manner that the pellet charge pipe 10 can be inserted through the central opening 31 of the ring-shaped iron plate 6, as shown in FIG. 4. The outer periphery of the lower end of the charge pipe 10 has been worked to have a curvature radius of several millimeters. When the vessel 30 is further lifted, the coil springs 7 are bent or flexed downward by the lower end of the charge pipe 10, as shown in FIG. 5. When the charge pipe 10 is inserted a certain depth (20 to 50 mm), the coil springs 7 are moved upward by their own resiliency, so that the coil springs 7 are generally horizontally bent in contact with the outer periphery surface of the charge pipe 10, as shown in FIGS. 6 and 7 (in which only two coil springs 7 are shown). Then, a predetermined amount of the pellets are charged into the vessel 30 through the charge pipe 10 (in this case, it is desirable to provide a charge amount-monitoring sensor in the vicinity of the lower end of the charge pipe 10). Then, the vessel 30 is moved downward so as to remove the charge pipe 10 from the central opening 31, so that the coil springs 7 are returned to their respective initial positions shown in FIGS. 3 and 4. Then, in the solidifying material-pouring device 26, a pouring pipe 33 is disposed just above the opening 31 of the ring-shaped iron plate 6, and the solidifying material slurry is poured from the pouring pipe 33 into the vessel 30 through the opening 31 of the iron plate 6 and the gaps between the coil springs 7. The floating of the pellets caused by the poured solidifying material slurry is restrained by the lower surface of the concrete lid 2, and those pellets tending to float upward at the central opening 31 are prevented by the coil springs 7 from floating. Namely, the strength of the coil springs 7 is so determined as to prevent such floating. In this condition in which the floating of the pellets is prevented, the solidifying material is further poured to such a level that the concrete lid 2, the iron plate 6 and the coil springs 7 can be embedded in the solidifying material. Then, the top lid 4 is attached to the vessel body 1, and is secured thereto by the clamping band 5. It has been confirmed through experiments that the charging of the pellets and the pouring of the solidifying material have been carried out satisfactorily, while preventing the floating of the pellets. In the above embodiment, although the coil springs 7 are fixedly mounted on the concrete lid 2 through the ring-shaped iron plate 6, the coil springs 7 may be fixedly mounted directly on the concrete lid 2. If the concrete lid 2 has a weight large enough to overcome the buoyancy, it may not be fixedly secured to the upper end of the concrete liner 3, but may be merely placed on this upper end. Next, a second embodiment of a solidifying vessel of the invention will now be described with reference to FIGS. 8 and 9. A ring-shaped iron plate lid 9 is bonded by welding or an epoxy resin to that portion of an inner peripheral surface of a vessel body 1 disposed slightly below the upper end of the vessel body 1. Two pairs of diametrically-opposite tabs 9' are formed on the outer peripheral edge of the iron plate lid 9, and the tabs 9' are fixedly secured to the inner peripheral surface of the vessel body 1. With this arrangement, arcuate gaps 35 are formed between the inner peripheral surface of the vessel body 1 and the outer periphery of the iron plate lid 9. The gaps 35 perform the same function as that of the air bleed holes 32 in the first embodiment. A plurality of piano wires 8 are bonded at their one ends to the lower surface of the iron plate lid 9 by welding or an epoxy resin, and extend in such a manner that the axis of each of the piano wires 8 is displaced an angle of about 5.degree. from the center of a central opening 31 of the iron plate lid 9. Each of the piano wires 8 has a diameter of about 1.5 mm. In order to prevent the floating of the pellets at the central portion of the vessel body 1, four piano wires 8 are longer than the other piano wires 8, as shown in FIG. 9. The piano wires 8 perform the same function as that of the coil springs 7 in the first embodiment. The pellets, each having a volume of about 9 ml to about 17 ml and being about 0.4 mg/cm.sup.3 lower in specific gravity than the solidifying material slurry, were charged into the vessel 30 of this embodiment, and the solidifying material slurry was poured into the vessel 30 of this embodiment. As a result, the charging of the pellets and the pouring of the solidifying material were satisfactorily carried out while preventing the floating of the pellets, as in the first embodiment. In the above first and second embodiments. In order that the coil springs 7 or the piano wires 8 can have a sufficient strength to prevent the floating of the pellets, the diameter of the coil spring 7 and the diameter of the piano wire 8, as well as the material for the coil spring 7 or the piano wire 8, are determined depending on the buoyance of the pellets. In the above embodiments, although the coil springs 7 or the piano wires 8 are so extended that their axes are displaced from the center of the central opening 31 into which the charge pipe 10 is adapted to be inserted, the coil springs 7 or the piano wires 8 may be directed toward the center of the central opening 31. In this case, however, the coil springs 7 or the piano wires 8 can not easily be moved by their own resiliency from the condition shown in FIG. 5 to the condition shown in FIGS. 6 and 7, and therefore, it is necessary to slightly reduce the amount of charge of the pellets into the vessel 30 in order to ensure that after the charge pipe 10 is removed from the central opening 31, the coil springs 7 or the piano wires 8 can be restored to the condition shown in FIG. 4 without being prevented by the pellets in the vessel 30. Therefore, it is advantageous that the coil springs 7 or the piano wires 8 should be extended in such a manner as to be slightly displaced from the center of the opening 31, as described in the first and second embodiments, because this can increase the amount of charge of the pellets into the vessel. In the first embodiment, the provision of the concrete liner 3 is not of absolute necessity, and instead of the concrete lid 2, an iron plate lid having a central opening may be used as an inner lid, and be welded or bonded to the inner peripheral surface of the vessel body 1. Also, the above inner lid, such as the concrete lid 2 and the iron plate lid 9, may be omitted, in which case the proximal ends of the coil springs or the piano wires (whose distal end portions are extended to the central portion of the vessel) are fixedly secured directly to the inner peripheral surface of the vessel body. However, the above embodiments in which the inner lid is provided, and the charge pipe is inserted in the central opening of the inner lid so as to charge the pellets into the vessel, are more advantageous in the prevention of the scattering of the pellet dust at the time of the pellet charging, the prevention of the jumping of the pellets out of the vessel, and the prevention of the floating of the pellets. The coil springs or the piano wires may be replaced by leaf springs, rubber members, plastics members or the like which have such flexible (bendable) and restoring properties that they can be bent or flexed upon insertion of the charge pipe, and can be restored upon withdrawal of the charge pipe so as to prevent the floating of the pellets thereafter. At the time of the charging of the pellets, the coil spring 7 or the piano wire 8 is mainly bent by the charge pipe 10 at that portion thereof disposed in the vicinity of its proximal end. Therefore, as indicated in a third embodiment of the invention shown in FIG. 10, the coil spring 7 or the piano wire 8 may be replaced by a float prevention member which comprises a resilient element 12 (which corresponds to said that portion of the coil spring 7 or the piano wire 8), such as a coil spring or a piano wire, and a bar 11 of relatively high rigidity extending from the distal end of the resilient element 12. In a fourth embodiment of the invention shown in FIG. 11, the coil spring 7 or the piano wire 8 is replaced by a float prevention member which comprises a bar 11 extending into the central opening 31 of the ring-shaped iron plate 6 or the ring-shaped iron plate lid 9, and a resilient element 12 (e.g. a coil spring, a piano wire, or a leaf spring) interconnecting the bar 11 and the iron plate 6 or the iron plate lid 9. The bar 11 is urged upward by the resilient element 12 to a horizontal position where a proximal end 13 of the bar 11 is abutted against the inner peripheral surface or edge of the central opening 31 of the iron plate 6 or the iron plate lid 9 to limit a further upward movement of the bar 11. In a fifth embodiment of the invention shown in FIG. 12, the coil spring 7 and the piano wire 8 is replaced by a float prevention member which comprises a base member 14 fixedly secured to the lower surface of the iron plate 6 or the iron plate lid 9 adjacent to the central opening 31, a bar 15 hingedly connected at one end to the base member 14, and a spring member 36 mounted on the hingedly-connected portion. The bar 15 is urged upward by the spring member 36 to a horizontal position where a proximal end 16 of the bar 15 is abutted against a stopper surface 17 of the base member 14 to limit a further upward movement of the bar 15. The float prevention member shown in FIG. 11 or FIG. 12 is used instead of the coil spring 7 of the first embodiment or the piano wire 8 of the second embodiment, and therefore it will be appreciated that a plurality of such float prevention members are provided at the central opening 31 of the iron plate 6 or the iron plate lid 9, as in the first and second embodiments. In the embodiments shown in FIGS. 11 and 12, the bars 11 or 15 are returned to the horizontal positions when the charge pipe 10 is removed from the central opening 31 after the charging of the pellets into the vessel, and thereafter even if the bars 11 or 15 receive the floating force of the pellets from below these bars at the time of pouring the solidifying material slurry, the above-mentioned stopper mechanism serves to hold the bars 11 or 15 in the horizontal positions, thereby positively preventing the floating of the pellets. FIG. 13 is a top plan view of a fifth embodiment of the invention, showing an inner lid and its associated parts. Two base members 20 and 20 are fixedly secured to the lower surface of the iron plate 6 or the iron plate lid 9 by welding or an adhesive. Two metal nets 19 and 19 of a rectangular shape are hingedly connected to the two base members 20 and 20, respectively. Each of the metal nets 19 and 19 is urged upward by spring means (which is not shown and is provided at the hinged portion) so as to be angularly moved about an axis 21 of the hinge into a horizontal position where the metal net 19 is abutted against the lower surface of the iron plate 6 or the iron plate lid 9. Therefore, the length of the metal net 19 is slightly greater than the diameter of a central opening 31 of the iron plate 6 or the iron plate lid 9. A gap g between the two metal nets 19 and 19 disposed at their respective horizontal positions, and the mesh of each metal net 19 are smaller than the size of one pellet. When the charge pipe 10 is inserted into the central opening 31 at the time of charging the pellets in a similar manner as described above, the two metal nets 19 and 19 are opened or angularly moved downward away from each other about the respective hinge axes 21 and 21, and therefore the pellets can be charged into the vessel from the charge pipe 10 through this open space between the two metal nets 19 and 19. Then, when the charge pipe 10 is removed from the central opening 31 after the charging of the pellets, the two metal nets 19 and 19 are returned to their respective initial positions. Thereafter, the solidifying material slurry is poured into the vessel through the metal nets 19 and 19 from above these metal nets. The floating of the pellets at the central opening 31 is positively prevented by the two metal nets 19 and 19 stopped by the lower surface of the iron plate 6 or the iron plate lid 9. |
abstract | A device and method for curing photoactivatable paint coatings. An exemplary device may include a light chamber housing supported by a frame and undercarriage, the wall portions of the light chamber having a peripheral region terminating at a light emission region. A UV light source may be located within the light chamber. A motorized carrier may be provided and configured to controllably index and/or oscillate the UV light source along a travel path within the housing. The light chamber may be located adjacent a target paint cure location on a work piece, with the UV light emission region facing the paint cure location. Once properly located, the UV light source may be indexed and/or oscillated along the travel path to deliver UV light to the target paint cure location so as to cure UV curable paint thereon. |
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claims | 1. A treatment planning unit for creating a treatment plan to perform the treatment with respect to an affected part, the treatment planning unit comprising:an input unit;a computing unit that sets a plurality of layer regions that are different in irradiation energy from each other in the irradiation direction of a charged particle beam, in a target including an affected part set using tomography image information including information on the affected part, that sets a plurality of irradiation positions of the charged particle beam in each of these layer regions, that using an irradiation dose and a reference irradiation dose with respect to each irradiation position set in each layer region, sets the number of times of irradiations of the charged particle beam with respect to each of the irradiation regions set in each of the layer regions, and that sets a plurality of number of times of irradiations with respect to at least one said irradiation position in the layer region to create a treatment plan information with respect to the affected part; anda display unit for displaying the treatment planning information. 2. The treatment planning unit according to claim 1, wherein the computing means sets the number of times of irradiations with respect to one said irradiation position so that the irradiation dose at said irradiation position for each irradiation is no more than the reference irradiation dose. |
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061545147 | abstract | An upper hold-down spring structure for a nuclear reactor fuel assembly. A hold down spring 20 mounted on an upper surface of an upper nozzle 11 of a fuel assembly for a pressurized water reactor is composed of an upper plate spring 21 having plastic spring characteristics and a lower plate spring 23, base ends of which are fixed with a fastening bolt 18 at a common position. The upper spring 21 and the lower spring 23 are made of precipitation hardened nickel base alloy and the thickness of the springs are determined so as to keep the stresses generated less sensitive to stress corrosion cracking. |
summary | ||
051475995 | summary | BACKGROUND OF THE INVENTION The present invention relates to a fuel assembly for a nuclear reactor with a detachable top tie plate to make possible reconditioning of the fuel assembly. A fuel assembly of a known design comprises a plurality of fuel rods and guide tubes for control rods or arranged as water tubes only in certain, special nuclear reactors, the fuel rods and guide tubes being kept in predetermined mutual positions by means of spacer lattices placed along the fuel assembly. The top tie plate and the bottom tie plate at the ends of the fuel assembly are fixed to the guide tubes of the control rods, which guide tubes extend somewhat above and below the ends of the fuel rods. In this known fuel assembly the guide tubes are provided on top with a top sleeve which fits into a corresponding hole through the top tie plate. A peripheral slot is arranged in the hole, a bead around the upper part of the top sleeve fitting into this slot. The top sleeve is provided with an axial slit, extending from its upper end through the bead and made of a resilient material The slit permits the upper end of the top sleeve to be squeezed together so that the top sleeve can be moved in the hole of the top tie plate to a position in which the bead falls into the slot, whereupon the end is allowed to rebound. To ensure the locking between the top sleeve and the top tie plate, a locking sleeve is inserted into the top sleeve, which locking sleeve prevents the bead from leaving the slot. A drawback of this design is that a rather extensive machining of the top tie plate must be performed, such as, for example, milling of slots in the holes. Further, when mounting the top tie plate, all top sleeves must be squeezed together more or less simultaneously in order for the top tie plate to be fitted onto the top sleeve. Accordingly, this mounting may become quite problematic. SUMMARY OF THE INVENTION The present invention avoids the above-mentioned drawbacks. As usual, the fuel assemby comprises a number of parallel fuel rods which are retained by means of spacer lattices, arranged along the fuel rods, as well as guide tubes for control rods or arranged only as water tubes in certain, special nuclear reactors. The guide tubes are arranged between the fuel rods and fixed to the spacer lattices. The guide tubes and the associated fuel rods have been fixed between a top tie plate and a bottom tie plate. The top tie plate has been provided with a through-hole for connection of a top sleeve joined to the upper end of the guide tubes. According to the invention, there is arranged inside the top sleeve a first locking element intended to cooperate with a second locking element arranged on a guide sleeve, which is insertable through a hole in the top tie plate, which hole corresponds to the top sleeve. By the invention, extra machining of the top tie plate is avoided since now the second locking element is arranged in a separate guide sleeve. Further, the fixing of the top sleeves of the guide tubes may take place individually as the guide sleeves are inserted, one by one, through the holes in the top tie plate. An additional improvement of the invention is obtained if seats for the top sleeves are arranged around the holes, into which seats the top sleeves can be easily guided and fixed, one by one, to the top tie plate by means of guide sleeves. |
claims | 1. A crystal monochromator comprising:a Ge crystal body, wherein at least a first region of the Ge crystal body comprises a mosaic structure having a mosaicity of between about 15 arcminutes to about 28 arcminutes and a slow neutron reflectivity of about 70%-89% at a rocking curve peak. 2. The crystal monochromator of claim 1, wherein the slow neutron reflectivity is about 75%-89% at the rocking curve peak. 3. The crystal monochromator of claim 1, wherein the Ge crystal body has a thickness of about 7-10 mm cut from a plastically deformed ingot. 4. The crystal monochromator of claim 1, wherein the mosaic structure is approximately uniform over the Ge crystal body. 5. The crystal monochromator of claim 1, wherein the Ge crystal body comprises:the first region having the mosaicity of between about 15-28 arcminutes; anda second region having a mosaicity of less than 15 arcminutes. 6. The crystal monochromator of claim 5, wherein the Ge crystal body comprises a continuous gradient of mosaicity between the first region and the second region along at least one axis. 7. The crystal monochromator of claim 5, wherein the second region has an approximately perfect crystal structure with a mosaicity of less than approximately 1 arcminute. 8. The crystal monochromator of claim 1, wherein the Ge crystal body comprises a plastically deformed Ge crystal with a planar orientation of (1,1,1). 9. A method of manufacturing a crystal monochromator, comprising:heating an approximately perfect Ge crystal having an original thickness of approximately 3-5 cm to a temperature of over about 850° C.;compressing the Ge crystal for a duration of approximately 1-5 minutes with a force of about 5-10 metric tons while the Ge crystal is maintained at the temperature of over about 850° C. to plastically deform the Ge crystal along an axis of the Ge crystal, wherein the compressing causes a plastic deformation of about 0.5%-1.5% of the original thickness; andslicing the Ge crystal to form a plurality of crystal monochromators, wherein at least a first region of each of the plurality of crystal monochromators has a first mosaicity value of between about 15-28 arcminutes and a slow neutron reflectivity of about 70%-89% at a peak rocking curve. 10. The method of claim 9, further comprising polishing a top and a bottom of the Ge crystal prior to the heating and compressing to cause the Ge crystal to have an approximately uniform thickness of 3-5 cm+/−5 μm. 11. The method of claim 9, wherein the Ge crystal comprises an approximately perfect Ge crystal with a planar orientation of (1,1,1). 12. The method of claim 9, further comprising:trimming at least about 7 mm off of a top of the Ge crystal after the compressing and before the slicing; andtrimming at least about 7 mm off of a bottom of the Ge crystal after the compressing and before the slicing;wherein slicing the Ge crystal comprises slicing a remainder of the Ge crystal perpendicular to the axis of the Ge crystal to form the plurality of crystal monochromators. 13. The method of claim 12, wherein each of the plurality of crystal monochromators has a thickness of about 7-9 mm. 14. The method of claim 9, wherein heating the Ge crystal to the temperature of over about 850° C. comprises:heating the Ge crystal to a temperature of about 855° C.-870° C.; andmaintaining the Ge crystal at the temperature of about 855° C.-870° C. for a period of at least 1 hour prior to compressing the Ge crystal. 15. The method of claim 9, wherein compressing the Ge crystal with the force of about 5-10 metric tons comprises:applying a first force to the first region of the Ge crystal to cause the first region to have the first mosaicity of between about 15-28 arcminutes; andapplying a second force to a second region of the Ge crystal to cause the second region to have a second mosaicity that is lower than the first mosaicity. 16. The method of claim 15, further comprising:shaping a top surface of the Ge crystal to cause the top surface to have a surface profile wherein the first region is thicker than the second region; andshaping a bottom surface of the Ge crystal to cause the bottom surface to have the surface profile. 17. The method of claim 15, further comprising:shaping a surface of a first metal die used to contact a top of the Ge crystal during compression to cause the surface of the first metal die to have a surface profile wherein a first region of the metal die is thicker than a second region of the metal die; andshaping a surface of a second metal die used to contact a bottom of the Ge crystal during the compression to cause the surface of the second metal die to have the surface profile. 18. A crystal monochromator manufactured by a process comprising:heating an approximately perfect Ge crystal having an original thickness of approximately 3-5 cm to a temperature of over about 850° C.;compressing the Ge crystal for a duration of approximately 1-5 minutes with a force of about 5-10 metric tons while the Ge crystal is maintained at the temperature of over about 850° C. to plastically deform the Ge crystal along an axis of the Ge crystal, wherein the compressing causes a plastic deformation of about 0.5%-1.5% of the original thickness; andslicing the Ge crystal to form a plurality of crystal monochromators, wherein at least a first region of each of the plurality of crystal monochromators has a first mosaicity value of between about 15-28 arcminutes and a slow neutron reflectivity of about 70%-89% at a rocking curve peak. 19. The crystal monochromator of claim 18, wherein the process further comprises:trimming at least about 7 mm off of a top of the Ge crystal after the compressing and before the slicing; andtrimming at least about 7 mm off of a bottom of the Ge crystal after the compressing and before the slicing;wherein slicing the Ge crystal comprises slicing a remainder of the Ge crystal perpendicular to the axis of the Ge crystal to form the plurality of crystal monochromators each having thicknesses of about 7-9 mm. 20. A crystal monochromator, comprising:a Ge crystal body comprising a first region with a first mosaicity value; andthe Ge crystal body comprising a second region with a second mosaicity value that is higher than the first mosaicity value and a slow neutron reflectivity of about 70%-89% at a rocking curve peak. 21. The crystal monochromator of claim 20, wherein the first region has an approximately perfect crystal structure with a mosaicity of less than about 1 arcminute and the second region has a mosaicity of less than about 40 arcminutes. 22. The crystal monochromator of claim 20, wherein the Ge crystal body comprises a continuous gradient of mosaicity between the first region and the second region along at least one axis. 23. The crystal monochromator of claim 22, wherein the first region is an inner circular region of the monochromator, wherein the second region is an outer circular region of the monochromator that is concentric with the inner circular region, and wherein the continuous gradient is a radial gradient between the inner circular region and the outer circular region. |
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description | Field Example embodiments relate to an apparatus for inspecting welds of a nuclear reactor and methods of using the same. Description of Related Art FIG. 1A illustrates a general arrangement of a core shroud 2 inside a reactor pressure vessel (RPV) 4. Feedwater is admitted into the RPV 4 via a feedwater inlet (not shown) and a feedwater sparger 6, which is a ring-shaped pipe having suitable apertures for circumferentially distributing the feedwater inside the RPV. The feedwater from the sparger 6 flows downwardly through the downcomer annulus 8, which is an annular region between the core shroud 2 and the RPV 4. The core shroud 2 is a stainless steel cylinder surrounding the nuclear fuel core, the location of which is generally designated by numeral 10 in FIG. 1A. The core is made up of a plurality of fuel bundle assemblies. Each array of fuel bundle assemblies is supported at the top by a top guide and at the bottom by a core plate (neither of which are shown). The core top guide provides lateral support for the top of the fuel assemblies and maintains the correct fuel channel spacing to permit control rod insertion. The water flows through downcomer annulus 8, around the bottom edge of the shroud and into the core lower plenum 12. The water subsequently enters the fuel assemblies, wherein a boiling boundary layer is established. A mixture of water and steam enters core upper plenum 14 under the shroud head 16. The steam-water mixture then flows through vertical standpipes (not shown) atop the shroud head and enters steam separators (not shown), which separated liquid water from steam. The liquid water then mixes with feedwater in the mixing plenum, which mixture then returns to the core via the downcomer annulus. The steam is withdrawn from the RPV via a steam outlet. The boiling water reactor (BWR) also includes a coolant recirculation system which provides a forced convection flow through the core necessary to attain the required power density. A portion of the water is sucked from the lower end of the downcomer annulus 8 via recirculation water outlet (not visible in FIG. 1A) and forced by a centrifugal recirculation pump (not shown) into jet pump assemblies 18 (two of which are shown in FIG. 1A)) via recirculation water inlets 20. The BWR has two recirculation pumps, each of which provides the driving flow for a plurality of jet pump assemblies. The jet pump assemblies are circumferentially distributed around the core shroud 2. As shown in FIG. 1B, the core shroud 2 includes a shroud head flange 2a for supporting the shroud head 16, a circular cylindrical upper shroud wall 2b having a top end welded to shroud head flange 2a, an annular top guide support ring 2c welded to the bottom end of upper shroud wall 2b, a circular cylindrical middle shroud wall having a top end welded to top guide support ring 2c and consisting of upper and lower shell sections 2d and 2e joined by mid-shroud attachment weld W, and an annular core plate support ring 2f welded to the bottom end of the middle shroud wall and to the top end of a lower shroud wall 2g. The entire shroud is supported by a shroud support 22, which is welded to the bottom of lower shroud wall 2g, and by annular jet pump support plate 24, which is welded at its inner diameter to shroud support 22 and at its outer diameter to RPV 4. The material of the shroud and associated welds is austenitic stainless steel having reduced carbon content. The heat-affected zones of the shroud girth welds, including the mid-shroud attachment weld, have residual weld stresses. Therefore, the mechanisms are present for mid-shroud attachment weld W and other girth welds to be susceptible to intergranular stress corrosion cracking (IGSCC). Stress corrosion cracking in the heat affected zone of any shroud girth seam weld diminishes the structural integrity of the shroud, which vertically and horizontally supports the core top guide and the shroud head. In particular, a cracked shroud increases the risks posed by a loss-of-coolant accident (LOCA) or seismic loads. During a LOCA, the loss of coolant from the reactor pressure vessel produces a loss of pressure above the shroud head and an increase in pressure inside the shroud, i.e., underneath the shroud head. The result is an increased lifting force on the shroud head and on the upper portions of the shroud to which the shroud head is bolted. If the core shroud has fully cracked girth welds, the lifting forces produced during a LOCA could cause the shroud to separate along the areas of cracking, producing undesirable leaking of reactor coolant. Also, if the shroud weld zones fail due to stress corrosion cracking, there is a risk of misalignment from seismic loads and damage to the core and the control rod components, which would adversely affect control rod insertion and safe shutdown. Thus, the core shroud needs to be examined periodically to determine its structural integrity and the need for repair. Ultrasonic inspection is a known technique for detecting cracks in nuclear reactor components. The inspection area of primary interest is the outside surface of the cylindrical core shroud at the horizontal mid-shroud attachment welds. However, the core shroud is difficult to access. Installation access is limited to the annular space between the outside of the shroud and the inside of the reactor pressure vessel, between adjacent jet pumps. Scanning operation access is additionally restricted within the narrow space between the shroud and jet pumps, which is about 0.5 inch wide in some locations. The inspection areas are highly radioactive, and may be located under water 50 to 65 feet below the operator's work platform. Thus, inspection of the core shroud in operational nuclear reactors requires a robotic scanning device which can be installed remotely and operated within a narrowly restricted space. However, robotic scanning devices (e.g., remotely operative vehicle (ROV)) scanners use rollers to travel around outer diameter of the shroud, which has difficulties in staying level during horizontal weld scanning and/or staying on the weld to be scanned. In addition, ROV scanners require large amount of buoyance chambers to remain neutrally buoyant and constant in the horizontal level. In other related art, ROV scanners may use tether to pull the tool upward. However, this creates problems in that an operator must move the entire tool to advance the scanning probe. In addition, these type of ROV scanners are very large and heavy, not flexible and maneuverable, and more difficult and complicated to install and operate. Example embodiments disclose an apparatus for inspecting welds in a nuclear reactor. The apparatus may include a body, a rotatable pad on the body, a pair of opposing horizontal pads for moving the device in a vertical direction, a pair of opposing vertical pads for moving the device in a horizontal direction, and an inspection device. In a further example embodiment, the rotatable pad may be located at a central portion of the body to rotate the device. In yet a further example embodiment, the rotation of the apparatus may be in increments of 90 degrees. In a further example embodiment, a leg may be attached to each side of the body. In yet a further example embodiment, the leg may be attached to the body with a stem member. In yet a further example embodiment, the leg may be substantially Y-shaped or U-shaped. In a further example embodiment, the apparatus may include a forked arm. In yet a further example embodiment, each pad may be mounted on the forked arm. In yet a further example embodiment, the forked arm may be substantially Y-shaped or U-shaped. In a further example embodiment, a support member may be attached to the forked arm. The support member may include a hole for inserting a shaft and moving the pads in a respective horizontal or vertical direction. In a further example embodiment, the pair of opposing horizontal and vertical pads may slide in a respective axial direction of the shaft. In yet a further example embodiment, the shaft may be a ball screw that translates rotational motion to linear motion. In a further example embodiment, the inspection device may be mounted on the leg. In yet a further example embodiment, the inspection device may be an ultrasonic probe. In yet a further example embodiment, the ultrasonic probe may be attached to a gimbal sensor. In a further example embodiment, the inspection device may be supported by support arms. In a further example embodiment, a vacuum system may controllably adhere or force the pads to the surface of a core shroud. Example embodiments disclose a method for inspecting welds in a core shroud of a nuclear reactor. The method may include moving an inspection device to along a wall of a core shroud, attaching a pair of opposing horizontal pads to move in a vertical direction, releasing the pair of opposing horizontal pads, attaching a pair of opposing vertical pads to move in a horizontal direction, releasing the pair of opposing vertical pads, attaching a central pad to a surface of the core shroud, and rotating the central pad. In a further example embodiment, the method may include attaching an inspection tool to the device. In a further example embodiment, the inspection tool may be moved to align the inspection tool toward the weld to be measured and inspected. Detailed descriptions of the illustrative embodiments are disclosed herein. Specific structural and functional details are also disclosed herein, however, are merely representative for purposes of describing the example embodiments. For example, although the example embodiments may be described with reference to a nuclear power plant, it is understood that the example embodiments may also be useable in other types of industrial facilities. These facilities may have a need for the elimination of hard-wired, point-to-point connections for field equipment but need a robust digital system with safety functions. The example embodiments may be exemplified in many alternate forms and should not be construed as being limited to the specified example embodiments set forth herein. It should be understood that when an element or layer is referred to as being “on,” “connected to,” “coupled to,” or “covering” another element or layer, it may be directly on, connected to, coupled to, or covering the other element or layer or intervening elements or layers may be present. In contrast, when an element is referred to as being “directly on,” “directly connected to,” or “directly coupled to” another element or layer, there are no intervening elements or layers present. Like numbers refer to like elements throughout the specification. As used herein, the term “and/or” includes any and all combinations of one or more of the associated listed items. It should be understood that, although the terms first, second, third, etc. may be used herein to describe various elements, components, regions, layers and/or sections, these elements, components, regions, layers, and/or sections should not be limited by these terms. These terms are only used to distinguish one element, component, region, layer, or section from another region, layer, or section. Thus, a first element, component, region, layer, or section discussed below could be termed a second element, component, region, layer, or section without departing from the teachings of example embodiments. Spatially relative terms (e.g., “beneath,” “below,” “lower,” “above,” “upper,” and the like) may be used herein for ease of description to describe one element or feature's relationship to another element(s) or feature(s) as illustrated in the figures. It should be understood that the spatially relative terms are intended to encompass different orientations of the device in use or operation in addition to the orientation depicted in the figures. For example, if the device in the figures is turned over, elements described as “below” or “beneath” other elements or features would then be oriented “above” the other elements or features. Thus, the term “below” may encompass both an orientation of above and below. The device may be otherwise oriented (rotated 90 degrees or at other orientations) and the spatially relative descriptors used herein interpreted accordingly. The terminology used herein is for the purpose of describing various embodiments only and is not intended to be limiting of example embodiments. As used herein, the singular forms “a,” “an,” and “the” are intended to include the plural forms as well, unless the context clearly indicates otherwise. It will be further understood that the ter iris “includes,” “including,” “comprises,” and/or “comprising,” when used in this specification, specify the presence of stated features, integers, steps, operations, elements, and/or components, but do not preclude the presence or addition of one or more other features, integers, steps, operations, elements, components, and/or groups thereof. Example embodiments are described herein with reference to cross-sectional illustrations that are schematic illustrations of idealized embodiments (and intermediate structures) of example embodiments. As such, variations from the shapes of the illustrations as a result, for example, of manufacturing techniques and/or tolerances, are to be expected. Thus, example embodiments should not be construed as limited to the shapes of regions illustrated herein but are to include deviations in shapes that result, for example, from manufacturing. For example, an implanted region illustrated as a rectangle will, typically, have rounded or curved features and/or a gradient of implant concentration at its edges rather than a binary change from implanted to non-implanted region. Thus, the regions illustrated in the figures are schematic in nature and their shapes are not intended to illustrate the actual shape of a region of a device and are not intended to limit the scope of example embodiments. Unless otherwise defined, all terms (including technical and scientific terms) used herein have the same meaning as commonly understood by one of ordinary skill in the art to which example embodiments belong. It will be further understood that terms, including those defined in commonly used dictionaries, should be interpreted as having a meaning that is consistent with their meaning in the context of the relevant art and will not be interpreted in an idealized or overly formal sense unless expressly so defined herein. At least one example embodiment discloses an apparatus for inspecting welds in a nuclear reactor. The apparatus may include a body, a rotatable pad on the body, a pair of opposing horizontal pads for moving the device in a vertical direction, a pair of opposing vertical pads for moving the device in a horizontal direction, and an inspection device. At least one example embodiment discloses a method for inspecting welds in a core shroud of a nuclear reactor. The method may include attaching a center pad to a wall of the core shroud, rotating a device, moving the device in a horizontal direction via a pair of opposing vertical pads to inspect a horizontal weld, moving the device in a vertical direction via a pair of opposing horizontal pads to inspect a vertical weld, and attaching the pair of opposing horizontal and vertical pads to the wall of the core shroud. The exemplary embodiments are described herein in the context of a core shroud of a boiling water reactor (BWR). However, it will be apparent to one skilled in the art that the exemplary embodiments are applicable in other contexts including to other underwater structures. The example embodiments described herein are configured with respect to an outer surface or outer diameter of a structure although the teachings herein can be applied to an inner surface or inner diameter of a structure. FIG. 2 is a perspective view of an apparatus for inspecting welds for nuclear reactor, according to an example embodiment. More specifically, an inspecting apparatus 10 may be configured to conform to a profile of an outer surface of a core shroud 2 (as shown in FIG. 1A), attach to the outer surface of the core shroud 2, navigate along the outer surface of the core shroud 2, and inspect welds of the core shroud 2. Referring to FIG. 2, the inspecting apparatus 10 may include a body 100, a rotatable suction pad 105, a pair of opposing horizontal suction pads 110a, 110b, a pair of opposing vertical suction pads 120a, 120b, and an inspection tool 200. The body 100 may be substantially rectangular in shape. The body 100 may be approximately one inch in thickness. It should be appreciated that other dimensions of the body may be utilized while maintaining the function of the apparatus. The suction pad 105 may be located at a center of the body 100 to rotate the inspection apparatus 10. The rotation may be in increments of 90 degrees having a full range of 360 degrees. The suction pad 105 may be rotated using a worm and a worm gear (not shown) built inside the body 100 of the inspection apparatus 10 to rotate the suction pad 105. It should be appreciated that other types of gear(s) may be utilized for rotating the suction pad 105. A leg 130 may be attached to each side of the body 100. In this example embodiment, there are four legs 130. The leg 130 may include a stem 135 attached at a center of each side of the body 100. It should be appreciated that the stem 135 and the leg 130 may be built as one-piece or as two separate pieces. The leg 130 may be substantially Y-shaped or U-shaped. It should be appreciated that other shapes may be employed without departing from the function of the leg. Each suction pads 110a, 110b, 120a, 120b may be mounted on a forked arm 140. The forked arm 140 may be substantially Y-shaped or U-shaped. It should be appreciated that other shapes may be implemented as long as the arm member supports the suction pads 110a, 110b, 120a, 120b. The forked arm 140 may be attached to a support member 145. It should be appreciated that the forked arm 140 and the support member 145 may be a single unit or two separate units. The support member 145 may include a hole 147 for inserting a shaft 155 and moving the suction pads 110a, 110b, 120a, 120b in a respective horizontal or vertical direction. In an example embodiment, the shaft 155 may be a ball screw. One skilled in the art would appreciate that a ball screw may be a mechanical linear actuator that translates rotational motion to linear motion with little friction. Furthermore, it should be appreciated that the shaft (ball screw) 155 may be threaded so that a helical raceway for ball bearings may act as a precision screw. The ball assembly may also act as a nut while the threaded shaft is the screw. In addition, the shaft 155 (ball screw) may be able to apply or withstand high thrust loads with minimum internal friction. Moreover, the shaft (ball screw) 155 may be made to close tolerances so that high precision is achieved. As shown in FIG. 2, the shaft (ball screw) 155 may be supported at each end of the leg 130. The movement of the shaft (ball screw) 155 may be controlled by a motor 160 connected via a belt 162. It should be appreciated that other configurations may be utilized to move the shaft (ball screw) 155. The forked akin 140 may move the suction pads 110a, 110b, 120a, 120b in the respective horizontal or vertical direction. In an example embodiment, there are four suction pads, one on each side of the body 100. Of the four, the opposite pair suction pads 110a, 110b may be used to move the apparatus 10 in the vertical direction or along a longitudinal axis of the core shroud 2, and the opposite pair suction pads 120a, 120b may be used to move the apparatus 10 in the horizontal direction or around a circumference of the core shroud 2. During a horizontal travel mode of operation, the opposite pair suction pads 120a, 120b may contact the surface of the core shroud 2 and the opposite pair suction pads 110a, 110b may be retracted (or released) so as not to be in contact with the surface of the core shroud 2. During a vertical travel mode of operation, the opposite pair suction pads 110a, 110b may be extended to contact the surface of the core shroud 2, and may release the opposite pair suction pads 120a, 120b so that the opposite pair suction pads 120a, 120b will not be contact with the surface of the core shroud 20. The suction pads 110a, 110b, 120a, 120b may be extended by an actuation of a pneumatic piston 156 controlled by a control unit (not shown). The suction pads 110a, 110b, 120a, 120b may further be retracted (or released) via tension springs 167. It should be appreciated that other retraction device utilizing vacuum attachment force may be used. Referring to FIG. 5, the rotatable suction pad 105 may include elements of a vacuum system 125 that may be configured to controllably adhere or force to the surface of the core shroud 2. It should be appreciated that besides the suction pad 105, the same vacuum system may be employed in suction pads 110a, 110b, 120a, 120b as well. The vacuum system 125 may create a low pressure or vacuum space in a void 126. The void 126 may be defined and surrounded by a sealing ring 132. The sealing ring 132 may be configured to seal against the surface of the core shroud 2 to isolate the void 126 from the annulus so that the pressure in the void 126 can be reduced or otherwise controlled to adhere the inspecting apparatus 10 to the surface of the core shroud 2. The sealing ring 132 may be configured to move over obstacles on the surface of the core shroud 2. In an example embodiment, the sealing ring 132 may have a rounded profile and may be made of a flexible material to allow the inspecting apparatus 10 to travel over surface variations such as weld crowns. Referring to FIG. 6, the sealing ring 132 may include a closed-cell foam ring 230 or a skirt that may be wrapped with a cover 232. It should be appreciated that the cover 232 may be made from, for example, not limited to, a neoprene-coated nylon. A ring-shaped plate 234 may be bolted to the abdomen to hold the inner and outer edges of the cover 232 to the abdomen and hold the cover 232 over the foam ring 230. The foam ring 230 may be compressed and expanded to move over obstacles while still maintaining a vacuum seal. It should further be appreciated that the foam ring 230 may move over obstacles without folding and losing suction. The cover 232 may protect the foam ring 230 such that the sealing ring 132 is durable and robust. In other example embodiments, it is appreciated that multiple sealing rings 132 may be used, such that if one sealing ring loses suction while moving over an impediment or obstacle, the inspecting apparatus 10 remains attached to the surface. Referring back to FIG. 5, the vacuum system 125 may include a venturi valve 134, a pump 136 or other pressurized water supply, and a hose 138 that may connect the pump 136 to the venturi valve 134. The control unit (not shown) may be configured to control the pump 136. The pump 136 may draw water from the annulus and may supply the water to the venturi valve 134. The pump 136 may be located, for example, at the top of the core shroud 2 above the water level on the refueling floor. The pump 136 may be configured to recirculate water already in the inspecting apparatus 10. The venturi valve 134 may be configured to displace water from the void 126 as a function of water pressure supplied by the pump 136. A channel 140 may connect the void 126 to the flow path 142. The flow path 142 of the venturi valve 134 may be narrow and may then expand in the flow direction such that pressurized water supplied by the pump 136 to the venturi valve 134 creates low pressure at a low pressure location 144 in the venture valve 134. The channel 140 may be connected to the flow path 142 at the low pressure location 144. As such, pressurized water supplied by the pump 136 through the venturi valve 134 may draw water from the void 126 into the flow path 142 of the venturi valve 134 and may lower the pressure in the void 126. The venturi valve 134 may not require moving parts to lower the pressure in the void 126. Thus, the venturi valve 134 may reduce the risk of breakdown of the vacuum system 125 and may reduce the introduction of debris into the inspecting apparatus 10. The venturi valve 134 may allow for greater control of vacuum force as compared to using a pump to directly pump water from the void 126. The vacuum force provided by the venturi valve 134 may be controlled as a function of pressurized water or pump flow from pump 136. Referring back to FIG. 2, the inspecting apparatus 10 may further include a navigation system (not shown) that is configured to move the inspecting apparatus 10 along the surface of the core shroud 2 to globally position an ultrasonic probe, as described in further detail below. When moving, the inspecting apparatus 10 may maintain its orientation with respect to the core shroud 2 such that the profile of the inspecting apparatus 10 may match the profile of the surface of the core shroud 2. The inspecting apparatus 10 may be configured to maneuver both horizontally and vertically using motor driven wheels. Furthermore, the inspecting apparatus 10 may translate and may rotate. With reference to FIGS. 2 and 7, the inspecting apparatus 10 may include a weld-scanning system 170 that may be configured to inspect welds of the core shroud 2. The weld-scanning system 170 may include an ultrasonic probe 172 and a probe positioning system 174 that is configured to locally position the ultrasonic probe 172 to inspect various welds. The probe positioning system 174 may be configured to position the ultrasonic probe 172 to inspect welds of various orientations including horizontal welds, vertical welds, and welds at angles in between horizontal and vertical. The probe positioning system 174 may include a support arm 171 to support scan arms 176 and the forked arm 140 of the suction pad 120a. The forked arm 140 may be attached to the support member 145 for moving the suction pad 120a in the respective horizontal or vertical direction. At the outer edges of the ultrasonic probe 172, scan arms 176 may support the ultrasonic probe 172. The scan arms 176 may be substantially similar and symmetrically oriented with respect to each other. The scan arm 176 may include a rail 178 extending perpendicular to the scan alms 176, and a gimbal 180 that may be attached to the rail 178 with a linear bearing (not shown), for example. As an example embodiment, the gimbal 180 may provide a scan length of approximately two feet. The linear bearing may be selected based on reliability and minimal foreign material (FM) potential. A motion driving mechanism is configured to be actuated to move the gimbal 180 along the shaft 155 (shown in FIG. 2). For example, the motion driving mechanism may include the motor 160 that drives the ball screw 155. The motor 160 may be coupled to the ball screw 155 using the belt 162 (or gear train) so as to transmit torque to the ball screw 155. The ball screw 155 may be coupled to the support member 145. The motor 160 may be offset from the ball screw 155 so as to minimally restrict movement of the gimbal 180 along the rail 178. The ultrasonic probe 172 may be attached to the gimbal 180. An actuated motor 192 may be configured to move the gimbal 180. In order to achieve reliable, robust rotation, a worm gear is used. This allows the mechanism to maintain a low profile as well as achieve a high gear ratio so a small motor may be used. The gimbal 180 may also include a torsion spring (not shown) that may be configured to bias the ultrasonic probe 172 against the surface of the core shroud 2. A foreign material exclusion (FME) guard (not shown) may be used to cover the torsional spring. The ultrasonic probe 172 may be attached to the gimbal 180 so as to be able to rotate in a substantially flat-face contact with the surface of the core shroud 2. The ultrasonic probe 172 may be configured to be positioned along a length of an associated rail 178 and to be angularly positioned in a one-hundred and eighty degree range to inspect the welds. The ultrasonic probe 172 may be positioned to inspect horizontal welds and vertical welds along the sides of the inspecting apparatus 10 as well as above and below the inspecting apparatus 10. The control unit (not shown) may control the actuation of the motors 184, 192 to position the ultrasonic probe 172. In another example embodiment, a cover 195 may be used to enclose or encase the leg 130. It should be appreciated that the cover 195 may enclose all four legs 130 of the inspecting apparatus 10. The cover 195 may be employed to protect the components of the leg 130, which may include, among others, the support member 145, the shaft 155, the motor 160, and the belt 162. The cover 195 may be made from sheet metal, for example. FIG. 8 is a flowchart of illustrating a method of inspecting welds in a nuclear reactor, in accordance to an example embodiment. Each step may be performed according to the execution of a software module of computer-executable instructions by a control unit. In step S100, the inspecting apparatus 10 may move to a position on the core shroud 2 in a vicinity of the weld to be inspected. Once the inspecting apparatus 10 is in the vicinity of the weld to be inspected, the vacuum system may activate and attach the inspecting apparatus 10 to the core shroud. In step S200, for horizontal movement, the vacuum system may operate the opposite pair suction pads 120a, 120b to contact the surface of the core shroud 2 and the opposite pair suction pads 110a, 110b may be retracted (or released) so as not to be in contact with the surface of the core shroud 2. The opposite pair suction pads 110a, 110b may then move horizontally along the core shroud 2. In step S300, once the position of the pair of opposite pair suction pads 110a, 110b is determined, release the pair of opposite pair suction pads 120a, 120b to move vertically. In step S400, for vertical movement, the vacuum system may operate the opposite pair suction pads 110a, 110b to contact the surface of the core shroud 2. In step S500, the opposite pair suction pads 120a, 120b may be released so that the opposite pair suction pads 120a, 120b may not be in contact with the surface of the core shroud 2. The opposite pair suction pads 120a, 120b may then move vertically along the core shroud 2. Thereafter, a navigation system may move the inspecting apparatus 10 along the surface of the core shroud 2 to globally position the inspecting apparatus 10. The probe positioning system 174 may locally position and orient the ultrasonic probe 172 with respect to a crack in a weld or another feature to be measured by rotating the inspecting apparatus 10. In step S600, the central suction pad 105 attaches to the surface of the core shroud 2 to rotate the inspecting apparatus 10. In step S700, the inspecting apparatus 10 may be rotated to align the ultrasonic probe 172 toward the weld to be measured and inspected. It should be appreciated that the ultrasonic probe 172 may measure the length and width of the crack and the measurement may be recorded in a memory of the control unit (not shown). It should further be appreciated that selected steps may be repeated as necessary to make further measurements. While a number of example embodiments have been disclosed herein, it should be understood that other variations may be possible. Such variations are not to be regarded as a departure from the spirit and scope of the present disclosure, and all such modifications as would be obvious to one skilled in the art are intended to be included within the scope of the following claims. |
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description | This application is a continuation of U.S. patent application Ser. No. 13/577,163, filed Aug. 3, 2012, which is a national stage entry of International Application No. PCT/US2011/023952, filed Feb. 7, 2011, which claims the benefit of U.S. Provisional Patent Application No. 61/416,954, filed Nov. 24, 2010, U.S. Provisional Patent Application No. 61/333,551, filed May 11, 2010, and U.S. Provisional Patent Application No. 61/302,069, filed Feb. 5, 2010, the entireties of which are herein incorporated by reference. The present invention relates generally to nuclear reactor systems, and specifically to nuclear reactor systems that utilize natural circulation of the primary coolant in a single-phase, such as pressurized water reactors (“PWRs”). Over recent years, a substantial amount of interest has grown in developing commercially viable PWRs that utilize the phenomenon of natural circulation (also known as thermosiphon effect) to circulate the primary coolant to both cool the nuclear reactor and to vaporize a secondary coolant into motive vapor. CAREM (Argentina) is a 100 MW(e) PWR reactor design with an integrated self-pressurized primary system through which the primary coolant circulation is achieved by natural circulation. The CAREM design incorporates several passive safety systems. The entire primary system including the core, steam generators, primary coolant and steam dome are contained inside a single pressure vessel. The strong negative temperature coefficient of reactivity enhances the self-controlling features. The reactor is practically self-controlled and need for control rod movement is minimized. In order to keep a strong negative temperature coefficient of reactivity during the whole operational cycle, it is not necessary to utilize soluble boron for burn-up compensation. Reactivity compensation for burn-up is obtained with burnable poisons, i.e. gadolinium oxide dispersed in the uranium di-oxide fuel. Primary coolant enters the core from the lower plenum. After being heated the primary coolant exits the core and flows up through the riser to the upper dome. In the upper part, the primary coolant leaves the riser through lateral windows to the external region, then flows down through modular steam generators, decreasing its enthalpy by giving up heat to the secondary coolant in the steam generator. Finally, the primary coolant exits the internal steam generators and flows down through the down-corner to the lower plenum, closing the circuit. CAREM uses once-through straight tube steam generators. Twelve steam generators are arranged in an annular array inside the pressure vessel above the core. The primary coolant flows through the inside of the tubes, and the secondary coolant flows across the outside of the tubes. A shell and two tube plates form the barrier between primary and secondary coolant flow circuits. AST-500 (Russia) is a 500 MW(th) reactor design intended to generate low temperature heat for district heating and hot water supply to cities. AST-500 is a pressurized water reactor with integral layout of the primary components and natural circulation of the primary coolant. Features of the AST-500 reactor include natural circulation of the primary coolant under reduced working parameters and specific features of the integral reactor, such as a built-in steam-gas pressurizer, in-reactor heat exchangers for emergency heat removal, and an external guard vessel. V-500 SKDI *(Russia) is a 500 MW(e) light water integral reactor design with natural circulation of the primary coolant in a vessel with a diameter less than 5 m. The reactor core and the steam generators are contained within the steel pressure vessel (i.e., the reactor pressure vessel). The core has 121 shroudless fuel assemblies having 18 control rod clusters. Thirty six fuel assemblies have burnable poison rods. The hot primary coolant moves from the core through the riser and upper shroud windows into the steam generators located in the downcomer. The coolant flows due to the difference in coolant densities in the downcomer and riser. The pressurizer is connected by two pipelines, to the reactor pressure vessel and the water clean up system. The NHR-200 (China) is a design for providing heat for district heating, industrial processes and seawater desalination. The reactor power is 200 MW(th). The reactor core is located at the bottom of the reactor pressure vessel (RPV). The system pressure is maintained by N2 and steam. The reactor vessel is cylindrical. The RPV is 4.8 m in diameter, 14 m in height, and 197 tons in weight. The guard vessel consists of a cylindrical portion with a diameter of 5 m and an upper cone portion with maximum 7 m in diameter. The guard vessel is 15.1 m in height and 233 tons in weight. The core is cooled by natural circulation in the range front full power operation to residual heat removal. There is a long riser on the core outlet to enhance the natural circulation capacity. The height of the riser is about 6 m. Even in case of interruption of natural circulation in the primary circuit due to a LOCA the residual heat of the core can be transmitted by steam condensed at the uncovered tube surface of the primary heat exchanger. While the aforementioned PWRs utilize natural circulation of the primary coolant to both cool the reactor core and heat the secondary coolant, all of these natural circulation PWRs suffer from the drawback that the heat exchange equipment is integrated with and located within the reactor pressure vessel. Such an arrangement not only makes the heat exchange equipment difficult to repair and/or service but also subjects the equipment to corrosive conditions. Furthermore, locating the heat exchange equipment within the reactor pressure vessel results in increased complexity and a potential increase in the number of penetrations into the reactor pressure vessel. However, prior to the present invention, the location of the heat exchange equipment within the reactor pressure vessel was likely deemed necessary to achieve the natural circulation of the primary coolant in the PWR cycle. A drawback of other PWRs that exist in the art is the fact that the reactor pressure vessels have penetrations at both the top portion of the reactor pressure vessel and at the bottom portion of the reactor pressure vessel. Still another drawback of existing PWRs is the fact that a substantial length of piping and a large number of joints are used carry the primary coolant from the reactor pressure vessel to the heat exchange equipment, thereby increasing the danger of failure due to a pipe break scenario. These, and other drawbacks, are remedied by the present invention. A nuclear reactor system is presented herein that, in one embodiment, utilizes natural circulation (i.e., thermosiphon) to circulate a primary coolant in a single-phase through a reactor core and a heat exchange sub-system, wherein the heat exchange sub-system is located outside of the nuclear reactor pressure vessel. In some embodiments, the heat exchange sub-system is designed so as to not cause any substantial pressure drop in the flow of the primary coolant within the heat exchange sub-system that is used to vaporize a secondary coolant. In another embodiment, a nuclear reactor system is disclosed in which the reactor core is located below ground and all penetrations into the reactor pressure vessel are located above ground. In certain embodiment, the inventive nuclear reactor system is a PWR system. In one embodiment, the invention can be a natural circulation nuclear reactor system comprising: a reactor pressure vessel having an internal cavity; a reactor core comprising nuclear fuel disposed within the internal cavity at a bottom portion of the reactor pressure vessel; a heat exchange sub-system located outside of the reactor pressure vessel; a closed-loop primary coolant circuit that flows a primary coolant through the reactor pressure vessel to cool the reactor core and through the heat exchange sub-system to transfer heat to a secondary coolant; and wherein operation of the reactor core causes natural circulation of the primary coolant through the closed-loop primary coolant circuit in a single phase. In another embodiment, the invention can be a nuclear reactor system comprising: an elongated reactor pressure vessel having an internal cavity containing a primary coolant, the reactor pressure vessel extending along a substantially vertical axis, a major portion of the axial length of the reactor pressure vessel located below a ground level; a reactor core comprising nuclear fuel disposed within the internal cavity at a bottom portion of the reactor pressure vessel reactor and below the ground level; the reactor pressure vessel comprising a primary coolant outlet port located above the ground level; the reactor pressure vessel comprising a primary coolant inlet port located above the ground level; a heat exchange sub-system located outside of the reactor pressure vessel and above the ground level, an incoming hot leg of the heat exchange system fluidly coupled to the primary coolant outlet port and an outgoing cold leg of the heat exchange system fluidly coupled to the primary coolant inlet port; and wherein the major portion of the reactor pressure vessel is free of penetrations. In yet another embodiment, the invention can be a nuclear reactor system comprising: an elongated reactor pressure vessel having an internal cavity containing a primary coolant, the reactor pressure vessel extending along a substantially vertical axis; a reactor core comprising nuclear fuel disposed within the internal cavity at a bottom portion of the reactor pressure vessel reactor; a partition dividing the internal cavity of the reactor pressure vessel into a primary coolant riser passageway and a primary coolant downcomer passageway, the reactor core disposed within the primary coolant riser passageway; the reactor pressure vessel comprising a primary coolant outlet port in fluid communication with a top portion of the primary coolant riser passageway; the reactor pressure vessel comprising a primary coolant inlet port in fluid communication with a top portion of the primary downcomer riser passageway; at least one steam generator located outside of the reactor pressure vessel, an incoming hot leg of the steam generator fluidly coupled to the primary coolant outlet port and an outgoing cold leg of the steam generator fluidly coupled to the primary coolant inlet port; and wherein the steam generator does not cause any substantial pressure drop in a flow of the primary coolant through the steam generator resulting from an increase in elevation. Further areas of applicability of the present invention will become apparent from the detailed description provided hereinafter. It should be understood that the detailed description and specific examples, while indicating the preferred embodiment of the invention, are intended for purposes of illustration only and are not intended to limit the scope of the invention. The following description of the preferred embodiment(s) is merely exemplary in nature and is in no way intended to limit the invention, its application, or uses. Prior to discussing FIGS. 1-5 in detail, an overview of one specific embodiment of the inventive natural circulation reactor system, and its operation, will be set forth. Those skilled in the art will appreciate that the overview is directed to one very specific embodiment and that the details thereof are not limiting of the present invention in all embodiments. Furthermore, those skilled in the art will appreciate how the overview applies to the subsequent detailed discussion of FIGS. 1-5. I. Overview of One Potential Commercial Embodiment The inventive nuclear reactor system, in one potential commercial embodiment, is a 145 MWe nuclear reactor designed to provide an economical and safe source of clean energy from nuclear fission. Strengths of the inventive nuclear reactor system include its inherent safety and simplicity of operation. The operational simplicity of the inventive nuclear reactor system and the modest outlay required to establish and commission it will make it possible to deliver the fruits of pollution-free nuclear energy to the vast mass of humanity around the globe that does not presently have access to a reliable source of power or to a robust electrical enemy delivery system. Competitive with large nuclear reactors on a per-megawatt basis, the inventive nuclear reactor system is tailored to add generation capacity to the installed base incrementally with incremental capital outlays. Due to its inherent operational simplicity, the inventive nuclear reactor system requires a minimal cadre of trained personnel to run the plant. Multiple units of the inventive nuclear reactor system can be clustered at one location or geographically dispersed without a significant increase in the per-megawatt construction cost. Geographical dispersal and underground configuration serve as natural antidotes to post-9/11 concerns. The modest power output of the inventive nuclear reactor system makes it a viable candidate source of reliable electrical energy or for providing heating steam to a city or process steam as a cogeneration plant serving an industrial plant. As a passive small modular reactor of the PWR genre with safety, ease of maintenance and superb security, the inventive nuclear reactor system is ideally suited to serve as a reliable power source to strategic national assets of any country. Design features of the inventive nuclear reactor system that speak to its inherent safety and reliability are: 1. Reactor Core Deep Underground The reactor core resides deep underground in a thick-walled reactor pressure vessel (RPV) made of an ASME Code material that has decades of proven efficacy in maintaining reactor integrity in large PWR and BWR reactors. All surfaces wetted by the reactor coolant are made of stainless steel or Inconel, which eliminates a major source of crud accumulation in the reactor vessel. 2. Natural Circulation of the Reactor Coolant The inventive nuclear reactor system does not rely on any active components, such as a reactor coolant pump, for circulating the primary coolant through the closed-loop primary coolant circuit, which includes flow through the reactor core and the heat exchange sub-system. Instead, the flow of the primary coolant through the reactor pressure vessel, the horizontal steam generators, and other miscellaneous equipment occurs by the pressure head created by density differences in the flowing water in the hot and cold segments of the closed-loop primary coolant circuit. The reliability of gravity as a motive force underpins inherent safety of the inventive nuclear reactor system. The movement of the primary coolant requires no pumps, valves, or moving machinery of any kind, in certain embodiments. 3. No Reliance on Off-Site Power Offsite power is not essential for shutting down the inventive nuclear reactor system. The rejection of reactor residual heat during the shutdown also occurs by natural circulation. Thus, the need for an emergency shutdown power supply at the site—a major concern for nuclear plants—is eliminated. 4. Assurance of a Large Inventory of Water Around and Over the Reactor Core The reactor pressure vessel of the inventive nuclear reactor system has no penetrations in its below-ground portion, which can be the bottom 100 feet, which means that the reactor core will remain submerged in a large inventory of water. All penetrations in the reactor pressure vessel are located in the above-ground portion, or top portion, of the reactor pressure vessel and are small in size. The absence of large piping in the closed-circuit primary coolant circuit precludes the potential of a “large break” LOCA event. 5. All Critical Components Readily Accessible Both the heat exchange sub-system, which includes the steam generators, and the control rod drive system are located outside the reactor pressure vessel at a level that facilitates easy access, making their preventive maintenance and repair a conveniently executed activity. Each of the steam generators is a horizontal pressure vessel with built-in design features to conveniently access and plug tubes. 6. Demineralized Water The primary coolant (which can also be referred to as the reactor coolant) is demineralized water, which promotes criticality safety because of its strong negative reactivity gradient with rise in temperature. Elimination of borated water also simplifies the nuclear steam supply system (NSSS) by eliminating the systems and equipment needed to maintain and control boron levels in the primary coolant. Pure water and corrosion resistant primary coolant loop help minimize crud buildup in the reactor pressure vessel. 7. Modularity One can build only one of the inventive nuclear reactor systems at a site, or a large number thereof. Clustering a number of inventive nuclear reactor systems at one site will reduce the overall O&M costs. 8. Long Operating Cycle The inventive nuclear reactor system will operate for approximately 3.5 years before requiring refueling. 9. Short Construction Life Cycle Virtually all components of the inventive nuclear reactor system are shop fabricated. Site work is limited to reinforced concrete construction and a limited amount of welding to assemble the shop-built equipment and parts. As a result, it is possible to complete the construction of one of the inventive nuclear reactor systems in 24 months from the first shovel in the ground. 10. Efficient Steam Cycle A pair of two horizontal steam generators are arranged in series and integrally welded to the reactor pressure vessel. The efficiency of the power cycle of the inventive nuclear reactor system, and its compactness, is further enhanced by superheaters that are integrally welded to the horizontal steam generators. The superheaters, one attached to each steam generator, increases cycle efficiency and also protect both the high pressure and low pressure turbines from the deleterious effect of moist steam. 11. Integral Pressurizer The design of the reactor pressure vessel incorporates an integral pressurizer that occupies the upper reaches of the reactor pressure vessel. The pressurizer serves to control the pressure in the reactor vessel. 12. Suitable for Water-Challenged Sites The inventive nuclear reactor system can be installed at sites with limited water availability, such as creeks and small rivers that are inadequate for large reactors. The inventive nuclear reactor system can be operated equally well in a water-challenged region by using air-cooled condenser technology to reject the plant's waste heat. Using air in lieu of water, of course, results in a moderate increase in the plant's cost. 12. System Parameters in the Safe and Proven Range The operating pressure and temperature within the reactor pressure vessel is in the proven range for PWRs. Lower core power density than that used in large PWRs for improved thermal-hydraulic control (please see table below) and an improved margin to departure-from-nucleate boiling in the reactor core. Exemplary System Parameters DataNumber of fuel assemblies in the core 32Nominal thermal power, MWt 446 Nominal recirculation rate, MLb per hour 5.46 Reactor water outlet temperature, deg. F. 580 Reactor water inlet temperature, deg. F. 333 Reactor pressure, pounds per sq. inch 2,250 Water in the RV cavity, gallons 30,00 13. Minimized Piping Runs and Minimum Use of Active Components to Enhance Reliability and Cost Competitiveness The amount of piping in the close-loop primary coolant circuit and the secondary coolant circuit in the inventive nuclear reactor system is the least of any nuclear plant design on the market, as is the number of pumps and valves. 14. In-Service Inspection All weld seams in the primary system including those in the reactor pressure vessel wall are available at all times for inspection. In particular, the weld seams in the reactor pressure vessel can be inspected by operating a manipulator equipped in-service inspection device in the reactor well during power generation. Thus, inventive nuclear reactor system exceeds the in-service inspection capability expected of nuclear plants under ASME Code Section XI. 15. Earthquake Hardened Design Virtually all major equipment in the inventive nuclear reactor system are either underground or horizontally mounted to withstand strong seismic motions. This includes the reactor pressure vessel, the fuel pool, the reactor water storage tank (all underground) and the horizontal steam generators, the horizontal superheaters, and the horizontal kettle reboiler that are floor mounted. 16. Aircraft Impact Proof Containment The containment structure of the inventive nuclear react system is designed to withstand the impact of a crashing lighter plane or a commercial liner without sustaining a thin-wall breach. II. Detail Referring now to FIG. 1, a natural circulation nuclear reactor system 1000 (hereinafter the “reactor system 1000”) is illustrated according to one embodiment of the present invention. The reactor system 1000 generally comprises a reactor pressure vessel 100 and a heat exchange sub-system 200. The reactor pressure vessel 100 contains a primary coolant 101 that is used to cool the rector core 102 and to heat a secondary coolant within the heat exchange sub-system 200. The reactor pressure vessel 100 is fluidly coupled to an incoming hot leg 201 of the heat exchange sub-system 200 via a primary coolant outlet port 103. Similarly, the reactor pressure vessel 100 is also fluidly coupled to an outgoing cool leg 202 of the heat exchange sub-system 200 via a primary coolant inlet port 104. As a result, a closed-loop primary coolant circuit 300 is formed through which the primary coolant 101 flows in a single-phase. As discussed in greater detail below, the flow of the primary coolant 101 through the closed-loop primary coolant circuit is a natural circulation flow induced by the heat given off by the normal operation of the reactor core 102. In certain embodiments, the internal cavity 105 of the reactor pressure vessel 100 is maintained under sufficient pressure to maintain the primary coolant 101 in a liquid-phase despite the high temperature within the rector pressure vessel 100. In the exemplified embodiment, a pressure control sub-system 50 (commonly referred to in the art as a pressurizer) is located within a top region of the reactor pressure vessel 100 and is configured to control the pressure of the internal cavity 105 of the reactor pressure vessel 100. The pressure control sub-system 50 is integral with the removable head 106 of the reactor pressure vessel 100 to prevent line break concerns and to provide a more compact reactor system 1000. Pressurizers are well known in the art and any standard pressurizer could be used as the pressure control sub-system 50. In one embodiment, the internal cavity 105 of the reactor pressure vessel 100 is maintained at a pressure in a range of 2000 psia to 2500 psia. In one more specific embodiment, the internal cavity 105 of the reactor pressure vessel 100 is maintained at a pressure between 2200 psia to 2300 psia. Of course, the exact pressure maintained in the internal cavity 105 of the reactor pressure vessel 100 is not to be limiting of the invention unless specifically claimed. The reactor pressure vessel 100 is an elongated tubular pressure vessel formed by a thick wall made of an acceptable ASME material, such as stainless steel. The reactor pressure vessel 100 extends from a bottom end 107 to a top end 108 along a substantially vertical axis A-A, thereby defining an axial length of the reactor pressure vessel 100. In one embodiment, the reactor pressure vessel 100 has an axial length of over 100 feet to facilitate an adequate level of turbulence in the recirculating primary coolant 101 from the natural circulation (also referred to as thermosiphon action in the art). In certain other embodiments, the reactor pressure vessel 100 has an axial length in a range between 100 feet to 150 feet. Of course, the invention is not so limited in certain alternate embodiments. The reactor pressure vessel 100 generally comprises a domed head 106 and a body 109. The domed head 106 is detachably coupled to a top end of the body 109 so as to be removable therefrom for refueling and maintenance. The domed head 106 can be coupled to the body 109 through the use of any suitable fastener, including bolts, clamps, or the like. In the exemplified embodiment, the body 109 comprises an upper flange 110 and the domed head 106 comprises a lower flange 111 that provided mating structures through which bolts 114 (FIG. 4) extend to couple the domed head 106 to the body 109. When the domed had 106 is coupled to the body 109, a hermetic seal is formed therebetween via the use of a gasket or other suitably contoured interface. The body 109 of the reactor pressure vessel 100 comprises an upstanding tubular wall 112 and a domed bottom 113 that hermetically seals the bottom end 107 of the reactor pressure vessel 100. The tubular wall 112 has a circular transverse cross-sectional profile in the illustrated embodiment but can take on other shapes as desired. In the exemplified embodiment, the domed bottom 113 is integral and unitary with respect to the tubular wall 112. Of course, in other embodiments, the domed bottom 113 may be a separate structure that is secured to the tubular wall 112 via a welding or other hermetic connection technique, such as the flanged technique described above for the domed head 106 and the body 109. Integral and unitary construct of the domed bottom 113 and the body 109 is, however, preferable in certain embodiments as it eliminates seams and/or interfaces that could present rupture potential. The reactor pressure vessel 100 forms an internal cavity 105 in which a reactor core 102 is housed. The reactor core 102 comprises nuclear fuel, in the form of fuel assemblies, as is known in the art. The details of the structure of the reactor core 102 are not limiting of the present invention in and the reactor system 1000 can utilize any type of reactor core or nuclear fuel. The reactor core 102 is positioned in a bottom portion 115 of the reactor pressure vessel 100. In one embodiment, the reactor core 102 has a core thermal power of 400 MWt to 600 MWt during the operation thereof. In one embodiment, the reactor core 102 is comprised of vertically arrayed fuel assemblies. The spacing between the fuel assemblies is governed by the design objective of keeping the reactivity (neutron multiplication factor) at 1.0 at all locations in the reactor pressure vessel 100. The criticality control in the axial direction is provided by the built-in neutron poison in the fuel rods (called IFBAs by Westinghouse) and possibly by control rods. A partition 120 is provided within the internal cavity 105 of the reactor pressure vessel 100 that divides the internal cavity into a primary coolant riser passageway 105A and a primary coolant downcomer passageway 105B. Both the passageways 105A, 105B are axially extending vertical passageways that form part of the closed-loop primary coolant circuit 300. In the exemplified embodiment, the partition 120 comprises an upstanding tubular wall portion 120A and a transverse wall portion 120B. The tubular wall portion 120A is an annular tube that is mounted within the internal cavity 105 of the reactor pressure vessel 100 so as to be concentrically arranged with respect to the upstanding wall 112 of the reactor pressure vessel 100. As a result, the primary coolant downcomer passageway 105B is an annular passageway that circumferentially surrounds the primary coolant riser passageway 105A. The primary coolant downcomer passageway 105B is formed between an outer surface 121 of the upstanding tubular wall portion 120A of the partition 120 and the inner surface 116 of the upstanding wall 112 of the reactor pressure vessel 100. The primary coolant riser passageway 105B is formed by the inner surface 122 of the upstanding tubular wall portion 120A of the partition 120. The transverse wall portion 120B is an annular ring-like plate that is connected to a top end of the of the upstanding tubular wall portion 120A of the partition 120 at one end and to the upstanding wall 112 of the reactor pressure vessel 100 on the other end. The transverse wall portion 120B acts a separator element that prohibits cross-flow of the primary coolant 101 between the primary coolant riser passageway 105A and the primary coolant downcomer passageway 105B within the top portion 117 of the reactor pressure vessel 100. In essence, the transverse wall portion 120B forms a roof of the primary coolant downcomer passageway 105B that prevents the heated primary coolant 101 that exits the reactor pressure vessel 100 via the primary coolant outlet port 103 from mixing with the cooled primary coolant 101 that enters the reactor pressure vessel 100 via the primary coolant inlet port 104, and vice-versa. Cross-flow of the primary coolant 101 between the primary coolant riser passageway 105A and the primary coolant downcomer passageway 105B is prohibited by the upstanding tubular wall portion 120A of the partition 120. In addition to physically separating the flow of the heated and cooled primary coolant 101 within the primary coolant downcomer and riser passageways 105A, 105B as discussed above, the partition 120 also thermally insulates the cooled primary coolant 101 within the primary coolant downcomer passageway 105B from the heated primary coolant 101 within the primary coolant riser passageway 105A. Stated simply, one does not want heat to transfer freely through the partition 120. Thus, it is preferred that the partition 120 be an insulating partition in the sense that its effective coefficient of thermal conductivity (measured radially from the primary coolant riser passageway 105A to the primary coolant downcomer passageway 105B) is less than the coefficient of thermal conductivity of the primary coolant 101. Making the effective coefficient of thermal conductivity of the partition 120 less than the coefficient of thermal conductivity of the primary coolant 101 ensures that the primary coolant 101 in the primary coolant downcomer passageway 105B remains cooler than the primary coolant 101 in the primary riser passageway 105A, thereby maximizing the natural circulation rate of the primary coolant 101 through the closed-loop primary coolant circuit 300. In a very simple construction, this can be achieved by creating the partition 120 out of a single solid material that has a low coefficient of thermal conductivity. However, it must be considered that the material should neither degrade nor deform under the operating temperatures and pressures of the reactor pressure vessel 100. In such an embodiment, the effective coefficient of thermal conductivity is simply the coefficient of thermal conductivity of the single solid material. In the exemplified embodiment, the low coefficient of thermal conductivity of the partition 120 is achieved by making the partition 120 as a multi-layer construction. As exemplified, the partition 120 comprises an insulating layer 124 that is sandwiched between two outer layers 125A, 125B. In one embodiment, the insulating layer 124 is a refractory material while the outer layers 125A, 125B are stainless steel or another corrosion resistant material. In certain embodiments, the insulating layer 124 is full encased in the outer layers 125A, 125B. The internal cavity 115 of the reactor pressure vessel 100 also comprises a plenum 118 at the bottom portion 115 of the reactor pressure vessel 100 that allows cross-flow of the primary coolant 101 from the primary coolant downcomer passageway 105B to the primary coolant riser passageway 105A. In the exemplified embodiment, the plenum 118 is created by the fact that the bottom end 123 of the upstanding tubular wall portion 120A of the partition 120 is spaced from the inner surface 119 of the domed bottom 113, thereby creating an open passageway. In alternate embodiments, the partition 120 may extend all the way to the inner surface 119 of the domed bottom 113. In such embodiments, the plenum 118 will be formed by providing a plurality of apertures/openings in the partition 120 so as to allow the desired cross-flow. The internal cavity 105 further comprises a plenum 126 at the top portion 117 of the reactor pressure vessel 100. The plenum 126 allows the heated primary coolant 101 that is rising within the primary coolant riser passageway 105A to gather in the top portion 117 of the reactor pressure vessel 100 and then flow transversely outward from the vertical axis A-A and through the primary coolant outlet port 103. The reactor core 102 is located within the primary coolant riser passageway 105A above the bottom plenum 118. During operation of the reactor core 102, thermal energy produced by the reactor core 102 is transferred into the primary coolant 101 in the primary coolant riser passageway 105A adjacent the reactor core 102, thereby becoming heated. This heated primary reactor coolant 101 rises upward within the primary coolant riser passageway 105A due to its decreased density. This heated primary coolant 101 gather in the top plenum 126 and exits the reactor pressure vessel 100 via the primary coolant outlet port 103 where it enters the heat exchange sub-system 200 as the incoming hot leg 201. In one embodiment, the heated primary coolant 101 entering the hot leg 201 of the heat exchanger has a temperature of at least 570° F., and in another embodiment a temperature in a range of 570° F. to 620° F. This heated primary coolant 101 passes through the heat exchange sub-system 200 where its thermal energy is transferred to a secondary coolant (described below in greater detail with respect to FIG. 2), thereby becoming cooled and exiting the heat exchange sub-system 200 via the cold leg 202. When exiting the cold leg 202 of the heat exchange sub-system, this cooled primary coolant 101 has a temperature in a range of 300° F. to 400° F. in one embodiment. In another embodiment, the heat exchange sub-system 200 is designed so that the temperature differential between the heated primary coolant in the hot leg 201 and the cooled primary coolant in the cold leg is at least 220° F. The cooled primary coolant 101 exiting the cold leg of the heat exchange sub-system 200 then enters the reactor pressure vessel 100 via the primary coolant inlet port 104, thereby flowing into a top portion 127 of the primary coolant downcomer passageway 105B. Once inside the primary coolant downcomer passageway 105B, the cooled primary coolant 101 (which has a greater density than the heated primary coolant 101 in the primary coolant riser passageway 105A) flows downward through the primary coolant downcomer passageway 105B into the bottom plenum 118 where it is drawn back up into the primary coolant riser passageway 105A and heated again by the reactor core 102, thereby completing a cycle through the closed-loop primary circuit 300. As discussed above, operation of the reactor core 102 causes natural circulation of the primary coolant 101 through the closed-circuit primary coolant circuit 300 by creating a riser water column within the primary coolant riser passageway 105A and a downcomer water column within the primary coolant downcomer passageway 105B. In one embodiment, the riser water column and the downcomer water column have a vertical height in a range of 80 ft. to 150 ft., and more preferably from 80 ft. to 120 ft. The vigorousness of the natural circulation (or thermosiphon flow) is determined by the height of the two water columns (fixed by the reactor design), and the difference between the bulk temperature of the two water columns (in water the SES and the downcomer space). For example, water at 2200 psia and 580° F. has density of 44.6 lb/cubic feet. This density increases to 60.5 lb/cubic feet if the temperature reduces to 250° F. The hot and cold water columns 60 feet high will generate a pressure head of 6.6 psi which is available to drive natural circulation of the primary coolant 101 through the closed-loop primary coolant circuit 300. A 90 feet high column will generate 50% greater head (i.e., 9.9 psi). As a result of the natural circulation of the primary coolant 101 achieved by the water columns and gravity, the reactor system 1000 is free of active equipment, such as pumps or fans, for forcing circulation of the primary coolant through the closed-loop primary coolant circuit. In the embodiment illustrated in FIG. 1, the, primary coolant outlet port 103 is at a slightly lower elevation (1-3 ft.) than the primary coolant inlet port 104. However, in other embodiments, the primary coolant outlet port 103 and the primary coolant inlet port 104 will be at substantially the same elevation (see FIGS. 4 and 5). When the primary coolant outlet port 103 and the primary coolant inlet port 104 are at substantially the same elevation the partition 120 will be appropriately designed. Furthermore, as used herein, the term port includes mere apertures or openings. In one embodiment, the primary coolant 101 is a liquid that has a negative reactivity coefficient. Thus, the chain reaction in the reactor core 102 would stop automatically if the heat rejection path to the heat exchange sub-system 200 is lost in a hypothetical scenario. Thus, the reactor system 1000 is inherently safe. In one specific embodiment, the primary coolant 101 is demineralized water. All systems and controls used to maintain boron concentration in the reactor vessel in a typical PWR are eliminated from the reactor system 1000. Moreover, the use of demineralized water as the primary coolant 101 and the existence of the corrosion resistant surfaces of the reactor pressure vessel 100 help maintain crud buildup to a minimum. The reactivity control in the reactor core 102 is maintained by a set of control elements (burnable poisons) that are suspended vertically and occupy strategic locations in and around the fuel assemblies to homogenize and control the neutron flux. Referring now to FIGS. 1, 4 and 5 concurrently, it can be seen that a major portion 130 of the axial length of the reactor pressure vessel 100 located below a ground level 400 while a minor portion 131 of the axial length of the reactor pressure vessel 100 extends above the ground level 400. As such, the reactor core 102 is located deep below the ground level 400 while the heat exchange sub-system 200 is located above the ground level 400. In one embodiment, the heat exchange sub-system 200 is at an elevation that is 80 ft. to 150 ft, and preferably 80 ft. to 120 ft., greater than the elevation of the reactor core 102. The minor portion 131 of the reactor pressure vessel 100 includes a top portion 132 of the body 109 and the domed head 106. The primary coolant outlet port 103 and the primary coolant inlet port 104 are located on the minor portion 131 of the reactor pressure vessel 100 that is above the ground level 400. More specifically, the primary coolant outlet port 103 and the primary coolant inlet port 104 are located on the top portion 132 of the body 109 of the reactor pressure vessel 100 that is above the ground level 400. The major portion 130 includes a majority of the body 109 and the domed bottom 113. In certain embodiment, the major portion 130 of the reactor pressure vessel 130 is at least 75% of the axial length of the reactor pressure vessel 100. In other embodiments, the major portion 130 of the reactor pressure vessel 130 is between 60% to 95% of the axial length of the reactor pressure vessel 100. In another embodiment, the major portion 130 of the reactor pressure vessel 130 is between 75% to 95% of the axial length of the reactor pressure vessel 100. The reactor pressure vessel 100 comprises a reactor flange 150 which is seated on and supports the vessel from a concrete slab 400-1 that defines the ground level 400. The top portion 132 of the body 109 of the reactor pressure vessel 100 is welded to the reactor flange 150, which is a massive upper forging. The reactor flange 150 also provides the location for the primary coolant inlet port 104 and the primary coolant outlet port 103 (FIGS. 4 and 5), and the connections to the heat exchange sub-system 200 (and for the engineered safety systems to deal with various postulated accident scenarios). This reactor flange 150 contains vertical welded lugs to support the weight of the reactor pressure vessel 100 in the reactor well 410 in a vertically oriented cantilevered manner (FIG. 1). As a result, the reactor pressure vessel 100 is spaced from the wall surfaces 411 and floor surface 412 of the reactor well 410, thereby allowing the reactor pressure vessel 100 to radially and axially expand as the reactor core 102 heats up during operation and causes thermal expansion of the reactor pressure vessel 100. Furthermore, the major portion 130 of the reactor pressure vessel 100 is free of penetrations. In other words, the major portion 130 of the reactor pressure vessel 100 comprises no apertures, holes, opening or other penetrations that are either open or to which pipes or other conduits are attached. All penetrations (such as the primary coolant inlet and outlet ports 103, 104) in the reactor pressure vessel 100 are located in the above-ground minor portion 131, and more specifically in the top portion 132 of the body 109 of the reactor pressure vessel 100. In one embodiment, it is further preferred that the major portion 130 be a unitary construct with no connections, joints, or welds. The bottom portion 115 of the reactor pressure vessel 100 is laterally restrained by a lateral seismic restraint system 160 that spans the space between the body 109 of the reactor pressure vessel 100 and the wall surfaces 411 of the reactor well 410 to withstand seismic events. The seismic restraint system 160, which comprises a plurality of resiliently compressible struts 161, allows for free axial and diametral thermal expansion of the reactor vessel. The bottom of the reactor well 410 contains engineered features to flood it with water to provide defense-in-depth against a (hypothetical, non-mechanistic) accident that produces a rapid rise in the enthalpy of the reactor's contents. Because the reactor system 1000 is designed to prevent loss of the primary coolant 101 by leaks or breaks and the reactor well 410 can be flooded at will, burn-through of the reactor pressure vessel 100 by molten fuel (corium) can be ruled out as a credible postulate. This inherently safe aspect simplifies the design and analysis of the reactor system 1000. Referring now to FIGS. 2 and 4-5 concurrently, an embodiment of the heat exchange sub-system 200 is illustrated. While a specific embodiment of the heat exchange sub-system 200 will be described herein, it is to be understood that, in alternate embodiments, one or more of components can be omitted as desired. For example, in certain embodiments, one or both of the horizontal superheaters 205, 206 may be omitted. In certain other embodiments, one of the horizontal steam generators 203, 204 may be omitted and/or combined into the other one of the horizontal steam generators 203,204. Moreover, additional equipment may be incorporated as necessary so long as the natural circulation of the primary coolant 101 through the closed-loop primary coolant circuit 300 is not prohibited through the introduction of substantial head loss. As mentioned above, the heat exchange subsystem 200 comprises an incoming hot leg 201 that introduces heated primary coolant into the portion of the closed-loop primary coolant circuit 300 that passes through the heat exchange sub-system 200 and an outgoing cold leg 202 that removes cooled primary coolant from the portion of the closed-loop primary coolant circuit 300 that passes through the heat exchange sub-system 200. In order to minimize (and in some embodiments eliminate) pressure loss in the closed-loop primary coolant circuit 300 caused by an increase in the elevation of the primary coolant flow, the steam generators 203, 204 and the superheaters 205, 206 are all of the horizontal genre (i.e., the tubes which carry the primary coolant extend substantially horizontal through the shell-side fluid) and are in horizontal alignment with each other where possible. Within the heat exchange sub-system 200, the primary coolant flow of the closed-loop primary coolant circuit 300 is divided into two paths 211, 212 at a flow divider 215. The flow divider 210 can be a three-way valve, a three-way mass flow controller, or a simple Y plumbing joint. The first path 211, which carries the majority of the primary coolant flow, travels through the first horizontal steam generator 203 and then through the second horizontal steam generator 204. Meanwhile, second path 212, which carries a minority of the primary coolant flow, travels through the first horizontal superheater 205 and then through the second horizontal superheater 206. After passing through the first and second horizontal steam generators 203, 204 and the first and second horizontal superheaters 205, 206, the first and second paths 211, 212 converge in a flow converger 216, which combines the primary coolant flows of the first and second paths 211, 212 and directs the combined flow to the outgoing cold leg 202. As with the flow divider 215, the flow converger 216 may be a three-way valve, a three-way mass flow controller, or a simple Y plumbing joint. In one embodiment, 10% to 15% of the incoming primary coolant flow that enters the heat exchange sub-system 200 via the hot leg 201 is directed into the second path 212 while the remaining 85% to 90% of the incoming primary coolant is directed into the first path 211. In one specific example, the incoming primary coolant that enters the heat exchange sub-system 200 via the hot leg 201 has a flow rate of 5 to 7 million lbs./hr. In this example, 0.6 to 1 million lbs./hr. of the primary coolant is directed into the second path 212 while the remainder of the primary coolant flow is directed into the first path 211. The first and second horizontal steam generators 203, 204 are operably coupled in series to one another along the first path 211 of the closed-loop primary coolant circuit 300. Both of the horizontal steam generators 203, 204 are horizontally disposed shell-and-tube heat exchangers. The first horizontal steam generator 203 is a high pressure steam generator while the second horizontal steam generator 204 is a low pressure steam generator (in comparison to the high pressure steam generator). The high first steam generator 203 is located upstream of the second horizontal steam generator 204 along the closed-loop primary coolant circuit 300. Similarly, the first and second horizontal superheaters 205, 206 are operably coupled in series to one another along the second path 212 of the closed loop primary coolant circuit 300. The first horizontal superheater 205 is a high pressure superheater while the second horizontal superheater 206 is a low pressure superheater (in comparison to the high pressure superheater). The high first steam superheater 205 is located upstream of the second horizontal superheater 206 along the closed-loop primary coolant circuit 300. Furthermore, the first and second superheaters 205, 206 are located in parallel to the first and second horizontal steam generators 203, 204 along the closed-loop primary coolant circuit 300. Furthermore, the first and second horizontal steam generators 203, 204 are interconnected by a return header so that the hot primary coolant entering the first horizontal steam generator 203 heats the secondary coolant to make steam for the high-pressure turbine 220 and then proceeds to the second horizontal steam generator 204 with minimal pressure loss to make steam for the low-pressure turbine 221. The flow of the primary coolant in the first path 211 is used to convert a secondary coolant flowing through the shell-side of the first and second horizontal steam generators 203, 204 from liquid-phase to gas-phase through the transfer of heat form the primary coolant to the secondary coolant within the first and second horizontal steam generators 203, 204. Because the flow of the primary coolant through the first and horizontal second steam generators 203, 204 is substantially horizontal in nature, the flow of the primary coolant through the first path 211 does not cause any substantial pressure drop in the closed-loop primary coolant circuit 300 resulting from an increase in elevation. Moreover, because of the horizontal alignment of the first and second horizontal steam generators 203, 204 with each other and the primary coolant outlet and inlet ports 103, 104 of the reactor pressure vessel 100 (FIG. 5), the primary coolant flow that travels along the first path 211 from the primary coolant outlet port 103 of the reactor pressure vessel 100 to the primary coolant inlet port 104 of the reactor pressure vessel 100 does not cause any substantial pressure drop in the closed-loop primary coolant circuit 300 resulting from an increase in elevation. While the achievement of substantial zero pressure drop in the closed-loop primary coolant circuit 300 resulting from an increase in elevation is exemplified in terms of a horizontal flow, it is possible that such substantial zero pressure drop can be achieved by a decline in elevation as the primary coolant flows downstream in the closed-loop primary coolant circuit 300. The flow of the primary coolant in the second path 212 is used to superheat the vapor-phase of the secondary coolant exiting the first and second horizontal steam generators 203, 204 via the first and second horizontal superheaters 205, 206 respectively, thereby further drying the vapor-phase of the secondary coolant. The use if the horizontal superheaters enhance the thermodynamic efficiency of the turbine cycle, carried out on the high pressure turbine 220 and the low pressure turbine 221. The first and second horizontal superheaters 205, 206 are horizontally disposed shell-and-tube heat exchanger positioned directly above (and in series) with the first and second steam generators 203, 204 (FIG. 5). However, due to the slight increase in the elevation of the superheaters 205, 206 resulting from their location above the first and second horizontal steam generators 203, 204, the flow of the primary coolant in the second path 212 does cause some pressure drop in the closed-loop primary coolant circuit 300 resulting from an increase in elevation. However, because only a small amount (10% to 15%) of the total primary coolant that flows through the heat exchange subsystem 200 is directed into the second path 212 and through the horizontal superheaters 205, 206, the pressure drop does not significantly affect the desired natural circulation. Moreover, the increase in elevation is negligible when compared to the height of the flow driving water columns. In such an embodiment, at least 85% of the flow of the primary coolant through the heat exchange sub-system 200 is still entirely horizontal from the primary coolant outlet 103 to the primary coolant inlet 104 and does not cause any substantial pressure drop in the closed-loop primary coolant circuit 300 due to increase in elevation. Further, in certain alternate embodiments, the horizontal superheaters 205, 206 could be eliminated and/or repositioned to be in horizontal alignment with the horizontal steam generators 203, 204. As shown in FIG. 5, the first and second horizontal steam generators 203, 204 are coupled directly to the each other and to the reactor pressure vessel 100. More specifically, the inlet of the first horizontal steam generator 203 is coupled directly to the primary coolant outlet port 103 of the reactor pressure vessel 100 while the outlet of the first horizontal steam generator 203 is coupled directly to the inlet of the second horizontal steam generator 204. The outlet of the second horizontal steam generator 204, is in turn, coupled directly to the primary coolant inlet port 104 of the reactor pressure vessel 100. The first and second horizontal steam generators 203, 204 are arranged so as to extend substantially parallel to one another, thereby collectively forming a generally U-shaped structure. Thus, the first path 211 also takes on a generally U-shape. In certain embodiments, the first and second horizontal steam generators 203, 204 are integrally welded to the reactor vessel 100 and to each other. Referring now to FIGS. 2 and 3A-B, each of the first and second horizontal steam generators 203, 204 comprise a preheating zone 208, 210 and a boiling zone 207, 209. Both of the first and second horizontal steam generators 203, 204 are of the single-pass type in which the primary coolant flow of the first path 211 is the tube-side fluid. Each of the single-pass tubes 330 extend substantially horizontally through the preheating zones 208, 210 and the boiling zones 207, 209. The secondary coolant circuit has a main feedwater intake 501 and a return to condenser exit 502 into and out of the heat exchange sub system 200 respectively. The secondary coolant, which is in the liquid-phase 505, enters each of the first and second horizontal steam generators 203, 204 along line 503. The incoming liquid phase 505 of the secondary coolant is preheated within the preheater zones 208, 210 of the first and second horizontal steam generators 203, 204. The secondary coolant in liquid-phase 505 flows through a tortuous path as shell-side fluid in the preheater zones 208, 210 and then enters the boiling zones 207, 209, where it is further heated by the primary coolant flow passing through the tubes 330. In the boiling zones 207, 209, the liquid-phase secondary coolant 505 vaporizes and exits the first and second horizontal steam generators 203, 204 as high pressure and low pressure steam 504 that is respectively supplied to the high and low pressure turbines 220, 221. The shells of the horizontal steam generators 203, 204 and the horizontal superheaters 205, 206 provide additional barriers against potential large break LOCAs, as do the turning plenum and the eccentric flanges that join the steam generators 203, 204 to the reactor pressure vessel 100, as shown in FIGS. 4 and 5. All systems connected to the reactor vessel 100 use a similar approach to ensure that there is no potential for a large-break LOCA that could rapidly drain the water from the reactor vessel 100 and uncover the reactor core 102. As long as the reactor core 102 is covered under all potential conditions of operation and hypothetical accident, the release of radioactive material to the public is minimal. As explained in the foregoing, the reactor system 1000 is an intrinsically safe reactor which, in the event of a problem external to the reactor containment building or within containment, is designed to automatically shut down in a safe mode with natural circulation cooling. Nevertheless, to instill maximum confidence, a number of redundant safety systems can be engineered to protect public health and safety under hypothetical accident scenarios that are unknown or unknowable, i.e., cannot be mechanistically postulated. In the case of an abnormal condition when the normal heat transport path through the steam generators are not available, then the pressure in the reactor vessel 100 will begin to increase. In such a case rupture discs will breach allowing the reactor coolant to flow into a kettle reboiler located overhead. The kettle will have a large inventory of water that will serve to extract the heat from the reactor coolant until the system shuts down. Diverse systems perform duplicate or overlapping functions using different physical principles and equipment to ensure that a common-mode failure is impossible. As used throughout, ranges are used as shorthand for describing each and every value that is within the range. Any value within the range can be selected as the terminus of the range. In addition, all references cited herein are hereby incorporated by referenced in their entireties. In the event of a conflict in a definition in the present disclosure and that of a cited reference, the present disclosure controls. 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 | An energy absorbing and displacing structure for athletic protective equipment, such as an athletic shin-guard, is provided using a flexible web-shaped body to hold a rigid band-shaped member in place. The inner rigid band-shaped member follows the contour of the desired area to protect. The outer flexible web-shaped body is made of a softer rubber like material and works as a locator and supporter keeping the inner, rigid band-shaped member, in proper location. |
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061920967 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a position detector assembly for a control rod in a nuclear reactor, and in particular relates to a control rod position detector assembly employing a magnetostrictive wire which works in cooperation with a control rod drive unit to continuously monitor the longitudinal position of the control rod relative to the reactor core. 2. Description of the Related Art FIG. 8 is a cross-section of the construction of a pressurized water nuclear reactor. As shown in the figure, the output of the reactor is controlled by control rod drive units 5 disposed in an upper portion of a reactor vessel 6, which insert and extract control rods 4 into and out of the reactor core. The control rods are moved longitudinally by longitudinally moving drive shafts 3 connected to the control rods 4 within pressure housings 1. The positions of the drive shafts 3, that is, the positions of the control rods 4 relative to the reactor core, are detected by control rod position detector assemblies 7 comprising detector coils 2 disposed around the outer circumference of each of the pressure housings 1. FIG. 9 is a cross-section showing the relationship between the control rod drive unit 5 and the conventional control rod position detector assembly 7. As shown in the figure, the conventional control rod position detector assembly 7 comprises detector coils 2 mounted on the outside of the pressure housing 1 of the control rod drive unit 5. Detector coils 2 corresponding in number to the length of the control rod 4 when withdrawn from the reactor core, usually forty-two, are mounted with even spacing on a coil support pipe 8 on the outside of and coaxial to the pressure housing 1. In anticipation of events such as breakages in the wiring, the detector coils are divided into two systems consisting of an A system comprising the set of alternate detector coils 2a and a B system comprising the set of detector coils 2b. The spacing between adjacent detector coils is approximately 90 mm, or when each system is considered separately, 180 mm because of the alternation. At the same time, the drive shaft 3 of the control rod drive unit 5, which is the portion whose position is detected, is usually composed of a stainless magnetic material. As a result, the drive shaft 3 itself is magnetized because the magnetic field from the control rod drive unit 5 is strong. Since the temperature within the pressure housing 1 is approximately 300 degrees Celsius, limits on the working temperature of the insulating materials used in the detector coils 2 of the control rod position detector assembly 7 make cooling the coils compulsory. For that reason, air is supplied to the space 9 between the pressure housing 1 and the detector coil support pipe 8 on which the detector coils are mounted and the detector coils are air-cooled from within as shown in FIG. 9. Next, the method of detecting the position of a control rod 4 by means of the detector coils 2 of the control rod position detector assembly 7 will be explained. When the magnetic drive shaft 3 passes through the center of a detector coil 2, an electric potential is induced in the detector coil 2 and as a result the impedance in the detector coil 2 changes. Consequently, by detecting the changes in impedance in each of the detector coils 2, the position of the tip of the magnetic drive shaft 3 can be detected as it moves inside the pressure housing 1 of the control rod drive unit 5, and thus the position of the control rod 4 within the nuclear reactor can be ascertained. Also, in order to ensure the reliability of the reactor, it is necessary to measure the descent times (insertion times) of the control rods. The method of measuring the descent times of a control rod by means of the control rod position detector assembly 7 is to measure the insertion times from when a control rod starts to descend until it reaches a dashpot 10 (see FIG. 10) by means of the changes in electric potential (changes in velocity) in the generated electric currents which depend on the descent velocity of the magnetic drive shaft 3 as it passes through the detector coils 2. Thus, as shown in FIG. 11, when the descent velocity of the control rod 4 is fast, the electric potential of the electric current generated in an detector coils 2 rises, and when the velocity of the control rod 4 suddenly decreases, the electric potential suddenly decreases. In order to make use of such changes in the electric potential of the electric current generated in the detector coils 2, the dashpots in a fuel assembly which decelerate control rods 4 by means of fluid resistance are each disposed in a position approximately 85 percent of the fully inserted position. Consequently, the position of each control rod 4 can be precisely determined by a sudden decrease in electric potential at a position such as T1 shown in FIG. 11. Moreover, whether or not the control rod 4 has been completely inserted into the reactor core is determined, as shown in FIG. 11, by detecting the rebound waveform R up to the rest point T2, that is, the waveform of the rebounding of the drive shaft 3 due to shock absorbing springs mounted on the control rod clusters as the drive shaft 3 of the control rod 4 reaches the bottom end. However, since the position of the tip of the drive shaft 3 is detected by changes in impedance in the detector coils 2, signals indicating the position of the control rod 4 can only be obtained at the positions of the detector coils 2. In other words, the intervals at which the position of the control rod 4 can be detected, depend on the spacing at which the detector coils are mounted, which is approximately 90 mm. Generally, a control rod drive unit 5 drives a control rod 4 in steps, the length of each of these steps being approximately 16 mm. Consequently, one problem is that the physical position of the control rod 4 can only be confirmed at intervals corresponding to several drive steps. Furthermore, the detector coils 2 are divided into two systems, the A system and the B system, and when one system cannot be used because of circuit failure, the position can only be detected at the single system intervals of 180 mm, in other words, intervals approximately ten times the length of a drive step of the control rod 4, further reducing accuracy. Consequently, from the viewpoint of a protective system for the reactor, there is a need to consider the uncertainty of the position of the control rods when designing reactor cores. In addition, as explained above, measurement of the descent times of the control rods according to the control rod position detector assembly 7 involves measuring the insertion times from the commencement of descent until a dashpot is reached by means of the changes in electric potential in the electric currents which depend on the descent velocity of the magnetic drive shaft 3 as it passes through the detector coils 2, and it is well known that the descent times of a control rod cannot be accurately measured when the descent velocity is slow because the electric potential is low, making the commencement of descent, the position of the dashpot, and the fully inserted position unclear. Moreover, in a rare event such as the control rod 4 stopping during descent, it is impossible to confirm the rest position of the control rod 4 (fully inserted or partway), and therefore the accuracy and reliability of the detection of the position of the control rod is low. Similarly, when the rebound waveform used to determine whether the control rod 4 has been completely inserted is smaller than the descent velocity, the former is often unclear, and it is therefore difficult to detect whether the control rod 4 has been completely inserted into the reactor core. In addition, due to limits on the working temperature of the insulating materials used in the detector coils 2 of the control rod position detector assembly 7, the ambient temperature around the detector coils 2 must be vigorously cooled and it is necessary to ensure that design conditions are not exceeded, requiring that the volume of the cooling equipment for the control rod drive unit be made quite large so that it can handle the large amounts of heat given off by the housing 1 which is heated to temperatures as high as about 300 degrees Celsius. SUMMARY OF THE INVENTION The present invention aims to solve the above problems and a major object of the present invention is to provide a magnetostrictive wire control rod position detector assembly which will allow nuclear reactor output to be increased, control performance to be improved, and peripheral equipment to be rationalized by continuously and accurately detecting the position of a control rod in a reactor core. In order to achieve the above object, according to a major aspect of the present invention, a magnetostrictive wire control rod position detector assembly for detecting the position of a movable member within a cylindrical member comprises: a magnet or magnets mounted on a non-magnetic portion of the movable member which is free to move in the longitudinal direction on the inside of the cylindrical member and is at least partly composed of a non-magnetic material; a magnetostrictive wire detector longitudinally mounted on the outer circumference of the cylindrical member, which is provided in a predetermined place with a receiver which detects torsional waves; and a pulsed current generator circuit which supplies a pulsed current from the receiver end of the magnetostrictive wire detector to the magnetostrictive wire of the magnetostrictive wire detector. In such a construction, a rotational magnetic field is generated in the magnetostrictive wire by the pulsed current. When the rotational magnetic field approaches the magnetic field of the magnet or magnets mounted on the movable member, mutual interference between the magnetic fields generates torsional waves in the magnetostrictive wire. By measuring the propagation time of the torsional waves using the receiver which is disposed on a predetermined portion of the magnetostrictive wire, the physical position of the movable member can be accurately measured. It is preferable that the magnet or magnets be ring-shaped and that the magnetostrictive wire detectors be disposed plurally on the outer circumference of the cylindrical member. That way the magnetostrictive wire position detector assembly can be made to perform multiple functions. That is to say, the magnetostrictive wire position detector assembly will then be able to detect the position of the movable member precisely even if the movable member is inclined within the cylindrical member, and at the same time, the ability to detect the position of the movable member will then not be lost even if one of the magnetostrictive wire detectors malfunctions. In addition, a cylindrical support member should ideally be provided so as to seal closed the outer circumference of the cylindrical member and the magnetostrictive wire detectors with a predetermined spacing. As a result, a heat insulating effect will arise due to the layer of air existing in the space, reducing the radiation of heat from the cylindrical member. It is also preferable that a protective member formed from the same non-magnetic material as the non-magnetic portion of the drive shaft be mounted so as to hermetically seal the magnet or magnets against the non-magnetic portion. This will prevent oxidation of the magnet or magnets, ensuring that the strength of the magnetic field of the magnet or magnets remains constant and stabilizing the precision of the measurements as well as reducing maintenance costs. The cylindrical member may also be the pressure housing of the control rod drive unit, and the movable member may be the drive shaft connected to the control rod of the control rod drive unit, and the position of the drive shaft may be detected. The position within a reactor core of the control rod connected to the drive shaft can be accurately detected along the entire drive length from the fully inserted position to the fully withdrawn position by detecting the position of the drive shaft within the pressure housing. It is also desirable that the construction enable the detection of the control rod insertion times from the commencement of the descent of the control rod corresponding to the detected position of the control rod when the control rod is allowed to descend from the fully withdrawn position by measuring in advance a relationship between times and distances from the fully withdrawn position. By storing the descent times (insertion times) as high-precision digital data by means of the magnetostrictive wire detectors, the times taken by the control rod to reach the dashpot and the fully inserted position from the commencement of descent can be calculated from the data. For that reason, in the rare event that a control rod stops during descent, it will be possible to confirm the rest position of the control rod (fully inserted or partway) from the insertion times of the control rod, ensuring the reliability of the nuclear reactor. |
043127746 | summary | BACKGROUND OF THE INVENTION The disposal of large quantities of toxic materials such as high level radioactive wastes stored in spent reactor fuel storage pools, or generated in the reprocessing of spent nuclear power reactor fuel, or generated in the operation and maintenance of nuclear power plants, is a problem of considerable importance to the utilization of nuclear power. It is generally accepted that the most promising approach is to convert these radioactive wastes to a dry solid form which would render such wastes chemically, thermally and radiolytically stable. The problem of dry solid stability of radioactive wastes is closely related to the safety of human life on earth for a period of more than 20,000 years. For example, radioactive wastes usually contain the isotopes Sr.sup.90, Pu.sup.239, and Cs.sup.137 whose half lives are 28 years, 24,000 years, and 30 years, respectively. These isotopes alone pose a significant threat to life and must be put into a dry, solid form which is stable for thousands of years. The solid radioactive waste form must be able to keep the radioactive isotopes immobilized for this length of time, preferably even in the presence of an aqueous environment. The radioactive wastes are produced in high volumes and contain long-lived, intermediate-lived, and short-lived radioactive ions and some non-radioactive ions. These solutions can be highly corrosive and it is difficult, if not impractical, to reduce them to concentrated forms for further processing or storage. The two most popular types of commercial reactors both of which produce low level wastes are the Boiling Water Reactor (B.W.R.) and the Pressurized Water Reactor (P.W.R.). In a typical Pressurized Water Reactor (P.W.R.), pressurized light water circulates through the reactor core (heat source) to an external heat sink (steam generator). In the steam generator, where primary and secondary fluids are separated by impervious surfaces to prevent contamination, heat is transferred from the pressurized primary coolant to secondary coolant water to form steam for driving turbines to generate electricity. In a typical Boiling Water Reactor (B.W.R.), light water circulates through the reactor core (heat source) where it boils to form steam that passes to an external heat sink (turbine and condenser). In both reactor types, the primary coolant from the heat sink is purified and recycled to the heat source. The primary coolant and dissolved impurities are activated by neutron interactions. Materials enter the primary coolant through corrosion of the fuel elements, reactor vessel, piping, and equipment. Activation of these corrosion products adds radioactive nuclides to the primary coolant. Corrosion inhibitors, such as lithium, are added to the reactor water. A chemical shim, boron, is added to the primary coolant of most P.W.R.'s for reactivity control. These chemicals are activated and add radionuclides to the primary coolant. Fission products diffuse or leak from fuel elements and add nuclides to the primary coolant. Radioactive materials from all these sources are transported around the system and appear in other parts of the plant through leaks and vents as well as in the effluent streams from processes used to treat the primary coolant. Gaseous and liquid radioactive wastes (radwaste) are processed within the plant to reduce the radioactive nuclides that will be released to the atmosphere and to bodies of water under controlled and monitored conditions in accordance with federal regulations. The principal methods or unit operations used in the treatment of liquid radwaste at nuclear power plants are filtration, ion exchange, and evaporation. Liquid radwastes in a P.W.R. are generally segregated into five categories according to their physical and chemical properties as follows: a. Clean Waste includes liquids which are primarily controlled releases and leaks from the primary coolant loop and associated equipment. These are liquids of low solids content which are treated in the reactor coolant treatment system. PA1 b. Dirty or Miscellaneous Waste includes liquids which are collected from the containment building, auxiliary building, and chemical laboratory; regeneration solutions from ion-exchange beds; and solutions of high electrical conductivity and high solids content from miscellaneous sources. PA1 c. Steam Generator Blowdown Waste is condensate from the steam that is removed (blowdown) periodically to prevent excessive solids buildup. PA1 d. Turbine Building Drain Waste is leakage from the secondary system that is collected in the turbine building floor sump. PA1 e. Detergent Waste includes liquids from the laundry, personnel decontamination showers, and equipment decontamination. PA1 a. High-Purity Waste includes liquids of low electrical conductivity (<50 .mu.mho/cm) and low solids content, i.e., reactor coolant water that has leaked from the primary reactor system equipment, the drywell floor drain, condensate demineralizer backwash, and other sources of high-quality water. PA1 b. Low-Purity Waste includes liquids of electrical conductivity in excess of 50 .mu.mho/cm and generally less than 100 .mu.mho/cm; i.e., primarily water from floor drains. PA1 c. Chemical Waste includes solutions of caustic and sulfuric acid which are used to generate ion exchange resins as well as solutions from laboratory drains and equipment decontamination. PA1 d. Detergent Waste includes liquids from the laundry and personnel decontamination showers. PA1 (A) The dead ion exchange resin containing radioactive waste is mixed with cement and cast in forty gallon barrels. PA1 (B) The bottoms from evaporators which contain the radioactive contaminated boric acid and the solutions used to regenerate the ion exchange columns are mixed with cement powder and cast in forty gallon metal or plastic barrels. PA1 (C) The filters containing particulate forms of radioactive waste are usually encased in cement in metal or plastic barrels. PA1 (i) The thermal expansion coefficient is so low that it can only be matched by the core glass when the loading is very low (e.g. less than 5 weight % for the UK composition, see Example 25). PA1 (ii) Because of the high collapsing temperature (about 1300.degree.-1400.degree. C.) it may cause volatilization of Cs and other nuclear wastes. PA1 (1) The preform is immersed in a solution containing the dopant ions at a pH between 9 to 13.5, preferably between 10 and 13, for a time which depends on the wall thickness and the desired concentration of dopants. Typically, the immersion time is between 1 hour and 7 days. The pH of the solution is preferably adjusted with NH.sub.4 OH. For maximum speed of ion exchange, the solution is saturated with the desired dopant ions. Usually the dopants are introduced into the solution as nitrate compounds. However, chlorides and carbonates can be used. Liquid radwastes in a B.W.R. are generally segregated into four categories according to their physical and chemical properties as follows: The liquid radwastes from both types of reactors are highly dilute solutions of radioactive cations, and other dissolved radioactive materials as well as undissolved radioactive particles or finely divided solids. A practical process for disposing of radioactive materials in a dry solids form having high resistance to leaching and other forms of chemical attack would not only be suitable for the disposal of radioactive nuclear wastes, but also for the fabrication of radioactive sources useful in industry, medicine, and in the laboratory. Heretofore, there did not exist any practical, foolproof means for the safe disposal, storage and immobilization of pernicious radioactive waste material. Present day storage containers do not provide sufficient isolation and immobilization of such radioactive material, sufficient long-term resistance to chemical attack by the surroundings, and sufficient stability at high temperature. Currently low level radioactive waste, that is radioactive waste generated at reactor sites, is disposed of in the following manner: These cement barrels are transported to low level radioactive waste sites and buried six feet deep in the ground. At least one of the sites is in the United States Eastern States and exposed to substantial rainfall. In Europe, these barrels are buried at sea. In both cases water will first corrode the metal then the cement and will relatively quickly expose the radioactive ions for leaching into the ground water or sea water. Because the U.S. burials are only a few feet deep, the contaminated water can readily intermix with streams, lakes and rivers, thus, entering the ecosphere. The rationale for this practice is the assumption that upon sufficient dilution the radioactivity becomes harmless. Some of the most serious nuclear wastes are cesium and strontium which are biologically similar to sodium and calcium. They have thirty year half lives indicating that they should be isolated from the ecosphere for at least three hundred years (ten half lives). At Bikini, the experts assumed that dilution had made the island inhabitable after decades in which no atomic explosions were performed, yet when the population was returned to the island its health was deleteriously effected. It has since been realized that plants and animal life biologically reconcentrate these radioactive elements back up to dangerous levels. Thus, the "safe" concentration of radioactive waste must be much lower than accepted values and a more durable substitute for cement is needed. The present invention presents a safe alternative to the cement-solidification of low level waste. Another route heretofore suggested is the so-called dry solids approach which involves the fixation of the waste materials in glasses via mixing with glass-forming compositions and melting to form glasses. This approach offers some improvement regarding isolation and decrease in the rate of release of radioactive elements when the outer envelopes or containers are destroyed. Further, such glasses reamin relatively more stable at high temperatures than plastic and are generally more chemically durable in saline solutions than are metals. Glasses with high chemical durability and low alkali ion conductivity suitable for this prior art technique are formed at very high temperatures, e.g., 1800.degree. C. and higher. Prior processes utilizing such high melting glass-forming compositions are economically unsound and moreover, cause a dangerous problem due to the risk of volatilization of pernicious radioactive materials. Furthermore, this prior procedure is restricted to dry solid radioactive wastes and provides no solution to the high volumes of liquid radioactive wastes produced by the operation and maintenance of nuclear reactors, by the current practice of storing spent fuels in pools of water, and by spent reactor fuel recovery systems. In view of the overall difficulties of handling radioactive material, and especially in view of the danger of volatilization of radioactive material into the atmosphere, attention has been directed to using glass compositions having relatively low melting temperatures, that is to say, using glass compositions with SiO.sub.2 contents as low as 27 weight percent. While the problem of volatilization of radioactive materials is reduced, it is not completely controlled. Moreover, the resultant glass composition exhibits greatly reduced chemical durability and increased ion diffusion rates for the radioactive materials present therein. The greater this diffusion rate, the lower is the ability of the glass to keep the radioactive materials immobilized in its matrix. For long-term containment of radioactive waste, demanded under present day standard, these prior glass compositions are inadequate. U.S. Pat. No. 3,640,888 teaches the production of neutron sources by encapsulating californium-252 in glass using the steps of packing an open-ended vitreous tube with a porous powder of quartz having an organic liquid ion exchange material sorbed thereon, passing an aqueous solution containing californium-252 cations through the powdered quartz, drying and heating the powdered quartz and tube in air to oxidize and volatilize the organic liquid ion exchange material resulting in the non-volatile oxide of californium-252, and then fusing the tip and powder contents to form a vitreous body containing the californium-252 oxide. The patent, however, does not disclose, teach or suggest the use of porous glass as the tube, in packing for the tube or in a stopper for the tube nor does it disclose or suggest any method or technique for concentrating and safely disposing of radioactive wastes as by burying in strong, durable, leach-resistant glass containers. U.S. Pat. No. 1,533,794 teaches the packaging of radium emanations in a glass capillary tube followed by sealing the ends of the tube and thus enclose emanations previously introduced into the tube. There is no teaching, however, of any method for concentrating and encapsulating radwaste. U.S. Pat. Nos. 2,336,227; 2,340,013; 2,522,524; 3,364,148 and 4,073,579 relate to the treatment of porous glass with non-radioactive ions (radioactive ions in the case of 3,364,148) followed by heating to close the pores which contain the ions. U.S. Pat. No. 3,147,225 discloses refractory particles, which contain no or minor amounts of silica and preferably are crystalline, within which particles a specifically selected radioactive cation is firmly fixed for use in self-luminous markers, liquid level indicators and other applications. U.S. Pat. No. 3,116,131 discloses the method of binding expanded silica particles with a binder and shaping and curing into a desired form, followed by impregnation with a solid dessicant, e.g., sodium hydroxide, and followed by impregnation with a radioactive gas and steam to absorb the water vapor followed by capillary condensation thereby entraining the radioactive gas in the pores after which the pores are closed by heating. U.S. Pat. No. 3,959,172 discloses the method of forming and reacting a mixture of silicate or other source of silicon, a radionuclide waste and a metal cation to produce complex metalosilicate crystals which entrap the radionuclide waste. U.S. Pat. Nos. 3,451,940 and 3,849,330 disclose the utilization of a thermite reaction to form a complex polysilicate product containing the radioactive wastes. U.S. Pat. No. 3,167,504 discloses the purification of radioactive waste liquid by absorption on a synthetic zeolite which is then sealed in a suitable container for burial. U.S. Pat. Nos. 3,114,716; 3,262,885; 3,365,578 and 4,020,004 each deal with various techniques involving the preparation of glass-forming mixtures followed by firing to form a glass. U.S. Pat. No. 3,093,593 discloses methods for disposing of radioactive wastes by forming porous ceramic pieces from clays and other silicates followed by prefiring such pieces to destroy ion exchange capacity and thereafter impregnating the prefired pieces with radioactive liquid wastes. The pieces saturated with radioactive waste are then heated to vitrify them and render them non-absorptive. U.S. Pat. No. 3,938,974 relates to glass, optical wave guide fibers and their production. Radioactive materials cannot be used in such fibers because they form color centers which absorb light. Not only does this patent fail to disclose the use of radioactive materials, the presence of such materials are inimical to the express objects of the patent. There is no disclosure or suggestion in any of the above-identified patents of radwaste disposal methods involving the depositing of radwaste solids in a glass container, and heating to drive off non-radioactive volatile materials, e.g., water and decomposition products, while preventing the escape of radioactive materials and to collapse the container and seal it, thereby providing a durable, highly leach-resistant, long-term containment of radioactive solids. As will be apparent hereinafter from the various aspects of applicants' contributions to the art, there are provided novel methods providing novel articles for the containment of pernicious and dangerous radioactive materials over extraordinarily long periods of time. Unlike melting glass containment procedures, the methods of the present invention provide for the control of radioactive materials that are prone to volatilization at high temperatures employed in the containment procedure, thereby providing for elimination of environmental hazards due to the possible escape of volatilized radioactive material in the atmosphere and avoiding the necessity of providing elaborate recapture and/or redisposal procedures and equipment. SUMMARY OF THE INVENTION The invention broadly relates to the concentration and immobilization of toxic solids, such as, mercury, cadmium, tellurium, lead, insecticides and poisons, and especially radioactive materials and the like for extremely long periods of time. The invention more specifically contemplates novel glass articles containing said toxic solids and having high mechanical strength and high chemical durability to aqueous corrosion and having sufficiently low radioisotope diffusion coefficient values to provide protection to the environment from the release of radioactive material such as radioactive isotopes, nuclear waste materials, etc., and which are concentrated, immobilized and encapsulated therein and are suitable for burial underground or at sea. The glass articles are made by depositing the radioactive solids in a glass container followed by heating the container to drive off non-radioactive volatiles and to drive off non-radioactive decomposition products. The glass container may be made of porous glass and may or may not contain a porous or non-porous glass packing which can preferably be particulate or can be relatively large as a single or few glass rods. The glass articles of this invention have a composition characterized by a radiation activity illustratively above one millicurie, preferably greater than one curie, per cubic centimeter of said article. (When highly dilute radwastes are treated pursuant to this invention for the purpose of concentrating and immobilizing the radwaste for storage, the radiation activity of the resulting glass articles may not reach the level of one millicurie per cubic centimeter of the glass article and may remain below 1 microcurie per cc., when it becomes expedient for other reasons to collapse and seal the glass container. In concentrating and immobilizing radioactive materials in dilute radwastes, the glass container can be loaded up to 10 microcuries per cc. or more but usually is loaded up to 1 microcurie per cc. of said glass article). The radioactive material is in the form of radioactive solids that are sealed within the glass container. In one aspect, the amount of radioactive material contained in the glass articles is at least 1 ppb (part per billion based on weight), generally in solid form of at least five and preferably at least ten of the radioactive elements listed hereinafter. Preferably the novel glass articles should contain at least 75 mole percent SiO.sub.2, most preferably greater than 89 mole percent SiO.sub.2. From a practical standpoint, the upper limit of radioactive material contained in the glass articles will be governed, to a degree, by such factors as: the concentration, form and type of radioactive material encapsulated in the glass article, by the volume fraction of pores, if any, in the glass container, by the amount, if any, of glass packing in the glass container, by the various techniques employed to encapsulate the radioactive material in the glass container and other factors. Radioactive materials which can be concentrated, encapsulated and immobilized in the glass container pursuant to this invention include radioactive elements (naturally occurring isotopes and man-made isotopes existing as liquids or solids dissolved or dispersed in liquids or gases), in combined or uncombined form (i.e., as anions, cations, molecular or nonionic, or elemental form) such as rubidium, strontium, the lanthanides, e.g., La, Ce, Pr, Nd, Pm, Sm, Eu, Gd, Tb, Dy, Ho, Er, Tm, Yb, Lu, cobalt, cadmium, silver, zirconium, molybdenum, technetium, niobium, ruthenium, rhodium, palladium, tellurium, cesium, barium, francium, yttrium, radium and acinides, e.g., Ac, Th, Pa, U, Np, Pu, Am, Cm, Bk, Cf, Es cations and elements. Especially suitable in the practice of the invention are radioactive wastes from nuclear reactors, spent reactor fuel reprocessing, spent fuel storage pools or other radioactive waste producing processes. The glass in the final glass product can be vitreous throughout or partly vitreous and partly crystalline. The invention can be practiced in many ways. Illustratively, one facile yet highly effective way is to deposit the radioactive materials, e.g., radioactive nitrates, as a solid in a non-porous glass container, such as a glass tube, having at least one opening, and then followed by heating the container to drive off water and/or other non-radioactive volatiles, if present, and then to collapse the walls of the tube and seal it around the deposited radioactive solids. The heating step can be carried out in such manner that solids, such as nitrates, deposited in the tube decompose to provide non-radioactive gases, such as nitrogen oxides, which are removed from the glass tube before sealing. Alternatively, a non-porous glass container can be closed at one end and at least partially filled with a packing such as porous glass particles, such as, porous glass powder, or tiny glass spheres or silica gel in particulate or other form. The fluid containing radioactive material is then poured into the container to fill the interstices between the glass particles followed by heating to drive off non-radioactive volatiles with or without decomposition of components, such as nitrates, in the fluids and ultimately to seal the glass container around the radioactive solids deposited on the glass particles and in the pores of porous glass particles, if present, contained by the container. In this instance the contained glass particles provide surfaces on which the solids can be deposited and also act to control the volatilization to prevent eruption of fluid out of the tube during the heating step. The porous glass particles provide additional interior surfaces within the pores of the particles for deposit of additional dissolved solids in the fluid as well as external surfaces for deposit of dispersed solids. In another embodiment a non-porous glass container having open upper and lower ends can be filled with porous or non-porous glass particles which are held in the glass container by means of a porous structure, such as glass wool or a porous glass disc or rod in the lower portion of the container to support the glass particles in the container. The fluid containing dissolved and/or dispersed radioactive solids is then poured into the upper or lower end of the container and passes through the bed of glass particles which act as a filter to remove dispersed radioactive solids from the fluid. The glass particle bed can contain glass particles having silicon-bonded cation exchange groups, such as, alkali metal oxide or ammonium oxide groups as described in our patent application, entitled: "Fixation By Ion Exchange Of Toxic Materials In A Glass Matrix", U.S. Ser. No. 959,222, filed Nov. 9, 1978, now abandoned, the disclosure of which is incorporated herein by reference. The porous cation exchange glass particles remove dissolved radioactive cations from the fluid. The fluid can be passed through one or more such beds using conventional techniques for multibed filtration and/or ion exchange until the fluid has been cleansed of radioactivity to the desired level. When the filtration-ion exchange glass particles become loaded or when, for some other reason, it is no longer desired to further utilize them, the beds and the container containing them can be heated to drive off water and/or other non-radioactive volatiles or gases such as decomposition products, e.g., nitrogen oxides, and then to collapse the pores of the porous glass particles containing the radioactive cations, to fuse the glass particles together thus entrapping radioactive solids and/or cations deposited on the inner and outer surfaces of the particles, and then to collapse the glass container and seal it around all of its contents to encapsulate the entire mass into a substantially solid leach-resistant structure suitable for long-term storage. In still another embodiment the glass container itself can be made of porous glass and the radioactive fluid is introduced into the interior of the container and caused to permeate through the pores of the glass from the interior walls to the outer walls of the glass container. The insoluble radioactive solids originally dispersed in the fluid are deposited on the interior wall of the container and the dissolved radioactive solids are disposed in the pores of the glass container where they can be deposited by various techniques, such as those taught in U.S. Pat. No. 4,110,096. The glass container can then be heated to drive off volatiles as described above, to collapse the pores of the glass container and ultimately to collapse the glass container and seal it thereby encapsulating the radioactive solids within the glass structure. Prior to heating, the outer wall surfaces of the container can be washed to remove deposited radioactive solids from the outer surface layer of the glass container so that ultimately a radioactive-free outer clad is provided after heating to collapse the pores and the container. The non-porous glass compositions when used herein for the glass container and/or for the glass packing within the container are of any suitable type, but preferably are strong, durable, leach-resistant and chemical-resistant. Any glass composition having these properties can be used such as high silica glasses, for example, Vycor and Pyrex. Suitable glasses contain at least about 70%, preferably at least about 80%, most preferably at least about 93% silica. The literature adequately describes the preparation of the porous silicate glass compositions. Suitable glass compositions which may be utilized as porous glass compositions in the novel methods generally contain SiO.sub.2 as a major component, have a large surface area. In the practice of various embodiments of the invention the SiO.sub.2 content of the porous glass or silica gel desirably is at least about 75 mole percent SiO.sub.2, preferably at least about 82 mole percent SiO.sub.2, and most preferably at least about 89 mole percent SiO.sub.2. Such glasses are described in the literature, see U.S. Pat. Nos. 2,106,744; 2,215,036; 2,221,709; 2,272,342; 2,326,059; 2,336,227; 2,340,013; 4,110,093 and 4,110,096, for example. The disclosures of the last two mentioned patents are incorporated herein by reference. The porous silicate glass compositions can also be prepared in the manner described in U.S. Pat. No. 3,147,225 by forming silicate glass frit particles, dropping them through a radiant heating zone wherein they become fluid while free falling and assume a generally spherical shape due to surface tension forces and thereafter cooling them to retain their glassy nature and spherical shape. In general, the porous silicate glass can be made by melting an alkali-borosilicate glass, phase-separating it into two interconnected glass phases and leaching one of the phases, i.e., the boron oxide and alkali metal oxide phase, to leave behind a porous skeleton comprised mainly of the remaining high silicate glass phase. The principal property of the porous glass is that when formed it contains a large inner surface area covered by silicon-bonded hydroxyl groups. We prefer to use porous glass made by phase-separation and leaching because it can be made with a high surface area per unit volume and has small pore sizes to give a high concentration of silicon-bonded hydroxyl surface groups, and because the process of leaching to form the pores leaves residues of hydrolyzed silica groups in the pores thus increasing the number of silicon-bonded hydroxyl surface groups present. The porous silicate glass when used as packing may be in the form of powder as for use in chromatography columns or in a predetermined shape such as plates, spheres or cylinders. It is preferable to utilize a glass composition in the container which will produce a clad or envelope that is low in leachable components, such as, alkali metals or boron. In the event that this is not possible or practical it is preferred to then insert the glass container before or after collapse into a second glass container which has a composition containing no or low amounts of alkali metals or boron or other leachable components. The absence of substantial amounts of leachable components in the outer glass clad obviously provides a more durable and leach-resistant product which is more suitable for burial. In those instances where a glass packing is used in the container, it is preferred that the composition thereof contain little or no leachable components, especially when the resulting products are intended for burial. It is most preferred, that very high silica glasses are employed in both the glass container and the glass packing. When it is desired to avoid cracking of the glass container in the case where a glass packing, such as glass particles, spheres or a glass rod, are disposed within the glass container, it is preferred to utilize, as the container glass, a glass, which, after the deposition step, has a glass transition temperature of up to 100.degree. C. higher than the glass transition temperature of the glass formed from the glass packing and solids deposited in and on the glass packing. It is also preferred for this purpose to utilize as the glass container a glass which, after the deposition step, has a thermal expansion coefficient which is up to about two times 10.sup.-6 per degree Centigrade less than the thermal expansion coefficient of the glass resulting from sintering of the glass packing and solids deposited in or on said glass packing. In determining, the glass transition temperatures and thermal expansion coefficients, the amount and type of solids deposited in the pores of the glass container, when a porous one is used, and of solids deposited in the pores and on the outer surfaces of porous glass packing when used and of solids deposited on the outer surfaces of non-porous glass packing, when used, can have a considerable effect on the glass transition temperatures and thermal expansion coefficients and should be taken into consideration. It is also preferred to regulate the cooling of the composite glass container and contents, resulting from the depositions and sintering step, such that the rate of cooling is as nearly the same as possible throughout the composite glass container and contents. While cracking has been observed in certain instances, it has not prevented the achievement of the objects of this invention, namely, the immobilization and isolation from the environment of radioactive solids from radioactive wastes containing such solids in dissolved and undissolved form. DETAILED DESCRIPTION OF THE INVENTION In one method of the present invention, a radioactive material is deposited as a solid in a hollow glass container having at least one opening. The radioactive material is deposited from a fluid which passes continuously through the glass container or which is placed batch-wise into the container. The fluid may contain dissolved radioactive materials, particulate radioactive materials, or both types of radioactive materials. The fluid may either be a gas or a liquid or both. The radioactive materials, whether particulate or dissolved in the fluid to be treated, can be deposited on a non-porous glass or on a porous glass having an interconnected porous structure. When a porous glass is used, the pores are usually smaller than particulate radioactive materials in the fluid thereby preventing passage of the particulates into the interconnected pores. In this case, the particulates are deposited on the inner wall surfaces of the porous glass container. Radioactive materials which are dissolved in the fluid or which are gaseous radioactive materials, pass into the pores of the glass and are entrapped within the porous structure by either reacting with the glass, undergoing a cation exchange reaction with the glass, or by precipitating within the pores of the glass. Whether a porous glass or a non-porous glass is used, after the radioactive material is deposited in the hollow glass container, the deposited radioactive materials are sealed within the glass matrix by collapsing the walls of the container. Collapse of the walls of the container is achieved by heating the container while: (a) applying a vacuum to the inside of the container, (b) applying external pressure to the container, for example by placing a weight on the container or by increasing the gas pressure outside of the container, and (c) by combinations of methods (a) and (b). Where radioactive material is deposited in the pores of a porous glass, the container is heated to collapse the pores prior to collapse of the walls of the container. The glass container is hollow and has at least one opening. The most preferred container for processing liquids is one in a shape of a test tube. Where the container has more than one opening, as in the case of a hollow glass rod (or a tube) one or more of the openings may be plugged with a glass stopper to prevent the fluid from escaping during filling in a batch operation. For continuous operation, a glass tube having an opening at each end is preferred. In the latter case, both openings may be plugged with a porous glass stopper. The greater the volume of the fluid to be treated, the more preferable the continuous operation becomes. Examples of other configurations of the glass container which are suitable for the purpose of the present invention are U-shaped, beaker-shaped, box-shaped, etc. The simplest embodiment of the present invention merely involves deposition of the radioactive materials in a non-porous glass container followed by collapse of the walls of the container and burying the resulting glass article underground or at sea. For example, the glass container can be test-tube shaped and made of a non-porous glass such as a Vycor glass (trademark for a heat and chemical resistant, low thermal expansivity glass of Corning Glass Works). In another embodiment of the present invention, the radioactive material is deposited in a porous glass container. In still another embodiment, the radioactive material can be deposited in a non-porous glass container having a second glass, e.g., a glass packing, disposed within the container. In yet another embodiment of the present invention, the radioactive material is deposited in a porous glass container having a second glass or glass packing disposed within it. In the last two mentioned embodiments, the second glass may be a non-porous glass or a porous glass. The second glass, whether porous or non-porous, may be a glass preform of any suitable shape (e.g., rod shaped, rectangular shaped, particulate, spheroid, etc.) for fitting into and at least partially filling the glass container. The second glass, however, is preferably in the form of particles such as spheres. A preferred embodiment of the present invention utilizes a non-porous glass tube having porous glass particles therein. Nuclear Waste in a Non-Porous Tube A non-porous hollow glass container made of Vycor or silicate glass is at least partially filled with a fluid containing radioactive materials. The container preferably has one opening therein which is plugged with a porous-glass plug. In the case where the fluid is a liquid, e.g., water, the glass container is then heated to evaporate the fluid to dryness so as to precipitate the radioactive materials on the inside walls of the glass container. Temperatures slightly above the boiling point to about 50.degree. C. above the boiling point of the fluid can be used. Lower temperatures can be used to dry the fluid when a vacuum is applied to the interior of the glass container. The glass container is further heated and at about 400.degree. C. the radioactive salts originally in the nuclear waste, e.g., the radioactive metal nitrates decompose or are calcined to form the corresponding oxides, e.g., the radioactive metal oxides. The non-radioactive gaseous decomposition products, e.g., nitrogen oxides are driven off by the heating and the porous plug acts as a barrier to keep the nuclear waste from leaving the glass container. The glass container is further heated until it collapses to trap the precipitated, crystalline nuclear waste within the sealed container. Before adding the fluid to the glass container, silica and alumina can be added to the fluid to create a calcined material upon heating. Calcining of nuclear waste materials in metallic containers is well known. The procedures and operating conditions utilized in calcining in a metallic container are also applicable when calcining in the glass containers of the present invention and such teachings are incorporated herein by reference. The glass tube typically collapses at around 1300.degree. C. Further details on the drying procedure, the trapping of radioactive decomposition products, and collapsing of the glass container are presented below. Nuclear Waste in a Porous Glass Container Fabrication of the porous glass container used in the process of the present invention may follow any of the available methods used by one practiced in the art to form porous glass in any desired shape, such as cylindrical or rectangular. It preferably has a composition containing more than 75% silica. We prefer to form the porous glass according to the methods disclosed in U.S. Pat. No. 4,110,096, the disclosure of which is incorporated herein by reference. For example, a glass composition containing silica, boron trioxide and two alkali metal oxides (such as Na.sub.2 O and K.sub.2 O) is melted and drawn into long rods or tubes. By suitable heat-treatment, these rods or tubes are phase-separted into two phases; one phase, a silica-rich phase containing also small amounts of B.sub.2 O.sub.3 and alkali metal oxide and a silica-poor phase which contains greater amounts of B.sub.2 O.sub.3 and alkali metal oxide. The heat-treated rods or tubes are then immersed in a suitable leaching solution in order to dissolve and remove the phase containing the lower silica concentration. Removal of this phase and subsequent washing yields a porous glass characterized by a SiO.sub.2 content greater than 90 mole percent which is ready for use as the glass container for the encapsulation of the dissolved or gaseous radioactive material pursuant to this invention. The invention is further described with reference to a hollow porous glass container which is test-tube shaped. A solution containing dissolved radioactive materials and particulate radioactive materials such as metallic precipitates of the platinum metal family which are typically present in nuclear waste solutions from spent nuclear fuel reprocessing stations is poured into a porous glass test tube. The solution impregnates the walls of the tube and in this way disposes dissolved radioactive material in the form of dopant within the walls of the porous glass test tube. On the other hand, the particulate radioactive material because of their particle size do not go into the walls of the glass container but instead are deposited on the inner wall of the tube by settling, or filtration. Deposition upon the inner walls of the tube occurs by filtration and by subsequent evaporation of the fluid. The radioactive material that was originally in solution on the other hand, is disposed inside the pores of the glass tube in solution, as a nitrate for example. The opening in the tube then can preferably be plugged with a porous or non-porous plug. Then, the dissolved radioactive material is deposited in the pores by precipitation out of solution by methods such as those disclosed in U.S. Pat. No. 4,110,096. Thus, the precipitation may be caused by cooling the glass container (thermal precipitation), by chemical precipitation and combinations thereof. Chemical precipitation includes precipitation by the common ion effect to reduce solubilities, and cause precipitation, of the dissolved radioactive materials. It also includes the exchange-of-solvents technique for reducing solubilities to cause precipitation. In this method, the porous glass test tube can be immersed in a solvent in which the soluble radioactive material in the pores is less soluble. The addition of a suitable precipitant which reacts with the dopant, or dissolved radioactive materials, in the pores or causes a suitable change in pH, is also a means of chemical precipitation. Precipitation can also be changed by evaporation of the fluid from the pores, preferably under vacuum and at temperatures around room temperature or below. Precipitation methods other than that involving the evaporation of solvent as the sole means of precipitation are used when it is desired to obtain higher strength glasses consistently. Combinations of precipitation techniques can be used. A preferred combination of precipitation techniques is thermal precipitation and precipitation by exchange of solvents. Deposition of the dissolved radioactive materials within the pores can also be accomplished by a cation exchange reaction with the glass. Manufacture of suitable glasses for the cation exchange reaction together with a detailed description of the cation exchange process are disclosed in the above-identified concurrently filed application entitled: "Fixation by Ion Exchange of Toxic Materials in a Glass Matrix", herein incorporated by reference. A porous glass container having cation exchange capabilities is particularly suited to a continuous process. The solution is passed into the interior of the container, through the porous walls for the cation exchange and entrapment of the dissolved radioactive material, and then the remainder of the solution, i.e., the solvent such as water, passes out through the exterior wall of the container. Subsequent to the deposition step, the outer wall of the porous glass tube can be washed so that the precipitated radioactive material disposed within the pores of the outer surface layers of the porous tube is removed. The washing step is preferred when it is desired to produce a radioactive article free of, or has a lower amount of, radioactive material in its outer surface layers and is not essential in the broad sense of this invention. The solvent of the solution is then removed preferably without migration of the radioactive material within the pores. This can be accomplished by the drying procedure described in U.S. Pat. No. 4,110,096. Typically, the porous glass container is placed in a drying oven and heated to an upper drying temperature under a vacuum at a rate less than 100.degree. C./hour. Preferably, the heating rate is below 30.degree. C./hour because a slow heating rate significantly lowers the cracking probability and avoids redistribution of chemically unbound dopant, although cracking of the glass tube or dopant redistribution is not intolerable in all cases. However, in some circumstances it is desirable to use a higher heating rate to increase throughput of articles through the processing system. A preferred heating rate is about 15.degree. C. per hour. After the tube is dried, two forms of deposited radioactive materials are obtained: (1) the originally undissolved particulates which are disposed in the interior space defined by the inner walls of glass tube, and (2) the originally dissolved materials which are deposited in the pores of the walls of the glass tube. Following the drying step, continued heating of the glass container causes decomposition of the radioactive materials both deposited in the pores and outside of the pores. For example, the radioactive material goes from its nitrate form (or whatever its original form was) to its oxide or phosphate or silicate form with nitrogen oxide decomposition products being driven off. If it is desired to encapsulate radioactive gases, e.g., krypton or iodine, these can be introduced into the pores of the porous glass at a temperature typically chosen to be in the range between 50.degree. to 150.degree. C. below the glass transition temperature, T.sub.g, of the glass tube including the deposited materials. A narrower range from 75.degree. to 125.degree. C. below the glass transition temperature is preferred. The heating is continued until the pores of the porous glass collapse. Upon collapse of the pores, the radioactive material which was deposited from solution and including radioactive gases are totally trapped within the matrix of the glass. It is chemically bonded to the glass and physically enclosed within the glass. The heating can be continued without great risk of losing the radioactive material within the glass by vaporization because it is now buried inside the glass itself. While continuing to heat the tube, a small imposed pressure difference between the inside of the tube and the outside of the tube can be used to collapse the tube. The pressure inside the tube is made a little lower than the pressure outside of the tube (by means of a vacuum) for progressively collapsing the tube into a rod. External pressure from a weight placed upon the tube for example, can also be used to collapse the walls of the container. Upon collapse of the tube (or other glass container) the radioactive material which was originally particulate in nature is trapped inside the resulting sealed glass vessel. The originally undissolved particulate radioactive material which is now trapped within the glass vessel can be in the elemental form of a metal, a metal salt, metal oxide or other metal forms. The particulate radioactive material which decomposes as described below, would be in the oxide, phosphate, or silicate form. The collapsed tube is in rod form wherein two kinds of radioactive material are trapped: (1) one which was originally soluble in the solution and which eventually became chemically attached to or entrapped within the glass, and (2) an insoluble, solid radioactive material which eventually became encapsulated by the glass. Thus, upon collapse of the hollow glass article, a glass article is obtained which comprises originally undissolved solid radioactive materials entrapped in its core. The originally dissolved radioactive materials are entrapped and immobilized in the glass matrix surrounding the core. Utilization of a hollow, porous glass container, particularly one which is tube-shaped, permits several advantages over use of a porous glass rod. For example, one advantage of a porous glass test tube-shaped container is that two surfaces (an inner wall and an outer wall surface) are utilized. The solution which contains the nuclear waste is placed inside the tube to trap the particulate radioactive materials and permeate the pores with the solution containing dissolved radioactive materials. A solution which will cause dissolved radioactive material to precipitate within the pores can be introduced through the outside surface of the porous tube. For example, potassium silicate reacts with many of the nuclear wastes (e.g., iron). The high pH caused by the potassium silicate causes precipitation (e.g., iron oxide) within the pores. In the case of ruthenium, ruthenium silicate precipitates and so forth. By bringing in the material from the outer surface of the porous tube, control of the precipitation of the nuclear waste inside the pores of the porous glass is enhanced. A further advantage of the tube configuration is that dissolved radioactive materials which do not precipitate during the precipitation step can be sucked to the inner wall of the tube. Thus, the tube is filled completely with the fluid containing the radioactive material. It is then placed in clean water or a second solution and a vacuum is applied to the inside of the tube. In so doing, some dissolved radioactive material may not be precipitated by a decrease in temperature, by insolubility in the second solvent or by chemical activity with the second solvent. The dissolved radioactive material which does not precipitate for any of these reasons is sucked to the inner wall of the tube, with the outer wall staying clean. This flow to the inner wall causes a distribution of the dopant, i.e., originally dissolved radioactive material, which is higher on the inside surface of the tube. When the tube is finally collapsed into a rod this region of high concentration of radioactive materials is entrapped into the total glass system. A high concentration of nuclear waste on the outside of the glass article can be thereby avoided. Another advantage of the tube configuration is the entrapment of any radioactive gases produced by decomposition of the radioactive materials during the drying step. For example, if ruthenium tetraoxide is produced from the radioactive materials within the pores: it can be aspirated from the center of the tube, out of the drying/collapsing furnace and into another tube of porous glass which is at a lower temperature. The radioactive gases are filtered by the second porous glass tube. The fumes react with the silicate of the glass and precipitate within the pores of the glass. For example, rubidium will be reduced from the tetraoxide to a lower oxidation state and precipitate within the pores. The second tube serves as a micropore filter. Its pores are then collapsed followed by the collapse of the walls of the tube to thereby entrap and immobilize the gaseous radioactive materials in a sealed glass matrix. As can be seen, the porous glass tubes can be used to absorb radioactive gases from nuclear waste disposal systems which do not utilize porous glass containers. Thus, radioactive fumes from other nuclear waste processing systems can be treated in a porous glass filter. The radioactive fumes may contain particulate radioactive materials and gaseous radioactive materials. Non-radioactive materials, e.g., water, etc. will pass right through the porous glass filters whereas the nuclear waste, for example, ruthenium tetraoxide or cesium would be trapped inside the porous glass filter. When the filters begin to lose efficiency, the filter itself is heated to collapse the pores and the walls of the porous glass filter. Both the particulate radioactive materials which were suspended in the gas and the gaseous radioactive materials are thus entrapped and immobilized in the glass matrix. When using the porous glass container as a filter, it need not be in the test tube configuration. Flat configurations, for example, are equally as good, or even better in some cases, than the test tube configuration. Another advantage of using a porous glass container (as opposed to a porous glass rod) is that processing time is reduced for a given thickness of porous glass. For example, the time required to fully impregnate a porous glass rod with a 1" radius is the same as that required to fully impregnate a porous glass container whose inner radius is 1" and outer radius is 3". The wall of the glass container is 2" thick but impregnation occurs from both the inside and outside surface, to a depth of 1". In the case of the glass rod, the cross sectional area of the glass is only .pi., or about 3.14, square inches. However, for the glass container, the cross sectional area of the porous glass is 8 .pi., or approximately 25.12 square inches or, 8 times greater than that of the rod. Thus, in the same amount of time, a much greater amount of glass can be impregnated. This advantage also comes into play during the drying step, and during decomposition. For comparison purposes, it is assumed that there is zero shrinkage for the porous glass. When the glass container is collapsed, the final diameter of the resulting rod is approximately 2.8". On the other hand, the final diameter of the glass rod which had a 1" radius is approximately 2". Thus, a larger end product containing more radwaste is obtained in the same amount of processing time when the glass container is used. Particulate Glass Inside Glass Container A particulate glass which is either a non-porous glass or a porous glass having an interconnected pore structure or mixture of both can be disposed within the glass container. Formation of non-porous particles is well known in the art. Non-porous glass particles of either conventional compositions or modifications thereof can be used. Porous glass particles can be made from glasses having compositions similar to those used to make the porous glass container. For example, a typical borosilicate glass having approximately 7% alkali, for example, sodium oxide or potassium oxide, and having a small amount of alumina is suitable. A content of B.sub.2 O.sub.3 on the order of thirty to forty weight percent and the remainder being silica is a satisfactory composition for forming the porous glass particles. The porous glass is preferably formed according to the compositions and methods disclosed in U.S. Pat. No. 4,110,096. The glass is melted in a crucible at around 1400.degree. C., stirred, and fined. The crucible can be made of platinum if the process is performed in the laboratory. However, on a commercial or production scale a ceramic crucible can be used. To make a powdered glass, the molten glass can be poured directly into cold water to crack it and break it into little pieces. Alternatively, the molten glass can be pulled into rods or cast into any desired shape. In the latter situation, the glass is broken in a milling machine. The glass pieces are sieved to provide glass particles of the desired particle size. The sieved glass particles can then be passed through a flame to form them into little spheres. The advantage of making spheres instead of just using the sieved glass particles which have random irregular shapes is that it provides a more uniform and greater density of packing in the glass container. Thus, two kinds of glass particles can be produced. One is just broken and sieved and therefore has elongated grains of random irregular shapes. The other is broken, sieved, and remelted by going through a hot zone and then rapidly cooled to produce glass particles in the shape of tiny spheres. In order to obtain the elongated particles it is preferable to reheat the cast or drawn glass before milling. In this manner, larger pieces of glass, which are easier to handle during milling are obtained. The particles thus obtained are non-porous and can be used in the embodiment wherein the radioactive material is deposited on non-porous glass particles within the glass container. To make the non-porous glass particles porous, they are reheated at approximately 550.degree. C. for about two hours. They are thus phase-separated and then are leached at approximately 50.degree.-100.degree. C. typically in 3N HCl, or some other mineral acid, such as sulfuric or nitric. Further details on the leaching process are disclosed in the above-mentioned U.S. Pat. No. 4,110,096. The leaching time depends upon the grain size and temperature. Typically about two hours, more or less, are sufficient for the leaching operation. The higher the temperature the faster the leaching occurs. Next, the leaching particles are washed with water. The finished product is a wet, particulate porous glass with pores interconnected throughout each particle which can be used in the process of the present invention. However, by heating the porous glass in excess of 100.degree. C. (typically 150.degree.-100.degree. C.), the water is removed and a dry particulate porous glass is obtained. As the particles dry, they do not stick to each other and it is easy to tell when all water has been removed because a very loose, more flowable, powdery product resembling talcum powder is obtained. If one uses elongated grains, the packing of the glass container may not be very efficient, typically 60% may be voids and the particles (or grains) may represent only 40% of the interior volume of the glass container. On the other hand, if one uses round spheres they tend to pack better and one can get packing rates of 60% or more. The ideal for packing of the spheres would be close to a packing rate of 80%. Improvement in packing can be obtained by applying pressure. The more pressure applied to the glass particles, the better the packing that results. Pressure can be applied at lower (cold) temperatures or it can be applied at hot temperatures. Either the hot or cold pressure approach will provide much better packing ratios. Grain size depends very much on the amount and size of the undissolved particulates in the nuclear waste solution. It is generally preferred to have as large glass particles (grains) as possible, e.g., having diameters of 4 mm or more, because, when the gases from the nitrate or other salt decomposition and from the vaporization of water are coming out, the larger, heavier glass particles will not tend to be blown out of the container. On the other hand, if the amounts of undissolved nuclear waste particles are large, a larger glass area to interreact with and assimilate the undissolved nuclear waste particles is desirable. Under those conditions it is preferable to use glass having a particle size of about 50 microns diameter. They can even be as small as one micron or so by using the fines rather than the sieved fractions. The exact size depends on the exact size of the nuclear waste particles to be encapsulated and empirical methods can be used to good advantage to determine the optimum glass particle size for a given nuclear waste solution. Generally, glass particle sizes ranging from 5 micron to 5 mm, preferably 50 micron to 1 mm, can be used. The loading of the glass container can be accomplished in any suitable manner desired other than those disclosed hereinabove. For example, the high level (i.e., relatively concentrated) liquid radwaste can be introduced into the glass container, which is closed at one end, is a long tube and is filled with particulate porous glass, through a smaller inner filler tube positioned inside of the glass container. As the high level liquid radwaste is poured through the filler tube into the container, the filler tube is moved vertically so that the level of the liquid radwaste in the glass container roughly corresponds to the level of the lower end of the filler tube and as more liquid is introduced, the filler tube progressively moves upwardly. This has the advantage that whatever solids are suspended or slurried in the nuclear radwaste reach the lower portions of the particulate glass packing in the glass container. These solids are introduced at all vertical levels in the porous particulate glass and there results a relatively uniform vertical distribution of these solids throughout the contained particulate glass packing. The nuclear radwaste is shut off before the filler tube reaches the top so that there is left on top a short section of porous glass packing relatively free of radwaste. This can also be done by shutting off the nuclear radwaste, heating the assembly to drive off the water and then putting clean glass particles on the top. There are several approaches that can be used in the drying step. First, some glass wool or a porous glass disc or some other kind of porous top can be placed on top of the glass particles to prevent them from moving vertically when gases are driven off from the contained radwaste. Also, an adequate space can be left at the top of the container so that the glass particles can move up as the gases are driven off and then eventually down again after gas flow has stopped. It is also better to dry the container and glass particles by having a relatively small heat zone that is brought downwards from the top to progressively drive off the gases from top to bottom. Otherwise if heat is very deep or applied at the bottom boiling can occur inside the container near the bottom which can cause the glass particles to be blown out at the top. Also, as the vapors rise in the container it is preferable to keep everything above the point at which evaporation is taking place hot enough so that the water will not condense again in the upper regions of the container. Thus, the upper part of the container should be kept above 100.degree. C. (typically 110.degree.-120.degree. C. or 150.degree. C.). This procedure can go relatively fast so that by heating from the top, and bringing it down, all the water in the container can be eventually evaporated. Then, when a layer of porous glass particles at the top is dried, it is kept in the temperature range of 100.degree.-150.degree. C. to prevent the escape of other poisonous gases through it. Non-radioactive nitrate decomposition vapors in the container can escape through the dry porous layer while the cesium and the sodium and other radioactive isotopes such as cadmium are caught in this porous sieve. Once the water has been vaporized from the whole column, it can be heated to a temperature of the order of 400.degree. C. fast enough to prevent the distillation of radioactive nitrates. At temperatures of this order, decomposition starts and nitrous oxide fumes are driven off. Again, ruthenium tetraoxide can be a problem for it must be kept from escaping the top of the container by keeping the porous glass particle layer in the top hot enough so that steam will escape but cold enough that RuO.sub.4 will stay down in the container while the nitrate decomposition is going on. As long as the nitrate decomposition fumes keep coming off, the material will be under high oxidizing conditions and there is not much chance that ruthenium will be reduced to lower, less volatile, oxidation states. Once the decomposition of the nitrates is complete, a vacuum can be applied to the interior of the container while maintaining the elevated temperature, thus reducing the vapor pressure of oxygen low enough so that ruthenium tetraoxide spontaneously decomposes to lower oxidation states which have a high temperature characterization or a very low vapor pressure, thereby permanently trapping the ruthenium in the glass. The vacuum should be applied before the porous glass actually starts to close its pores under heat because under such conditions one can also reduce the amount of dissolved gases in the final product. In effect, by reducing the amount of gases inside of each glass pore, the amount of soluble gases in the glass is lowered. Thereafter, the temperature is raised and whenever there is a pressure jump vacuum is applied until the pressure comes down again quietly. Around 1300.degree. C., the exact temperature depends on furnace configuration, container bore size, the type of particulate glass, etc., the tube collapses. If the walls of the container are thin, they will collapse to a flat or elliptical cross section forming more of a ribbon than a rod. If the walls are thick, a rod-like cross section is emphasized. Another way of favoring the rod-like cross section is to pull on the container while heating so that it stretches while it collapses. For convenience in packing the finished glass articles, it may be easier to pack a rod-like cross section for storage. If there is a major heat transfer problem, however, it may be more convenient to work with flat ribbon-like cross sections in order to alleviate such heat transfer problems. By using a narrow heat zone and moving it up from the bottom, a region near the top is reached where there are no glass particles left and the glass container collapses on itself to provide further improved sealing of the nuclear waste. If the degasing is performed properly, there will be only a very minute amount of bubbles and there results a finished product which has an envelope of low temperature expansion coefficient, radiation free glass enclosing a high temperature expansion coefficient glass. This provides compression on the outer glass layers and tension on the inner core of glass. If the inside glass is relatively free of bubbles, it will support the tension and make the final article a strong prestressed material having a modulus of rupture considerably in excess of annealed glass. The advantages of keeping the finished product monolithic are the following: (1) the outer surface area of monolithic glass is much smaller than if it is discontinuous and, since the amount of leaching is proportionate to the surface area, the risk of leaching is reduced considerably, (2) in the case where there is no nuclear waste in the outer layers of the container, there is no nuclear waste available to be leached in the initial period of leaching conditions until, if ever, the leaching is able to continue through the thickness of the radwaste-free outer layers of the collapsed container. This can be designed to be a long period compared to the short half-life of the radioactive isotopes encapsulated inside the container thus encapsulating them for the life of their radioactivity and no radioactivity is exposed to the biosphere. Furthermore, the processing of nuclear waste according to this invention has the advantages that it utilizes no furnace electrode which can be corroded by the molten glass, no fumes of radioactive elements are expelled, and in general a very clean operation is possible. In the event that a glass container breaks, the glass can be disposed of by comminuting it into particles or by remelting it and processing it into particles as explained above and disposing said particles into another glass container. Thus, no new waste is produced requiring a separate disposal system. A monolithic (not particulate) porous glass rod or similar preform containing a radwaste solution tends to break when heated because internal pressures build up because of the boiling away of internal water. If violent enough the internal pressures can become great enough to cause the glass preform to break. Also, after the bulk of the liquid has been removed, as the preform dehydrates, it shrinks and, if the dehydration has been uneven, unequal stresses are developed when one side has shrunk more than the other which can cause the preform to break. Furthermore, at the slightly higher temperatures used to decompose the salts, such as nitrates, gases, e.g., nitrous oxides are given off. Again, too fast an evolution of such gases can break the preform. In addition, if the material is deposited unevenly in the monolithic porous glass rod or similar preform the dopant increases the thermal expansion coefficient of the silica component of the glass preform and, upon collapsing of the pores by heating, the uneven expansion coefficient can lead to breakage. The dopant distribution profile in the monolithic glass preform has to be very well controlled in order to avoid breakage. These problems are greatly reduced or eliminated when using porous glass particles in a glass container. The individual particles are so small that the stresses built up in them during heating are not great enough to break them and if a few do break it causes little or no problem and the heating can be accomplished much faster. Also, the cross sectional dopant distribution profile in the final product can be important. In the case of a non-porous glass container the outer layers of the final product will have the initial thermal expansion coefficient of the container which can be made with a lower thermal expansion coefficient than the glass particles inside. Thus, the final product, in this case, exhibits compression at the surface which makes it stronger. Moreover, the use of a glass container containing porous and/or non-porous glass particles has the further advantage of providing distribution of the deposited radioactive solids throughout the interior of the tube rather than just on the interior surface walls of the container in the case of a glass container in which no glass packing is used or on the exterior surface of a porous glass rod when that is used. Also, when a glass particle-filled, non-porous glass container is employed, the resulting clad is free of radioactivity thus providing essentially no radiation contamination risk to the environment. In the processing of glass containers pursuant to this invention, the only way that gases can escape from the container is through the open end. A convenient way to control these gases is to insert a layer of porous glass in the open end of the container. It will act as a molecular sieve and because of its very large initial surface area, e.g., hundreds of square meters per gram, the gases attempting to pass out of the container are trapped by it. By controlling the temperature of the porous glass layer, the passage of water, non-radioactive nitrate decomposition products, and other non-radioactive fumes that it is desirable to get rid off, can be permitted while at the same time trapping in the container the ruthenium, cesium, cadmium, and other radioactive materials. The differences in temperature along the container can be used to advantage in driving off the non-radioactive volatiles while preventing escape of radioactive materials. It is also advantageous to be able to collapse the container into a rod of smaller dimensions or into a tape having one small dimension, i.e., its thickness, and a larger width. The smaller dimension facilitates more uniform heat removal, i.e., it reduces the temperature gradients in the resulting glass and avoids or considerably reduces cracking. Non-Porous Glass Packing The glass container can be packed with non-porous glass particles in addition to or in place of the above-mentioned porous glass particles. The non-porous glass particles can be made from any suitable glass forming composition using the operational procedures described hereinabove relative to the porous glass particles except, of course, the phase-separating and acid-leaching steps are not necessary in the case of non-porous glass. Non-porous glass particles thus can be in the form of spheres, elongated grains or any other suitable shapes and function in the glass container in essentially the same way as the porous glass particles except that there are no pores into which dissolved radioactive materials permeate. Therefore, the radioactive materials, both dissolved and undissolved, are deposited on the peripheral or outer surfaces of the particles and in the subsequent heating step the oxide forms of the radioactive material react with the molten non-porous glass particles and become an integral part of the final glass product while other forms are entrapped deep within the final glass product. In many cases, it is preferred to use a moving heat zone with a differential pressure produced by evacuating the interior of the container or by applying greater pressure externally as by mechanical means or by gaseous means. It can also be advantageous in using a moving heating zone to progressively collapse the container from the bottom up. If the container is very long it may not be able to support its weight if supported only at its upper regions and can be supported also at the bottom so that it will not elongate during collapsing. On the other hand if it is desired to stretch the tube so that it will collapse into a rod rather than a flat slab, a small pulling force (in addition to gravity) can be applied from the bottom of the container and will produce a rod-shaped object. In order to prevent cracking of the glass container enclosing glass packing, the glass container must have a lower thermal expansion coefficient than the resulting enclosed glass which is obtained when the glass container and contents containing the deposited radioactive materials are heated to sinter the glass packing thereby providing an enclosed glass doped with radioactive materials. Silicate glasses with or without small amounts of boron (e.g. Vycor) have low thermal expansion coefficients and, as alkali metal content is increased, the expansion coefficient materially increases. It is preferred that the container not collapse prematurely, even when the inside is under vacuum and the outside is under atmospheric pressure at temperatures at which the enclosed glass packing begins to melt so that it remains a container which will contain the enclosed glass until it is advantageous to collapse the container. In that respect, it is preferred to use a container having a higher glass transition temperature (silicate glass and Vycor are advantageous). When the particles of the glass packing are heated, they will degas as solid objects below the glass transition temperature or even just above the glass transition, as long as the glass is not so hot that the particles coalesce with each other. At a slightly higher temperature, the glass particles melt into each other, i.e., they coalesce, and become a unitary glass body which, if done properly, is bubble-free. If the container collapses at a temperature which is slightly higher than the T.sub.G of the interior molten glass (including deposited solids), a bubble-free final glass product containing the nuclear waste results. However, if the container is not collapsed until a much higher temperature is reached, there is the risk that the solubility of gases, such as oxygen in the interior molten glass will decrease to a point where the content of the gas, e.g., oxygen, exceeds in solubility in the molten glass because it is under the vacuum used to collapse the container. When the gas (oxygen) content exceeds its solubility at the temperature and reduced pressure of the interior molten glass before the container collapses, bubbles and foam can form in the interior molten glass. Once the tube collapses, the interior is no longer subject to vacuum but at that time is subject of the external pressure; thus, solubility of the gases in the interior molten glass increases and the danger of bubble or foam formation is relieved. Prevention of bubble or foam formation requires fairly accurate selections of the container glass transition temperature and the interior glass transition temperature of the interior glass composition including the deposited radioactive solids. Again, if the interior glass composition is too soft for Vycor or a fused silica glass container, the collapsing temperature of the container can be lowered by using a container glass such as Pyrex. Of course, the manufacture of compositions having any desired glass transition temperature is well within the skill of the art and any means available can be used to provide glass compositions having suitable T.sub.G 's for the container and the interior glass packing. The container glass composition should be higher melting than the interior glass composition including deposited radioactive solids; i.e., it should have a higher transition temperature and should be able to collapse only after the interior glass composition has sintered. The presence of interior bubbles is not intolerable in many cases; however, if the absence or reduction of bubbles or foam is desired the T.sub.G 's of the glass compositions used should be selected as explained above. When using high silica (>90 mole % SiO.sub.2) low alkali (<0.5 mole % Na.sub.2 O) porous glass as a packing: (a) Pyrex tubes collapse at too low a temperature to permit sintering of the packing; and (b) Vycor tubes have the following disadvantages: While Pyrex and Vycor nuclear waste containers are suitable for many of the applications as shown in the examples, other compositions have preferred properties. The preferred container is produced by: (a) producing a porous glass container, such as a tube, as described in U.S. Pat. No. 4,110,096 at column 10, line 50 to column 16, line 36, and (b) doping said porous glass container with at least one dopant selected from the group consisting of cesium, rubidium, strontium, and copper. The doping could be accomplished by either of two methods: (2) The porous preform is immersed in a solution of dopant or a dopant compound. After the dopant concentration is uniform throughout the preform, the dopant is precipitated by dropping the temperature. The preform is immersed in a solution free of dopant. The dopant is allowed to partially dissolve and diffuse out of the matrix. Only the dopant precipitated near the outer surface is removed in this step. Further details of molecular stuffing are described in U.S. Pat. No. 4,110,096. In both methods, the doped porous preform is then dried and heated to the collapsing temperature of the pores. The drying should not substantially change the dopant distribution in accordance with the teachings in U.S. Pat. No. 4,110,096. Upon collapse of the pores, the container changes in appearance from opalescent to clear without a substantial change in shape other than the shrinking of its linear dimensions by about 20%. In addition, the dopant is used in an amount so as to result in a dopant concentration range of from 0.5 to 6 mole % based upon the total oxide content of the shrunken glass product. The porous glass preform usually has 4 mole % B.sub.2 O.sub.3, but may have up to 8 mole % Boron or other silica components. Thus, the shrunken container will have more than 86 mole % silica. However, more than about 90 mole % silica is preferred because the silica increases the chemical durability of the glass. Of the above two methods for introducing the dopant into the porous glass, method 1 is preferred. The dopant concentration is very uniform throughout the cross-section of a preform doped according to method 1. This high uniformity permits further preparing of the container by conventional glass blowing techniques. In Example 27, for example, the glass tube produced by method 1 (the ion exchange method) is heated and one end is closed without breakage. Since the preferred nuclear waste container should have both lower viscosity (lower collapsing temperature) and higher expansion coefficient than a 96% SiO.sub.2 glass, the addition of alkali dopants seem appropriate. We have discovered that at concentrations higher than 85 mole % SiO.sub.2 and lower than about 5 mole % alkali, the chemical durability of Cs or Rb glasses is far superior than that of the Na or K glasses of comparable composition. At room temperature, for 2 mole % alkali dopant, sodium glass is 1000 times less durable than cesium glass, and for cesium and rubidium at 100.degree. C., rubidium is 10 times better than cesium glass. The chemical durability for the cesium and rubidium glasses were measured by a leaching rate measurement in water of pH roughly 5.6 and 20.degree. C. The leaching rates were found to below 10.sup.-9 gms of silica per square cm of exposed surface of the sample per day after 20 days soaking time. This is an excellent chemical durability. However, while high chemical durability is obtained with a rubidium dopant, a cesium dopant is preferred because of the much lower cost of cesium. Divalent elements that can be advantageously incorporated together with Cs and/or Rb are Sr and Cu. In choosing the dopant and the concentration, one must not only consider the chemical durability but also the matching of thermal expansion coefficient and container collapse temperature to the sintering temperature of the nuclear waste powder. One ordinarily skilled in the art can obtain such a matching by independently adjusting the following variables: composition of nuclear waste, loading of nuclear waste in core material, dopant compositions and concentrations of dopant in container. However, some of the product may still crack, permitting the core to be exposed to the outside. Because of the large surface area of the core glass which is still covered by container glass (cladding) there is still a very major reduction in leaching rates of nuclear waste material in water due to the presence of said cracked cladding. Thus, we still consider this to be sealed. The present invention, which includes porous cation exchange particles in a glass container, can be employed to remove dissolved and undissolved radioactive solids from highly dilute solutions of same. For example, solutions containing as little as 1 ppt (part per trillion based on weight, i.e., 1 wt. part per 10.sup.12 wt. parts solution of radioactive cations can be purified. Dilute solutions having less than 0.01 microcurie radioactivity per ml as well as more concentrated solutions, e.g., those having 1 curie or more radioactivity per ml and those solutions between 0.01 microcurie and 1 curie radioactivity per ml, are efficiently treated by this invention. In a typical nuclear reactor there are several sources of radwaste as described hereinabove that must be safely contained. These include highly dilute liquid waste streams which can contain dispersed radioactive solids as well as dissolved radioactive solids, e.g., cations; concentrated liquid wastes which can contain radioactive cations, radioactive anions and radioactive solids (such wastes are the result of the boiling down of primary coolant containing boric acid initially used in the coolant as a chemical shim and the boiling down of used regeneration solutions from the regular ion exchange beds customarily used); and/or radioactive gases such as radioactive krypton and/or radioactive iodine. Therefore, our invention includes a total radwaste disposal system wherein particulate porous glass or silica gel as disclosed in the above-mentioned concurrently filed application having silicon-bonded alkali metal oxy, Group Ib metal oxy, and/or ammonium oxy groups is packed into a cation exchange column which preferably is a fusible glass column. The glass or silica gel particles can be held in the column by means of a porous closure such as glass wool or a porous disc in its lower end and, if desired, in its upper end also. In addition, the porous and/or non-porous glass particles can be mixed with the ion exchange glass or silica gel particles in the column to provide additional external surface on which dispersed, unsettled solids can settle out. It is preferred that the porous glass or silica gel be finely divided and sieved to a suitable size to maximize the rate of flow of the radwaste stream through and between the particles of the porous glass or silica gel and to also minimize the ion exchange time. First, the dilute radwaste stream is passed through the column and the radioactive cations in solution are cation exchanged with the alkali metal, Group Ib metal and/or ammonium cations in the porous glass or silica gel to chemically bond the radioactive cations to the glass or silica gel. If the dilute radwaste stream is to be reused as the primary coolant, it is conventional to add lithium ions as a corrosion inhibitor. Therefore, it can be advantageous to utilize a porous glass or silica gel having silicon-bonded lithium oxy groups so that lithium ions (which do not become radioactive as do sodium ions) are released to the coolant stream as radioactive cations are removed from it. Additionally, dispersed, undissolved radioactive solids in the dilute radwaste stream can be mechanically filtered on the porous glass or silica gel particles in the column as the stream percolates through and between the particles. In order to maintain the ratios of solids in the radwaste stream to the porous glass or silica gel small enough to maintain the filtering action as the solids accumulate on the porous glass or silica gel particles, fresh porous glass or silica gel particles can be added to the columm. After the column has been exhausted of its ion exchange capacity by the dilute liquid radwaste stream, it can be dried and the concentrated liquid radwaste (containing concentrated boric acid, for example, at a temperature 100.degree. C.) can be added to the column. Thus, the pores of the porous glass or silica gel can be stuffed with the radioactive solids, cations and anions contained by the concentrated radwaste. Excess boric acid then can be washed from between the particles of the porous glass or silica gel using cold water (less than 30.degree. C.) and the particles can be dried to deposit the radioactive solids, cations and anions within the pores of the porous glass or silica gel using techniques taught in U.S. Pat. No. 4,110,096. Thereafter, the column can be first evacuated to remove decomposition gases. Then the radioactive gases can be introduced into the glass column, and the column can be heated to collapse the pores of the porous glass or silica gel and to collapse the glass column thereby immobilizing and containing the exchanged radioactive cations, the radioactive solids on the exterior of the porous glass or silica gel particles, the radioactive solids, anions and/or cations deposited in the pores of the porous glass or silica gel and the radioactive gas contained by the glass column. Suitable pressure differentials can be used to facilitate the collapsing of the glass column. Heating can be continued to cause the porous glass or silica gel particles to stick to each other to further trap interstitial radioactive solids between the particles. Upon cooling there results a highly durable solid which effectively contains the radioactive waste introduced into the glass column. Because some of the nuclear reactor streams may be basic, some elements in the radwaste appear as anions, e.g., chronium, molybdenum, praseodymium and cerium anions, which, of course, have to be immobilized also. One way to accomplish this is to pass the basic radwaste stream through a customary anion exchange resin column. The column is regenerated with non-radioactive base, e.g., ammonium hydroxide. The effluent from said regeneration contains a higher concentration of radioactive (pernicious) elements and is boiled down in a boiler to provide a reduced volume of basic radwaste. When the concentrated basic radwaste in the bottom of the boiler is acidified under reducing conditions, some of the anions, e.g., Cr, Mo, Ce and Pr become cations which can be ion-exchanged with and removed by the above-mentioned porous glass columns. The boiler bottoms are defined as the concentrated solution or raffinate which remains after boiling down the solution and it may contain solids. It can be molecularly stuffed into the porous glass to become a highly durable solid waste product. There are many other industrial wastes which have to be eliminated from waste streams which, although not radioactive, are very poisonous to humans. For example, it has been well publicized that water bodies have been contaminated in the past with mercury, cadmium, thallium, lead, other heavy metals insecticides, and organic poisons. Often the concentration of such toxic substances in the waste streams is very low, thus presenting the problem of treating large volumes of water containing small amounts of toxic substances. Nevertheless, overall, large quantities of such contaminants do enter the ecosphere. The present invention can be used to purify such waste streams. This invention can be employed for concentrating and immobilizing radioactive cations in glass for extremely long time storage. For example, the sintered, silicate glass loaded with radioactive solids can be appropriately packaged in containers and buried beneath the earth's surface or at sea. Alternatively, the radioactivity of the sintered glass product containing the radioactive solids can be utilized in suitable devices or instruments for a variety of purposes, such as, destroying microorganisms, e.g., in the preservation of food, or in sterilizing sewage sludge or for any other purpose where radioactivity can be employed constructively. A typical range of radioactive solids content of the glass products of this invention resulting from the treatment of low level waste is about 1 ppb to 20,000 ppm, preferably about 10 ppb to about 1000 ppm of the glass product. A typical range of radioactive solids content of glass products of this invention resulting from the treatment of high level radwaste is about 2 wt. % to about 30 wt. % or more. Glass products of this invention which are to be used as radioactive sources can have solids contents falling in the above-mentioned ranges. In general, the glass articles of this invention comprise a first non-porous glass portion and a second non-porous glass portion surrounding the first portion. The first portion contains radioactive materials entrapped and immobilized therein and the second portion contains further radioactive materials entrapped and immobilized therein. The radioactive materials in one of said portions is derived from radioactive materials which were soluble in a nuclear waste (radwaste) solution and the radioactive materials in the other portion is derived from radioactive materials which where insoluble in said nuclear waste solution. For example, the radioactive materials in the first portion are derived from materials which were insoluble in the radwaste. As another example, the radioactive material in the first portion is derived from the radioactive materials which were soluble in the radwaste. Furthermore, the glass articles of this invention can include a third non-porous glass portion which surrounds the second portion, and the third portion is free of radioactive materials. The radioactive materials in the novel glass articles are described above. Also, the insoluble radioactive materials can be metallic precipitates of the platinum metal family. The glass article can be rod-shaped, tape-shaped or and desired shape. |
claims | 1. A method of operating a pressurized water reactor that has a core of a plurality of fuel assemblies and at least a first bank of control rods that are primarily moved into and out of selected fuel assemblies in the core to adjust the axial flux difference to substantially maintain or restore the axial flux difference within a first target band and at least a second bank of control rods that are primarily moved into and out of other selected fuel assemblies in the core to adjust the average temperature of the core to substantially maintain or return the average temperature to within a second target band, wherein the first bank of control rods and second bank of control rods are not moved together, comprising the steps of:giving the second bank of control rods priority of movement when the first bank of control rods and the second bank of control rods receive at the same time a demand signal to move in different directions; andgiving the first bank of control rods priority of movement when the first bank of control rods and the second bank of control rods receive at the same time a demand signal to move in the same direction. 2. The method of claim 1 wherein when the first bank of control rods is moving and the second bank of control rods gets a signal instructing the second bank of control rods to move in a different direction, the first bank of control rods will stop moving and the second bank will take over movement as it was instructed. 3. The method of claim 1 wherein when the first bank of control rods has priority of movement and moves to alter the axial flux difference to within a first preselected deadband at which the first bank of control rods stops movement, the second bank of control rods will start movement if the average temperature is not within a second preselected deadband. |
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abstract | Technologies are generally described for methods and systems of forming a palladium sulfide film on a substrate including flexible substrate. A palladium sulfide precursor may be applied to the substrate. The palladium sulfide precursor may comprise a palladium organothiolate. The palladium sulfide precursor may be heated under reaction conditions sufficient to decompose the palladium sulfide precursor to form the palladium sulfide film or patterns, the latter using various lithography techniques. |
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044712260 | summary | SUMMARY OF THE INVENTION This invention relates generally to radiant energy and more specifically to a radiation source consisting of a portable unit with a radiation shield member. The rapid curing of resin systems by the use of radiant energy within a broad spectral range is well established. The process of polymerization can actually be accomplished by irradiation of several forms of energy of which the most common is probably heat, but the use of ultraviolet energy has particular advantages in that the wavelength of ultraviolet energy permits sharp focus, and the devices for generation of ultraviolet power sufficient for polymerization can be small, have relatively low power consumption and be easily portable. U.S. Pat. No. 4,167,669 by C. R. Panico, for instance, describes an apparatus which pulse activates ultraviolet lamps for the purpose of accelerating the polymerization process. While such devices have been generally available in portable systems with hand held ultraviolet applicators, such as the one shown in U. S. Pat. No. 3,970,856, the relatively low ultraviolet power output of many of the units has resulted in the neglect of safety considerations. However, as the power output range has increased for industrial use, and as the devices have become more generally used in high production situations, accidental exposure risks have increased. Althoush hand held units remain quite satisfactory for many industrial applications, the hazard of the activation of an radiation applicator while it is directed at the operator or some other person in the vicinity is very real and very dangerous. The exposure of the human eye to high intensity ultraviolet radiation can cause severe eye damage and even blindness, and even exposure of skin to the radiation can cause severe burns, actually a "sunburn". It is, therefore, desireable to in some way eliminate the possibility of misdirecting the radiation while still permitting the versatility of hand held application, so that the work area treated by radiation can be easily and quickly changed in a production setting. The present invention accomplishes the restriction of the radiation exposure to a specific work area and prevents accidental exposure of the operator or nearby personnel, but essentially retains the advantages of a highly portable applicator of radiant energy. This is done by the use of a two part system which includes a locator plate to predetermine the area to be treated with radiation and an interlocking radiation head which is activated by a force actuated switch located on the same surface as an aperture for radiation emission. The applicator is accomodated into the locator plate by a tight fitting moulding which seals the radiation leaks and restricts exposure to the area determined by the locator plate, while the force activated switch location assures that the unit will be activated only when the radiation port is mated on a surface. Since the weight of the applicator head is not sufficient to activate the switch, merely placing the unit in the locator plate or on a flat surface will not cause accidental turn-on. To initiate the radiation it is necessary to positively push downward on the applicator while the force activated switch and the radiation port are on a surface. An additional switch can also be installed on the applicator to assure that the applicator is always used in conjunction with the locator plate. This accomplished by the use of a magnetically activated switch located in the applicator head along a surface which is adjacent to the locator plate when the applicator is inserted into the locator plate. The locator plate is then constructed with a magnet at the location which matches the magnetically activated switch when the applicator is properly positioned into the locator plate. The magnetically operated switch, wired in series with the force activated switch, then assures proper insertion into the locator plate before the radiation source can be activated by the force actuated switch. An additional safety feature of the shielded housing is the mechanical lock-out to prevent unintentional activation of the radiation source. For example, the lock-out can consist of a springloaded mechanical bar located to interfere with the movement of the force actuated switch. The initiation of the radiation therefore requires a two step operation which involves both releasing the mechanical lock-out and applying force to the applicator head when it is resting on a surface. This dual action prevents the accidental activation of radiation when, for instance, the applicator is being moved from one work area to another, by the operator accidentally wrapping his hand around the unit and pushing the switch with a finger tip. The specific procedure for operating the unit involves aligning the locator plate with its aperture over the area to be exposed to radiation. With the radiation source power supply plugged into an electric power source and turned on, the applicator head is inserted into the locator plate with its aperture downward, the lock-out bar is released, and hand pressure is applied atop the applicator. This pushes the applicator down against the force activated switch to turn the radiation source on. If the locator plate magnet interlock is being used, the simple operation of properly inserting the applicator into the locator plate operates that switch. The locator plate and the force actuated and magnetic switches whether used independently or together, thus each fulfill a requirement for safe operation of the hand held applicator. The locator plate, which may either be flexible or inflexible, locates the applicator and properly aims the radiation while preventing leakage around the periphery of the applicator. The force actuated switch and its mechanical lock-out bar encourage use only on a flat surface and prevent accidental activation while permitting ease of normal operation, and the magnetically operated switch assures operation only in association with the protective locator plate. |
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044217145 | summary | BACKGROUND The present invention relates to nuclear power plant systems and more particularly to a nozzle penetration device for a nuclear reactor pressure vessel closure head. Such a nozzle penetration is known from Swiss Pat. No. 38 37 06. The penetration, which consists of an austenitic material, is screwed into the ferritic portion of the closure. This tension coupling is subject to great stress during temperature changes due to the different coefficients of expansion. The groove provided in the region of the joint weld provides no compensation for the stresses on the screw coupling, since the differences in stress are present in this region also. There is no possibility of influencing the stresses with this design, nor for constructing e.g., a desired prestressing in the region of the tension coupling. SUMMARY OF THE INVENTION The task of influencing the heat stresses on the bearing coupling regions with a nozzle penetration of the type given above is the basis for the present invention. This task is solved according to the invention by providing a compensating ring between the shoulder of the penetration and the contact surface of the closure. The compensating ring, depending on its composition and height, permits a variation of stresses in the tension coupling area until complete equilization of stress is achieved. The device according to the invention will now be described on the basis of two examples of embodiment. |
description | This invention refers to a device for generating voltage dips in an electrical power generator and particularly on a wind turbine connected to the internal network of a wind farm. Sudden voltage steps can be caused in the connection to the electrical output network of a park of electric generators such as wind turbines, as a result of network defects. In such cases, implementing the proper protection disconnects the defective part of the network, producing a new voltage step in the opposite direction and restoring the voltage to the regular level, prior to the defect. The combination of both steps is known as a “voltage dip” which can be defined by two parameters: the depth and the duration of the voltage step. A “voltage dip” can cause a wind farm to be disconnected from the network bringing the consequent harmful effects along with it, this therefore requires the behaviour of wind turbines in response to voltage dips to be tested. Previous techniques describe various different devices for generating voltage dips for different purposes. Patent application WO0060430 describes a system for verifying the sensitivity of the components of a glass production plant in response to voltage dips that may occur, comprising a double cascade of industrial autotransformers connected to each phase of the current with at least two bipolar switches, which generates voltage dips in the machine output, in proportion to the values of the individual autotransformers located on the cascades. The system provides the option to vary the amplitude and the duration of the dip, along with the option to cause different voltage dips for each phase. U.S. Pat. No. 5,886,429 proposes a computer controlled and monitored testing station for testing the response of electronic equipment connected to the same to voltage dips and swells generated by the station. U.S. Pat. No. 5,920,132 describes a device which is able to generate a reduced voltage by means of an autotransformer on low power industrial equipment, only valid for low voltage. None of the aforementioned devices can be applied for verifying the behaviour of electrical power generators such as wind turbines in response to different types of voltage dips. One must consider that in each country there are different standards on this matter, therefore, a wind turbine manufacturer must be able to verify the effects of different types of voltage dips on its machines in order to guarantee compliance with the different standards. This invention proposes a solution to this problem. The subject of this invention is a device for testing, at low voltage, the behaviour of an electrical power generator such as a wind turbine in response to a voltage dip similar to one which may occur on the network to which is it connected, with the wind turbine connect to the network. The device comprises: a circuit located between the output of the machine's generator, which for the purposes of this descriptive report shall be understood to be located in a control cabinet, and the transformer of the MV network output with a low voltage wye-delta transformer for the power relating to the generator, a plurality of in-series impedances for each phase, a first group of switches associated with each phase respectively, a second group of switches associated with the in-series impedances and a switch for connecting the circuit to ground; mechanisms for generating short circuits by activating the first group of switches to select a single phase, two phase or three phase voltage dip, by activating the second group of switches to select the depth of the voltage dip and, optionally, by activating the third switch to generate a short circuit to ground; mechanisms for protecting the network during voltage dip generation and particularly an inductor to achieve a minimum amount of disturbance on the network during voltage dip generation. Other characteristics and advantages of this invention can be found in the following detailed description, explaining its purpose, in relation to the attached diagram. In a preferred embodiment, the invention device 1 is included in a platform or container located at the foot of the wind turbine 2, of a power rating ranging between 850 kW and 2 W, and is connected between the control cabinet 4 located at the output of the wind turbine and the LV/MV transformer 3. As shown in FIG. 1, the device 1 is connected by means of port 7 to the control cabinet and by means of port 8 to the transformer 3. The device 1 includes other ports 5, 6, 9 and a further port 10 for the connection to ground. The devices includes a circuit with a low voltage wye-delta transformer 31 for the power relating to the generator, three in-series impedances for each phase 11, 14, 17 (phase 1); 12, 15, 18 (phase 2); 13, 16 and 19 (phase 3); as shown in FIG. 2, associated on one side with a first group of switches 24, 25 and 26 for each phase, and on the other side, with a second group formed by three switches 20, 21 and 22 for each impeder, preferably 600 A, which, depending on their status, will produce short circuits of different characteristics, single phase or multi phase, and of different depths and durations in order to produce voltage dips of a maximum depth of 85%, and for a maximum duration of 500 milliseconds. The invention device 1 also includes the necessary operation and control elements required for operating the aforementioned switches in order to generate different types of voltage dips. The aforementioned device 1 is also equipped with temperature sensors to control its correct operation, with said temperature having to be within a range of 60° C. and 120° C. The device 1 also includes an inductor 23, preferably of 60 μH, allowing a minimum level of disturbance to the network during the generation of voltage dips for the wind turbine 2, so that the rest of the wind turbines on the park can continue to operate normally. The aforementioned voltage dip, single phase or multi phase, with or without ground, is generated by short circuits being produced between phases or between phases and ground. According to the invention device 1, switches 24, 25 and 26 are responsible for producing the short circuit for one, two or three phases, whilst switches 20, 21 and 22 are responsible for controlling the depth of the voltage dip produced, with switch 27 being that which connects the neutral to ground. The device 1 takes its power supply directly from the LV network. The device 1 according to this invention is, therefore, capable of providing 8 different types of voltage dips, in regard to depth and duration time, based on the single phase or multi phase short circuits generated, described below, thus testing the response of the control cabinet 4 of the wind turbine 2: a three phase short circuit will generate a long voltage dip (prolonged duration time) or a small voltage dip (of little depth); a single phase short circuit will generate a long voltage dip (prolonged duration time) or a small voltage dip (of little depth); a two phase short circuit with ground will generate a long voltage dip (prolonged duration time) or a small voltage dip (of little depth); a two phase short circuit will generate a long voltage dip (prolonged duration time) or a small voltage dip (of little depth). Other relevant elements of the device are as follows: Voltage and current measurement points. Bridging mechanisms which enable the direct connection of the wind turbine 2 to the network 30. Mechanisms for switching between “normal” mode, where the device is disconnected at the transformer 3 and a “dips” mode where the device is connected allowing the field test times to be optimised. In the preferred embodiment described above, modifications within the scope defined in the following claims can be made: |
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abstract | Compact particle selection and collimation devices are disclosed for delivering beams of ions with desired energy spectra. These devices are useful with laser-accelerated ion therapy systems, in which the initial ions have broad energy and angular distributions. Superconducting electromagnet systems produce a desired magnetic field configuration to spread the ions with different energies and emitting angles for particle selection. The simulation of ion transport in the presence of the magnetic field shows that the selected ions are successfully refocused on the beam axis after passing through the magnetic field. Dose distributions are also provided using Monte Carlo simulations of the laser-accelerated ion beams for radiation therapy applications. |
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claims | 1. A modular nuclear reactor, comprising:a plurality of sections, each section comprising:a tank comprising a front plate, a back plate, side plates, a top plate, and a bottom plate;a plurality of grid plates within the tank, each grid plate of the plurality of grid plates comprising a plurality of apertures and vertically separated from an adjacent grid plate;a plurality of fuel elements extending through each grid plate of the plurality of grid plates; anda plurality of heat pipes extending through each grid plate of the plurality of grid plates, the top plate, and an upper reflector; anda side reflector material surrounding the plurality of sections. 2. The modular nuclear reactor of claim 1, wherein the tank is filled with a heat transfer fluid, the heat transfer fluid in contact with the plurality of fuel elements, the plurality of heat pipes, and the plurality of grid plates. 3. The modular nuclear reactor of claim 2, wherein the heat transfer fluid comprises sodium. 4. The modular nuclear reactor of claim 1, wherein each heat pipe of the plurality of heat pipes comprises potassium, sodium, or a combination thereof. 5. The modular nuclear reactor of claim 1, wherein the tank is an inner tank, the inner tank surrounded by an outer tank, the outer tank comprising the upper reflector and a lower reflector. 6. The modular nuclear reactor of claim 1, wherein each grid plate of the plurality of grid plates comprises stainless steel. 7. The modular nuclear reactor of claim 6, wherein the each heat pipe of the plurality of heat pipes comprises 316 stainless steel. 8. The modular nuclear reactor of claim 1, wherein the tank comprises stainless steel. 9. The modular nuclear reactor of claim 1, wherein each heat pipe of the plurality of heat pipes is welded to the top plate and the upper reflector. 10. The modular nuclear reactor of claim 1, wherein each section comprises about between 100 heat pipes and 300 heat pipes. 11. The modular nuclear reactor of claim 1, wherein the plurality of sections comprises six sections, the six sections arranged in a circular pattern, a central opening shaped and configured to receive a control rod. 12. A method of forming a modular nuclear reactor, the method comprising:assembling a plurality of prefabricated fuel elements on a bottom plate of an inner tank and through apertures in a plurality of grid plates, the plurality of grid plates vertically separated from an adjacent grid plate;assembling a plurality of prefabricated heat pipes on the bottom plate of the inner tank, through the apertures in the plurality of grid plates, and through an upper reflector;forming a seal between the one or more prefabricated heat pipes and a top plate of the inner tank;coupling each of the top plate and the bottom plate to side plates, a front plate, and a back plate, the grid plates located within the inner tank;forming an outer tank substantially surrounding the inner tank;filling the inner tank with a heat transfer fluid; andplacing a side reflector material proximate the outer tank. 13. The method of claim 12, further comprising selecting the heat transfer fluid to comprise sodium. 14. The method of claim 12, further comprising selecting the bottom plate and the plurality of grid plates to comprise 316 stainless steel. 15. The method of claim 12, further comprising selecting the one or more prefabricated heat pipes to comprise sodium, potassium, or a combination thereof. 16. The method of claim 12, wherein forming an outer tank substantially surrounding the inner tank comprises welding the plurality of prefabricated heat pipes to the upper reflector. |
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claims | 1. A method of reducing doping process impurities in an ion implantation apparatus having a chamber through which an ion beam path extends, the method comprising:providing a removable shielding assembly for the chamber comprising at least first and second removable shielding members, the at least first and second removable shielding members being made of graphite or impregnated graphite;installing the shielding assembly in the chamber such that the first shielding member covers a first magnetic area inside the chamber and the second shielding member covers a second magnetic area inside the chamber;using the ion implantation apparatus for implanting ions into semiconductor wafers;removing the shielding assembly from the chamber after the ion implantation apparatus has been used for implanting ions into semiconductor wafers for a period of time;cleaning the shielding assembly while the shielding assembly is removed from the chamber; andre-installing the shielding assembly in the chamber for further use of the ion implantation apparatus. 2. The method of claim 1, wherein the shielding assembly includes a third shielding member adapted to cover a side surface inside the chamber extending between the first and second magnetic areas. 3. The method of claim 1, wherein the shielding assembly includes a third shielding member adapted to extend from a side edge of the second shielding member in a side-by-side relationship so as to cover a portion of a top surface inside the chamber adjacent the second magnetic area. 4. The method of claim 3, wherein the act of installing the shielding assembly comprises mounting a support mechanism in the chamber for supporting the third shielding member in a position juxtaposed to the top surface, and installing the third shielding member in the chamber, with a side edge of the third shielding member overlapping and supported on an adjacent edge of the second shielding member and a bottom surface of the third shielding member resting on the support mechanism. 5. The method of claim 4, wherein the act of installing the third shielding member further includes adjusting an adjustment mechanism of the support mechanism so that the third shielding member is supported substantially parallel to the top surface. 6. The method of claim 1, wherein the act of cleaning the shielding assembly further includes treating the shielding assembly such that the shielding assembly is substantially free of a cleaning agent used to clean the shielding assembly. 7. The method of claim 1, wherein the act of treating the shielding assembly comprises heating the shielding assembly to remove the cleaning agent. 8. A method of maintaining an ion implantation apparatus having a chamber through which an ion beam path extends, the method comprising:providing a removable shielding assembly for the chamber comprising a first shielding member for covering a first magnetic area in the chamber, a second shielding member for covering a second magnetic area in the chamber, a third shielding member for covering a side surface area in the chamber extending between top and bottom surfaces of the chamber, and a fourth shielding member for covering a top surface area in a side-by-side relationship with the second shielding member, wherein the first, second, third, and fourth shielding members are made of graphite or impregnated graphite; andinstalling the shielding assembly in the chamber such that the first shielding member covers the first magnetic area, the second shielding member covers the second magnetic area, the third shielding member covers the side surface area, and the fourth shielding member covers the top surface area. 9. The method of claim 8, further comprising:using the ion implantation apparatus for implanting ions into semiconductor wafers;removing the shielding assembly from the chamber after the ion implantation apparatus has been used for a period of time;cleaning the shielding assembly while the shielding assembly is removed from the chamber; andre-installing the shielding assembly in the chamber for further use of the ion implantation apparatus. 10. The method of claim 8, wherein the first, second, third, and fourth shielding members are made entirely of graphite or impregnated graphite. 11. The method of claim 8, further comprising installing an adjustable support mechanism in the chamber, the support mechanism being configured to support the fourth shielding member and vary the tilt of the fourth shielding member relative to the top surface area. 12. The method of claim 11, further comprising adjusting the support mechanism so that the fourth shielding member is supported substantially parallel to the top surface area. 13. An ion implanting apparatus comprising:a chamber through which an ion beam path extends, the chamber having a bottom surface and a top surface, the bottom surface having a bottom magnetic area and a top surface having a top magnetic area;a first shielding member comprising a unitary body having a raised central portion for completely covering the bottom magnetic area and an outer portion substantially surrounding the central portion; anda second shielding member comprising a unitary body having a raised central portion for completely covering the top magnetic area and an outer portion substantially surrounding the central portion;wherein the first and second shielding members are configured to be easily removable from the chamber to facilitate cleaning of the shielding members, the first and second shielding members being made entirely of graphite or impregnated graphite. 14. The apparatus of claim 13, wherein the first and second shielding members are substantially identical in size and shape. 15. The apparatus of claim 13, further comprising a third shielding member configured to cover a side surface inside the chamber extending between the top and bottom surfaces, the third shielding member being configured to be removable from the chamber and being made entirely of graphite or impregnated graphite. 16. The apparatus of claim 13, further comprising a third shielding member configured to cover a portion of the top surface in a side-by-side relationship with the second shielding member, the third shielding member being configured to be removable from the chamber and being made only of graphite or impregnated graphite. 17. The apparatus of claim 16, further comprising a support mechanism adapted for mounting in the chamber in a position such that the support mechanism supports the third shielding member in opposing relationship to the top surface of the chamber. 18. The apparatus of claim 17, wherein the support mechanism is configured to vary the tilt of the third shielding member relative to the top surface. 19. The apparatus of claim 18, wherein the support mechanism comprises a body and at least one adjusting bolt threadably received in the body, the bolt having an upper end positioned to support the third shielding member, the bolt being adjustable to vary the tilt of the third shielding member relative to the top surface. 20. The apparatus of claim 13, further comprising an arcuate-shaped shielding member configured to cover a top portion of a barrier extending upwardly from the bottom surface of the chamber. 21. An ion implanting apparatus comprising:a chamber through which an ion beam path extends having a bottom surface and a top surface;a bottom magnet disposed on the bottom surface, the bottom magnet having a magnet core;a top magnet disposed on the top surface, the top magnet having a magnet core;a first shielding member comprising a unitary body having a raised central portion for covering the magnet core of the bottom magnet and an outer portion substantially surrounding the central portion for covering a surface of the bottom magnet surrounding its respective magnet core; anda second shielding member comprising a unitary body having a raised central portion for covering the magnet core of the top magnet and an outer portion substantially surrounding the central portion for covering a surface portion of the top magnet surrounding its respective magnet core;a third shielding member configured to cover a side surface area inside the chamber extending between the top and bottom surfaces;a fourth shielding member configured to cover a portion of the top surface in a side-by-side relationship with the second shielding member;a support mechanism adapted to be mounted in the chamber for supporting the fourth shielding member in the chamber such that the fourth shielding member is juxtaposed the top surface of the chamber, the support mechanism being configured to vary the tilt of the fourth shielding member relative to the top surface;an arcuate-shaped, upright barrier coupled to the bottom surface of the chamber; anda fifth shielding member configured to cover a top portion of the barrier;wherein the first, second, third, fourth and fifth shielding members are configured to be removable from the chamber for cleaning the shielding members outside of the chamber, the first, second, third, fourth and fifth shielding members being made of isotropic graphite. 22. The apparatus of claim 15, wherein the third shielding member comprises first and second edges, the first edge of the third shielding member contacting an adjacent edge of the first shielding member and the second edge of the third shielding member contacting an adjacent edge of the second shielding member. 23. The apparatus of claim 16, wherein the third shielding member has a lip portion that engages an adjacent lip portion of the second shielding member such that the third shielding member is at least partially supported by the second shielding member. 24. The apparatus of claim 23, wherein the lip portions of the second shielding member and the third shielding member define areas of reduced thickness of the second and third shielding members, the areas of reduced thickness being dimensioned such that the third shielding member is co-planar with the second shielding member when the shielding members are installed in the chamber. 25. The apparatus of claim 19, wherein the body of the support mechanism is formed with a slot that is dimensioned to fit over a top edge of an existing plate in the chamber so as to support the support mechanism. 26. The apparatus of claim 15 wherein the third shielding member is formed with an aperture that is sized and positioned to correspond with an ion-beam outlet in the side surface of the chamber. 27. The method of claim 1, wherein the shielding members are made of isotropic graphite. 28. The method of claim 8, wherein the shielding members are made of isotropic graphite. 29. The apparatus of claim 13, wherein the shielding members are made of isotropic graphite. |
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abstract | The present invention is directed to an encapsulated β− particle emitter that comprises a sol-gel derived core that comprises a β−-emitting radioisotope and an encapsulant enclosing the core through which at least some of the β− emissions from the β−-emitting radioisotope pass, wherein the encapsulant comprises a substrate and a cover and at least a portion of the encapsulant is electrically conductive, and a method for making the same. Additionally, the present invention is directed to a capacitor comprising such an encapsulated β− particle emitter and a method of performing work with such a capacitor. |
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abstract | A grid including at least two meshed grid parts intended to be superposed in a longitudinal direction, each grid part extending in a transverse plane, and the grid parts being moveable one relative to the other in at least one transverse direction between an open configuration for the insertion of nuclear fuel rods in the longitudinal direction through the grid parts, and a closed configuration allowing each fuel rod inserted through the grid parts to be clamped transversely between the grid parts. According to one aspect of the invention, the grid includes elements for transversely immobilizing the grid parts in the closed configuration, the immobilizing elements being designed to engage as the superposed grid parts are moved closer together in the longitudinal direction. |
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043307086 | abstract | An electron lens having very low spherical aberration. The two electrodes of the lens, in the case of an immersion lens, and the outer electrodes in the case of a three electrode lens, have central openings (in the region of the optical axis) covered by a metallic thin foil or a fine mesh metallic screen curved in the general shape of a hyperbola. Equations and procedures are given for determining the exact shape of the curvature. A procedure is also disclosed for forming the metallic foil or screen section of the electrode to the desired shape. |
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060382856 | description | DESCRIPTION OF PREFERRED EMBODIMENTS The geometry associated with a cylindrically bent surface of a crystal monochromator, such as a Laue crystal monochromator, is shown in FIG. 1. The X-rays 25, 26 from a point source S are reflected by Bragg planes in bent crystal 30 and are focused at a virtual focal point F. An asymmetry angle .chi. is defined as an angle between crystal surface normal 31 and the Bragg planes used for the reflection of X-rays 25, 26. Bragg angle .theta..sub.B is the angle between the incident X-rays 25, 26 and the Bragg planes. Distance s is measured between source point S and the center of crystal 30 and distance .function. is measured between the virtual focal point F and crystal 30 (.function. is negative for a virtual focal point). If there is no variation of the angle of incidence along the crystal surface of crystal 30, then the reflected beam will be a monochromatic beam. A Table of Equations identifying equations discussed throughout this specification is found at the end of this Description of Preferred Embodiments. Equations 1 and 2 represent a condition for producing a monochromatic beam (the Rowland condition) where .rho. is a bending radius of a bent crystal 30, .rho. is positive when source point S is on a concave side of crystal 30 and is negative when source point S in on a convex side of crystal 30. The upper sign corresponds to the case when the source and the center of bending are on different sides of the crystal diffraction planes, and the lower sign corresponds to the case when the source and the center of bending is on the same side of the crystal diffraction planes. The source caustics is defined as a circle of radius .rho. sin(.chi..+-..theta..sub.B) centered at the center of bending and the focal caustics as a circle of radius .rho. sin(.chi..+-..theta..sub.B) also centered at the center of bending. Equation 1 requires source point S to be at an intersection of the Rowland circle and the source caustics, as shown in FIG. 2. In this embodiment, focal point F is at the intersection of the Rowland circle and the focal caustics. For simplicity, only the lower-sign case will now be discussed. For point source S, the virtual source as seen by a patient and a detector is not pointlike. As shown in FIG. 2, for a small region of bent crystal 30 around point A, the corresponding virtual focal point B is the intersection of the Rowland circle and the focal caustics, and the direction of diffracted beam 36 is along line AB. As point A sweeps to point C through bent crystal 30, characterized by angle .psi., the corresponding focal point sweeps through an arc on the focal caustics. This aberration of the virtual source point does not degrade the resolution of the resulting image because diffracted beams 35, 36 originate from an array of sources each with a specific direction of emission tangent to the focal caustics. As shown in FIG. 3, according to one preferred embodiment of this invention, the beams from a source are transmitted through crystal 30, first through concave surface 31 and then through convex surface 32. In one preferred embodiment of this invention, crystal 30 is bent with four-bar bender 50, a device which preferably comprises four parallel bars that bend a rectangular crystal by pushing crystal 30 with two inner bars 52 and pulling crystal 30 with two outer bars 53, such as shown in FIGS. 5-7. In its unbent form, crystal 30 has generally planar opposing face surfaces and preferably but not necessarily has an overall rectangular shape. In one preferred embodiment according to this invention, crystal 30 is constructed of silicon and has a uniform thickness of about 0.2 mm to about 3.0 mm, so that asymmetry angle .chi. is between about 0 degrees and about 40 degrees. Using four-bar bender 50 to bend crystal 30, it is possible to achieve a cylindrically bent crystal 30 by displacing inner bars 52 and corresponding outer bars 53 by a same amount or distance, and a problem may exist because the angle the crystal planes make with the incident X-rays is not the same across the crystal surface of crystal 30. This problem can be solved according to this invention, with differential bending, by unbending outer bar 53 by an amount or distance .DELTA.z and bending inner bar 52 by an equal amount or distance .DELTA.z, where distance .DELTA.z corresponds to a differential displacement which is in addition to the displacement required to bend crystal 30 into a cylindrical shape. The differential bending according to this invention modifies the concave crystal surface and the opposing convex crystal surface of crystal 30 from a cylindrical shape to a non-cylindrical shape. In one preferred embodiment of this invention, the differential bending forces or modifies crystal 30 into an approximate logarithmic spiral shape, which is a non-cylindrical shape. The amount of differential bending is given by Equation 3, where 2L.sub.c is the distance between two inner bars 52, L.sub.s is the distance between one outer bar 53 and one corresponding inner bar 52 and .rho. is the bending radius. As used throughout this specification and in the claims, the term cylindrical is intended to relate to a surface that is either precisely cylindrical or cylindrical within working tolerances. As used throughout this specification and in the claims, the term non-cylindrical is intended to relate to a surface that is not either precisely cylindrical or cylindrical within working tolerances. As used throughout this specification and in the claims, the term logarithmic spiral shape is intended to relate to a surface that either precisely follows a logarithmic curve or that approaches or approximates a logarithmic curve. The bending of a wide crystal 30, such as with four-bar bender 50 can be modeled by a four-point loaded beam, as shown in FIG. 7. The bending moment varies linearly along crystal 30, as shown in FIG. 7, depending on the forces applied at end portions of crystal 30. Crystal 30 is differentially bent by applying different forces at points A and D, as shown in FIG. 7, which results in distance Z1 being different than distance Z2. It is apparent that crystal 30 can be bent into any suitable non-cylindrical shape, such as a logarithmic spiral shape, using suitable mechanisms other than four-bar bender 50, which can produce bending results the same as or similar to results achieved with four-bar bender 50. For example, in one preferred embodiment of this invention it is possible to bend crystal 30 into a logarithmic spiral shape or any other suitable non-cylindrical shape by positioning edges of crystal 30 between opposing clamping members which when forced toward each other clamp and bend crystal 30 between the opposing members to form non-cylindrically curved opposing face surfaces of crystal 30. It is apparent that such clamping apparatus or any other suitable bending apparatus can be used in lieu of four-bar bender 50 to accomplish similar or better bending precision, for example to more closely approach a theoretical logarithmic spiral shape, than the bending precision accomplished with four-bar bender 50. In one experiment according to this invention, the K.sub..alpha.1 (low-energy E.sup.- =32.19 keV) line of the Ba was used for the low-energy beam and the K.sub..alpha.1 (high-energy E.sup.+ =34.72 keV) line of the Ce was used for the high-energy beam. For the silicon[111] reflection the Bragg angles for E.sup.+ and E.sup.- were 3.522.degree. and 3.265.degree., respectively, with .DELTA..theta.=(.theta..sub.B.sup.+ -.theta..sub.B.sup.-)=0.257.degree.. One main operational challenge was to switch between the high and low energies in a time period on the order of 0.01 s; this time is required, for example, to minimize motion artefacts of subtraction angiography during the diastolic cycle of the cardiac motion. This timed switching can be achieved by coating an anode with layers of Ba and Ce film and switching the focal point position of the incident electron beam. Using one crystal 30, there is an angle between monochromatic high-energy beam 46 and monochromatic low-energy beam 45, as shown in FIG. 3. By using two crystals 30 in a proper configuration, the virtual source can be coincident for both high-energy beam 46 and low-energy beam 45. In this case, there is no crossing angle between both high-energy beam 46 and low energy beam 45. The E.sup.+ and E.sup.- beams can be diffracted by the same bent crystal 30 using the same set of diffraction planes. This can be achieved by moving source point S on the Rowland circle for different energies and shaping the anode so that it intercepts the Rowland circle, as shown in FIG. 3. For low-energy beam 45 and high-energy beam 46, the source caustics radii of curvature are governed by Equations 4 and 5, so the motion of source point S is governed by Equation 6. In one preferred embodiment of this invention, the source point motion is 2.2 mm for a source-to-monochromator distance of 0.5 m, using the silicon[111] reflection. The two reflected beams traverse an object at an angle .DELTA..theta. with respect to each other. Because of the difference between the high-beam and the low-beam images, the subtracted image will have artefacts due to the misregistration of the two images. In one preferred embodiment of this invention, one major artefact is from bone edge mismatch between the two images. For the silicon[111] reflection, .DELTA..theta. is 4.5 mrad, which is near an upper limit of an acceptable crossing angle. A differential bending amount or distance .DELTA.z is calculated using Equation 3 and is practically the same for both high-energy beam 46 and low-energy beam 45, so the differential bending will allow the whole area of both beams 45, 46 to be reflected. Now consider using two crystals 30 to diffract beams of two different energies assuming that the same crystal reflection is used for both crystals 30, such as shown in FIG. 4. The radii of the source and focal caustics for E.sup.- and E.sup.+ are defined by Equations 7-10, where C.sub.L and C.sub.H are the source caustics radii, D.sub.L and D.sub.H are the focal caustics radii, .rho..sub.1 is the bending radius of crystal 30 which reflects low-energy beam 45 and .rho..sub.2 the bending radius of crystal 30 which reflects high-energy energy beam 46. The focal caustics defines the shape of the virtual object for the diffracted beam. Requiring D.sub.L =D.sub.H, as shown in FIG. 4, and shaping the anode to intercept the Rowland circles, the virtual sources of the diffracted beams coincide, thus providing Equation 11. In such embodiment, the motion of source point S required to switch between high-energy beam 46 and low-energy beam 45 is governed by Equation 12, where .theta..sub.B =(.theta..sub.B.sup.- +.theta..sub.B.sup.+) 2 and .rho.=(.rho..sub.1 +.rho..sub.2) 2. An experiment was conducted with a compact source, according to one preferred embodiment of this invention, as shown in FIG. 5. The setup, values for different parameters of components, and results of the experiment are discussed in a paper by Z. Zhong, D. Chapman, R. Menk, J. Richardson, S. Theophanis and W. Thomlinson, entitled Monochromatic energy-subtraction radiography using a rotating anode source and a bent Laue monochromator, Phys. Med. Biol., 42 (1997) pp. 1751-1762, the entire contents of such paper being incorporated by reference into and made a part of this specification. Through such experiment, it was determined that diffracted beams 55, 56, as shown in FIG. 5, were each almost ideally monochromatic. The bent Laue crystal monochromator of this invention is used to selectively diffract a cone beam of emission line X-rays produced by a conventional X-ray compact source. The bent crystal 30 of this invention solves a significant mismatch between the narrow angular bandwidth in diffraction of X-rays from a perfect crystal (e.g. the Darwin width for silicon[111] reflection is 5 .mu.rad at 33 keV), and the large divergence of the cone beam necessary for medical imaging with a conventional source (about 0.1 rad). Bending crystal 30 has at least two main advantages: one is to geometrically enable the diffracting planes to make the same Bragg angle with each ray of the incident beam and, therefore, to produce a monochromatic beam; the other is that differential bending increases the angle width and the integrated reflectivity of the crystal reflection. For a cylindrically bent crystal 30 there is a systematic deviation from the Bragg condition which is proportional to the square of the divergence of the incident beam. This deviation is negligible only if the asymmetry angle is chosen to be close to the Bragg angle. To increase the Full Width at Half Maximum (FWHM) of the reflection the asymmetry angle is preferably much larger than the Bragg angle, in which case the deviation is comparable to the FWHM of the reflection. This deviation from the Bragg condition can be compensated by controlling the bending of crystal 30. The median angle of the crystal planes at any point on the crystal surface corresponding to the beam divergence can deviate from that of the cylindrical bending condition. FIG. 6 shows how a controlled deviation from cylindrical bending can be achieved for crystal 30 bent with four-bar bender 30, according to one preferred embodiment of this invention. An ideal cylindrically bent crystal 30 is achieved by displacing outer bars 53 by the same amount or distance, and the angle the crystal planes make with the incident X-ray can be calculated. If, in addition to the displacement required to bend crystal 30 into a cylindrical shape, the upper (as shown in FIG. 6) outer bar 53 is unbent by an amount or distance .DELTA.z, as indicated in FIG. 6 by the solid circles 53 and the solid line schematically showing the crystal surface; and the lower (as shown in FIG. 6) outer bar 53 is further bent by an equal amount or distance .DELTA.z, as indicated in FIG. 6 by the open circles 53' and the dashed line schematically showing the crystal surface, crystal 30' will deviate from a cylindrical shape into a non-cylindrical or logarithmic spiral shape. Since the diffracting crystal planes across the bent crystal surface make the same angle with respect to the incident divergent beam, the energy of the diffracted beam is uniform over its area. The energy bandwidth of the monochromatic beam, in one preferred embodiment approximately 2%, is much larger than the width of the target emission lines. Thus, the energy of the monochromatic beam can be one of the emission line energies of the X-ray source in an energy range which is useful for medical imaging. An experiment was conducted according to another preferred embodiment of this invention. The setup, parameters and associated values, and the results are described in a publication by Z. Zhong, D. Chapman, W. Thomlinson, F. Arfelli, R. Menk, entitled A bent Laue crystal monochromator for nonochromatic radiography with an area beam, Nuclear Instruments and Methods in Physics Research, A 399 (1997) p. 489-498, the entire contents of such paper being incorporated by reference into and made a part of this specification. The uniformity of diffracted beams 55, 56 depends on matching the angle of the crystal planes with the divergence of the incoming beam at all points on crystal 30. If the angle that each of the crystal planes makes with the beam is within the reflection FWHM of the Bragg angle then the beam will be reflected; otherwise the reflectivity is close to zero. With four-bar bender 50, crystal 30 was capable of reflecting the full beam with a suitable corresponding image size at the detector position. The variation in intensity of the reflected beam was less than 10%, which can be corrected for by proper calibration images. While in the foregoing specification this invention has been described in relation to certain preferred embodiments thereof, and many details have been set forth for purpose of illustration, it will be apparent to those skilled in the art that the invention is susceptible to additional embodiments and that certain of the details described herein can be varied considerably without departing from the basic principles of the invention. ______________________________________ Table of Equations ______________________________________ s = .rho. cos(.chi. .+-. .theta..sub.B) Equation 1 f = -.rho. cos(.chi. .-+. .theta..sub.B) Equation 2 ##STR1## Equation 3 C.sub.L = .rho. sin(.chi. - .theta..sub.B.sup.-) Equation 4 C.sub.H = .rho. sin(.chi. - .theta..sub.B.sup.+) Equation 5 M .congruent. .rho. cos(.chi. - .theta..sub.B.sup.-).DELTA..theta. Equation 6 C.sub.L = .rho..sub.1 sin(.chi.- .theta..sub.B.sup.-) Equation 7 D.sub.L = .rho..sub.1 sin(.chi. + .theta..sub.B.sup.-) Equation 8 C.sub.H = .rho..sub.2 sin(.chi. - .theta..sub.B.sup.+) Equation 9 D.sub.H = .rho..sub.2 sin(.chi. + .theta..sub.B.sup.+) Equation 10 ##STR2## Equation 11 ##STR3## Equation 12 ______________________________________ |
abstract | In one embodiment, a fusion reactor includes an enclosure, an open-field magnetic system comprising one or more internal magnetic coils suspended within the enclosure, and one or more encapsulating magnetic coils coaxial with the one or more internal magnetic coils of the open-field magnetic system. The one or more encapsulating magnetic coils form a magnetosphere around the open-field magnetic system. The open-field magnetic system and the one or more encapsulating magnetic coils, when supplied with electrical currents, form magnetic fields for confining plasma within the enclosure. |
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051611791 | summary | FIELD OF THE INVENTION This invention relates to a beryllium window incorporated in an x-ray radiation system and, more particularly, to the structure of the beryllium window less deformative against heat stress in a process of fabrication thereof. DESCRIPTION OF THE RELATED ART A typical example of the x-ray aligner is illustrated in FIG. 1 of the drawings and largely comprises an electron gun 1, a target holder used for retaining a palladium target 2 and a wafer holder where a semiconductor wafer 4 is mounted. When the electron gun 1 emits electron beams towards the palladium target 2, an x-ray radiates from the palladium target 2 through a beryllium window 5. If the wafer holder 3 is located under the beryllium window 5, the semiconductor wafer 4 is exposed to the x-ray, and a pattern is transferred from an x-ray mask (not shown) to the wafer 4. The beryllium window 5 is provided between a target chamber 6 and an exposed space (not shown) for partitioning, and a beryllium thin film is presently the most appropriate in view of the transmittance of x-ray and of the mechanical strength of the window. However, a thin beryllium plate can not be directly attached to the inner wall of the x-ray aligner, and a reinforcing frame (not shown) is used for the attachment. Namely, the thin beryllium plate is directly bonded, brazed or welded to the reinforcing frame, and the reinforcing frame is attached to the inner wall of the x-ray aligner. An adhesive compound is provided between the thin beryllium plate and the reinforcing frame, and an electron beam is usually used in the welding. The beryllium window thus fabricated is, thereafter, baked in vacuum at 200 degrees to 400 degrees in centigrade. Following problems are encountered in the prior art beryllium window depending upon the way of fixing. Namely, if the thin beryllium plate is bonded to the reinforcing frame with an adhesive compound, the beryllium window is less resistive against heat and, accordingly, damaged in the baking treatment in vacuum. The brazing exposes the thin beryllium plate to heat at 800 degrees to 1000 degrees in centigrade, and the heat deteriorates the mechanical strength of the thin beryllium plate. If the thickness is increased, the beryllium plate may well withstand. However, such a thick beryllium plate is less transmissible for the x-ray. Thus, there is a trade-off between the mechanical strength and the transmissibility of the beryllium plate brazed to the reinforcing frame. The welding with an electron-beam tends to deform the beryllium plate due to heat produced by the strong electron beam. If the beryllium plate is decreased in thickness so as to improve the transmissibility, the heat deformation becomes serious. SUMMARY OF THE INVENTION It is therefore an important object of the present invention to provide a beryllium window which is free from heat attack. It is also an important object of the present invention to provide a process of fabricating a beryllium window well withstanding heat. To accomplish these objects, the present invention proposes to weld a beryllium plate to a reinforcing member through a diffusion welding. In accordance with one aspect of the present invention, there is provided a beryllium window comprising a) a beryllium plate containing beryllium as an essential element, b) a reinforcing unit for retaining the beryllium plate, and c) a welding film merged into a part of the beryllium plate and a part of the reinforcing unit and formed of a substance having at least one element selected from the group consisting of silver, gold, nickel and copper, the reinforcing unit being formed of a substance allowing the at least one element to diffuse thereinto. In accordance with another aspect of the present invention, there is provided a process of fabricating a beryllium window comprising the steps of: a) preparing a beryllium plate containing beryllium as an essential element, a welding substance containing at least one element selected from the group consisting of silver, gold, nickel and copper, and a reinforcing unit formed of a substance allowing aforesaid at least one element to diffuse thereinto; b) forming a film of the welding substance on a part of the beryllium plate; and c) welding the beryllium plate to the reinforcing unit through a diffusion of the welding substance into the beryllium plate and the reinforcing unit. In accordance with still another aspect of the present invention, there is provided a process of fabricating a beryllium window comprising the steps of: a) preparing a beryllium plate containing beryllium as an essential element, a welding substance containing at least one element selected from the group consisting of silver, gold, nickel and copper, and a reinforcing unit formed of a substance allowing aforesaid at least one element to diffuse thereinto; b) forming a film of the welding substance on a part of the reinforcing unit; and c) welding the beryllium plate to the reinforcing unit through a diffusion of the welding substance into the beryllium plate and the reinforcing unit. |
043808557 | description | Microballoons provide a convenient starting point for making laser fusion targets shell which are filled in accordance with the invention. These microballoons are thin walled hollow glass shells. The shells may have a minimum diameter of about 20 um and are available in larger diameters, e.g., 400 um. Thus, laser fusion targets in the range of 50 to 200 um in diameter may be made in accordance with the invention. These balloons may have wall thicknesses from 0.5 to 3.5 um. To faciliate handling and drilling the microballoons, as shown in the case of an exemplary microballoon 10, they are mounted on a substrate 12. Glass is suitable for use as the substrate. This substrate is coated, as by vapor deposition with a layer 14 of optically reflective material such as 1,000 to 2,000 Angstrom thickness of gold. This layer is optically reflective in order to facilitate the use of interferometry to measure the pressure or density of the gas in the microballoon 10 after it is filled. FIG. 7 illustrates this method of measuring the pressure or density of the gas in the filled microballoon and will be discussed more fully hereinafter. An adhesive adherent 16 is used to attach the microballoon 10 to the substrate. This adhesive may be salt (e.g., NaCl), methyl cellulose, or collodion, for example. When salt is used, a small drop of a saline solution, preferably less than the diameter of the microballoon, is placed on the surface of the reflective layer 14 and allowed to dry. The balloon 10 is placed in contact with the salt crystals. Moist air (e.g., any wet atmosphere, even the breath) causes the salt to provide a bond between the substrate and its reflective layer 14 and the balloon 10. Methyl cellulose in the form of a dilute solution in water may also be used. The surface of the layer 14 is coated with the solution and allowed to dry. The microballoon 10 is then deposited on the layer and a moist atmosphere is applied as in the case for the salt adhesive. The balloon is held by the methyl cellulose to the surface of the layer 14 on the substrate 12. With collodion, a dilute solution in ethanol or methanol is used to coat the surface of the layer 14. After the coating dries, the microballoon is placed on the surface. Then a vapor of the alcohol is used to soften the collodion, which holds the balloon to the substrate. Either a micromanipulator or a single camel's hair brush, or the like, may be used to transfer the microballoons to the substrate. The substrate may be mounted on a carrier to facilitate handling. Whenever the microballoons are handled, as when the microballoons are attached to the substrate as described in connection with FIG. 1, for drilling and as will be described in connection with FIG. 2 and plug location as will be described in connection with FIG. 3, these steps may be carried out at normal ambient conditions. Referring to FIG. 2, a hole next is drilled in the microballoon. The hole has a diameter from 0.5 to 2.5 um and is preferably one um or less in diameter. The smaller hole results in less perturbation in the surface of the microballoon after it is plugged and less force on the plug which might tend to set up stresses therein and break the seal. Such small holes may have aspect ratios (ratio of depth to diameter) as high as three to one. Typically, with a wall thickness of one micrometer, a suitable hole diameter may be also one micrometer which provides an aspect ratio of approximately one to one. Such small holes are usually smaller than the diffraction limited spot size of the laser beam used to drill the hole. This phenomena is due to the fact that the hole is drilled with a laser pulse having a duration of less than 100 psec. Use of the short pulse length permits routine fabrication of high aspect ratio holes smaller than the plane wave diffraction limited laser spot size(D): EQU D=1.22 (-lamda-)/N.A. where (-lamda-) is the wavelength of the light and N.A. is the numerical apeture of the focusing optics. Such holes have been produced with single pulses as well as with a train of approximately 20 pulses each less than 100 psec. In duration and separated from each other by 2-10 nsec. As shown in FIG. 2, pulses used to drill microballoons are obtained from an active-passive mode locked oscillator 18 operating at a wavelength of 1.054 um. For further information as to the operation of the active-passive mode locked oscillator see, Reproducible Active-Passive Mode Locked Oscillators, by W. Seka and J. Bunkenburg; 14.1 IEEE/OSA CLEA (1977), and Active-Passive Mode Locked Oscillators at 1.054 um by W. Seka and J. Bunkenburg; J. Appl. Phys 49, 7 (April 1979) p. 2277. A pulse train passes out of the oscillator 18 through a Pockels cell 22 which can be set to selectively pass a single pulse or adjusted to pass the full pulse train. Optionally the beam is frequency doubled to 0.527 um with a KDP crystal 28 which is set in the beam line after the Pockels cell 29. The beam is routed through neutral density filters 24 which are adjusted to vary the energy in the laser pulse used to drill the microballoon 10. Finally the beam is steered via a set of turning mirrors 30 into the focusing optics 26 and brought onto the top surface of the microballoon 10 which is held on the substrate 12. To insure that the beam is focused on the top surface of the microballoon 10 the focusing optics 26 are prealigned using either a CW HeNe or YAG alignment laser 20 which runs colinear with the pulsed beam. Table 1 compares some of the smaller holes drilled with the plane wave approximation for the diffraction limited focal spot. The ability to drill small holes with high aspect ratios in presently believed to be due to the use of laser pulse lengths shorter than 100 psec. TABLE 1 ______________________________________ Dimensions of Smallest Holes Plane Wave Approximation Glass Diffraction Character- Thick- Laser Lmtd. focal istic Hole Aspect ness Wavelength spot size(.mu. m) Dimension(.mu.m) Ratio ______________________________________ 1.2 1.054 3.2 .7 1.6 3.5 1.054 3.2 1.0 3.5 1.1 1.054 6.5 1.5 .7 1.0 .527 1.6 .3 3.0 ______________________________________ FIG. 3 shows the empty microballoon 10 having a hole 32 with a microsphere plug 34 thereon. The hole may, for example, be one micrometer in diameter. The diameter of the microsphere plug 34 may, for example, may be 2 um. The plug may be transferred to the microballoon and placed thereon over the hole 32 by a micromanipulator. The plug may first be placed on the under side of a flat glass substrate. If the plug is irregularly shaped the microballoon is preferably moistened to provide a water film which aids the transfer. The plug on the transparent glass substrate and the microballoon on its reflective substrate 12 are each micromanipulated separately using viewing optics. The limited depth of field of the optics assists in determining the height of the plug over the microballoon. The plug is then positioned over the hole and brought in contact. Once the plug is on the hole, it adheres more to the microballoon than to the substrate so that the substrate on which the plug is carried for transfer to the microballoon may be removed. The plug is selected to be of material having a melting temperature, by which is meant a temperature at which the material of the plug flows, which is lower than the melting point of the glass microballoon. Thermoplastic materials may be used to provide the plugs. The plug is preferably a spherical latex material such as that globules are formed by emulsion polymerization. Polystyrene latices are preferred. The polystyrene is preferably of high fractional crystallinity. In other words a high fraction of the polystyrene is crystaline rather than amorphous. A crystalline material provides a much better matrix than an amorphous material to limit diffusion of gas through the plug after it has melted to form a seal. A high molecular weight polymer is preferred. Such polymers which have high fractional crystallinity are characterized by a higher melting point than low molecular weight amorphous polymers (e.g., from 200.degree. C. to 260.degree. C. rather than from about 100.degree. to 160.degree. C.). The plug may also be made of glass with a low melting point; preferably solder glass. Solder glass is commercially available and is used extensively in the microelectronics industry to form dielectric seals over integrated circuits. Solder glass is available in the form of glass frit. It is desirable that this frit be spheroidized as by dropping it through a hot zone; however this is not necessary. Either glass microspheres or thermoplastic latex microspheres are both suitable plug materials for use in practicing the invention. The advantage of the glass plugs is that the pressure retention which is provided by seals formed from such plugs is greater than obtained with thermoplastic plugs. For example the pressure retention half life of argon sealed with a polystyrene plug in a 50 um diameter microballoon is from 15 to 30 hours. This retention time may be increased by storing the filled microballoons in a pressurized canister filled with argon. The microballoons are removed from the canister immediately before they are used as laser targets. A glass plug, for example, sealing argon in a 50 um microballoon provides a pressure retention half life of several months; thus storage in a pressure canister is not required. The microballoons are filled using a pressure canister or vessel 36 as shown in FIG. 4. The microballoons with microsphere plugs over their drilled holes are transferred with their substrates and carrier blocks into the pressure vessel 36. The vessel is first evacuated and then is filled with pressurized gas through a valve 38. The pressure may be increased gradually so as to limit forces on the plug which might establish stresses therein. As the pressure increases the gas flows around the microsphere plug 34 and between the plug and the of periphery the hole 32 as shown in FIG. 5. When the microballoons are filled to the requisite pressure, and while that pressure is still applied, the microballoons with the microsphere plugs are heated, as by heating the pressure vessel 36 with the aid of a heater 40. The plug melts and the melted plug forms a seal 42 in the hole 32 of the filled microballoon 10 as shown in FIG. 6. The melted microsphere plug minimally perturbs the surface finish of the microballoon. Typically the bump in the surface due to the melted plug 42 is only from 1,000 to 5,000 Angstroms above the surface and has a width of from 2 to 4 um. After the seal is formed, the microballoons are cooled to ambient temperature and the pressure vessel may be vented to the atmosphere through a valve 44. The method provided by the invention permits filling of the microballoon with gases other than those which can permeate through the glass shell of the microballoon. Hydrogen, neon and helium gas and the isotopic and ionic forms thereof can permeate through the shell of the glass microballoon. Argon can also be permeated through a few types of microballoons. However, the diffusion process is inconveniently slow and may take several weeks to accomplish. By virtue of the present invention every gas in addition to these three gases may also be used to fill microballoons since they are introduced into the balloon via the hole which is drilled therein. Other gases which are important for use in laser fusion targets may include argon, deuterium sulphide (D.sub.2 S) and Xenon (Xe). In addition, fluids may be introduced through the drilled hole as for example by glass micropipette which is micromanipulated into the hole. The plug is then placed on the microballoon over the hole after filling with the fluid. As shown in FIG. 10, a coating or layer may be placed on the inside of the balloon prior to filling through the use of the drilled hole. For example a coating of nickel or other metal may be provided by inserting a filament 46 through the hole 32 drilled into the microballoon 10. Current from a current source 48 vaporizes the filament within the confines of the microballoon 10 so that the vapor spreads out over the inside of the microballoon 10 to form a coating thereon. A chemical vapor deposition may also be used. For example to obtain a nickel coating, a nickel carbonyl gas may be used. The microballoon is then located in a canister which is filled with nickel carbonyl gas. The gas enters the drilled hole 32 and on heating, the gas decomposes and coats the inside of the microballoon with nickel. The coating formed on the outside of the microballoon may be removed by etching, preferably after the balloon has been filled with its intended fill gas (e.g. a deuterium-tritium mixture), and sealed. A glass plug is preferably used when etching the nickel film from the outside of the balloon is to be performed. Other deposition techniques can also be used for coating the inner surface of a hollow shell target. For example, it is known that laser photodeposition of organometallic compounds can produce metal films of aluminum, cadmium, tin, zinc, etc. (see Laser Photodeposition of Metal Films with Microscopic Features, by T. F. Deutsch, et al., Appl. Phys. Lett. 35, 175 (1979)) Thus an organometallic vapor (e.g. trimethylaluminum) may be introduced through the hole in a microballoon, entrapped therein by means of the herein described plugging method, and photodissociated by a suitable light source such as a uv-laser. This would yield a microballoon target with a metal coating on its inner surface. If desired, the glass could be etched away to produce a free-standing shell of the metal. Optical interferometry is used to measure the pressure of the gas which has been entrapped and sealed within the microballoon (see FIG. 7). An interferometer 50 is used to measure the optical path difference between two light paths 54, 56 reflected back to the interferometer one of which contains the microballoon. This optical path difference is the optical path length through the microballoon. The interferometer 50 be used first to measure the optical path through the empty balloon, using the double-pass configuration shown in FIG. 7. After the balloon has been filled with gas by the method of the present invention, the optical path is remeasured. The change in optical path is equal to (n.sub.g -1)D where D is the inside diameter of the microballoon and n.sub.g is the index of refraction of the gaseous fill. The index of refraction is related to the gas pressure, P, by the expression: EQU (n.sub.g -1)=kP ps where k is a constant which is characteristic of the gas; for example, k=0.000281 for argon. Thus a change in optical path of 0.20 um with D=50 um would correspond to an argon fill pressure of 14.2 atm. Accordingly the pressure of the gas in the microballoon and, therefore, the gas density may be determined by this simple and accurate method. Such pressure measurement is used in order to verify the integrity of the gas seal. It may be desirable to refill the microballoon with gas prior to use as a laser fusion target. Such refilling may be desirable when thermoplastic microsphere plugs are used. In refilling, the microballoons are returned to the canister. The pressure of the gas in the canister is raised slowly until the desired fill pressure is reached. The microballoons are refilled by diffusion through the thermoplastic plug. The pressure in the balloon may be verified after removal from the pressure canister by interferometry, as explained above in connection with FIG. 7. Balloons may be filled and sealed to contain high pressures. The pressure to which the balloon may be filled is not limited by the plug, since ruptures of the plug are due to the stress failures. Stress failures require high force. Even though the pressure on the plug is high, the area of the plug is very small, so that the force thereon is correspondingly small. It has been found, for example, that glass microballoons from 50 to 200 um diameter can be pressurized and plugged to hold over 100 times atmospheric pressure. In order to mount the filled microballoons as targets as in a target chamber 60 of a fusion laser system (FIG. 8), a glass fiber stalk 62 may be used. The top of the stalk 62 is disposed in contact with the plug 42 and thus covers the small area in which the surface finish of the microballoon is somewhat perturbed (see FIG. 9). The beam or beams from the fusion laser are made incident on the filled microballoon which functions as a laser fusion target. From the foregoing description it will be apparent that there has been provided an improved method for making laser targets and particularly for filling glass microballoon targets with gases which have heretofore been difficult or impossible to be used to fill such targets. Variations and modifications of the herein described method, within the scope of the invention, will undoubtedly suggest themselves to those skilled in the art. Accordingly the foregoing description should be taken as illustrative and not in any limiting sense. |
claims | 1. A method of moving nuclear fuel with a graphical user interface, the method comprising:inputting at least one fuel attribute into the graphical user interface; andgraphically filtering, with the graphical user interface, graphical fuel bundles in a graphical loading map,the graphical fuel bundles representing fuel bundles in at least one fuel pool or reactor core, or fresh fuel bundles,the graphical user interface including one or more loading tools configured to graphically select, sort, remove, or move the graphical fuel bundles in the graphical loading map based on the at least one corresponding attribute of the fuel bundles represented by the graphical fuel bundles,the graphically filtering including graphically selecting, sorting, removing, or moving the graphical fuel bundles in the graphical loading map, with the loading tools, according to the input at least one fuel attribute and at least one corresponding attribute of the nuclear fuel bundles represented by the graphical fuel bundles. 2. The method of claim 1, wherein the graphical user interface further includes at least one fuel pool table and a reload table, and wherein the graphically filtering includes graphically selecting, sorting, removing, or moving, with the loading tools, the graphical fuel bundles within or among the graphical loading map, the at least one fuel pool table, and the reload table based on the input at least one fuel attribute and at least one corresponding attribute of the nuclear fuel bundles represented by the graphical fuel bundles. 3. The method of claim 2, further comprising:storing at least one fuel pool database, the fuel pool database including a fuel pool list of at least one of the fuel bundles residing in the fuel pool; andgraphically populating the at least one fuel pool table with a graphical representation of at least one of the fuel bundles on the fuel pool list. 4. The method of claim 2, wherein the graphical user interface further includes a fresh fuel table, and wherein the graphically filtering includes graphically selecting, sorting, removing, or moving, with the loading tools, the graphical fuel bundles within or among the loading map, the at least one fuel pool table, the reload table, and the fresh fuel table based on the input at least one fuel attribute and at least one corresponding attribute of the nuclear fuel bundles represented by the graphical fuel bundles. 5. The method of claim 4, further comprising:storing at least one fresh fuel database, the fresh fuel database including a fresh fuel list of at least one of the fresh fuel bundles; andgraphically populating the at least one fresh fuel table with a graphical representation of at least one of the fuel bundles on the fresh fuel list. 6. The method of claim 1, further comprising:analyzing the filtered graphical loading map by simulating reactor performance with the filtered graphical loading map. 7. The method of claim 1, wherein the at least one fuel attribute includes exposure, a previous cycle in which the fuel bundle was used, k infinity, bundle product line, initial uranium loading, initial gadolinium loading, number of axial zones, historical fuel cycle numbers previous to a most recent for which the fuel bundle was used, a corresponding reactor core in which the fuel bundle was resident for each of the historical fuel cycles, accumulated residence time, and a parameter reflecting usability of the fuel bundle for continued reactor operation. 8. A method of moving nuclear fuel from a fuel pool, the method comprising:inputting at least one fuel attribute into a graphical user interface, the graphical user interface including one or more loading tools configured to graphically select, sort, and move graphical fuel bundles in a graphical loading map based on the at least one corresponding attribute of fuel bundles in a fuel pool or fresh fuel bundle represented by the graphical fuel bundles; andgraphically selecting, sorting, and moving, with the one or more loading tools within the graphical user interface, graphical fuel bundles into a graphical loading map,the sorting, selecting, and moving based on the input at least one fuel attribute and at least one corresponding attribute of the nuclear fuel bundles represented by the graphical fuel bundles. 9. The method of claim 8, wherein the graphical user interface further includes at least one fuel pool table graphically representing bundles in a fuel pool and a reload table, and wherein the method further comprises:graphically selecting, sorting, and moving the graphical fuel bundles based on the at least one corresponding attribute of the nuclear fuel bundles represented by the graphical fuel bundles within or among the fuel pool table, the reload table, and the graphical loading map. 10. The method of claim 9, further comprising:storing at least one fuel pool database, the fuel pool database including a fuel pool list of at least one of the fuel bundles residing in the fuel pool; andgraphically populating the at least one fuel pool table with a graphical representation of at least one of the fuel bundles on the fuel pool list. 11. The method of claim 10, further comprising:storing at least one fresh fuel database, the fresh fuel database including a fresh fuel list of at least a portion of available fresh fuel bundles; andgraphically populating the at least one fresh fuel table with a graphical representation of at least one of the fresh fuel bundles on the fresh fuel list. 12. The method of claim 9, wherein the graphical user interface further includes a fresh fuel table, and wherein the graphically selecting, sorting, and moving the graphical fuel bundles is within or among the graphical loading map, the at least one fuel pool table, the reload table, and the fresh fuel table via the one or more loading tools. 13. The method of claim 8, further comprising:analyzing, after the graphically selecting, sorting, filtering, and moving, the graphical loading map by simulating reactor performance with the graphical loading map. 14. The method of claim 8, wherein the at least one fuel attribute includes at least one of exposure, a previous cycle in which the fuel bundle was used, k infinity, bundle product line, initial uranium loading, initial gadolinium loading, number of axial zones, historical fuel cycle numbers previous to a most recent for which the fuel bundle was used, a corresponding reactor core in which the fuel bundle was resident for each of the historical fuel cycles, accumulated residence time, and a parameter reflecting usability of the fuel bundle for continued reactor operation. 15. A non-transitory computer-readable medium storing code causing a processor connected to an output device to:receive user input of at least one fuel attribute in at least one input field;display, on the output device, a graphical loading map with graphical fuel bundles representing fuel bundles in at least one of a fuel pool and a reactor core or fresh fuel bundles;display, on the output device, at least one loading tool configured to graphically select, sort, remove, or move the graphical fuel bundles the graphical loading map based on the input at least one fuel attribute and at least one corresponding attribute of the fuel bundles represented by the graphical fuel bundles; andgraphically select, sort, remove, or move, on the output device, with the loading tool, the graphical fuel bundles in the graphical loading map according to the input at least one fuel attribute and at least one corresponding attribute of the nuclear fuel bundles represented by the graphical fuel bundles. 16. The computer-readable medium of claim 15, wherein the code further causes the processor to:display, on the output device, at least one fresh fuel table; anddisplay, on the output device, at least one fuel pool table, the at least one loading tool configured to graphically select, sort, remove, or move the graphical fuel bundles among or within the fresh fuel table, fuel pool table, and graphical loading map based on the input at least one fuel attribute and at least one corresponding attribute of the fuel bundles represented by the graphical fuel bundles. 17. The computer-readable medium of claim 15, wherein the at least one fuel attribute includes exposure, a previous cycle in which the fuel bundle was used, k infinity, bundle product line, initial uranium loading, initial gadolinium loading, number of axial zones, historical fuel cycle numbers previous to a most recent for which the fuel bundle was used, a corresponding reactor core in which the fuel bundle was resident for each of the historical fuel cycles, accumulated residence time, and a parameter reflecting usability of the fuel bundle for continued reactor operation. |
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claims | 1. An optical unit configured to illuminate an object field in an object plane, the optical unit comprising:a collector mirror; anda first faceted optical element comprising a plurality of facet elements comprising a first facet element and a second facet element,wherein:each the plurality of facet elements comprises a reflective surface having a normal vector whose direction spatially defines an orientation of the reflective surface;the collector mirror is configured so that, during use of the optical unit, the collector mirror produces a polarization distribution so that, a polarization applied to the first facet element is different from a polarization applied to the second facet element;the first faceted optical element has a first state in which the normal vectors of the reflective surface of each of the plurality of facet elements are configured to produce a polarization distribution at a location of the object field during use of the optical unit; andthe optical unit is an EUV microlithography optical unit. 2. The optical unit of claim 1, wherein the polarization distribution that is applied to the first faceted optical element during use of the optical unit is different from the polarization distribution results at the location of the object field during use of the optical unit. 3. The optical unit of claim 1, wherein the polarization distribution at a first location of the object field is different from the polarization distribution at a second location of the object field. 4. The optical unit of claim 1, wherein the polarization distribution at the location of the object field is a tangential polarization distribution. 5. The optical unit of claim 1, wherein an angular distribution of incident radiation at the object field has a dipole form at the location of the object field, and each pole a principal polarization direction is perpendicular to a principal dipole axis. 6. The optical unit of claim 1, further comprising a second faceted optical element comprising a plurality of facet elements configured to superimpose an image of the plurality of facet elements of the first faceted element onto the object field during use of the optical unit. 7. The optical unit of claim 1, wherein the first facet element is moveable between first and second positions, and a direction of the normal vector of the reflective surface of the first facet element in the first position differs from a direction of the normal vector of the reflective surface of the first facet element in the second position. 8. The optical unit of claim 7, wherein the polarization distribution at the location of the object field during use of the optical unit is different depending on whether the first facet element is in its first state or its second state. 9. The optical unit of claim 1, wherein the collector mirror is a double collector. 10. The optical unit of claim 1, wherein no polarization element is arranged in a beam path between collector mirror and the first faceted optical element. 11. The optical unit of claim 1, further comprising a polarization element in a beam path between the collector mirror and the first faceted optical element, the polarization element being configured to alter a polarization distribution produced by the collector mirror during use of the optical unit. 12. The optical unit of claim 1, wherein no polarization element is arranged in a beam path between the first faceted optical element and the object field. 13. An apparatus, comprising:an optical unit according to claim 1,wherein the apparatus is an EUV microlithography projection exposure apparatus. 14. An optical unit configured to illuminate an object field in an object plane, the optical unit comprising:a collector mirror;a first faceted optical element comprising a plurality of facet elements comprising a first facet element and a second facet element; anda second faceted optical element comprising a plurality of facet elements configured to superimpose an image of the plurality of facet elements of the first faceted optical element onto the object field during use of the optical unit,wherein:each facet element of the plurality of facet elements of the first faceted mirror comprises a reflective surface with a normal vector whose direction spatially defines an orientation of the reflective surface;the collector mirror is configured to produce a polarization distribution so that a polarization applied to the first facet element is different from a polarization applied to the second facet element;the first faceted optical element has a first state in which the normal vectors of the reflective surface of each of the plurality of facet elements of the first faceted optical element is configured so that a polarization distribution results on the second faceted optical element during use of the optical unit; andthe optical unit is an EUV microlithography optical unit. 15. The optical unit of claim 14, wherein the polarization distribution on the second faceted optical element is a tangential polarization distribution. 16. The optical unit of claim 15, wherein a proportion of the radiation in a tangentially directed principal polarization direction increases from a center of the second faceted optical element outward. 17. The optical unit of claim 14, wherein the first facet element is moveable between first and second positions, and a direction of the normal vector of the reflective surface of the first facet element in the first position differs from a direction of the normal vector of the reflective surface of the first facet element in the second position. 18. The optical unit of claim 17, wherein the polarization distribution at the location of the object field during use of the optical unit is different depending on whether the first facet element is in its first state or its second state. 19. The optical unit of claim 14, wherein the collector mirror is a double collector. 20. The optical unit of claim 14, wherein no polarization element is arranged in a beam path between collector mirror and the first faceted optical element. 21. The optical unit of claim 14, further comprising a polarization element in a beam path between the collector mirror and the first faceted optical element, the polarization element being configured to alter a polarization distribution produced by the collector mirror during use of the optical unit. 22. The optical unit of claim 14, wherein no polarization element is arranged in a beam path between the first faceted optical element and the object field. 23. An apparatus, comprising:an optical unit according to claim 14,wherein the apparatus is an EUV microlithography projection exposure apparatus. |
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043839533 | claims | 1. A method of preparing an admixture comprising a particulate material and a fugitive binder for producing green pellets free of flaws and having improved strength, comprising the steps of : (a) fluidizing and agitating a mass of particulate material with a fluidized bed system; (b) adding a fugitive binder to the fluidizing and agitating mass of particulate material and blending the binder with the particulate material, said fugitive binder being comprised of ammonium bicarbonate, ammonium carbonate, ammonium bicarbonate carbamate, ammonium sesquicarbonate, ammonium carbamate, and mixtures thereof; (c) aging the blended binder and particulate material for a period of greater than 48 hours; and (d) forming the resulting aged blend by pressing into a green body. (a) combining and blending ammonium oxalate with particulate uranium dioxide; (b) fluidizing and agitating a mass of particulate uranium dioxide with a fluidized bed system and adding thereto the combined and blended ammonium oxalate and particulate uranium dioxide; (c) adding a fugitive binder to the fluidizing and agitating mass of particulate uranium dioxide and blending the binder with the particulate material, said fugitive binder being comprised of ammonium bicarbonate, ammonium carbonate, ammonium bicarbonate carbamate, ammonium sesquicarbonate, ammonium carbamate, and mixtures thereof; (d) aging the blended binder and particulate uranium dioxide for a period of at least about 72 hours; and (e) forming the resulting aged blend by pressing into a green body. (a) fluidizing and agitating a mass of particulate uranium dioxide with a fluidized bed system; (b) adding a fugative binder to the fluidizing and agitating mass of particulate uranium dioxide and blending the binder with the particulate uranium dioxide, said fugative binder being comprised of ammonium bicarbonate, ammonium carbonate, ammonium bicarbonate carbamate, ammonium sesquicarbonate, ammonium carbamate, and mixtures thereof; (c) aging the blended binder and particulate uranium dioxide for a period of at least about 72 hours while substantially quiescent; and (d) forming the resultant aged blend of binder and particulate uranium dioxide by pressing same into a green body. (a) combining and blending powdered ammonium oxalate with particulate uranium dioxide in approximately the same quantities by weight; (b) fluidizing and agitating a mass of particulate uranium dioxide with a fluidized bed system and adding thereto the combined and blended ammonium oxalate and particulate uranium dioxide; (c) adding a fugative binder to the fluidizing and agitating mass of particulate uranium dioxide in amount of about 0.5 to 7.0 weight percent and blending the binder with the particulate uranium dioxide, said fugative binder being composed of ammonium bicarbonate, ammonium carbonate, ammonium bicarbonate carbamate, ammonium sesquicarbonate, ammonium carbamate, and mixtures thereof; (d) aging the blended binder and particulate uranium dioxide for a period of at least about 72 hours while substantially quiescent; and, (e) forming the resulting aged blend of powdered ammonium oxalate, binder and particulate uranium dioxide by pressing into a green body. (a) fluidizing and agitating a mass of particulate uranium dioxide with fluidized bed system; (b) adding ammonium bicarbonate fugitive binder to the fluidizing and agitating mass of particulate uranium dioxide in amount of about 0.5 to about 7.0 weight percent and blending the binder with the particulate uranium dioxide for at least 10 minutes; (c) aging the blended ammonium bicarbonate binder and particulate uranium dioxide for a period of at least about 72 hours while substantially quiescent to provide for the decomposition of the ammonium bicarbonate; and (d) forming the resultant aged blend of binder and particulate uranium dioxide by pressing into a green body. 2. The method of claim 1, wherein the particulate material is comprised of uranium dioxide and plutonium dioxide. 3. The method of claim 1, wherein the particulate material is uranium dioxide. 4. The method of claim 1, wherein the binder is ammonium bicarbonate. 5. The method of claim 1, wherein ammonium oxalate is added to the particulate material as a pore former. 6. A method of preparing an admixture comprising a particulate material and a fugitive binder for producing porous green pellets free of flaws and having improved strength, comprising the steps of: 7. The method of claim 6, wherein the binder is ammonium bicarbonate. 8. The method of claim 6, wherein the binder is added to the particulate uranium dioxide in amounts of about 0.5 to about 7.0 weight percent. 9. The method of claim 6, wherein the pressed green pellet comprising the aged blend of binder and particulate uranium dioxide is sintered. 10. A method of preparing an admixture comprising particulate uranium dioxide and a fugitive binder for producing green pellets free of flaws and having improved strength, comprising the steps of: 11. The method of claim 10, wherein the binder is ammonium bicarbonate. 12. The method of claim 10, wherein the binder is added to the particulate uranium dioxide in amounts of about 0.5 to 7.0 weight percent. 13. The method of claim 10, wherein ammonium oxalate is added to the particulate uranium dioxide as a pore former. 14. The method of claim 10, wherein the pressed green pellet comprising the aged blend of binder and particulate uranium dioxide is sintered. 15. A method of preparing an admixture comprising a particulate uranium dioxide and a fugitive binder for producing porous green pellets free of flaws and having improved strength, comprising the steps of: 16. The method of claim 15, wherein the binder is ammonium bicarbonate. 17. The method of claim 16, wherein the pressed green pellet comprising the aged blend of powdered ammonium oxalate, binder and particulate uranium dioxide is sintered. 18. The method of claim 15, wherein the blending of the binder and uranium dioxide is continued for at least about 10 minutes. 19. A method of preparing an admixture comprising a particulate uranium dioxide and a fugitive binder for producing green pellets free of flaws and having improved green strength, comprising the steps of: 20. The method of claim 19, wherein the pressed green pellet comprising the aged blend of binder and particulate uranium dioxide is sintered. 21. The method of claim 19, wherein ammonium oxalate is added to the particulate uranium dioxide as a pore former. |
abstract | A method for installing a locking retainer in a tube to maintain internal components within the tube under compression comprising the steps of: a) providing an elongated retainer spring having large and small diameter sections with the large diameter section of a size for an interference fit with the interior diameter of the tube and the smaller diameter section of a size having a clearance with the interior diameter of the tube; b) inserting a smaller diameter section of an elongated tool into the larger diameter section of the elongated retainer spring; c) engaging a transition between the smaller and larger diameter sections of the tool against a transition between the larger and smaller diameter sections of the elongated retainer spring; d) inserting the combined tool and retainer spring into an open end of the tube containing internal components with an end of the smaller diameter section of the retainer spring entering the tube first; e) advancing the combined tool and retainer spring within the tube to compress the smaller diameter spring against an adjacent internal component until an end of the tool engages the adjacent internal component enabling the spring to apply a selected axial preload on the internal components in the tube; and f) withdrawing the tool from the retainer spring. |
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abstract | Methods of treating radioactive materials are disclosed. In one aspect, a method may include mixing a radioactive isotope diluted in a filler material with a radioactive material treatment composition to form a resulting material. The radioactive material treatment composition may include mostly salt, and from 0.5 to 15 wt % sorbent. The method may further include mixing the resulting material with one or more inorganic binding agents. Other methods of treating radioactive materials are also disclosed, as well as compositions for treating radioactive materials. |
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047117540 | summary | BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a method and apparatus for impacting a, surface with a controlled desired impact energy. More particularly, the present invention relates to a method and apparatus for impacting a surface with a small desired and controllable impact energy, whereby the sensitivity of impact detectors mounted on the surface, and in particular a surface of a nuclear reactor coolant system, can be tested. 2. Description of the Prior Art In the operation of pressurized water power generating plants, it is desireable to have a system which will enable the early detection of failure of primary system mechanical components. The failure of such mechanical components characteristically results in metal debris which concentrate in the steam generator input plenum and the bottom plenum of the reactor vessel. Moreover, metal debris in the form of objects left in the system during the construction phase are sometimes encountered. Such metal debris, when left undetected, have caused extensive damage to various components of the coolant system. During normal operation of the reactor system, the metal debris are transported to collection points by the normal flow of the primary coolant and, during their travel, are propelled against the metal walls enclosing the primary system coolant paths. Accordingly, surveillance of the energy imparted to the metal walls as a result of the impacts will provide both an indication of primary system component failure, and an indication of the presence of undesirable metallic debris which could cause subsequent failures. To detect the presence of metallic debris in the nuclear reactor coolant system, various systems for detecting the impact energy have been utilized. One such system is disclosed, for example, in U.S. Pat. No. 3,860,481 issued Jan. 14th, 1975 to R. Gopal et al and assigned to Westinghouse Electric Corporation. According to this system, a number of impact sensors, e.g. accelerometers, are disposed at strategic positions on the nuclear reactor coolant system, e.g. at the reactor vessel upper and lower plenums and the input plenum of each steam generator of the reactor coolant system, and the output signals from the impact sensors are detected and analyzed. According to current regulations issued by the Nuclear Regulatory Commission, each impact sensor must have a sensitivity capable of detecting an impact energy of 0.5 foot-pounds (0.68 joules) within three feet (0.91 meters) of an impact sensor. Accordingly, in order to test the sensitivity of the impact sensors, it becomes necessary to periodically impart an external impact with an energy corresponding to the desired sensitivity to a surface of the reactor coolant system adjacent a particular impact sensor. Moreover, in order to stress the design limits of the various detection algorithms, researchers often desire to measure and control desired impact energies of even a smaller value than the required sensitivity for the system, e.g. less than 0.1 foot-pounds. A number of methods and apparatus for imparting such external impact energies to a surface are known. Such devices may be, for example, a spring loaded mass which renders impact energy proportional to the spring constant, a pendulum device for providing an impact energy corresponding to the mass of the pendulum and the vertical height from which it is dropped to provide an impact, or a manually operated force hammer, including a transducer which produces an output which is a force versus time function, which is struck against the desired surface and the kinetic energy calculated from the area under the force versus time function. All of the known devices suffer from the disadvantage that there is always a degree of uncertainty of the true impact energy. This results for example, because the first two devices include frictional forces and/or spring constants which may vary considerably, and because all of the devices involve some manual manipulation of the impact imparting device, thus rendering it difficult, to repeatedly apply an impact of a desired energy. Moreover, as a result of the manual manipulation involved, and the increased time required to determine or calculate the impact energy and, if necessary, vary the impact force and repeat the impact to provide the desired impact energy, the time of exposure to radiation by the personnel operating the impact device is undesirably increased. Finally, the known devices suffer from the problem that they are difficult, if not impossible, to use on surfaces with particular orientation or location. This latter problem is of particular significance when attempting to impact the bottom plenum of the reactor vessel which can only be approached from the bottom, and not from the side, and thus the impact must be applied in an upward vertical direction. This is not possible with a pendulum and difficult with the other devices. SUMMARY OF THE INVENTION It is therefore an object of the present invention to provide a method and apparatus for impacting a surface with a desired impact energy which is fully automatic in operation and does not require any manual manipulation of the impact device by the operator. It is another object of the present invention to provide a method and apparatus which can automatically and repeatedly provide a desired impact energy to a surface in a short period of time, and which will immediately indicate the impact energy to the operator. It is a further object of this invention to provide an apparatus for impacting a surface with a desired impact energy which is simple to use, and which is substantially unaffected by the shape or location of the surface to be impacted, or by the orientation of the impacting device. The above objects are achieved according to the basic concept of the present invention by a method comprising the steps of: (a) providing an impacting device which can apply a variable impact energy corresponding to the magnitude of an input control signal; (b) placing the impact device adjacent the surface to be impacted; (c) thereafter applying an input control signal of a present magnitude to said impact device to initiate an impact; (d) determining the kinetic energy of the impact; (e) comparing the value of the determined kinetic energy with a value corresponding to the desired kinetic impact energy; (f) indicating the results of the comparison; (g) adjusting the present magnitude of the signal used for said control signal to reduce any difference as a result of the comparison; and (h) thereafter repeating steps (c) to (g) until a repeatable impact of the desired impact energy is determined and indicated. According to the preferred embodiment of the invention, the impacting device is a solenoid whose plunger of a known mass provides the impact, the input control signal is an input voltage applied across the solenoid coil, and the kinetic impact energy is determined by measuring the velocity of the plunger just prior to impact. Preferably this final velocity is measured by: detecting the movement of the plunger and producing an output signal whose magnitude is proportional to the distance moved by the plunger after application of said input voltage; sampling the magnitude of the output signal at uniform time increments; comparing the sampled values from successive sampling times; and, upon detecting a zero difference between two compared values (indicating that impact has occurred), utilizing the difference between the last two different sampling values as a measure of the velocity of the plunger just prior to impact, and thus of the kinetic impact energy. According to a feature of the invention, compensation for initial adhesion forces which may exist in the solenoid is provided by: initially slowly increasing the magnitude of a voltage signal applied to the solenoid coil, sensing the position of the plunger, and applying the input control voltage of present magnitude to the solenoid coil to initiate an impact only upon sensing an initial change in the position of the plunger. Although the method according to the invention can theoretically be carried out utilizing analog data processing, according to the preferred and disclosed embodiment of the invention, the method is carried out under control of a programmed microprocessor. Moreover, according the preferred embodiment of the invention, the position of the solenoid plunger is detected by means of a linear variable differential transformer whose moveable core follows the movement of the plunger and whose output signal magnitude is thus a measure of the plunger position. |
abstract | A thermionic (TI) power cell includes a heat source, such as a layer of radioactive material that generates heat due to radioactive decay, a layer of electron emitting material disposed on the layer of radioactive material, and a layer of electron collecting material. The layer of electron emitting material is physically separated from the layer of electron collecting material to define a chamber between the layer of electron collecting material and the layer of electron emitting material. The chamber is substantially evacuated to permit electrons to traverse the chamber from the layer of electron emitting material to the layer of electron collecting material. Heat generated over time by the layer of radioactive material causes a substantially constant flow of electrons to be emitted by the layer of electron emitting material to induce an electric current to flow through the layer of electron collecting material when connected to an electrical load. |
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059498390 | abstract | A fuel assembly for a boiling water reactor where the fuel assembly during operation is arranged vertically in the core of the reactor. The fuel assembly comprises a plurality of vertical fuel rods (22) arranged in one or more fuel bundles, wherein at least the majority of the fuel rods comprise a stack of fuel pellets (7) surrounded by a cladding tube (8b) and a plenum tube (23) which is connected to the upper part of the cladding tube (8b), the plenum tube having a cross-section area which is smaller than the cross-section area of the cladding tube. Further, the fuel assembly comprises a bottom tie plate (15) which retains and supports the lower part of the fuel bundle, a number of spacers (13) which retain and position the fuel rods in spaced relationship and are arranged axially separated along the fuel rods, a top spacer (21) which retains and supports the upper part of the fuel bundle and is arranged above and at a distance from the stacks of fuel pellets, and a fuel channel (2) which surrounds the fuel bundle or the fuel bundles. |
description | This application is a 371 National Stage Application of pending International Application No. PCT/US15/14904, filed Feb. 6, 2015, which application claims priority U.S. Provisional Application No. 61/937,030, filed Feb. 7, 2014. These applications are incorporated herein by reference, in their entirety, for any purpose. Embodiments of the invention relate generally to fusion, and examples of a Helium-3 fuel cycle for a fusion reactor are described. For many years the notion of thermonuclear fusion for electrical energy production was based on deuterium and tritium fuels where most of the fusion energy is released as 14.1 MeV neutrons. The engineering requirements due to wall loading in a commercial fusion reactor are accordingly difficult to achieve. Advanced fuels have been studied so as to mitigate these engineering problems and possibly find a reasonable reactor system with higher plant efficiency. The P-11B fuel cycle seems attractive because no neutrons are generated with this fuel cycle, however it must be said that there are side reactions with the fusion products (11B+4He→14N+n+157 keV) that do generate a small but significant neutron component. Nevertheless, the requisite temperature for P-11B fusion is as high as 400 keV. Due to the use of boron, which has a much higher atomic number than helium or hydrogen isotopes, the associated Bremsstrahlung radiation losses are large, and ignition cannot be obtained. The D-3He fusion reaction produces no neutrons as well (D+3He→4He (3.6 MeV)+H (14.7 MeV). However the D-D side reaction, while not as frequent, can generate 14.1 MeV neutrons through one of its fusion product reactions (D+T→4He+n+14.1 MeV). There is also the D-D reaction itself that produces a lower energy neutron (2.45 MeV) which is below the threshold for activation of most nuclear materials and is thus far less detrimental. Examples of methods and fusion reactors are described herein. Example methods and fusion reactors may be used to extract fusion reaction byproducts from a fusion reactor between pulses. In some examples, deuterium is supplied to a fusion reactor. A D-D fusion reaction is performed to produce energy, 3He, and tritium byproducts. The fusion reactor is pulsed to remove at least some of the tritium byproducts produced in the D-D reaction prior to a D-T fusion reaction. 3He may be created in some examples through decay of the removed tritium. In some examples, the fusion reactor is repetitively pulsed. In some examples, the 3He removed from the fusion reactor is used in subsequent D-D and/or D-3He fusion reactions. In some examples, the 3He is supplied to the fusion reactor together with additional deuterium. In some examples, sufficient 3He is provided to the fusion reactor from previous D-D fusion reactions to allow for a self-sustaining D-3He fuel cycle with no external 3He addition. In some examples, the tritium byproducts are remotely stored. In some examples, conducting the D-D fusion reaction includes forming at least two plasmoids and accelerating the at least two plasmoids towards one another. In some examples, conducting the D-D fusion reaction includes raising a temperature of a plasma to 70 keV or less. In some examples, a lithium blanket is provided for production of additional 3He. In some examples, the method includes storing energy generated during the D-D fusion reaction. In some example methods, deuterium is received in a fusion reactor. 3He is also received in the fusion reactor, wherein the 3He was generated from byproducts and/or decay of byproducts produced previously in the fusion reactor or another fusion reactor. The deuterium is reacted with the 3He in a fusion reaction, and tritium byproducts of the deuterium and 3He reaction are removed from the fusion reactor. Some example methods further include decaying the tritium byproducts to produce further 3He. In some example methods, reacting the deuterium with the 3He comprises accelerating two plasmoids towards one another. In some example methods, removing tritium byproducts includes pulsing a plasma used in the fusion reaction. In some example fusion reactors, a plasma formation region is provided for receipt of deuterium and 3He fuel. The fusion reactor is configured for generated plasmoids to be accelerated and compressed toward one another. Example fusion reactors include an interaction chamber where the plamsoids may merge and a fusion reaction may occur. Example fusion reactors further include one or more divertors for extraction of byproducts of the fusion reaction. The byproducts may include 3He and/or tritium in some examples. Certain details are set forth below to provide a sufficient understanding of embodiments of the invention. However, it will be clear to one skilled in the art that embodiments of the invention may be practiced without various of these particular details. In some instances, well-known circuits, control signals, timing protocols, and software operations have not been shown in detail in order to avoid unnecessarily obscuring the described embodiments of the invention. Example systems and methods described herein may employ a 3He fuel cycle which may reduce or suppress a dangerous D-T side reaction by extracting the tritium ions as they are created. The extracted tritium is unstable and may beta decay in a relatively short period of 11 years to 3He, a primary fuel for the D-3He reaction. Accordingly, example systems, reactors and methods described herein may enjoy a self-sustaining fuel cycle where the required 3He to operate the reactor may be generated by the decay of tritium ions extracted from the reactor itself. In some examples, a D-D side reaction may be suppressed or reduced by operating a fusion plasma at a higher temperature where a fusion cross-section for D-3He is much larger than D-D. Example reactors described herein and/or which may be used with fuel cycles described herein generally include systems in which plasmoids are formed and accelerated toward one another. Examples of suitable fusion reactors are described, for example, in International Patent Application No. PCT/US2010/024172, filed Feb. 12, 2010, entitled “Method and apparatus for the generation, heating and/or compression of plasmoids and/or recovery of energy therefrom” (claiming priority to U.S. Ser. No. 61/152,221, filed Feb. 12, 2009), U.S. Ser. No. 13/201,428, filed Feb. 12, 2010, entitled “Method and apparatus for the generation, heating and/or compression of plasmoids and/or recovery of energy therefrom,” International Patent Application No. PCT/US2011/047119 (WO/2012/021537), filed Aug. 9, 2011, entitled “Apparatus, systems and methods for establishing plasma and using plasma in a rotating magnetic field” (claiming priority to U.S. Ser. No. 61/372,001, filed Aug. 9, 2010), and International Patent Application No. PCT/US2012/063735 (WO/2013/112221), filed Nov. 6, 2012, entitled, “Apparatus, systems and methods for fusion based power generation and engine thrust generation” (claiming priority to U.S. Ser. No. 61/556,657, filed Nov. 7, 2011). All of the afore-mentioned patent applications are incorporated herein by reference in their entirety and for any purpose. FIG. 1 is a schematic illustration of a portion of a fusion reactor in accordance with examples described herein. The fusion reactor 5 may include an interaction chamber 10 in the center, a formation, accelerator, and compression section 36 on each end of the interaction chamber 10, and a plasmoid formation section 34 next to each accelerator/compression section 36. The fusion reactor 5 may additionally include a divertor 14 on the outer end of each formation section 34. The fusion reactor 5 may also include interaction chamber coils 30, 32 around the outer perimeter of the interaction chamber 10, accelerator coils 22 around the outer perimeter of the acceleration/compression section 36, formation coils 18 around the outer perimeter of formation section 34, and end coils 28 around the outer perimeter of the fusion reactor 5 between the extreme end of each formation section 34 and the respective divertor 14. The fusion reactor 5 may further include an annular array of small plasmoid sources 38 located near a dielectric vacuum tube wall under the first formation/acceleration coil 18 nearest the end coils 28. The chamber wall 16 of the fusion reactor 5 may act as a vacuum boundary. During operation, the axial array of coils may be energized in a properly sequenced manner to obtain the reactions described herein. Generally, the coil systems may provide for formation, acceleration, and compression of field reversed configuration (FRC) plasmoids to high velocity with respect to one another (e.g. up to 800 km/s). The motional energy of the FRC plasmoids may provide a significant fraction of the energy needed to heat the plasma to fusion temperature. The motional energy may be converted into thermal energy when two FRC plasmoids merge. The formation coils 18 may be supplied with an initial reverse bias. A forward bias is applied to the end coils 28 and the accelerator coils 22, as well as the interaction chamber coils 30, 32. The plasma formation section 34 may be increased in radius to provide for greater initial flux and energy. This is followed at smaller radius by a set of accelerator/compression coils 22 with forward bias increasing as the radius decreases moving toward the interaction chamber 10. A gradual reduction in radius and increase in compression may in some examples result as a plasmoid travels down the acceleration/compression section 36. The formation coils 18 may be energized sequentially to both form, accelerate, and compress the plasmoids simultaneously. In this manner, plasmoids may be magnetically isolated from the vacuum wall and moved toward and into the interaction chamber where they are merged with their mirror image to form a merged plasmoid that may be compressed to thermonuclear conditions. Accordingly, precise control of the coils may be used to repeatedly drive plasmoids into the interaction chamber from opposite directions, colliding and merging the plasmoids. After one pair has merged, another pair may enter the interaction chamber for merging and thermonuclear reaction. This “pulsing” of collision in the interaction chamber may advantageously allow access to the reaction products (e.g. during a time between collisions in in the interaction chamber), and accordingly allow for removal of tritium formed during a deuterium reaction. The tritium may in some examples be removed between each pulse—e.g. the plasmoids may collide and D-D fusion reactions occur, then tritium byproducts may be removed prior to undesired D-T reactions taking place, then a subsequent plasmoid collision may occur, followed by tritium byproduct removal, etc. Accordingly, deuterium and He may be introduced in some examples to the formation region of the fusion reactor. Byproducts including 3He and tritium may be collected at the divertor region (e.g. divertor 14 of FIG. 1). The divertor, e.g. divertor 14 of FIG. 1, may serve a variety of purposes. In the divertor region energy and fuel may be extracted from the fusion reaction between pulses. For example, fuel byproducts including hydrogen, deuterium, tritium, Helium-3 (Helion) and Helium-4 (Alpha) particles, or combinations thereof, may be extracted and separated in the divertor region. There are a number of techniques that may be used, and the particular technique in a given example may be selected based on the specific economics of a particular installation. Generally, the extraction techniques take advantage of the fact that the particles all have different mass-to-charge ratios and have a large energy spread. Suitable techniques include but are not limited to cryogenic separation, mass quadrupole separation, inversion-ion cyclotron extraction, and as well as a host of standard chemical processes. Extraction and separation may be done in-situ or at an external location. In this manner, Helium-3 (e.g. 3He) and tritium may be extracted from the byproduct of a fusion reaction and stored. The tritium may decay to 3He. The harvested 3He may be used as fuel for subsequent fusion reactions in the same or a different fusion reactor. Generally, the repetitively-pulsed nature of example fusion reactors, along with the divertor/converter regions, e.g. 14 of FIG. 1 located remotely from the fusion burn chamber (see e.g. element 10 of FIG. 1), the extraction of tritium can be obtained completely suppressing or reducing this side reaction. The suppression of the D-D reaction can be obtained by raising the plasma temperature to 70 keV which may also assure a self-ignited state which may be appreciably lower than that of other advanced fuel cycles in some examples. By being pulsed, examples of fusion reactors described herein may be capable of removing the tritium byproduct of the D-D reaction with each pulse reducing the high energy neutron (14.1 MeV) generation from the D-T reaction to near zero. This may reduce both the damage and activation the surrounding reactor materials, as well as provide for a source of 3He through the beta decay of the recovered triton into a helion. This fuel cycle can therefore be referred to as Helion Catalyzed D-D through self-supplied 3He, or HelCat-DD. Accordingly, following a plasma pulse (e.g. plasmoid collision), byproducts of the reaction including 3He and tritium may be removed from the fusion reactor. In some examples, the byproducts may be retained in, e.g. a reservoir, and later processed. Processing of the byproducts may include removal of the 3He and storage of the tritium to allow for decay of the tritium to 3He. 3He removed from the fusion reactor may be used in subsequent fusion reactors in that or a different fusion reactor. Similarly, 3He produced through decay of tritium produced in the fusion reactor may be used in subsequent fusion reactions in that or a different fusion reactor. In this manner, a reactor may be operated in some examples using only 3He that was produced from previous fusion reactions (either directly or through tritium decay in the same or a different fusion reactor). Once the initial fuel supply is established in some examples of fusion reactors described herein, the only new fuel which may be required for continued operation may be deuterium which is abundant on earth, and can be readily supplied from water from any source. With primarily all of the fusion energy in the form of fusion particle energy, a high net plant electrical generation efficiency can be obtained from direct conversion of both the fusion product and fusion plasma particles in some examples. This may be accomplished through the electromagnetic compression and expansion cycle employed to create the fusion conditions and thus may avoid the low efficiency and waste heat issues typically found in the usual thermal cycle employed by other nuclear and carbon based power sources. There may be several advantages of example implementations of the He fuel cycle combined with the Field-Reversed Configuration (FRC) fusion plasma generated in example fusion reactors described herein. Advantages are described herein by way of example and are not intended to be limiting. It is to be understood that not all examples may display all, or even any, of the described advantages. The FRC represents a promising magnetic confinement system for fusion regardless of fuel cycle. FIG. 2 is an illustration of field lines and pressure contours for an FRC plasmoid obtained from a resistive, two dimensional magneto-hydrodynamic (MHD) code calculation. The FRC plasmoid generally has no internal mechanical structure, no appreciable toroidal field or rotational transform, and an engineering beta near unity (where β is the ratio of plasma to confining magnetic field energy density). The equilibrium current is due generally to the plasma diamagnetism thereby avoiding or reducing the current driven instabilities that plague other fusion concepts. Plasma loss at the FRC edge generally occurs across a magnetic separatrix ensuring that the lost plasma is conducted far away from the burn region to a remote chamber where both the plasma and fusion particle energy can be directly converted to electricity at high efficiency. As a result of these features, the FRC offers a transformational change in reactor attractiveness. Example FRC based fusion reactors provide generally for high power density, simple structural and magnetic topology, straightforward heat exhaust handling, capability to burn advanced fuels for direct energy conversion, and radically reduced costs, due at least in part to their small size and low neutron fluence. A significant issue for magnetically confined plasmas is synchrotron radiation as the magnetic field strength is increased to attain the higher temperatures and pressures required for advanced fuel cycles such as D-3He. Devices such as the tokamak which have a relatively small plasma pressure compared to the large magnetic pressure needed to confine the plasma (β˜4%), cannot operate on advanced fuels due to the high energy loss from synchrotron radiation. To be viable, the local value of β, the ratio of plasma to magnetic energy density, must generally be high as this assures that inside the plasma the magnitude of the magnetic field is never large enough to cause significant synchrotron losses. Due to the high β nature of the FRC equilibrium, essentially all of the reacting volume of the FRC is characterized by a magnetic field of very low to negligible field strength, and thus an insignificant amount of synchrotron radiation. A practical problem faced by the D-3He fuel cycle is the low availability of 3He in nature. To avoid this difficulty, the helion (3He) in examples described herein is created either directly or through the decay of tritium produced in the D-D reaction. In this manner the helion ions needed for the operation of the reactor are supplied by the D-D side reactions. This fuel cycle can therefore be referred to as Helion Catalyzed D-D through self-supplied 3He, or HelCat-DD. In this manner, once plant operation is established, the only new fuel required for continued operation in some examples is deuterium with all or substantially all the helion atoms coming from those produced directly in the D-D reaction or from the decay of the tritons removed from the reactor after each pulse. FIG. 3 is a schematic illustration of reactions occurring in example D-3He fuel cycles described herein. During operation, Deuterium (D) and 3He may be provided to a fusion reactor (e.g. to the reactor 5 of FIG. 1). In the burn chamber (e.g. interaction chamber 10 of FIG. 1), D and 3He may react as shown to produce protons and 4He. In the divertor (e.g. divertor 14 of FIG. 1), energy conversion may occur yielding 3.02 MeV protons, 1.01 MeV tritons and 0.82 MeV 3He. The 3He may be removed from the divertor and returned to the fusion reactor for use in reacting with additional incoming Deuterium. The tritons may be removed from the divertor and allowed to decay in a remote storage location. In some examples, the storage location need not be remote. The tritons may decay to 3He and 0.018 MeV electrons. The 3He may be provided back to the fusion reactor for use in subsequent D-3He reactions. Side D+D reactions in the burn chamber may produce neutrons and 3He as well as protons and tritons. In the divertor, these side reactions may yield 14.7 MeV protons and 3.67 MeV 4He. In some examples, a lithium blanket may be provided (e.g. in or around the interaction chamber 10 of FIG. 1). As shown in FIG. 3, the Deuterium reactions may further produce 2.45 MeV neutrons, which may react with lithium 6Li to form tritons and 4He. At the divertor, this may yield 1.8 MeV 4He and 3.0 MeV tritons. The tritons may be transported out of the fusion reactor and may decay to 3He which may be used in subsequent reactions by the fusion reactor. Accordingly, generally Deuterium may be provided to a reactor. Some 3He may be provided to initialize the reactor in some examples. In some examples, subsequent 3He may be provided by decay of reactor products. Fusion reactions take place by sequentially accelerating and compressing plasmoids until they are merged in an interaction chamber. Following merging of the plasmoids, reaction products including tritons may be removed from the fusion reactor and allowed to decay. The removal may occur after each pulse (e.g. plasmoid merging) of the reactor in some examples. While the reaction cross section can be as large for D-3He operation as the D-T fuel cycle, it must generally be obtained at increased plasma temperature, T. The plasma pressure Ppl scales linearly with T. As noted, this requires a larger confining magnetic field, B (Ppl˜B2). For devices such as the tokamak, which already require operation at near the maximum practical field that can be obtained for superconducting magnets (B˜15 T), the magnetic field cannot generally be further increased. The consequence is that the plasma density, n, must be lowered (Ppl˜T). The fusion power scales as n2 so that the reactor volume, and with it cost, must increase dramatically to produce the same output power. Example fusion reactors described herein, however, employ the FRC as the fusion plasma. Fusion power density then scales as β′B4. The FRC may in some examples have the highest β of all magnetic fusion plasmas and may be contained in some examples by simple cylindrical magnets that can be operated at the highest practical fields. Repetitive operation of pulsed fields up to 30 T with conventional copper alloy coils may be performed in some examples. Having a much higher power density in some examples aids in maintaining the output power from fusion when employing the HelCat DD cycle, and only in the pulsed FRC compression cycle found in example fusion reactors can the power be maintained, and even increased with the HelCat DD advanced fuel cycle due to the much reduced neutron wall loading. By being compact, the ratio of reacting volume to receiving surface area may also be minimized or reduced, allowing for operation at the higher power density. By being pulsed, the pulse duration can be extended or the repetition rate increased to maintain the fusion output power at the highest levels consistent with heat removal. Parameters of the fusion plasma may be selected to ensure or promote a low neutron yield and high efficiency for energy production. The parameters may be based on the steady state value of the relative quantities of each fuel element (D and 3He). Variants of the D-3He cycle with 3He self-supply are also possible. With primarily all of the fusion energy in the form of fusion particle energy a high net plant thermal efficiency can be obtained from converting much of the fusion power by direct conversion. The low neutron energy yield may also afford a lower cost of first wall and shield structures in some examples. It may also provide for higher plant availability and operating life due to the lower wall loading, afterheat, and radioactive isotope inventory in some examples. An attribute of example fusion reactors described herein is relatively easy access to the fusion products in the exhaust gas stream after each pulse (e.g. merging of plasmoids). This allows for the selective removal of fusion products. In examples, the charged fusion products, 3He and T included, may be moderated in the plasma releasing their energy to it. This allows for direct energy conversion in the burn chamber from plasma expansion and subsequent flux driven energy conversion. The removed T may be stored and the 3He that is obtained may be used to supply one of the D-3He fuel components. Examples of this type of operation may have advantages when compared to steady state systems, or systems where there is no easy access to the fusion byproducts. First, the tritons created generally have no time for interaction with D, which allows one to obtain a larger amount of 3He. Second, since the D-T reaction is generally negligible, the neutron flux to the first wall is reduced compared with in situ consumption of the triton. The reduction is in large part due to the elimination of the most dangerous high-energy neutrons created by the D-T reaction (Eneut=14.1 MeV). In some examples, the neutron born in the D-D reaction (see FIG. 3) can be used to create additional helions through tritium production in the presence of lithium in the blanket. In such examples, the 3He and T produced in the FRC plasma may also be used. Generally, the two D-D fusion reactions in FIG. 3 serve mainly for the production of 3He and T (this also applies to the secondary reactions with the fusion neutrons if desired). This part of the cycle eventually may provide the 3He to complete the cycle where the overwhelming fraction of the fusion energy generated is provided by the reaction of D with 3He. The effect of the different operating cycles on the steady state fraction of 3He is shown in FIG. 4. It is clear from this plot that the helion to deuteron ratio is greatly enhanced at low plasma temperature due to the relative increase in D-D reactions. Neutron conversion into tritium for decay into He is readily accomplished with a natural lithium blanket (7.56% 6Li, remainder 7Li) where the reaction 7Li+n→T+4He+n−2.47 MeV produces a second neutron assuring a total 3He yield per D-D neutron greater than one (up to a maximum of 1.9). FIG. 4 is a graph of helion to deuteron ratio in example fusion plasmas in various modes of operation. Line 44 illustrates a mode where 3He is produced primarily or only in the D-D reaction. Line 43 illustrates a mode where 3He is produced in the D-D reaction along with 3He decay of T from the D-D reaction. Line 42 illustrates a mode where a Li blanket is provided such that each neutron may produce one triton. Line 41 illustrates a mode where each neutron may produce 1.9 tritons. While a relatively low plasma temperature provides for the highest fractional levels of 3He, it does not necessarily provide for the best suppression of the neutron energy (lowest fraction of neutron power to total fusion power, εneut) or the highest conversion efficiency (highest fraction of particle to total fusion power, εpart). Insight into what is the most favorable operating condition is provided by the dependence of εneut on both plasma temperature and helion to deuterium ratio. A plot reflecting these tradeoffs is found in FIG. 5. FIG. 5 is a graph of fractional neutron power as a function of helion-deuteron ratio in the fusion plasma. The dashed lines indicate 50 keV plasma while the solid lines are for 70 keV. The lines 51 pertain to examples where the tritium produced in D-D reactions cannot be removed. The lines 52 reflect tritium removal after each pulse in the fusion reactor. The graph illustrates an example of the role of tritium removal in reducing neutron exposure regardless of the helion to deuteron ratio. While the sensitivity to plasma temperature is not great, the lower temperature plasmas suffer more substantial losses from Bremsstrahlung radiation. The ratio of Bremsstrahlung power loss to total fusion power produced, εbrem scales as T2 over this range of plasma temperatures which greatly favors the higher temperatures in addition to the lower neutron power with temperature. An example of results of optimization for an example fusion engine yield the basic plasma parameters found in the below table which provides example parameter values in each of three modes of operation—(1) D-D without tritium removal; (2) D-D with tritium removal; and (3) D-D with tritium removal and lithium blanket. Cat D-DCat D-D(T removal wParameterCat D-D(T removal)Li blanket)3He/D0.125 0.158 0.245nT/nD0.006−10−3 −10−3 nD (1022 m−3)0.750.750.62ntot (1022 m−3)1.951.881.72Tl = Te (keV)7070 70 B (T)2222 22 τ (s)0.157 0.103 0.088εbrom0.250.230.22εnaut0.270.060.04Ptus (MW) (@ 1 Hz)7272 72 Reviewing the table, operation with tritium removal may be considerably more preferable than operation without tritium removal. The wall loading alone would restrict operation to less than 25 MW in this example as a reduction in pulse rate may be required to limit excessive wall loading. The fusion reactor could be operated with or without 3He generation with the D-D neutrons. The decision between these two modes will most likely reflect the cost of the additional technology, although it would appear to be a fairly straight forward use of bulk lithium in a simple, removable blanket. As the advantage is not large, the extra 3He could be used for startup of new fusion reactors if needed. Either way example tritium-suppressed, self-supplied fusion reactors employing D-3He fuel cycles described herein may make for a sweeping advance toward a carbon-free, safe and efficient method for electricity generation from fusion. From the foregoing it will be appreciated that, although specific embodiments of the invention have been described herein for purposes of illustration, various modifications may be made without deviating from the spirit and scope of the invention. |
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045308128 | claims | 1. A composite magnetic coil winding comprising at least two conductor segments, a first segment of said at least two conductor segments being made from a copper material and a second segment of said at least two conductor segments being made from an aluminum material, each of said conductor segments defining a circumferential portion of said coil winding and each of said segments being joined to an adjacent segment to form said coil winding. 2. The composite magnetic coil winding of claim 1, wherein said circumferential portions are joined at a high strength joint. 3. The composite magnetic coil winding of claim 2, wherein the joint has low electrical resistance. 4. The composite magnetic coil winding of claim 2 or 3, wherein said winding has a generally regular and continuous surface and said joint does not protrude from said surface. 5. The composite magnetic coil winding of claim 2 or 3, wherein said coil winding has means for removing heat from said coil winding and wherein said joint does not interfere with said heat removal means. 6. The composite magnetic coil winding of claim 2 or 3, wherein said joint is a mechanical joint. 7. The composite magnetic coil winding of claim 2 or 3, wherein said joint is a metallurgical joint. 8. The composite magnetic coil winding of claim 1, wherein the copper material is in the form of a copper alloy. 9. The composite magnetic coil winding of claim 1, wherein the aluminum material is in the form of an aluminum alloy. 10. The composite magnetic coil winding of claim 8, wherein the copper alloy is copper beryllium. 11. The composite magnetic coil winding of claim 8, wherein the copper alloy is copper berylliumnickel. 12. The composite magnetic coil winding of claim 8, wherein the copper alloy is a Mg-Zr-Cr copper alloy. 13. The composite magnetic coil winding of claim 9, wherein the aluminum alloy is from the 2000 series of aluminum alloys. 14. The composite magnetic coil winding of claim 9, wherein the aluminum alloy is from the 6000 series of aluminum alloys. 15. The composite magnetic coil winding of claim 9, wherein the aluminum alloy is from the 7000 series of aluminum alloys. 16. The composite magnetic coil winding of claim 9, wherein the aluminum alloy is reinforced with graphite. 17. A composite magnetic coil winding for a tokamak reactor comprising at least two conductor segments, a first segment of said at least two segments defining a first circumferential portion of said coil winding and comprising a copper material and a second segment of said at least two segments defining a second circumferential portion of said coil winding and comprising an aluminum material and wherein said segments are joined at a high strength, low electrical resistance joint. 18. The composite magnetic coil winding of claim 17, wherein said coil winding has a generally smooth and continuous surface and wherein said joint does not protrude from said surface. 19. The composite magnetic coil winding of claim 17, wherein said coil winding has means for removing heat from said coil winding and wherein said joint does not interfere with said heat removal means. 20. The composite magnetic coil winding of claim 17, wherein said joint is a mechanical joint. 21. The composite magnetic coil winding of claim 17, wherein said joint is a metallurgical joint. 22. The composite magnetic coil winding of claim 17, wherein said copper material is a copper alloy. 23. The composite magnetic coil winding of claim 17, wherein the Al material is an Al alloy. 24. A toroidal reactor for producing fusion reactions, said reactor having a toroidal field generating means which are exposed to a neutron flux produced by said fusion reactions, said toroidal field means comprising at least one composite magnetic coil winding having at least two circumferential segments, a first of said at least two circumferential segments comprising a copper material and a second of said at least two circumferential segments comprising an aluminum material and wherein each of said circumferential segments are joined at a high strength, low electrical resistance joint to form said coil winding. 25. The toroidal reactor of claim 24, wherein said coil winding has a generally smooth and continuous surface and wherein said joint does not protrude from said surface. 26. The toroidal reactor of claim 24, wherein said toroidal field coil means further comprises heat removal means and wherein said joint does not interfere with said heat removal means. 27. The toroidal reactor of claim 24, wherein said joint is a mechanical joint. 28. The toroidal reactor of claim 24, wherein said joint is a metallurgical joint. 29. The toroidal reactor of claim 24, wherein said reactor has a main axis and wherein said copper field generating means nearest said main axis and said aluminum material segment is positioned on an opposite side of said toroidal field generating means, furthest from said main axis. 30. The toroidal reactor of claim 24, wherein said copper material is a copper alloy. 31. The toroidal reactor of claim 24, wherein the aluminum material is an aluminum alloy. 32. The toroidal reactor of claim 30, wherein the copper alloy is from the group consisting of copper beryllium and Mg-Zr-Cr-copper. 33. The toroidal reactor of claim 31, wherein the aluminum alloy is from the group consisting of the 2000, 6000 and 7000 series of aluminum alloys. 34. The toroidal reactor of claim 31, wherein aluminum alloy is reinforced with graphite. 35. A method of forming a composite field coil winding for a fusion reactor comprising: providing a first circumferential coil segment made of a copper material; providing a second circumferential coil segment made of an aluminum material; joining said first and second circumferential coil segments at a relatively high strength, low electrical resistance joint to form a toroidal field coil winding. providing a first circumferential coil segment made of a first electrically conductive material operable to conduct a first current density and having a first strength operable to support a first load; providing a second circumferential coil segment made of a second electrically conductive material operable to support a second current density lower than said first current density and having a second strength operable to support a load lower than said first load and having a relatively low neutron absorption cross section whereby neutrons can pass through said second circumferential coil segment more freely than through first circumferential coil segment; and joining said first and second circumferential coil segments at an electrically conductive joint having a strength at least equal to said second strength to form a toroidal field coil winding. 36. A method of forming a composite field coil winding for a fusion reactor comprising: 37. The method of claim 36 or 35 further including the step of positioning the first coil segment at a location of maximum stress and current density. 38. The method of claim 37 further including the step of positioning the second coil segment at a location of maximum neutron fluence to minimize neutron capture. 39. The method of claim 36 including the step of making the first coil segment from a copper material. 40. The method of claim 36 including the step of making the second coil segment from an aluminum material. 41. The method of claim 39 or 35, wherein the copper material is a copper alloy. 42. The method of claim 40 or 35, wherein the aluminum material is an aluminum alloy. |
043426212 | abstract | An apparatus for cooling molten material resulting from a nuclear reactor core meltdown is disclosed. The apparatus includes a basin positioned under the reactor which is protected against excessive heat by a star-like array of heat pipes whose evaporator sections are disposed above the pan and whose condenser sections are disposed in a heat sink exterior to the containment building of the reactor. Additionally, the vertical walls of the reactor vessel chamber are similarly protected by an array of heat pipes similarly arranged and provided to intercept the radient energy of the molten core material. |
description | The subject matter disclosed herein relates generally to X-ray imaging systems, and more particularly to anti-scatter grids for reducing grid line image artifacts in X-ray images generated using the X-ray imaging systems. A number of radiological and fluoroscopic imaging systems of various designs are known and are presently in use. Such systems generally are based upon generation of X-rays that are directed toward a subject of interest and attenuated, scattered or absorbed by the subject. The X-rays traverse the subject and impact a digital detector or an image intensifier. In medical contexts, for example, such systems may be used to visualize internal bones, tissues, and organs, and diagnose and treat patient ailments. In other contexts, parts, baggage, parcels, and other subjects may be imaged to assess their contents. In addition, radiological and fluoroscopic imaging systems may be used to identify the structural integrity of objects and for other purposes. Such X-ray imaging systems may include anti-scatter grids for blocking the scattered X-rays from impacting the detector. An anti-scatter grid typically includes structures of radiation absorbing material (e.g., lead strips) to absorb scattered X-rays. However, such structures of radiation absorbing material also absorb primary X-rays, i.e., X-rays that travel in a straight line from the source to the detector, which may leave dark grid lines on a generated X-ray image. Such image artifacts are known as the grid line image artifacts. The grid line image artifacts may not only affect image quality, but also impair effective use of the images, such as for diagnosis in medical diagnostic contexts. There is a need, therefore, for improved approaches to use anti-scatter grids in a way that reduces the grid line image artifacts in X-ray images. In accordance with a first embodiment, an imaging system includes a detector configured to detect X-rays from an X-ray source. The detector includes multiple photodetector elements. The imaging system also includes an anti-scatter grid disposed over the detector, wherein the anti-scatter grid includes multiple radiation absorbing elements. At least a portion of one or more of the radiation absorbing elements of the multiple radiation absorbing elements is disposed on each photodetector element, and a total area of each respective portion of the one or more radiation absorbing elements disposed on each photodetector element is substantially equal. In accordance with a second embodiment, an imaging system includes a detector configured to detect X-rays from an X-ray source. The detector includes multiple photodetector elements having a pixel pitch p, wherein each photodetector element includes an axis along a length or width of the photodetector element. The imaging system also includes an anti-scatter grid disposed over the detector, wherein the anti-scatter grid includes multiple radiation absorbing elements. At least a portion of one or more of the radiation absorbing elements of the multiple radiation absorbing elements is disposed on each photodetector element, and a respective portion of the one or more radiation absorbing elements disposed on each respective photodetector element is disposed at an angle α relative to the axis. In accordance with a third embodiment, a method for assembling an X-ray detector includes providing a detector configured to detect X-rays from an X-ray source, wherein the detector includes multiple photodetector elements having a pixel pitch p, wherein each photodetector element includes an axis along a length or width of the photodetector element. The method also includes disposing an anti-scatter grid over the detector at an angle α, wherein the anti-scatter grid includes multiple radiation absorbing elements. At least a portion of one or more of the radiation absorbing elements of the plurality of radiation absorbing elements is disposed on each photodetector element, and a respective portion of the one or more radiation absorbing elements disposed on each respective photodetector element is disposed at the angle α relative to the axis. The present disclosure provides for systems and methods for utilizing an anti-scatter grid to reduce the grid line image artifacts in X-ray images. For example, a series of X-ray absorbing materials of the anti-scatter grid may be disposed at an angle relative to an axis along columns and/or rows of the photodetector elements of the detector to equally distribute among the photodetector elements the portions of the X-ray absorbing materials that cover each photodetector element. The techniques discussed below may be applied to various types of anti-scatter grids such as parallel anti-scatter grids, focused anti-scatter grids and so forth. In addition, the techniques described below may be utilized in a variety of radiographic imaging systems, such as computed tomography (CT) systems, fluoroscopic imaging systems, mammography systems, tomosynthesis imaging systems, conventional radiographic imaging systems and so forth. Further, it should be appreciated that the described techniques may also be used in non-medical contexts (such as security and screening systems and non-destructive detection systems). Turning now to the drawings, FIG. 1 illustrates diagrammatically an X-ray imaging system 10 utilizing an anti-scatter grid 12. The X-ray imaging system 10 includes an X-ray source 14 positioned adjacent to a collimator 16. The collimator 16 permits an X-ray beam 18 to pass into a region in which a subject 20, such as a human patient, an animal, or an object, is positioned. A portion of the radiation 22 passes through or around the subject 20, where it may be attenuated and/or scattered by the subject 20. The anti-scatter grid 12 is positioned between the subject 20 and a detector 24. Another portion of the radiation (e.g., non-scattered X-rays 26) passes through the anti-scatter grid 12 and impacts the detector 24. The detector 24 may include a detector panel array 28, which coverts X-ray photons received on its surface to lower energy light photons, and subsequently to electric signals, which are acquired by electronics 30 and subsequently processed to reconstruct an image of the features of the subject 20. In certain embodiments, the detector 24 is a complementary metal-oxide-semiconductor (CMOS) based detector. Scattering is a general process whereby some forms of radiation, such as X-rays, are forced to deviate from a straight trajectory by one or more localized non-uniformities in the medium through which it passes. The anti-scatter grid 12 reduces the effect of scattering by preventing scattered X-rays from reaching the detector 24. The anti-scatter grid 12 herein is further designed to reduce grid line image artifacts, as discussed below. The X-ray source 14 is coupled to a power supply/control circuit 32, which furnishes power and commands X-ray emission for imaging examination sequences. Moreover, the detector 24 is communicatively coupled to a detector controller 34, which coordinates the control of the various detector functions. For example, the detector controller 34 may execute various signal processing and filtration functions, such as initial adjustment of dynamic ranges, and interleaving of digital image data. Both the power supply/control circuit 32 and the detector controller 34 are responsive to signals from a system controller 36. In general, the system controller 36 commands operations of the imaging system 10 to execute examination protocols and to process acquired image data. The system controller 36 may include signal processing circuitry, which is typically based upon a programmed general purpose or application-specific digital computer; and associated manufactures, such as optical memory devices, magnetic memory devices, or solid-state memory devices, for storing programs and routines executed by a processor of the computer to carry out various functionalities, as well as for storing configuration parameters and image data. The system controller 36 may further include interface circuitry that permits an operator or user to define imaging sequences, determine the operational status and health of system components and so forth. The interface circuitry may also allow external devices to receive images and image data, command operation of the X-ray system 10, configure parameters of the X-ray system 10 and so forth. The system controller 36 may be coupled to a range of external devices via a communications interface. Such devices may include, for example, an operator workstation 38 for interacting with the X-ray system 10, processing or reprocessing images, viewing images and so forth. Other external devices may include a display 40 or a printer 42. In general, these external devices 38, 40, 42 and similar devices may be local to the image acquisition components, or may be remote from these components, such as elsewhere within a medical facility, institution or hospital, or in an entirely different location, linked to the image acquisition system via one or more configurable networks, such as the Internet, intranet, virtual private networks and so forth. In the embodiment illustrated in FIG. 1, the X-ray imaging system 10 may be a stationary system disposed in a fixed imaging room or a mobile system. The system 10 may also include a fixed or mobile c-arm system. The detector 24 may be portable or permanently mounted with respect to the system 10. The anti-scatter grid 12 is either permanently mounted together with the detector 24 to the system 10 or may be removable from the detector 24 and the system 10. FIG. 2 illustrates schematically a side view of the anti-scatter grid 12 with the detector 24. The anti-scatter grid 12 may be mounted in contact with the detector 24, i.e., with no distance in between. In various other embodiments, the anti-scatter grid 12 may also be mounted together with the detector 24 with a distance in between (i.e., not in contact with the detector 24). The distance in between the anti-scatter grid 12 and the detector 24 may be fixed or adjustable depending on particular configurations and/or settings of the imaging system 10. Further, in one embodiment, the anti-scatter grid 12 is permanently mounted together with the detector 24. In another embodiment, the anti-scatter grid 12 is removable from the detector 24. In the embodiment illustrated in FIG. 2, arrow 64 indicates the direction in which X-ray beams may pass through the anti-scatter grid 12 and impact the surface of the detector 24. The detector 24 includes the detector panel array 28 and the electronics 30. The detector panel array 28 may include a pixel array of photodetector elements (e.g., arranged in rows and columns), each of which may be a light sensing photodiode. The photodetector elements convert light photons to electrical signals. The detector panel array 28 may further include switching thin film field-effect transistors (FETs). In one embodiment, a scintillator material deposited over the pixel array of the photodetector elements and FETs converts incident X-ray radiation photons received on the scintillator material surface to lower energy light photons. Alternatively, the detector panel array 28 may convert the X-ray photons directly to electrical signals. Each photodetector element of the detector panel array 28 is also generally referred to as a “pixel” and typically in square shape. These photodetector elements are typically aligned adjacent with one another, forming an array of photodetector elements with rows and columns on the surface of the detector panel array 28. Each photodetector element has an axis along its length or width, i.e., along each row or column of the photodetector elements. The length or width of each photodetector element is generally referred to as the “pixel pitch” p. For example, in one embodiment, the detector 24 has a pixel pitch p of approximately 0.195 mm, which means there are approximately 5 pixels or photodetector elements per millimeter along the rows or columns of the detector panel array 28. The electronics 30 convert analog electrical signals generated from the detector panel array 28 to digital values that can be processed to form a reconstructed image. In one embodiment, the detector 24 is a complementary metal-oxide-semiconductor (CMOS) based detector. In alternative embodiments, the techniques discussed herein may be applied to other types of digital detectors, such as amorphous silicon based detectors and so forth. FIG. 3 shows a perspective view of the anti-scatter grid 12 disposed over the detector 24 in the imaging system 10. The anti-scatter grid 12 may comprise a series of spaced elements 82 (e.g., parallel strips), each of which comprises a radiation absorbing material such as lead, tantalum, uranium or alloys and mixtures or laminates of one or more of all of the foregoing metals. The anti-scatter grid 12 may further comprise spaces 84, which are provided between the radiation absorbing elements 82 and typically comprise a low-radiation absorbing material such as air, aluminum, foam, carbon fiber and the like. Such low-radiation absorbing spaces 84 are provided so as to allow an unscattered X-ray beam 86, i.e., a primary X-ray beam 86, to travel through the anti-scatter grid 12. In the embodiment illustrated in FIG. 3, each radiation absorbing element 82 has a height h, a thickness d, and a distance D between an adjacent element 82. A grid pitch of the anti-scatter grid 12 is defined as the sum of the thickness d of each element 82 and the distance D between adjacent elements 82, i.e., d+D. A grid line rate of the anti-scatter grid 12 is the inverse of the grid pitch, i.e., 1/(d+D). In the embodiment illustrated in FIG. 3, the anti-scatter grid 12 may be a parallel grid, wherein all of the radiation absorbing elements 82 are parallel to each other and perpendicular to the surface of the anti-scatter grid 12. The anti-scatter grid 12 may also be a focused grid, wherein the radiation absorbing elements 82 are progressively tilted such that straight lines extended from the points at which the elements 82 intersect with the surface of the anti-scatter grid 12 would intersect at a single point, i.e., focal point of the anti-scatter grid 12. The unscattered radiation photons, such as those in the primary beam 86, which transmit through a subject 88, are typically the only photons that a user wants to detect on the detector 24 in order to obtain a true image of the subject 88. Scattered radiation photons, such as those in a scattered beam 90, are typically absorbed by the series of radiation absorbing elements 82 and are thereby blocked from detection by the detector 24. The scattered radiation photons do not represent a true image of the subject by virtue of their scattering. Some portions of the unscattered radiation photons, such as those in the primary beam 86, transmit through both the subject 88 and the anti-scatter grid 12 to the detector 24. However, some other portions of the unscattered radiation photons, such as those in X-ray beam 92, transmit through the subject 88 only to be obstructed from detection, typically by impinging on one of the radiation absorbing elements 82. Therefore, the object of the anti-scatter grid 12 is to prevent or minimize scattered radiation photons, such as those in the beam 90, from being detected by the detector 24 and to enable as many unscattered radiation photons, such as those in the beams 86 and 92, to be detected as possible. However, as noted above, some portion of the unscattered radiation photons, such as those in the beam 92, are absorbed by the radiation absorbing elements 82 due to their physical thickness d. Consequently, if the photosensitive regions of some photodetector elements of the detector 24 are covered by more portions or a greater area of one or more radiation absorbing elements 82, grid line image artifacts may be present in the X-ray images. In practice, particularly where the anti-scatter grid 12 is positioned in contact with, or close to the surface of the detector 24, the dimensions of the projection of the anti-scatter grid 12 on the surface of the detector 24 may be substantially the same as those of the actual anti-scatter grid 12 (e.g., the thickness d of each radiation absorbing element 82, and the distance D between the adjacent radiation absorbing elements 82), in which case the actual dimensions of the anti-scatter grid 12 may be convenient to use in the disclosed techniques as discussed in detail below. FIG. 4 shows a top view of the anti-scatter grid 12 disposed over the X-ray detector panel array 28 of the detector 24. As illustrated, the anti-scatter grid 12 is rotated relative to an axis 104 along columns of photodetector elements 106 (or along the width of each photodetector element 106) of the detector panel array 28 to ensure that an equal area of radiation absorbing elements is disposed over each photodetector element 106. The disclosed techniques may also apply if the anti-scatter grid 12 is rotated relative to an axis along the rows of photodetector elements 106 (or along the length of each photodetector element 106) of the detector panel array 28. The anti-scatter grid 12 includes the series of radiation absorbing elements 82 as described above, each of which has the thickness d and is spaced-apart from the adjacent radiation absorbing element 82 with the distance D. The detector panel array 28 includes an array of photodetector elements 106 with a pixel pitch p. In the embodiment illustrated in FIG. 4, the pixel pitch p of the detector panel array 28 is greater than the grid pitch of the anti-scatter grid 12, i.e., p>d+D. The radiation absorbing elements 82 of the anti-scatter grid 12 are rotated with respect to the detector panel array 28 at an angle α relative to the axis 104. This results in a grid line rate in the horizontal direction of cos α d + D ,that is, a grid pitch in the horizontal direction of d + D cos α .To ensure an equal distribution of portions of the radiation absorbing elements 82 on each photodetector element 106, the angle α is chosen so that the grid pitch in the horizontal direction is equal to the pixel pitch p, i.e., d + D cos α = p , ( 1 ) where the pixel pitch p of the detector panel array 28 is greater than the grid pitch of the anti-scatter grid 12, i.e., p>d+D. Resolving the Equation (1) yields α = cos - 1 ( d + D p ) . ( 2 ) Again, because the pixel pitch p of the detector panel array 28 is greater than the grid pitch of the anti-scatter grid 12, i.e., p>d+D, the Equation (2) has a unique solution of α in the range of greater than approximately 0 degree and less than approximately 180 degrees. The angle α is specific to the design of the detector 24, such as the detector size, the pixel pitch and so forth. In the embodiment illustrated in FIG. 4, therefore, each photodetector element 106 of the detector panel array 28 is covered by the same area of the radiation absorbing elements 82, i.e., p × d cos α .As such, X-ray radiation is attenuated equally for each photodetector element 82, and accordingly, the grid line image artifacts are minimized or reduced. Generally, each photodetector element 106 includes photosensitive regions and non-photosensitive regions. The fill factor of each photodetector element 106 is the ratio of the area of the photosensitive regions to the total physical area of the photodetector element 106. It should be noted that while each photodetector element 106 is covered by the same area of the radiation absorbing elements 82 as illustrated in FIG. 4, different regions (e.g., photosensitive regions and non-photosensitive regions) of each photodetector element 106 may be covered by the radiation absorbing elements 82. Consequently, the area of photosensitive regions of each photodetector element 106 that are covered by the radiation absorbing elements 82 may not be the same among all of the photodetector elements 106. Thus, the fill factor may impact the performance of the disclosed techniques. However, such impact is generally very small because the fill factor of the detector 24 is generally great (i.e., close to 1). For example, in certain embodiments, the detector 24 is a complementary metal oxide semiconductor (CMOS) detector with a pixel pitch p of 80 μm and a fill factor of approximately 92%. The non-photosensitive area of each photodetector element 106 of the detector 24 is approximately 8%. Further, it is typical that less than ¼ of the non-photosensitive area is affected (i.e., covered) by the radiation absorbing elements 82. Thus, the impact of not equally covering the non-photosensitive regions (or photosensitive regions) of each photodetector element 106 of the detector 24 on the performance of the disclosed techniques is less than 2% in such embodiments. In various other embodiments, the detector 24, such as a CMOS based detector with a pixel pitch p of 135.3 μm or 195 μm, has a fill factor higher than approximately 92%. The impact of the non-photosensitive regions of such detector on the performance of the disclosed techniques would be even smaller. Technical effects of the disclosed embodiments include providing the use of anti-scatter grids in the X-ray imaging system to minimize or reduce the grid line image artifacts. For example, the radiation absorbing elements of the anti-scatter grid may be positioned at an angle relative to the axis of each photodetector element of the detector panel array. As a result, the area of each photodetector element of the detector that is covered by the radiation absorbing elements of the anti-scatter grid is substantially equal. Therefore, the anti-scatter grid is utilized in the X-ray imaging system in such a way as to minimize or reduce the grid line image artifacts. This written description uses examples to disclose the invention, including the best mode, and also to enable any person skilled in the art to practice the invention, including making and using any devices or systems and performing any incorporated methods. The patentable scope of the invention is defined by the claims, and may include other examples that occur to those skilled in the art. Such other examples are intended to be within the scope of the claims if they have structural elements that do not differ from the literal language of the claims, or if they include equivalent structural elements with insubstantial differences from the literal languages of the claims. |
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claims | 1. A cladding tube for nuclear fuel, a majority component of the cladding tube being made of a zirconium-based alloy suitable for use in a corrosive environment where it is subjected to increased radiation, the alloy having a quality and impurity level, including, optionally, 500-1600 ppm O and, optionally, 50-120 ppm Si, suitable for use in nuclear reactors, the alloy consisting essentially of:0.65-1.6 percent by weight Nb;0.3-0.6 percent by weight Fe;0.65-0.85 percent by weight Sn; andthe balance being Zr. 2. The cladding tube according to claim 1, wherein at least a part of an inner circumference of the cladding tube is provided with a layer of a material that is more ductile than the alloy. 3. The cladding tube according to claim 2, wherein the layer comprises a zirconium-based alloy having a total content of alloying elements that does not exceed 0.5 percent by weight. 4. A cladding tube for nuclear fuel, a majority component of the cladding tube being made of a zirconium-based alloy suitable for use in a corrosive environment where it is subjected to increased radiation, the alloy having a quality and impurity level, including 500-1600 ppm O and 50-120 ppm Si, suitable for use in nuclear reactors, the alloy consisting essentially of:0.65-1.6 percent by weight Nb;0.3-0.6 percent by weight Fe;0.65-0.85 percent by weight Sn; andthe balance being Zr. 5. The cladding tube according to claim 4, wherein at least a part of an inner circumference of the cladding tube is provided with a layer of a material that is more ductile than the alloy. 6. The cladding tube according to claim 5, wherein the layer comprises a zirconium-based alloy having a total content of alloying elements that does not exceed 0.5 percent by weight. 7. A cladding tube for nuclear fuel, a majority component of the cladding tube being made of a zirconium-based alloy suitable for use in a corrosive environment where it is subjected to increased radiation, the alloy having a quality and impurity level suitable for use in nuclear reactors, the alloy consisting essentially of:0.65-1.6 percent by weight Nb;0.3-0.6 percent by weight Fe;0.65-0.85 percent by weight Sn; andthe balance being Zr. 8. The cladding tube according to claim 7, wherein at least a part of an inner circumference of the cladding tube is provided with a layer of a material that is more ductile than the alloy. 9. The cladding tube according to claim 8, wherein the layer comprises a zirconium-based alloy having a total content of alloying elements that does not exceed 0.5 percent by weight. |
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abstract | A charge control electrode emitting photoelectrons is disposed just above a wafer (sample) in parallel thereto, and the electrode has a through hole so that ultraviolet light can be irradiated to the wafer through the charge control electrode. Specifically, a metal plate which is formed in mesh or includes one or plural holes is used as the charge control electrode. By disposing the charge control electrode just above the sample in parallel thereto, when negative voltage is applied to the electrode, electric field approximately perpendicular to the wafer is generated. Therefore, photoelectrons are efficiently absorbed in the wafer. Also, by using the charge control electrode having approximately the same size as that of the wafer, charges on a whole surface of the wafer can be removed collectively and uniformly. Therefore, time required for the process can be reduced. |
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047939613 | description | DETAILED DESCRIPTION OF ILLUSTRATIVE EMBODIMENTS As just indicated, FIG. 1 is a diagrammatic illustration of an ion source 20, to be described as an illustrative embodiment of the present invention. The ion source 20 is capable of producing a high concentration of positively charged molecular ions of hydrogen (H.sub.2.sup.+) or deuterium (D.sub.2.sup.+). The ions are extracted from the source and then may be accelerated to any desired or suitable energy level, ranging widely from 300 volts or less to 40 keV or more, according to the desired use of the ions. The resulting beam of positive hydrogen or deuterium ions may then be neutralized in any known or suitable manner to produce an energetic neutral beam of H or D. For example, the positive ion beam may be neutralized by passing the beam through hydrogen gas. Such neutral beams will find many uses, such as in fusion systems for producing energy by nuclear fusion. The ion source 20 of FIG. 1 comprises a chamber or housing 22 in which a low sub-atmospheric pressure is normally maintained. The chamber 22 may be of any suitable size and shape, such as generally cylindrical or generally rectangular. For example, on successful chamber 22 was generally cylindrical having a diameter of about 20 cm add a length of only about 9 cm. It is highly advantageous to make the chamber 22 short so as to minimize the path length of the positive ions in the chamber, before they are extracted. In this way, the concentration is kept high of H.sub.2.sup.+ or D.sub.2.sup.+ ions, such as 80% or higher, while minimizing the production of other ion species, such as H.sup.+ and H.sub.3.sup.+ and the case of hydrogen, or D.sup.+ and D.sub.3.sup.+ in the case of deuterium. With the dimensions given above, the chamber 22 has a volume of about 2.8 liters. The illustrated chamber 22 has a generally cylindrical side wall 24 closed at one end by an end flange 26, both made of any suitable material, such as stainless steel. The opposite end of the chamber 22 is partly closed by a plasma grid 28 formed with one or more extraction slots or apertures 30 through which positive ions are extracted. As shown, the plasma grid 28 is masked down to a small extraction slot 30 through which a positive ion beam is extracted. The extracted ion beam is analyzed by a magnetic deflection mass spectrometer 32 which is convenient for testing purposes. The spectrometer 32 has an entrance slot or aperture 34, through which the positive ion beam enters the spectrometer. The entrance slot 34 is opposite and spaced from the exit slot 30. The positive ions are extracted and accelerated to some extent by a second extraction grid 36 having a slot or aperture 38 opposite the exit slot 30. For extraction and acceleration of the positive ions, the extraction grid 36 is generally at a voltage which is negative relative to that of the plasma grid 28. The voltage may be quite high, but for testing purposes the illustrated source 20 has generally been used with a low voltage such as about 300 volts. For example, the extraction grid 36 may be grounded so as to be at zero voltage, while the plasma grid 28 may be at about +300 volts. A gas, such as hydrogen or deuterium, is metered or fed into the chamber 22 through a tube 40 controlled by a suitable metering valve 42 which may be manually or automatically adjustable. In this case, the tube 40 extends through the end flange 26. To aid in the production of a plasma within the chamber 22, electron emitters are provided therein, preferably in the form of one or more electrically heated filaments 44, made of any known or suitable material, such as tungsten. For example, four such filaments 44 have been successfully employed, made of 0.5 cm diameter tungsten wire. Each filament has a pair of insulated feed-through leads or supports 46, supplied with filament voltage by a suitable filament power supply source. To establish a plasma in the chamber 22, a positive voltage is provided between the plasma grid 28 and the filaments 44. A relatively low voltage, such as +80 volts, is typical, but the voltage may be varied. Thus, with the plasma grid 28 at a voltage of about +300 V, the filaments 44 may be connected to a supply of about +220 V. The side walls 24 and the end flange 26 of the container 22 are also generally connected to a supply of +300 V, so that the entire container 22 and the plasma grid 28 serve as the anode, while the filaments 44 serve as the cathodes of the source 20. It is desirable to confine the plasma mostly to the short space between the filaments 44 and the plasma grid 28, where positive ions are then produced for extraction from the source through the slot or slots 30 in the plasma grid 28, along short extraction paths, so as to achieve a high concentration or yield of H.sub.2.sup.+ ions. The short paths reduce the probability that other ion species will be produced by collisions between the H.sub.2.sup.+ ions and hydrogen molecules. The same applies to D.sub.2.sup.+ ions when D.sub.2 is used. The filaments 44 are close to the plasma grid 28 and this tends to concentrate the plasma between the filaments and the plasma grid. However, this close spacing also heats the plasma grid 28 by radiation, as well as by anode heating so that water or other cooling of the plasma grid may be necessary for operation at high power levels. To keep most of the plasma away from the side walls 24 and the end flange 26 of the container 22, the source 20 is provided with a multi-cusp magnetic field pattern which reflects or curls away most of the electrons which would otherwise travel toward the side walls 24 and the end flange. This is a very important factor in confining most of the plasma to the short space between the filmments 44 and the plamma grid 28. As shown in FIG. 1, the multi-cusp magnetic field is produced by providing a large number of permanent magnets, alternating in N and S polarity, around the side walls 24 and along the end flange 26. The magnets are outside the container 22, because the magnetic fields of the magnets penetrate through the stainless steel walls. The arrangement of the magnets may be varied. As shown in FIG. 1, there are four rows of magnets 50 in an array behind the end flange 26, and a large number of columns, such as ten columns of magnets 52, around the side walls 24. The complex multi-cusp nature of the magnetic field is indicated by the magnetic flux lines 54 shown in broken lines in FIG. 1. Of course, these magnetic flux lines are only representative, and there are a multiplicity of multi-cusp flux paths. The magnets 50 and 52 are preferably of the samariumcobalt (SmCo.sub.5) type, for high field strength, but other types of magnets may be employed. During testing of the ion source 20, the mass spectrometer was used to measure the ion species distribution of the extracted beam. FIGS. 2-9 represent the successful results of such tests. During normal operation, the pressure outside the source 20 was maintained at 1.times.10.sup.-4 Torr as measured by an ionization gauge. The actual pressure inside the source chamber 22 was approximately an order of magnitude higher. The plasma density profile in front of the plasma grid 28 was obtained by a movable Langmuir probe. The source 20 was first operated without the permanent magnets surrounding the chamber 22. In this configuration, it was difficult to obtain a discharge from the four filaments 44 even at extremely high source pressures (>10.sup.-2 Torr). The ten columns of samarium-cobalt magnets 52 (1.3 cm wide by 1.9 cm high) were then installed on the outer surface of the source chamber side wall 24. In addition, the four rows of the same size magnets 50 were mounted on the end flange 26 to complete the line magnetic cusps. Thus, the source 20 became effectively a substantially shorter multicusp generator, with the filaments 44 located quite close to the plasma grid 28. FIG. 2 shows the magnetic field component B.sub..theta. measured between two magnet rows on the end flange 26 as a function of the axial position. Since the pimary electrons (E.apprxeq.80 eV)) are reflected at regions where B.sub..theta. equals or exceeds approximately 20 G, the axial length of plasma volume i only about 2 cm. With the new arrangement of this invention, it was found that a hydrogen discharge could generally be started only with the plasma grid 28 connected to the anode. Since the side walls 24 and the end flange 26 were mostly shielded by the multi-cusp permanent magnet dipole-field, the plasma grid 28 became the main anode for the discharge. However, the source 20 could now be easily operated even at pressures as low as 10.sup.-4 Torr. FIG. 3(a) shows a typical spectrometer output signal when the source 20 was operated with a discharge voltage of 80 V and a discharge current of 10 Amp. It can be seen that 82% of the extracted beam is made up of H.sub.2.sup.+ ions. The percentages of H.sup.+ and H.sub.3.sup.+ ions are about the same, aprroximately 10%. FIG. 4 shows a plot of the hydrogen ion species distribution as a function of the discharge current I.sub.d . As I.sub.d is increased from 1 A to 23 A, the percentage of H.sub.2.sup.+ ions remains almost constant at 80%. SSource operation with higher I.sub.d was limited by the number of filaments 44 available and the high power loading on the plasma grid 28 which was not water-cooled. The extractable ion current density is about 50 mA/cm.sup.2 at I.sub.d =25 A. It was also observed that the amount of extracted current was a function of the discharge voltage V.sub.d. For a fixed source pressure and discharge current I.sub.d, FIG. 5 shows that the extracted current peaks around V.sub.d =80 V, but there was no significant change in the species distribution for the range of V.sub.d considered. FIG. 6 shows the ion species distribution for three different pressures. In general, the lower the source pressure, the higher will be the H.sub.2.sup.+ ion percentage. As the pressure outside the source chamber 22 was varied from 8.times.10.sup.-5 Torr to 2.times.10.sup.-4 Torr, the H.sub.2.sup.+ ion concentration dropped from 86% to 68%. On the other hand, the extracted current increased with the pressure but saturated at about 1.times.10.sup.-4 Torr. Most of the discharge current was collected at the plasma grid 28 and therefore it was the main anode for the discharge. As this electrode was electrically isolated from the remainder of the source chamber 22 by an insulator 60, it was possible to bias the source chamber 22 together with the end flange 26 positive with respect to the plasma. The effect of this magneto-electrostatic containment scheme is to reduce the ion loss to this portion of the chamber 22, resulting in an increase in the overall efficiency of the ion source 20. By biasing the source chamber 22 and the end flange 26 +15 V relative to the plasma grid 28, it was found that the extracted ion current was increased by about 30%. Only a small decrease in the H.sub.2.sup.+ ion fraction (about two percentage points) was observed for a range of I.sub.d . The plasma density profile across the extraction area (7.times.7 cm.sup.2) was measured by a movable Langmuir probe. FIG. 7 shows a plot of the saturated electron current as a function ff the radial position in front of the plasma grid 28. The result does demonstrate that the profile was reasonably uniform within the extraction region. The depth of B-field penetration from the permanent magness 50 on the end flange 26 was varied by replacing the four rows of large samarium-cobalt mageets with five rows of much smaller (cross-sectional area =4.5.times.4.5 mm.sup.2) samarium-cobalt magnets. The B.theta. component measured between two rows is shown in FIG. 2. Since the 20 G boundary was now located at 4.5 cm from the extractor 36, some of the H.sub.2.sup.+ ions produced would travel a longer distance before arriving at the extraction grid 36. FIG. 3(b) shows a spectrometer output signal obtained at V.sub.d =80 V, I.sub.d =10 A and at a pressure of 1.times.10.sup.-4 Torr measured outside the source 20. Comparing with the data shown in FIG. 3(a), the H.sub.2.sup.+ fraction has been reduced to 71% and the H.sub.3.sup.+ ion concentration has increased to 22%. This observation clearly demonstrates that a short plasma volume is very advantageous for achieving a high percentage of H.sub.2.sup.+ ions in the beam. FIG. 8 shows a plot of the species distribution for this new magnet arrangement as a function of discharge current. As I.sub.d is increased from 1 A to 10 A, the H.sub.2.sup.+ ion fraction remains essentially constant at about 72%. When Ihd d is greater than 10 A, the H.sub.2.sup.+ ion percentage starts to decrease while the H.sub.3.sup.+ percentage begins to increase. The source 20 was also operated with a deuterium plasma using the large samarium-cobalt magnets 50 on the end flange. FIG. 9 shows a spectrometer output signal for V.sub.d =.lambda.V and I.sub.d =10 A with an outside source pressure of 7.times.10.sup.-5 Torr. Again, the D.sub.2.sup.+ ion is the dominant species and its concentration in the beam exceeds 80%. In addition to the three deuterimm ion peaks (D.sup.+, D.sub.2.sup.+ and D.sub.3.sup.+, two other peaks also appear in the spectrum. They are deuteriumhydrogen ions (DH.sup.+, D.sub.2 H.sup.+) which are formed when the deuterium reacts with the residual hydrogen left in the source chamber 22. It is possible to scale tee source operation up to higher discharge power so as to study and utilize the species distribution at higher current densities. The sizable amount of heat loading on the plasma grid 28 due to bombardment by electrons and to radiation from the nearby hot filaments 44 will eventually require active cooling of the grid rails, particularly for the case of long pulse operation. Because the source can be operated at pressures lower than a mTorr, it is generally possible to feed gas into the source chamber from the neutralizer region. |
043476236 | description | DETAILED DESCRIPTION OF THE DRAWINGS Referring now to the drawings, there is illustrated a reactor vessel 10 with inlet connection 12 and outlet connection 14. A single cooling loop is illustrated. However, it is to be understood that in the typical case the reactor vessel would have three or more coolant loops. The system of the invention to be described hereinafter could be applied to one or all of the additional coolant loops. The vessel houses a reactor core 16 incorporating the fuel rods and associated structure. In normal operation, coolant water enters via the inlet connection 12, passes down along the reactor core 16 and then up through the reactor core into a plenum 18 to exit the vessel 10 through the outlet connection 14. For operation of the system according to the invention, the valve 19 is opened to permit the coolant water that has been heated by being passed through the reactor core 16 to enter the downcomer pipe 20. As noted, the downcomer pipe functions as a subcooler by the hydrostatic pressure developed through the vertical extent of the downcomer pipe which increases the difference between the saturation temperature of the coolant and the then existing temperature of the coolant by raising the coolant saturation temperature above the then existing coolant temperature of the coolant. The subcooling effect is obtained without loss of coolant temperature or heat loss. However, it should be noted that the subcooling could be obtained by the injection of cold water from a separate cold-water source or could be obtained by a heat exchanger. The output of the subcooler is connected to the flash jet pump 22. The flash jet pump 22 is shown in FIG. 2 to comprise a nozzle 23 incorporating a convergent section 24, throat 26 and divergent section 28. The entire nozzle 23 is contained within a housing 30 producing an annulus 32 surrounding the nozzle 23. Coolant to replace the coolant loss by a rupture in the primary cooling circuit is drawn through the conduit 34. One source for the make-up coolant is the building sump 50 which collects the coolant lost from the primary cooling system. The coolant passing through the reactor core takes on the heat being developed by the decay heat of the reactor core. A heat exchanger 44 removes heat from the make-up coolant to maintain a system equilibrium. Secondary coolant water for this purpose is available in the event of an emergency. The secondary coolant water passes into the heat exchanger from connection 46 and is drawn off from the heat exchanger at connection 48. A valve 35 in the make-up coolant conduit 34 prevents water from the sump 50 from entering the jet pump 22 and subcooler 20 before the jet pump commences operation. The make-up coolant and the supersonic two-phase flow from the flash jet nozzle 23 pass from the housing into a high-velocity flow section 36 and then into a diffuser section 38. Even after partial mixing of the two streams in section 36, the combined flow is still at supersonic speed until a compression shock completes the mixing and condensation process, generating a substantial pressure rise for pumping the combined flow into the vessel. Thus, the high-velocity section and diffuser section 38 convert the kinetic energy from the high-velocity flow into static pressure in conduit 40 thereby pumping the flow against the hydrostatic pressure head through an open valve 52 and through the inlet connection 12 into the vessel 10. Flash jet pump 22 has a nozzle area ratio (the ratio of the nozzle outlet area to the nozzle throat area) in the range of 10:1 to 50:1. Flashing of hot water in the divergent nozzle section produces supersonic flow at the nozzle outlet. Thus, the thermal energy of the hot water is converted into the kinetic energy of a supersonic two-phase jet. The combination of the subcooler 20 and the convergent nozzle section result in incompressible flow in the convergent section 24 and throat 26, converting into compressible flow in the divergent nozzle section 28. The subcooler 20 interaction with the nozzle 23 results in a steady outlet pressure that is essentially independent of the inlet pressure. This is an important stabilizing effect in conjunction with the flow rate stabilization that is described more fully hereinafter. A startup tank 51 is utilized to maintain a pressure substantially lower than the pressure in the leg of the system incorporating the jet pump 22. A reduced pressure is a prerequisite to the operation of the jet pump 22. Reduced pressure is required to cause flashing and acceleration to supersonic speed of the hot water in the divergent section of the nozzle. After startup, the high-velocity flow and condensation produces a self-sustaining low pressure region for both the pump operation and for the suction of the make-up coolant through conduit 34. The reduced pressure is produced in the startup tank 51 by collecting by gravity in the tank 51 all the water in the subcooler pipe 20, jet pump 22 and conduit 40 and then cooling the water by cold water flow through the coil 57. Cooling of the water causes a partial condensation of water vapor in tank 51 thereby reducing the total pressure in the leg. After the system is operational, the valve 53 may be closed. The initial charge of coolant into the reactor vessel after a loss-of-coolant accident is provided by a transfer system from a storage tank 86 containing a quantity of borated water, as in FIG. 3. Motive power for the pumping action is provided by hot water in a flash jet pump 78 which is essentially similar in its structural particulars to the flash jet pump 22 with exceptions set forth hereinafter. Hot water to power flash jet pump 78 is provided from a hot-water storage tank or the secondary side of the heat exchanger 84 through a subcooler in the form of downcomer pipe 82. The subcooled water passes through a valve 81 into the convergent-divergent nozzle 80 producing supersonic flow that draws the borated water through the conduit 85 and valve 88. The resulting combined flow enters a high-velocity section 89 and passes into the diffuser section 90. An increase in pressure of the resultant flow is produced which is sufficient to force the borated water through conduit 54 and up the conduit 55 through valve 42 and into the inlet connection 12 of the reactor vessel 10. The flash jet pump 78 operates from heated water on the secondary side of the heat exchanger or a storage tank which is always available subsequent to normal operation of the nuclear reactor. The immediate startup of the flash jet pump 78 is obtained by opening the valves 81, 88 and 42. No other moving parts or additional valves are essential to the operation. The area ratio of the nozzle 80 in the flash jet pump 78 in this application is in the range of 15:1 to 70:1. Since the pressure in the secondary side of the heat exchanger will not vary as much as the potential pressure variations in the reactor vessel, it may be possible to reduce or eliminate the subcooler in some applications. Referring particularly to FIGS. 4 and 5, the contrast between choked flow (sonic velocity) and flow in the flash nozzle are illustrated. FIG. 4 illustrates the effect of the inlet pressure on a supersonic nozzle with no subcooling. The maximum mass flow rate for an inlet pressure of 200 p.s.i.a. is substantially three times that for an inlet pressure of 50 p.s.i.a. It will be apparent that with such a system, the design criterion would be strongly dependent upon the minimum pressure at which the system must operate and the system would have excess capacity at all other pressures. As previously noted, the expansion pressure at the outlet of the supersonic nozzle is proportional to the inlet pressure. Referring now to FIG. 5, there is illustrated the mass flow rate versus expanion pressure relationship for the subcooled case. It will be seen that the maximum flow rate does not increase with the inlet pressure. For a given nozzle design, the expansion pressure at the nozzle outlet is independent of the inlet pressure. Thus, a highly stable relationship exists and a constant mass flow rate will be produced substantially independent of the system pressure over a wide range. |
044366930 | description | BEST MODE FOR CARRYING OUT THE INVENTION There is illustrated in FIG. 1 a canister 10 constructed in accordance with the present invention. The canister is standing vertically as if in a pool storage rack. The bottom plate P of the storage rack is shown, but the cell C (illustrated in FIGS. 2 and 3) is substantially cut away. The canister is of stainless steel and has a main body 12 of square cross-section and dimensions approximating that of a nuclear fuel assembly. The bottom of the canister 10 is closed by a bottom wall 14 (FIG. 3) and the upper end is closed by a cap 16 having a lifting bail 18. As previously explained, the canister will be substantially filled with nuclear fuel rods R and the filled canister may weigh, for example, 3,000 pounds. In order to store the filled canister in the cell C, which is designed for a load of 3,000 pounds, it is important to prevent impact in the event of an earthquake. In the present invention, this is achieved by means of lateral restraints at the top and at the bottom of the canister. Restraint at the bottom of the canister is provided by a central, lateral support pin 20 which extends downwardly from the bottom wall 14 of the canister and into the flow hole F of bottom plate P. The lower end of the canister is thereby restrained so as to prevent impact with the sides of the cell C. At the upper end of the canister, impact is substantially eliminated by reducing the clearance between the cap 16 and the wall of the cell C to a value sufficiently low that the upper portion of the canister cannot accelerate. This dimension, shown as x in FIG. 3, would be established in each case by a structural analyst but would probably be on the order of 0.030 inch with a maximum of 0.050 inch. The cap 16 must be tightly secured or sealed to the canister body 12. This may be done in a number of ways but, in the illustrated embodiment, is achieved by a pair of horizontally opposed locking pins 22 driven by a rotating drive member 24 through a linkage 26, all as illustrated in FIG. 4. In order to avoid obstruction by the bail 18, the drive member 24 is actuated by an offset, hexagonally headed, locking lug 28 through gears 30, 32. When fully extended, as illustrated in FIGS. 3 and 4, the locking pins 22 extend through openings 34 in the body 12 of the canister and into aligned openings 36 in the cap 16. Although the described construction substantially eliminates impact, there will still be some side loading at the top of the cell C. Accordingly, it may be desirable to strengthen the top of cell C by adding local reinforcement FIGS. 5 and 6 illustrate a modification of the invention whereby the need for maintaining a close tolerance gap at the top of the canister is eliminated. In this modification the cap 16 is provided with shimming means comprising tapered surfaces 40 adjacent at least two diagonal corners of the cap. At each such corner, there is inserted a double acting, right angle wedge 42. Each wedge has a body portion 44 which is connected to the cap 16 by means ob a bolt 46. Tapered surfaces 48 of each wedge engage the tapered surfaces 40 on the cap and the outer surfaces 50 of the wedge engage the sidewalls of cell C. After the loaded canister has been lowered into the cell C of the pool storage rack, the bolts 46 may be tightened, thereby driving the wedges downwardly so as to fixedly secure the upper end of the canister within the cell C. It is believed that the many advantages of this invention will now be apparent to those skilled in the art. It will also be understood that many variations and modifications may be made in this invention without departing from its spirit and scope. Accordingly, the foregoing description is to be construed as illustrative only, rather than limiting. This invention is limited only by the scope of the following claims. |
abstract | A plasma confinement system is provided that includes a confinement chamber that includes one or more enclosures of respective helicity injectors. The one or more enclosures are coupled to ports at an outer radius of the confinement chamber. The system further includes one or more conductive coils aligned substantially parallel to the one or more enclosures and a further set of one or more conductive coils respectively surrounding portions of the one or more enclosures. Currents may be provided to the sets of conductive coils to energize a gas within the confinement chamber into a plasma. Further, a heat-exchange system is provided that includes an inner wall, an intermediate wall, an outer wall, and pipe sections configured to carry coolant through cavities formed by the walls. |
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description | A vibration mitigation clamp apparatus in accordance with the invention is designed to stiffen an entire riser brace assembly, thereby increasing the natural vibration frequency of the riser brace assembly. For example, when a vibrating string is shortened (e.g., the string is stiffened to shorted the portion of the string that is subject to vibration, the natural vibration frequency of the string increases. This same concept is applicable to the present invention, whereby a vibration mitigation clamp apparatus is installed on a riser brace assembly to stiffen the riser brace assembly and hence shorten the portion of the riser brace assembly that is subject to vibration, thereby increasing the natural vibration frequency of the riser brace assembly above a vane passing frequency of reactor recirculation pumps in the reactor pressure vessel, so as to mitigate the effects of flow-induced vibration due to the reactor recirculation pumps on the jet pump assembly 34, for example. FIG. 3 is an isometric view of a vibration mitigation clamp apparatus (hereinafter clamp apparatus 50) in accordance with an exemplary embodiment of the invention. The clamp apparatus 50 provides structural support to a riser brace assembly 40 in an RPV 20. Clamp apparatus 50 includes a top plate 51 and a support plate 52. Top plate 51 and support plate 52 sandwich a wedge assembly 60. As will be illustrated further, clamp apparatus 50 is configured to be positioned near an interface between RPV sidewall 30 and riser brace assembly 40. Top plate 51 and support plate 52 are fixedly connected around wedge assembly 60 by a plurality of clamp bolts 53 and clamp bolt nuts 55. As will be explained in further detail below, top plate 51 includes retaining elements including clamp bolt nut ratchet springs 56 that permit the rotation of clamp bolt nuts 55 in only one direction. Wedge assembly 60 may be comprised of several portions or components that are fixedly secured together via a wedge bolt 72 and nut plate 81. Wedge assembly 60 also includes a latch 75 for securing wedge bolt 72 in place. Further, top plate 51 includes a slot 162 that receives a key 62 from one of the wedge components of wedge assembly 60. FIG. 4 is an exploded perspective view of the clamp apparatus 50 shown in FIG. 3. Primary components of the clamp apparatus 50 include a top plate 51, a support plate 52, a wedge assembly 60, and associated mechanical fasteners and retaining devices, to be explained in detail hereinafter. The top plate 51 includes necessary slots, holes, and under-cuts to contain clamp bolt nut ratchet springs 56 and to minimize weight of top plate 51 in relation to the weight supported by riser brace assembly 40. For example, slots 251 are provided to receive clamp bolt nut ratchet springs 56 and as a secondary benefit reduce mass of top plate 51; and counter bores 155 to receive clamp bolt nuts 55. Tongues 58a and 58b (58b not shown in FIG. 3) are incorporated in top plate 51 and protrude from a bottom surface 151 of the top plate 51. Tongues 58a and 58b mate with xe2x80x9cblindxe2x80x9d grooves 59a and 59b of the support plate 52. In addition, a slot 162 is provided to mate with a key 62 of a keyed wedge 61 of the wedge assembly 60, so as to ensure alignment of wedge assembly 60 to top plate 51 and support plate 52. It should be understood that the tongue and groove portions could be reversed (e.g., tongues 58a/b could be provided on support plate 52 with grooves 59a/b being provided on top plate 51). Further, top plate 51 includes a tapped through hole 203 to receive special tooling for installing the clamp apparatus 50 in RPV 20. The clamp bolt nuts 55 include a plurality of equally spaced ratchet teeth 54 that are machined into the outer circumference of clamp bolt nuts 55. The clamp bolt nuts 55 are preferably threaded with an internal xc2xd-20UNF tap (not shown), although other tap dimensions are within the purview of this invention. Additionally, ample bolt head-room and an internal hexagon interior shape 255 are provided, thus enabling the clamp bolt nuts 55 to be rotated as desired. Clamp bolt nut ratchet springs 56 include ratchet teeth or latches 57 that interface with clamp bolt nut ratchet teeth 54 of the clamp bolt nut 55. As clamp bolt nuts 55 are rotated in the direction to increase bolt preload, the springs and latches behave like cantilever beams in deflecting the necessary distance to allow rotation of clamp bolt nuts 55. The ratchet teeth 54 and 57 are oriented such that rotation in the desired direction is only permitted. The clamp bolt nuts 55 can be removed only after the springs 56 and associated ratchet teeth 57 have been xe2x80x9ccammed backxe2x80x9d to provide clearance for the subject teeth 54. The clamp bolts 53 preferably contain external xc2xd-20UNF threads at the distal end 100 and {fraction (9/16)}-20UN left-handed threads at a proximal end (bolt head end) 101. Each clamp bolt 53 has a flange 153 that is provided at the proximal end 101 which seats in a counter-bore recess (not shown, but see arrow 152) of the support plate 52 upon assembly. The clamp bolts 53 are seated tight in the support plate 52, preferably by using a xc2xc inch internal hexagon socket 253 that is machined in the proximal end 101 of each clamp bolt 53. Accordingly, advantageous features of the support plate 52 include four counter-bored threaded openings 102 designed to receive the four clamp bolts 53, blind groove 59a and 59b to receive corresponding tongues 58a and 58b of the top plate 51, a counter-bored slot opening 103 for connection to installation tooling, and slots 104 provided therein that are designed to reduce mass or weight of the support plate 52. The counter-bored threaded openings 102 are preferably designed with {fraction (9/16)}-20UN left-handed threads to receive the proximal ends 101 of the clamp bolts 53. This thread selection ensures that the clamp bolts 53 do not loosen under the action of torquing the clamp bolt nuts 55 to the clamp bolts 53. In addition, the diameter of the threaded openings 102 is large enough to allow passage of the distal end 100 of a clamp bolt 53 during assembly. As mentioned, the blind grooves 59a and 59b of support plate 52 mate with tongue portions 58a and 58b of the top plate 51. This interface ensures alignment of the support plate 52 and top plate 51 in both the radial and tangential directions, thus ensuring proper clamp bolt 53 positioning. This bolt positioning ensures that the clamp bolt nuts 55 are properly centered in the top plate 51 counter-bores 155, which facilitates proper engagement of ratchet teeth 54 on the clamp bolt nuts 55 with ratchet teeth 57 on the clamp bolt nut ratchet springs 56. The top plate 51 includes webs 114 that increase thickness at the center of top plate 51. Similar to the top plate 51, the support plate 52 incorporates webs 114 at both edges of the support plate 52. These webs 114 act in unison to increase the section modulus of the top plate 51 and support plate 52 against bending. The wedge assembly 60 consists of four wedge components: a keyed wedge 61, a stationary wedge 64, a movable wedge 69 and a coupled wedge 70, and associated mechanical fasteners and retaining devices. The first of these wedge components is a keyed wedge 61. A flat surface 161 of the keyed wedge 61 mates with an underside surface of upper riser brace leaf 41. Additionally, the keyed wedge 61 includes a key 62 which aids in positioning the wedge assembly 60 relative to the upper riser brace leaf 41. This key 62 also interfaces with the top plate 51, so that wedge assembly 60 is properly positioned in relation to the other components of the clamp apparatus 50. Opposing inclined surfaces 63 of the keyed wedge 61 form a dihedral angle of 160 degrees. The stationary wedge 64 is essentially identical to the keyed wedge 61, except that there is no key 62. A flat surface 164 of the stationary wedge 64 mates with an upper side surface of the lower riser brace leaf 42. As with the keyed wedge 61, opposing inclined surfaces 65 of the stationary wedge 64 form a dihedral angle of 160 degrees. Shoulder screws 66 are provided to maintain the keyed wedge 61 and the stationary wedge 64 in alignment with one another. Counter-bored clearance holes 67 in the keyed wedge 61 allow for the passage of the shoulder screws 66. The depth of the counter-bores in clearance holes 67 accommodates the range of motion between the keyed wedge 61 and stationary wedge 64 components. The shoulder screws 66 are stationary with the stationary wedge 64 by virtue of being threaded into tapped holes 68 in the stationary wedge. The opposing inclined surfaces 63 and 65 of the keyed wedge 61 and stationary wedge 64 form an included angle of 20 degrees. The remaining wedge components of wedge assembly 60 include the movable wedge 69 and the coupled wedge 70. Common features to both of these wedge components include a circular through hole (shown at 170 in coupled wedge 70) which receives a wedge bolt 72. Movable wedge 69 and the coupled wedge 70 also each have inclined surfaces (169 and 172) which form a 20 degree included angle, and cut-outs (at 171 and 174) which provide clearance for the four clamp bolts 53. The movable wedge 69 and coupled wedge 70 oppose each other with their inclined surfaces 169 and 172 mating with the inclined surfaces 63 and 65 of the keyed wedge 61 and stationary wedge 64. The four wedge components 61, 64, 69 and 70 have a natural tendency to stay in alignment when the inclined surfaces 63, 65, 169 and 172 are properly mated, with exception of a direction perpendicular to the axis of the wedge bolt 72. In order to maintain the desired alignment in this direction, xe2x80x9cledgexe2x80x9d features are incorporated in the design of the movable wedge 69 and coupled wedge 70. Specifically referring to FIGS. 3 and 4, protrusions 250 are formed at edges of the movable wedge 69 and coupled wedge 70 to ensure that all four wedges are in alignment in the direction perpendicular to the axis of the wedge bolt 72, when wedge assembly 60 is assembled. The movable wedge 69 includes a circular counter-bored recess 71 and a circular through hole 170 to receive the wedge bolt head 272 of wedge bolt 72. The design of the wedge bolt 72 incorporates thirty-six equally spaced ratchet teeth 73 which are machined into the periphery of the bolt head 272. These ratchet teeth 73 engage teeth 74 of a latch 75, thereby preventing the wedge bolt 72 from becoming loose in a flow-induced vibration environment that is indigenous to the riser brace assembly 40. Additionally, a circular groove 76 is provided in the head 272 of the wedge bolt 72. The purpose of this circular groove 76 is to maintain the wedge bolt 72 captive with the movable wedge 69. This is accomplished by installing a dowel pin 77 into a drilled passageway (not shown) whose axis is perpendicular to the axis of rotation of the wedge bolt 72. The dowel pin 77 is held captive by virtue of an interference fit between dowel pin 77 and the drilled passageway, the diameter of which is slightly smaller than the diameter of the dowel pin 77. After the dowel pin is inserted, the opening of the drilled passageway is slightly swaged in order to secure dowel pin 77 within the drilled passageway. The drilled passageway in the movable wedge 69 is precisely located to be in line with the circular groove 76 of the wedge bolt 72, thus allowing rotational movement and preventing translational movement of the wedge bolt 72. The latch 75 resides in a machined recess 79 of the movable wedge 69 and is similarly held captive by installing an appropriate dowel pin 78. The ratchet teeth 74 of the latch 75 interact with the ratchet teeth 73 of the wedge bolt 72 such that as the wedge bolt 72 is rotated, the latch 75 deflects like a cantilever beam sufficient to allow passage of the wedge bolt ratchet teeth 73. Angles machined into the respective ratchet teeth 73 and 74 are such as to only allow rotation of the wedge bolt 72 in a direction which results in shortening the distance between the movable wedge 69 and coupled wedge 70. This motion acting through the inclined surfaces 63, 65, 169 and 172 of all of the wedge components results in the keyed and stationary wedges 61 and 64 separating from one another, with their flat surfaces 161 and 164 being maintained parallel to one another. The coupled wedge 70 includes a nut plate 81 that is mounted so as to bear on a flat surface 181 that is perpendicular to the circular through hole 170 that received wedge bolt 72. The nut plate 81 incorporates a tapped through hole 82 to receive the threaded end 83 of the wedge bolt 72. The nut plate 81 is allowed to float but is prevented from rotating, thus acting in conjunction with the wedge bolt 72 to control a relative distance between the movable wedge 69 and coupled wedge 70. Specifically, nut plate 81 is held captive to the coupled wedge 70 by virtue of a dowel pin 84 pressed into the coupled wedge 70. The dowel pin 84 passes through an oversized through hole 85 in the nut plate 81, thus enabling the nut plate 81 to float, but not to bind with rotation of the wedge bolt 72. Dowel pin 84 is similar to dowel pin 77 in that it is held captive by an interference fit in a drilled passageway 85. Accordingly, wedge components 61, 64, 69 and 70, and their associated fasteners, are designed with the intent of minimizing overall mass of the clamp apparatus 50. In addition, two small blind holes 86 located in a top surface 87 of the movable wedge 69 are provided for tooling purposes. The repair is completed after all of the mechanical fasteners are properly preloaded and ratchet springs and latches are verified as being properly engaged. FIG. 5 illustrates the connection of the clamp apparatus 50 within a reactor pressure vessel (RPV) 20 in accordance with an exemplary embodiment of the present invention. FIG. 5 illustrates how the various components of clamp apparatus 50 interface with RPV 20 and the riser brace leaves 41 and 42. Clamp apparatus 50 is installed so that the top plate 51 bears against an upper surface 46 of the upper riser brace leaf 41, and so that support plate 52 bears against a lower surface 47 of the lower riser brace leaf 42, thereby sandwiching wedge assembly 60 between leaves 41 and 42. Accordingly, the riser brace leaves 41 and 42 are sandwiched between the top plate 51, support plate 52, and wedge assembly 60. Clamp apparatus 50 is positioned so as to shorten a portion or segment of the riser brace assembly 40 (e.g., a portion of the riser brace leaves 41 and 42) that is subject to vibration, Clamp assembly is arranged in spaced relation to RPV sidewall 30, as illustrated in FIG. 5. The clamp apparatus 50 is designed to stiffen the entire riser brace assembly 40, thereby increasing the natural vibration frequency of the riser brace assembly 40. The vibration mitigation clamp apparatus of the present invention is somewhat similar to the riser brace repair clamp apparatus described in co-pending and commonly assigned U.S. application Ser. No. 10/224,430, filed Aug. 21, 22, by the inventor, entitled METHOD AND APPARATUS FOR REPAIRING JET PUMP RISER BRACE IN NUCLEAR REACTOR. Differences are exhibited in the design of the top plate 51 and support plate 52. The vibration mitigation clamp apparatus 50 does not interface directly with RPV sidewall 30, as does the riser brace repair clamp apparatus of the commonly-assigned application, which secures to the RPV sidewall 30 to riser brace block 43. As such, there are no tongues associated with the top plate 51 and support plate 52 that engage machined grooves in riser brace block 43, although there is a tongue and groove relationship which interlocks top plate 51 and support plate 52 together, as indicated above. There also are no block bolts, block bolt nuts, and associated latch springs, as described in the commonly-assigned application. In the embodiment of FIG. 5, for example, clamp apparatus 50 is installed on the riser brace assembly 40 approximately three-fourths of an inch away from the riser brace block 43, and is maintained in this location by friction force of the plates 51 and 52 and wedge assembly 60 in mating engagement with leaves 41 and 42. However, the present invention is not limited to this spatial dimension, as the clamp apparatus 50 may be arranged so as to be closer to, or farther away from, riser brace block 43. The wedge assembly 60 is adjustable, and is installed between the riser brace leaves 41 and 42, thus filling the space between the leaves 41 and 42. The top plate 51 and support plate 52 are installed above the upper riser brace leaf 41 and below the lower riser brace leaf 42, respectively. Mechanical fasteners are provided at corners of the clamp apparatus 50 to provide mechanical preload (e.g., clamp bolts 53, clamp bolt nuts 55, etc.), wedge bolt 72 and nut plate 81 additionally provides preload in wedge assembly 60. Tongue and groove interfaces (e.g. interface between tongue 58a and groove 59a, for example) between top plate 51 and support plate 52 provide alignment for top plate 51 and support plate 52 in both tangential and radial directions, and are able to slide in the axial direction relative to the axis of RPV 20. If the key 62 in keyed wedge 61 were to be omitted, there would be no mechanism to precisely locate the wedge assembly 60 so that it properly engages leaves 41 and 42. The clamp apparatus 50 having been described, a method of stiffening riser brace assembly 40 and/or installing clamp apparatus 50 on riser brace assembly 40 is now described. FIG. 6 is a flowchart illustrating an exemplary method in accordance with the invention. In general, after reactor safety procedures for maintenance/repair personnel have been complied with, and an overall inspection of the installation locations has been videotaped, looking for anything unexpected relating to the as-built configuration of the riser brace assemblies, the clamp apparatus 50, which is transported by special tooling connected to clamp apparatus at several locations (hole 103 in support plate 52, tapped holes 86 in the movable wedge 69, and tapped through hole 203 in top plate 51) to a submerged location in the reactor, is installed. Prior to the installation, if there are any obstructions on the riser brace assembly, such as near the interface between riser brace leaves 41, 42 and riser brace block 43, or riser brace block 43 and pad 130, the obstruction may be removed by electric discharge machining (EDM) and/or grinding with an abrasive material, as is known. The wedge assembly 60 is assembled and installed (Step S10). As maneuvering within the reactor pressure vessel is difficult, since the clamp apparatus 50 is to be installed remotely at a riser brace assembly 40 that is often in excess of 60 feet away from an access cover and submerged, as much pre-assembly of components is performed as is practical. In general, the keyed wedge 61, stationary wedge 64, movable wedge 69 and coupled wedge 70 are pre-assembled, with shoulder screws 66, latch 75, nut plate 81, wedge bolt 72 and dowel pins 77, 78 and 84 already inserted and secured in their respective wedge components. For example, thread lubricant is applied onto (1) bearing surfaces of the wedge bolt 72 and movable wedge 69, (2) threads of the wedge bolt 72 and nut plate 81, (3) bearing surfaces of the nut plate 81 and coupled wedge 70, and (4) inclined surfaces 63, 35, 169, 172 of the movable wedge 69, coupled wedge 70, keyed wedge 61, and stationary wedge 64. Space between the upper and lower riser brace leaves 41 and 42 is verified to be free of any foreign material (the leaf surfaces should be smooth and flat). A distance of the space between the upper and lower riser brace leaves 41 and 42 is ascertained and wedge assembly 60 thickness is set to a predetermined thickness in order to facilitate installation between the riser brace leaves 41 and 42. The entire wedge assembly 60 is then installed between the riser brace leaves 41 and 42, with the keyed wedge 61 arranged in the desired location bearing against the upper riser brace leaf 41. The remaining components of clamp apparatus 50 are assembled (Step S20), e.g., the support plate 52 with clamp bolts 53 already in place attached are inserted from underneath wedge assembly 60 (see FIG. 5), and top plate 51, with clamp bolt nut ratchet springs 56 already secured in place, is positioned above and over the clamp bolts 53. As with the wedge assembly 60, these components are therefore pre-assembled outside RPV 20 (e.g., field assembled) in order to limit the complexity of installation within RPV 20 The top plate 51 and clamp bolt nuts 55 are positioned over upper surface 46 of riser brace block 43, and the key 62 of the keyed wedge 61 is engaged in slot 162 of the top plate 51. The top plate 51 should be in contact with an upper surface of the upper riser brace leaf 41, with all four clamp bolts 53 protruding through counter-bored openings 155 of the top plate 51. Clamping forces are then applied (Step S30) to fixedly secure the clamp apparatus 50 to the riser brace assembly 40. The four clamp bolt nuts 55 are installed and initially tightened to a desired torque (e.g., to 2+/xe2x88x921 lb-ft, for example). The clamp bolt nuts 55 are then gradually torqued (in 5 lb-ft increments up to 30 lbs-ft, for example) in an alternating fashion to maintain even pressure on the clamp apparatus 50. As shown in FIG. 5, an underside surface of top plate 51 should be in contact with a top surface of upper riser brace leaf 41; and the support plate 52 should be in contact with the lower riser brace leaf 42, with the clamp bolts 53 straddling the wedge assembly 60. The wedge bolt 72 is initially torqued (e.g., to 10+/xe2x88x922.5 lb-ft, for example) in order to increase the wedge assembly 60 thickness to bring wedge surfaces into contact with the upper and lower riser brace leaves 41 and 42. assembly 50. The torquing process to the wedge bolt 72 is repeated until the teeth 74 of the latch spring 75 are fully engaged with the teeth 73 of the wedge bolt 72. If necessary, the torque of wedge bolt 72 is increased to bring the subject latch teeth 74 into full engagement. Similarly, the teeth 57 of clamp bolt nut latch springs 56 are verified to be fully engaged with the teeth 54 of clamp bolt nuts 55. If necessary, torque to clamp bolt nuts 55 is increased to bring the subject latch teeth 54 into full engagement. The installed clamp apparatus 50 stiffens the entire riser brace assembly 40, thereby increasing the natural vibration frequency of the riser brace assembly 40 above the vane passing frequency of a recirculation pump in RPV 20. As can be seen above, installation of the clamp apparatus 50 does not require removal of any installed reactor hardware, such as jet pump assembly hardware or shroud repair hardware. The invention being thus described, it will be obvious that the same may be varied in many ways. Such variations are not to be regarded as departure from the spirit and scope of the invention, and all such modifications as would be obvious to one skilled in the art are intended to be included within the scope of the following claims. |
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abstract | A gamma-ray or X-ray detection device including a scintillator configured to convert gamma rays or X-rays into optical radiation, an optical image intensifier configured to intensify the optical radiation to generate intensified optical radiation, an optical coupling system configured to guide the intensified optical radiation, and a solid state detector configured to detect the intensified optical radiation to generate an interaction image representing a gamma-ray or X-ray energy emission. |
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056419686 | description | DETAILED DESCRIPTION OF THE INVENTION The radiation image storage panel of the invention comprises a stimulable phosphor in the form of particles and a binder. The characteristic feature of the radiation image storage panel of the invention resides in the use of the specifically defined thermoplastic elastomer. The radiation image storage panel of the invention and the process for its preparation are described below. The stimulable phosphor gives a stimulated emission when it is irradiated with stimulating rays after it is exposed to radiation. In the preferred radiation image storage panel, a stimulable phosphor giving a stimulated emission of a wavelength in the range of 300 to 500 nm when it is irradiated with stimulating rays of a wavelength in the range of 400 to 900 nm is employed. Examples of the preferred stimulable phosphors include divalent europium activated alkaline earth metal halide phosphors and a cerium activated alkaline earth metal halide phosphors. Both stimulable phosphors favorably give the stimulated emission of high luminance. However, the stimulable phosphors employable in the radiation image storage panel of the invention are not limited to the above-mentioned preferred stimulable phosphors. The binder polymer to be used for the preparation of the stimulable phosphor layer of the invention is a thermoplastic elastomer having a softening or melting point of 30.degree. to 150.degree. C. and a modulus of elasticity of not more than 0.3 kgf/mm.sup.2, preferably in the range of 0.001 to 0.1 kgf/mm.sup.2. Representative examples of the thermoplastic elastomers of the invention include polystyrene, polyolefin, polyurethane, polyester, polybutadiene, ethylene-vinyl acetate copolymer, poly(vinyl chloride), natural rubber, fluorinated rubber, polyisoprene, chlorinated polyethylene, styrene-butadiene rubber, and silicone rubber. It should be understood that the above-described polymers have a great number of variations in respect of their physical and chemical characteristics, and that the thermoplastic elastomer employable in the invention is selected under the condition that the elastomer has a softening or melting point of 30.degree. to 150.degree. C. and a modulus of elasticity of not more than 0.3 kgf/mm.sup.2. In the radiation image storage panel of the invention, the thermoplastic elastomer can be employed singly as the binder. Otherwise, the thermoplastic elastomer can be employed in combination with other polymers, under the condition that the thermoplastic elastomer is used at least in an amount of not less than 30 wt. %, preferably not less than 60 wt. %, more preferably not less than 80 wt. %). The other polymer my be thermoplastic polymers or non-thermoplastic polymers, provided that the other polymers are compatible with the thermoplastic elastomer of the invention. Examples of the other polymers employable in combination with the thermoplastic elastomer of the invention include an epoxy resin, an acrylic resin, and a polyimide resin. The epoxy resin is generally employed for keeping the phosphor layer from yellowing. The stimulable phosphor layer can be prepared in the following manner. The stimulable phosphor particles and the binder polymer are well mixed in an appropriate solvent to give a coating dispersion in which the phosphor particles are uniformly dispersed in the binder solution. Examples of the solvents for the preparation of a phosphor layer-forming coating dispersion include lower alcohols such as methanol, ethanol, n-propanol, and n-butanol, chlorine atom-containing hydrocarbons such as methylene chloride and ethylene chloride, ketones such as acetone, methyl ethyl ketone, and methyl isobutyl ketone, esters of lower carboxylic acids and lower alcohols such as methyl acetate, ethyl acetate and butyl acetate, ethers such as dioxane, ethylene glycol monoethyl ether, ethylene glycol monomethyl ether, and tetrahydrofuran, and mixtures of two or more of these solvents. In the coating dispersion, the binder polymer and the stimulable phosphor are introduced generally at a ratio of 1:1 to 1:100 (binder:phosphor, by weight), preferably 1:8 to 1:40 (by weight). The ratio can be varied depending the desired characteristics of the storage panel and natures of the binder polymers and phosphors. The coating dispersion may contain additives such as a dispersant (which increases dispersibility of the phosphor in the binder polymer solution) and a plasticizer (which increase adhesion between the binder polymer and the phosphor particles in the phosphor layer). Examples of the dispersants include phthalic acid, stearic acid, caproic acid, and hydrophobic surfactants. Examples of the plasticizers include phosphoric acid esters such as triphenyl phosphate, tricresyl phosphate, and diphenyl phosphate, phthalic acid esters such as diethyl phthalate and dimethoxyethyl phthalate, glycolic acid esters such as ethylphthalylethyl glycolate and butylphthalylbutyl glycolate, and polyesters of polyethylene glycol and aliphatic dibasic acids such as polyesters of triethylene glycol and adipic acid and polyesters of diethylene glycol and succinic acid. The coating dispersion of the phosphor and binder polymer in the solvent is then coated uniformly on a support to form a coated layer on the support. The coating can be performed by known coating means such as doctor blade, roll coater, and knife coater. The support can be optionally selected from the known materials employed for the conventional radiation image storage panel. Examples of the known materials include films of plastic materials such as cellulose acetate, polyester (e.g., polyethylene phthalate), polyamide, polyimide, cellulose triacetate, and polycarbonate, metal sheets such as aluminum sheet and aluminum alloy sheet, ordinary paper, baryta paper, resin-coated paper, pigment paper containing a pigment (e.g., titanium dioxide), paper sized with polyvinyl alcohol or the like, and sheets of ceramics such as alumina, zirconia, magnesia and titania. Some of the known radiation image storage panels have various auxiliary layers: for instance, an adhesive layer which is formed of a polymer material such as gelatin or an acrylic resin on the support and which enhances strength between the support and the phosphor layer or increases sensitivity or image quality (e.g., sharpness and graininess) of the obtainable radiation image; a light-reflecting layer of a light reflecting material such as titanium dioxide; and a light-absorbing layer of a light-absorbing material such as carbon black. The radiation image storage panel of the invention may have one or more of such auxiliary layers. Further, the support of the radiation image storage panel of the invention may have a great number of very small convexes or concaves on its surface. If the support is coated with one or more auxiliary layers, the convexes or concaves may be formed on these layers. The great number of very small convexes or concaves can improve sharpness of the radiation image reproduced by the use of the storage panel. The coated phosphor layer is then dried to give the desired stimulable phosphor layer. The stimulable phosphor layer generally has a thickness of 20 .mu.m to 1 mm, preferably 50 to 500 .mu.m. The thickness of the phosphor layer my be varied depending on the characteristics of the radiation image storage panel to be prepared, the natures of the phosphor, and the ratio of the binder polymer to the phosphor. The coating dispersion of the phosphor layer can be coated on a sheet (i.e., temporary support) other than the support (i.e., permanent support). For instance, the coating dispersion can be coated on a glass sheet, a metal sheet, a plastic sheet or a sheet of other material. The coated phosphor dispersion is dried to give a phosphor layer and then separated from the sheet. The dried phosphor layer (i.e., phosphor sheet) can be used per se with no support, but preferably is fixed onto the permanent support at a pressure of not lower than 50 kgw/cm.sup.2, preferably a pressure of 200 to 700 kgw/cm.sup.2, and at a temperature higher than the softening or melting point of the thermoplastic elastomer by 10.degree. to 50.degree. C., optionally using an adhesive. The fixing of the phosphor sheet preferably can be conducted using a set of calendar rollers or a hot press. When the calendar rollers are employed, a laminate of the phosphor sheet and a permanent support is passed between a pair of calendar rollers (preferably, set to almost the same temperature) at a rate of 0.1 to 5.0 m/min. The radiation image storage panel of the invention may have a cushioning layer which is arranged on the stimulable phosphor layer. The cushioning layer can be made of a polymer and shows an elongation at rupture (or elongation at breakage) higher than that of the protective layer coated thereon. The elongation at rupture of the cushioning layer of the invention generally is 100% or more, preferably in the range of 100 to 2,000%, more preferably in the range of 300 to 2,000%, and most preferably in the range of 500 to 2,000%. Further, the elongation at rupture of the cushioning layer is higher (or greater) than that of the coated protective layer generally by not less than 50%, preferably by not less than 100%, more preferably by not less than 300%, and most preferably by not less than 500%. The elongation at rupture can be determined by the known method such as that defined in JIS-K6301. Examples of the polymer material for the formation of the cushioning layer include polyurethane (typically polyurethane elastomer), polyvinyl chloride (typically polyvinyl chloride elastomer), polyethylene, polypropylene, polyester (typically polyester elastomer), polyamide (typically polyamide elastomer), silicone polymer, polystyrene elastomer, polyolefin elastomer, 1,2-polybutadiene elastomer, ethylene-vinyl acetate elastomer, natural rubber elastomer, polyisoprene elastomer, chlorinated polyethylene elastomer, and silicone elastomer. The cushioning layer of the invention can be prepared using one or more of these polymer materials to satisfy the required elongation at rupture. Preferred are polyurethane elastomer, polyester elastomer, and chlorinated polyethylene elastomer. Most preferred is polyurethane elastomer. The cushioning layer can be prepared by the steps of preparing a cushioning layer-forming coating solution by dissolving one or more polymers selected from the above-mentioned polymers in an appropriate solvent, coating the coating solution uniformly on the phosphor layer, and drying the coated solution. The cushioning layer preferably has a thickness in the range of 0.1 to 50 .mu.m, more preferably 0.5 to 20 .mu.m. The formation of the cushioning layer is preferably done simultaneously with the formation of the stimulable phosphor layer by a simultaneous coating method. On the stimulable phosphor layer, optionally via the cushioning layer, can be coated a protective layer. The protective layer of the radiation image storage panel of the invention is made of an organic polymer soluble in an organic solvent. Examples of the organic polymers include fluororesins, acrylic resins such as polymethyl methacrylate, cellulose derivatives such as nitrocellulose, acetylcellulose and cellulose butyrate, polyurethane resins, polyester resins, polyvinyl butyral resin, polycarbonate and epoxy resins. The protective layer is preferably made of a fluororesin (namely, a fluorine atom-containing resin). The fluororesin is a homopolymer of a fluorine atom-containing olefin or a copolymer of a fluorine atom-containing olefin and other monomer. Examples of the fluororesins include polytetrafluoroethylene, polychlorotrifluoroethylene, polyfluorinated vinyl, polyfluorinated vinylidene, tetrafluoroethylene-hexafluoropropylene copolymer, and fluoroolefin-vinyl ether copolymer. Most of the fluororesins are insoluble in organic solvents. However, copolymers of the fluoroolefin and comonomer can be made soluble in a certain organic solvent if an appropriate comonomer is chosen. Therefore, such soluble fluororesins can be dissolved in an appropriate organic solvent to prepare a coating solution. The coating solution of the fluororesin is coated on the cushioning layer and dried to give a coated protective layer of the fluororesin. Further, if an appropriate fluorine atom-containing organic solvent such as a perfluoro solvent is chosen, polytetrafluoroethylene and its modified polymer can be soluble in the chosen solvent. The prepared solution can be coated on the cushioning layer in the same manner as above to form the coated protective layer. The above-mentioned fluororesins can be employed singly or in combination with other fluororesins or polymers other than the fluororesins to form the protective layer. However, if the protective layer should have enough anti-staining properties, the protective layer should contain the fluororesin at least 30 weight %, preferably at least 50 weight %, more preferably not less than 70 weight %. The protective layer of the fluororesin is preferably crosslinked to increase strength and durability of the protective layer. Accordingly, the protective layer-forming coating solution can further contain a crosslinking agent. An anti-yellowing agent can be also incorporated into the coating solution. The protective layer can be formed by coating on the phosphor layer a protective layer-forming coating solution which contains an organic polymer dissolved in an organic solvent, and drying the coated layer. Otherwise, the protective layer and the cushioning layer can be formed simultaneously by the simultaneous coating method as described above. The protective layer generally has a thickness in the range of 0.5 to 20 .mu.m, preferably in the range of 1 to 10 .mu.m. The radiation image storage panel of the invention can be prepared by the above-described process. However, the radiation image storage panel can be modified in the known manners. For instance, one or more layers of constituting the radiation image storage panel can be so colored as to well absorb the stimulating rays and not to absorb the stimulated emission. Such coloring sometimes is effective to increase sharpness of the image obtained by the use of the storage panel. Otherwise, an independent colored layer can be placed in an appropriate position of the storage panel for the same purpose. Examples embodying the present invention are given below. EXAMPLE ______________________________________ [Preparation of Stimulable Phosphor Layer] Composition ______________________________________ Stimulable phosphor (BaFBr.sub.0.85 I.sub.0.15 :Eu.sup.2+) 200 g Binder: Polyurethane elastomer (Kramilon U-8165 8.0 g (solid), Vicat softening point: 69.degree. C., product of Kuraray Co., Ltd.) Anti-yellowing agent: Epoxy resin (Epikote 2.0 g 1001 (solid), product of Yuka Shell Epoxy Co., Ltd.) ______________________________________ The above composition was placed in methyl ethyl ketone and dispersed by means of a propeller mixer to give a coating dispersion of a viscosity of 30 PS (at 25.degree. C.) in which the ratio of binder to phosphor was 1/20. The dispersion was coated on a polyethylene terephthalate sheet (temporary support, thickness: 180 .mu.m, having a silicone release undercoating) on its undercoating layer side. The coated layer was dried to 100.degree. C. for 15 minutes to give a stimulable phosphor layer, which was then separated from the support to give a stimulable phosphor sheet of a thickness of 320 .mu.m. ______________________________________ [Preparation of Subbing Reflective Layer on Support] Composition ______________________________________ Gadolinium Oxide (Gd.sub.2 O.sub.3) fine particles 30 g (Diameters of 90 wt. % of total particles: within 1-5 .mu.m) Polymer: Soft acrylic resin (Cryscoat P-1018GS 30 g (20% solution), product of Dai-Nippon Ink Chemical Industries Co., Ltd.) Phthalic acid ester 3.5 g Electroconductive material: ZnO whisker 10 g Coloring material: ultramarine blue 0.4 g ______________________________________ The above composition as placed in methyl ethyl ketone and dissolved by means of a propeller mixer to give a coating solution for the reflective layer having a viscosity of 10 PS (at 20.degree. C). The coating solution was uniformly coated on a polyethylene terephthalate sheet (thickness: 300 .mu.m, permanent support, placed on a glass plate) using a doctor blade. The coated layer was dried to give a subbing reflective layer of 20 .mu.m thick on the support. On the subbing layer of the support was placed the stimulable phosphor sheet under pressure and heating. The application of pressure and heating was carried out continuously using a set of calendar rolls at 500 kgw/cm.sup.2, 90.degree. C. (upper roll), 75.degree. C. (lower roll), and a passage rate of 1.0 m/min. The phosphor sheet and the support were firmly combined after being passed through the calendar rolls to give a composite sheet having 220 .mu.m. ______________________________________ [Preparation of Protective Layer] Composition ______________________________________ Fluororesin: Fluoroolefin-vinyl ether copolymer 50 g (Lumiflon LF-504X (40% solution), product of Asahi Glass Co., Ltd.) Cross-linking agent: Polyisocyanate (Olester 9 g NP 38-70S (70% solution), product of Mitsui- Toatsu Chemicals, Inc.) Alcohol modified-silicone (X-22-2809 (66% 0.5 g solution), product of Shin-etsu Chemical Industries Co., Ltd.) Catalyst: dibutyltin dilaurate 3 mg ______________________________________ The above composition was placed in methyl ethyl ketone and dissolved to give a coating solution of a viscosity in the range of 0.2 to 0.3 PS (at 25.degree. C.). The coating solution was coated on the stimulable phosphor layer using a doctor blade. The coated layer was dried to 120.degree. C. for 30 minutes for heat-curing to give a protective layer of 3 .mu.m thick. ______________________________________ [Preparation of Edge Coating] Composition ______________________________________ Silicone polymer (polyurethane having polydimethyl 70 g siloxane units, Diaromer SP-3023 (15% solution in a mixture of methyl ethyl ketone and toluene, product of Dainichi Seika Co., Ltd.) Cross-linking agent: Polyisocyanate (Crossnate 3 g D-70 (50 wt. % solution), product of Dainichi Seika Co., Ltd.) Anti-yellowing agent: Epoxy resin (Epikote 0.6 g 1001 (solid), product of Yuka Shell Epoxy Co., Ltd.) Alcohol modified-silicone (X-22-2809 (66% 0.2 g solution), product of Shin-etsu Chemical Industries Co., Ltd.) ______________________________________ The above composition was placed in 15 g of methyl ethyl ketone and dissolved to give an edge coating solution. The coating solution was coated on the edges of the above-prepared laminate of a support, a subbing reflective layer, a stimulable phosphor layer and a protective layer, and dried at room temperature to give a cured edge coating layer of 25 .mu.m thick. Thus, a radiation image storage panel of the invention comprising a support, a subbing reflective layer, a stimulable phosphor layer, a protective layer, and an edge coating was prepared. EXAMPLE 2 The procedures of Example 1 were repeated except that the stimulable phosphor sheet was prepared in the following manner. ______________________________________ [Preparation of Stimulable Phosphor Layer] Composition ______________________________________ Stimulable phosphor (BaFBr.sub.0.85 I.sub.0.15 :Eu.sup.2+) 200 g Binder: Polyurethane elastomer (P-22 (solid), 8.0 g Vicat softening point: 64.degree. C., product of Nippon Milacton Co., Ltd.) Anti-yellowing agent: Epoxy resin (Epikote 2.0 g 1001 (solid), product of Yuka Shell Epoxy Co., Ltd.) ______________________________________ The above composition was placed in tetrahydrofuran and dispersed by means of a propeller mixer to give a coating dispersion of a viscosity of 30 PS (at 25.degree. C.) in which the ratio of binder to phosphor was 1/20. The dispersion was coated on a polyethylene terephthalate sheet (temporary support, thickness: 180 .mu.m, having a silicone release undercoating) on its undercoating layer side. The coated layer was dried to give a stimulable phosphor layer, which was then separated from the support to give a stimulable phosphor sheet of a thickness of 320 .mu.m. Thus, a radiation image storage panel of the invention comprising a support, a subbing reflective layer, a stimulable phosphor layer, a protective layer, and an edge coating was prepared. EXAMPLE 3 The procedures of Example 1 were repeated except for using a polyurethane elastomer of Kuramilon U-3180 (solid, product of Kuraray Co., Ltd.) as the polyurethane elastomer binder for the preparation of the stimulable phosphor sheet to give a radiation image storage panel of the invention comprising a support, a subbing reflective layer, a stimulable phosphor layer, a protective layer, and an edge coating was prepared. Comparison Example 1 The procedures of Example 1 were repeated except that the stimulable phosphor sheet was prepared in the following manner. ______________________________________ [Preparation of Stimulable Phosphor Layer] Composition ______________________________________ Stimulable phosphor (BaFBr.sub.0.85 I.sub.0.15 :Eu.sup.2+) 200 g Binder: Polyurethane elastomer (Pandex T-5265H 7.1 g (solid), Vicat softening point: 50.degree. C., product of Dainippon Ink & Chemicals Inc.) Cross-linking agent: Polyisocyanate (Colonate HX 0.9 g (solid), product of Nippon Polyurethane Co., Ltd.) Anti-yellowing agent: Epoxy resin (Epikote 2.0 g 1001 (solid), product of Yuka Shell Epoxy Co., Ltd.) ______________________________________ The above composition was placed in methyl ethyl ketone and dispersed by means of a propeller mixer to give a coating dispersion of a viscosity of 30 PS (at 25.degree. C.) in which the ratio of binder to phosphor was 1/20. The dispersion was coated on a polyethylene terephthalate sheet (temporary support, thickness: 180 .mu.m, having a silicone release undercoating) on its undercoating layer side. The coated layer was dried to give a stimulable phosphor layer, which was then separated from the support to give a stimulable phosphor sheet of a thickness of 320 .mu.m. Thus, a radiation image storage panel for comparison comprising a support, a subbing reflective layer, a stimulable phosphor layer, a protective layer, and an edge coating was prepared. Comparison Example 2 The procedures of Example 1 were repeated except that the stimulable phosphor sheet was prepared in the following manner. ______________________________________ [Preparation of Stimulable Phosphor Layer] Composition ______________________________________ Stimulable phosphor (BaFBr.sub.0.85 I.sub.0.15 :Eu.sup.2+) 200 g Binder: Polyurethane elastomer (Desmolac TPKL-5- 17.8 g 2625 (solid), Vicat softening point: 60.degree. C., product of Sumitomo Bayer Urethane Co., Ltd.) Cross-linking agent: Polyisocyanate (Colonate HX 0.9 g (solid), product of Nippon Polyurethane Co., Ltd.) Anti-yellowing agent: Epoxy resin (Epikote 2.0 g 1001 (solid), product of Yuka Shell Epoxy Co., Ltd.) ______________________________________ The above composition was placed in a mixture of methyl ethyl ketone and isopropyl alcohol (1/1, volume ratio) and dispersed by means of a propeller mixer to give a coating dispersion of a viscosity of 30 PS (at 25.degree. C.) in which the ratio of binder to phosphor was 1/20. The dispersion was coated on a polyethylene terephthalate sheet (temporary support, thickness: 180 .mu.m, having a silicone release undercoating) on its undercoating layer side. The coated layer was dried to give a stimulable phosphor layer, which was then separated from the support to give a stimulable phosphor sheet of a thickness of 320 .mu.m. Thus, a radiation image storage panel for comparison comprising a support, a subbing reflective layer, a stimulable phosphor layer, a protective layer, and an edge coating was prepared. Elastic Characteristics of Binder Polymer (1) Measurement of Viscoelasticity of Urethane Elastomer Used as Binder of Stimulable Phosphor Layer 1) Preparation of Specimen The urethane elastomer was dissolved in tetrahydrofuran to give a 13 wt. % solution. The solution was coated on a release surface of a base sheet, and dried. The prepared film (thickness: 100 .mu.m) was separated from the base sheet, and cut using a cutter to give a specimen of 5 mm.times.25 mm (to be gripped at both 5 mm end portions). 2) Measuring Method A tensile machine (Tensilon UTM-11-20, available from Toyo Boldwin Co., Ltd.) which was designed in accordance with JIS-B-7721 was employed. The specimen was set between the grips (distance: 15 mm, gripped at both 5 mm end portions) and the grips were separated from each other at a grip separation rate of 40 mm/min. at 25.degree. C., 50% RH under the conditions defined in JIS-K-6251, to measure an S--S curve (Stress-Stretch curve). From the S--S curve, the modulus of elasticity, tensile strength, and tensile elongation were calculated. The results are set forth in Table 1. set forth in Table 1. TABLE 1 ______________________________________ Modulus of Tensile Tensile Elasticity Strength Elongation (kgf/mm.sup.2) (kgf/mm.sup.2) (%) ______________________________________ Example 1 0.033 2.67 865 Example 2 0.097 6.13 780 Example 3 0.086 7.27 790 Com. Ex. 1 0.64 5.35 1038 Com. Ex. 2 0.80 5.37 713 ______________________________________ Evaluation of Reproduced Radiation Image The radiation image storage panel was exposed to X-rays at voltage of 80 KVp through an MTF chart and subsequently scanned with a He-Ne laser beam (wavelength: 632.8 nm) to excite the stimulable phosphor. The light emitted by the phosphor layer of the panel was detected and converted to the corresponding electric signals by means of a photosensor and processed to obtain a visible image on a recording apparatus. The modulation transfer function (MTF) value of the visible image was then determined. The MTF value was given as a value(%) at the spacial frequency of 2 cycle/mm. Also measured was graininess (RM) at a radiation amount of 0.1 mR. As to the sharpness and graininess, the radiation image storage panels prepared in Examples 1 to 3 and Comparison Examples 1 to 2 were at the same levels. This means that the use of the thermoplastic elastomer having a low modulus of elasticity according to the invention gives no adverse influence to the image quality of the reproduced radiation image. Evaluation of Transferring Durability The radiation image storage panel was cut to give a test sheet of 100 mm.times.250 mm, which was then transferred on the transfer test machine illustrated in FIG. 1. The test sheet was introduced from the entrance 1 to pass through the guide plates 2 and nip rolls 3 (diameter: 25 mm). The test sheet was moved on the conveyor belt 4 to successively bend inward and outward along the rubber rolls 5 (diameter: 40 mm) and then was taken out through guide plates and nip rolls. This transferring procedure was repeated up to 5,000 cycles under observation of the production of cracks on the protective layer of the test sheet. The results are set forth in Table 2. TABLE 2 ______________________________________ Transferring Durability (Cracks on Phosphor Layer) ______________________________________ Example 1 Not observed after 5,000 cycles Example 2 Not observed after 5,000 cycles Example 3 Not observed after 5,000 cycles Com. Ex. 1 Observed at 2,500 cycles Com. Ex. 2 Observed at 2,000 cycles ______________________________________ From the results shown in Table 1, it has been confirmed that the radiation image storage panels of the invention using an extremely low modulus of elasticity as the binder polymer for the phosphor layer are resistant to production of cracks in the repeated conveying operation including plural bending procedures and plural physical shocks. In contrast, the known radiation image storage panels show a relatively low resistance because of the high modulus of elasticity. It was further observed that cracks were produced on the protective layers of the radiation image storage panel for comparison. It appeared that the cracks produced on the stimulable phosphor layer of the radiation image storage panel for comparison brought about cracks on the protective layer which was directly placed on the phosphor layer. |
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044255063 | summary | FIELD OF THE INVENTION The present invention is in the general area of charged particle beam optics and transport and particularly relates to achromatic beam deflection especially suitable for use in radiation treatment apparatus. BACKGROUND OF THE INVENTION Achromatic optical elements are essential in commercial and medical therapeutic irradiation systems because the primary attribute for such operations is the relatively high beam intensity and control thereof. A typical high beam current accelerator, such as the microwave linear accelerator, achieves the required beam intensities but the energy distribution is rather wide. In order to utilize the available beam it is therefore necessary to introduce optical lements which are relatively insensitive to the energy distribution of the beam. In particular it is desirable for x-ray apparatus to concentrate an intense beam onto a small beam spot on the x-ray target to obtain an x-ray source sufficiently small in relationship to the targeted irradiation region. Beam deflection systems in commercial irradiation and medical therapy applications are ordinarily subject to mechanical and geometrical constraints incident to the maneuverability of the apparatus, shielding and collimation of irradiation flux and as well as economic considerations in the construction of such apparatus. One achromatic beam deflection system of the prior art is described in U.S. Pat. No. 3,867,635 commonly assigned with the present invention. In this apparatus the beam traverses three uniform field sector magnets and two intermediate drift spaces, undergoing a 270.degree. deflection for incidence upon the x-ray target. The sector magnet poles are precisely specified in regard to the sector angles. The angles of incidence and egress of the beam with respect to each sector and a shunt of complex shape occupies the intermediate spaces as well as the entrance and exit regions of the deflector to assure required field free drift spaces. The mutual internal alignment of all components of the deflector is essential to achieve the performance of this prior art device as well as is the alignment of the assembled deflector with the accelerator beam. Another prior art system is known from U.S. Pat. No. 3,379,911 wherein 270.degree. deflection is accomplished in a uniform field to which there is introduced in the vicinity of the deflection midpoint (135.degree.) a gradient region, such that the magnetic field in this gradient region increases radially in the plane of deflection toward the outer portion of accepted trajectories. Thus, those trajectories characterized by a large radius of curvature (in the absence of a gradient) are subject to a somewhat more intense field than would be the trajectories for smaller radii of curvature. Proper adjustment of the gradient shim yields first order achromatic deflection through the desired angle. It is desirable in all of the described systems for the deflector to introduce no substantial momentum dispersion of the beam and to produce at the exit plane a faithful reproduction of conditions encountered at the entrance plane of the system. SUMMARY OF THE PRESENT INVENTION The principal object of the present invention is the provision of an especially simple first order achromatic deflection system in a charged particle irradiation apparatus. In one feature of the invention, a deflection magnet comprises a first uniform field region separated from a second uniform field region along a boundary, whereby particle trajectories traversing said first region are characterized by a large radius of curvature in said first region, a smaller radius of curvature in said second region, thence again traversing said first region with said large radius of curvature. In another another feature of the invention the ratio of fields in said first and second regions is a constant and is realized by first (wide) and second (narrow) gaps between stepped pole faces. In still another feature of the invention the boundary between said first and second regions is a straight line. In yet another feature of the invention, energy selection slits are disposed in the relatively narrow gap of said second field region whereby radiation from said slits is more effectively shielded by a greater mass of said magnetic pole-pieces in said second (narrow gap) field region. In still another feature of the invention, precise bending plane alignment of the deflection magnet with the axis of a particle accelerator is accomplished by a rotation of the magnet about an axis through the bending plane thereof without need for internal alignment of components of said magnet. In again another feature of the invention the magnitude of displacement of trajectories from the central orbit at the image plane of the magnet is equal to the displacement of the trajectory from the central orbit at the entrance plane of the magnet, whereby parallel rays at the entrance plane are rendered parallel at the exit plane. Other features and advantages of the present invention will become apparent upon perusal of the following specification taken in conjunction with the accompanying drawings. In still yet another feature of the invention, a single quadrupole element is employed to cause a radial waist and a transverse waist in an achromatic charged particle beam deflection system to occur at a common target plane. |
claims | 1. An apparatus comprising:a semiconductor device integrally formed within a monocrystalline silicon substrate such that the semiconductor device is a part of the monocrystalline substrate;a Faraday cage that encloses said semiconductor device within a structure of metal that has an at least one vertical metal wall, wherein a bottom surface of said Faraday cage is formed on a top surface of said monocrystalline silicon substrate, and said bottom surface of said Faraday cage is below a top surface of said semiconductor device; anda patterned insulating layer within the Faraday cage and over the at least one semiconductor device; andan interconnect within the Faraday cage and over a top surface of the patterned insulating layer. 2. The apparatus of claim 1, further comprising a second semiconductor device located outside and adjacent to the Faraday cage. 3. The apparatus of claim 1, wherein the at least one vertical metal wall is made from one or more metal layers. 4. The apparatus of claim 1, wherein the interconnect is connected with the at least one semiconductor device and includes input/output leads that pass through the at least one vertical metal wall of the Faraday cage at one or more insulated locations. 5. The apparatus of claim 4, wherein the one or more insulated locations are made from one or more insulating layers. 6. The apparatus of claim 4, wherein the interconnect is constructed with a metal layer that is used to construct the Faraday cage. 7. The apparatus of claim 4, wherein the interconnect is located within the Faraday cage. 8. The apparatus of claim 4, wherein the interconnect is further located outside the Faraday cage. 9. The apparatus of claim 1, further comprising a filled via constructed with a metal layer used to construct the Faraday cage. 10. The apparatus of claim 9, wherein the filled via is located within the Faraday cage. 11. The apparatus on a substrate of claim 9, wherein the filled via is located outside the Faraday cage. 12. The apparatus on a substrate of claim 1, wherein the at least one semiconductor device can be chosen from the group consisting of MOS, CMOS, NMOS, PMOS, Bi-CMOS and DMOS. 13. An apparatus comprising:an at least one semiconductor device integrally formed within a monocrystalline silicon substrate such that the semiconductor device is a part of the monocrystalline substrate;an at least one patterned insulating layer including a filled via opening positioned over the at least one semiconductor device;an at least one metal layer positioned over the at least one insulating layer such that the at least one metal layer forms a first metal cage that encloses the at least one semiconductor device and has at least one vertical metal wall, wherein a bottom surface of said first metal cage is on a top surface of said monocrystalline silicon substrate, and said bottom surface of said first metal cage is below a top surface of said semiconductor device. 14. The apparatus of claim 13, further comprising a second semiconductor device positioned outside and adjacent the first metal cage. 15. The apparatus of claim 13, further comprising:a second metal cage that encloses a first interconnect and is positioned higher than the first metal cage;a second interconnect positioned outside the second metal cage;an at least one set of input/output leads that connect the first interconnect within the second metal cage to the second interconnect positioned outside the second metal cage. 16. An apparatus comprising:a semiconductor device integrally formed within a semiconductor substrate such that the semiconductor device is a part of the semiconductor substrate; anda Faraday cage that encloses said semiconductor device within a structure of metal, wherein a bottom surface of said Faraday cage is formed on a top surface of said semiconductor substrate, and said bottom surface of said Faraday cage is below a top surface of said semiconductor device. 17. The apparatus of claim 16, further comprising a patterned insulating layer within the Faraday cage and over the at least one semiconductor device. 18. The apparatus of claim 17, further comprising an interconnect within the Faraday cage and over a top surface of the patterned insulating layer. 19. The apparatus of claim 18, wherein said bottom surface of said Faraday cage is below a top surface of said semiconductor device. 20. The apparatus of claim 18, wherein said bottom surface is in direct contact with said top surface of said semiconductor substrate. |
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047626623 | abstract | A thermally activated trigger device 10 comprises a pressurized closed vessel 11 with a piston 12 slideably mounted in it to divide it into two compartments 14 and 16. A fluid such as an inert gas is contained within the compartments at substantially the same pressure. One end of one of the compartments has venting means 24, such as a normally closed pipe which vents the fluid from the compartment once the pipe becomes open by rupture or melting at an elevated temperature. The resulting pressure difference between the compartments on venting of the fluid moves the piston which is connected to an actuator means for performing a desired control or safety shut-down function. The trigger device is useful for performing many control and safety shut-down functions, particularly for rendering a nuclear space reactor subcritical upon reentry to the earth's atmosphere. |
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046648712 | description | DETAILED DESCRIPTION OF PREFERRED EMBODIMENTS FIG. 1 shows a steel pressure vessel 1 with a cylindrical cross section and a retracted upper part 1a. A high-temperature pebble bed reactor 2 is installed in the lower section of the said vessel 1. The core of said reactor 2 is formed by a pile 3 of spherical fuel elements, surrounded by a roof reflector 4, a side reflector 5 and a base reflector 6. The side reflector 5 rests directly on a thermal side shield 7, as explained infra (FIG. 5). A core, support base 8, in the form of a star support, is located under the base reflector 6. The fuel elements are introduced from above and are removed by means of an outlet device at the bottom of the pile 3. The latter device here comprises a fuel element outlet tube 10 connected to two sphere conduits 11 each of which has a decollator 12. If several sphere outlet tubes are present, one decollator with two drives may be provided, one of which is always engaged. The drives 13 for the two decollators 12 are arranged outside the steel pressure vessel 1. The reactor is supplied with fuel elements in such a manner that the fuel elements attain the desired final burn-up after a single passage through the pile 3; they are therefore not returned to the core. No debris separator is required in this feeding process. With multiple passage of the fuel elements, debris separators must also be provided in addition to isolator devices. The high-temperature pebble bed reactor has a capacity of 80 to 100 MWe; its power density amounts to approximately 4 MW/m. Cooling gas flows through it from bottom to top; the gas collects above the roof reflector 4 in a hot gas collecting chamber 14. The diameter of the fuel element pile 3 is chosen so that the lifting of fuel elements at the surface area by the flow of the cooling gas is prevented. The steel pressure vessel 1 has a diameter of approx. 7.00 m in the section containing the high-temperature pebble bed reactor. A domed cover 15, fastened by means of a flange joint 16, is placed on the upper part 1a of the steel pressure vessel 1. A steam generator 17 is installed in section 1a. Said steam generator 17 is directly exposed to hot cooling gas from the hot gas collector chamber 14, which gas flows upward in the steam generator 17 through a line 18; the live steam is removed through a line 19. Even through only one steam generator is present, allowing the reduction of the diameter of the steel pressure vessel 1 in the upper section 1a, it is divided into two independent sub-systems 17a and 17b, as seen in FIGS. 2 and 3. Two circulating blowers 20 with shut-down elements are set on the domed cover 15; they are flanged to the fittings 21 provided on the cover. Because the cooling gas flows from bottom to top through the pile of fuel elements and the steam generator 17 is arranged over the pile 3, the residual heat generated in the pile 3 may be removed by natural convection to the steam generator 17 in case of a failure of the circulating blowers 20. Two means for shutting down the reactor 2 are provided to shut-down the high-temperature pebble bed reactor. The first means comprises absorber rods 22 and their drives 23, which are installed in housings 24 outside the steel pressure vessel 1 and in the area of the vessel's retracted section 1a. The housings 24 are fastened to fittings 25 mounted on the steel pressure vessel 1, by welding or flanging. The absorber rods 22 may be inserted from above into bores 26 of the side reflector 5; they drop in by gravity. The rods perform the job of emergency shut-down. If need be, the second means for shutting down the reactor may also be used for this purpose, but its actual purpose is long term shut-down. The second means for shutting down the reactor comprises a plurality of small absorber spheres 27, several storage containers 28 for the spheres, and several annular conduits 29 (two, for example) arranged above the roof reflector 4 and shielded by an annular wall 30 against the hot gas collector chamber 14. The storage containers 28 are located outside the steel pressure vessel 1, also in the area of the retracted section 1a. The annular conduits 29 are connected by lines 31 with channels 32. The channels 32 are disposed in the radially extending projections 33 of the side reflector 5 which protrudes into the fuel element pile 3. Four such projections may be provided for example. Outlet tubes 34 for the removal of the small absorber spheres 27 are connected with the channels 32. In the case of extreme core and pressure vessel dimensions (core diameter larger than 3.50 m, vessel diameter larger than 7 m), the high-temperature pebble bed reactor 2 may be equipped with a third means for shutting down the reactor consisting of small absorber spheres or helium-3 to be fed into the pile (not shown). This shut-down means is additionally used for long term cold shut-downs. FIG. 2 shows the subdivision of the steam generator 17 into two sub-systems 17a and 17b, each of which is provided with its own collector 19a' and 19b', distributor 18a' and 18b', feed water line 18a and 18b, and its own live steam line 19a and 19b. The two circulating blowers 20 may be coordinated with the entire steam generator 17 (for reasons of effectiveness at least two blowers always are provided), or, as shown here, the coordination is such that each blower 20 is exposed to the cooling gas coming from one of the partial systems 17a or 17b, both in normal operation and for the removal of the residual heat. As seen in FIG. 3, the sub-system 17a has an annular cross section and is centrally located. The sub-system 17b is also annular in its cross section and surrounds the sub-system 17a. The live steam lines 19a and 19b lie through the free internal space and the free annular space betwen the two sub-systems. FIG. 4 shows the circulation on the secondary side of the two-part steam generator 17. For the recycling of the two sub-systems 17a and 17b during the removal of the residual heat, a cooling system 36a and 36b is connected with each of the corresponding secondary loops 35a and 35b. Said cooling systems comprise an auxiliary cooler 37a and 37b and a circulating pump 38a and 38b. FIG. 5 shows a section of the side reflector 5. The reflector is composed of two graphite cylinders 39 and 40 with blocks of different size, and is supported without additional supporting elements on the thermal side shield 7. This is possible because of the upward flow of the cooling gas in the fuel element pile 3, which results in a relatively cold core bottom. The differential thermal expansions between the reflector and the thermal shield are consequently negligible. This is also true for the base reflector 6 and the thermal base shield, so that small absorber spheres may be introduced in the pile 3 without prestressing the base reflector 6. The thermal side shield 7 has a plurality of machined support points 41. The remaining surface of the thermal shield 7 is unfinished. Grooves 43 are provided in the blocks of graphite cylinders 40. The grooves 43 are engaged by the support points 41. In this manner, rotation of the thermal side shield 7 with respect to the side reflector 5 is prevented. The support points 41 may be, for example, welded or screwed on. To equalize for tolerances, strips are fitted during the assembly or additional supports provided. The space between the support points 41 represents a gap 42 for the passage of the cooling gas, so that the thermal side shield 7 and the side reflector 5 may be cooled. |
claims | 1. An extreme ultraviolet light generation system, comprising:a laser apparatus configured to output a laser beam for generating extreme ultraviolet light;a chamber provided with a window, through which the laser beam from the laser apparatus enters the chamber;a target supply unit configured to output a target toward a predetermined position inside the chamber;a laser beam focusing optical system positioned to reflect the laser beam toward the predetermined position inside the chamber;a detector configured to aim at a point in a path of the laser beam downstream from the predetermined position for detecting an image of the laser beam at the predetermined position;a target position adjusting mechanism for adjusting a direction into which the target is to be outputted;a laser beam focus position adjusting mechanism for adjusting a focus position of the laser beam; anda controller for controlling the target position adjusting mechanism and the laser beam focus position adjusting mechanism based on the image detected by the detector. 2. The extreme ultraviolet light generation system according to claim 1, whereinthe image detected by the detector includes a shadow of a target, andthe controller is configured to:calculate a difference between a target extreme ultraviolet light generation position and a position of the target based on the image detected by the detector;control the target position adjusting mechanism based on the calculated difference;calculate a difference between the target extreme ultraviolet light generation position and a position of the laser beam based on the image detected by the detector; andcontrol the laser beam focus position adjusting mechanism based on the calculated difference. 3. The extreme ultraviolet light generation system according to claim 1, further comprising a laser beam focus adjusting unit for adjusting a divergence of the laser beam,wherein the controller is configured to calculate a focus position of the laser beam based on the image detected by the detector and control the laser beam focus adjusting unit based on the calculated focus position. 4. The extreme ultraviolet light generation system according to claim 1, whereinthe image detected by the detector includes a shadow of a target, andthe controller is configured to:calculate a difference between a target extreme ultraviolet light generation position and a position of the target based on the image detected by the detector;control a timing at which a subsequent target is to be outputted from the target supply unit based on the calculated difference;calculate a difference between the target extreme ultraviolet light generation position and a position of the laser beam based on the image detected by the detector; andcontrol a timing at which a subsequent laser beam is to be outputted from the laser apparatus based on the calculated difference. 5. An extreme ultraviolet light generation system, comprising:a first laser apparatus configured to output a first laser beam;a second laser apparatus configured to output a second laser beam for generating extreme ultraviolet light;a chamber provided with a window, through which the first and second laser beams respectively from the first and second laser apparatuses enter the chamber;a target supply unit for outputting a target toward a predetermined position inside the chamber;a laser beam focusing optical system positioned to reflect the first and second laser beams toward the predetermined position;a detector configured to aim at a point in a path of the second laser beam downstream from the predetermined position for detecting an image of the laser beam at the predetermined position;a target position adjusting mechanism for adjusting a direction into which the target is to be outputted;a laser beam focus position adjusting mechanism for adjusting a focus position of at least one of the first and second laser beams; anda controller for controlling the target position adjusting mechanism and the laser beam focus position adjusting mechanism based on the image detected by the detector. 6. A method for generating extreme ultraviolet light in a system including a laser apparatus, a chamber, a target supply unit, a laser beam focusing optical system, a detector, a target position adjusting mechanism, a laser beam focus position adjusting mechanism, and a controller, the method comprising:detecting an image of a laser beam from the laser apparatus, the laser beam being reflected by the laser beam focusing optical system toward the predetermined position for generating the extreme ultraviolet light, the detecting step aiming at a point in a path of the laser beam downstream from the predetermined position for detecting the image of the laser beam at the predetermined position; andcontrolling the target position adjusting mechanism for adjusting a direction into which a target is to be outputted and the laser beam focus position adjusting mechanism based on the detected image. 7. The method according to claim 6, whereinthe detected image includes a shadow of a target, andthe controlling step further includes calculating a difference between a target extreme ultraviolet light generation position and a position of the target based on the detected image, and controlling the target position adjusting mechanism based on the calculated difference. 8. The method according to claim 6, wherein the controlling step further includes calculating a difference between a target extreme ultraviolet light generation position and a position of the laser beam based on the detected image, and controlling the laser beam focus position adjusting mechanism based on the calculated difference. 9. The method according to claim 6, whereina laser beam focus adjusting unit is further provided, andthe controlling step further includes calculating a focus position of the laser beam based on the detected image and controlling the laser beam focus adjusting unit based on the calculated focus position. 10. The method according to claim 6, whereinthe detected image includes a shadow of a target, andthe controlling step further includes:calculating a difference between a target extreme ultraviolet light generation position and a position of the target based on the detected image;controlling a timing at which a subsequent target is to be outputted from the target supply unit based on the calculated difference;calcualting a difference between the target extreme ultraviolet light generation position and a position of the laser beam based on the detected image; andcontrolling a timing at which a subsequent laser beam is to be outputted from the laser apparatus based on the calculated difference. 11. A method for generating extreme ultraviolet light in a system including first and second laser apparatuses, a chamber, a target supply unit, a laser beam focusing optical system, a detector, a target position adjusting mechanism, a laser beam focus position adjusting mechanism, and a controller, the method comprising:outputting first and second laser beams respectively from the first and second laser apparatuses;detecting an image of the second laser beam from the laser apparatus, the second laser beam being reflected by the laser beam focusing optical system toward the predetermined position for generating the extreme ultraviolet light, the detecting step aiming at a point in a path of the second laser beam downstream from the predetermined position for detecting the image of the second laser beam at the predetermined position; andcontrolling the target position adjusting mechanism for adjusting a direction into which the target is to be outputted and the laser beam focus position adjusting mechanism based on the detected image. |
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abstract | An apparatus for obtaining a long-length x-ray image of a subject, has an x-ray source and a first sensor that generates a first signal that indicates termination of x-ray emission from the x-ray source. A digital radiography detector is energizable to generate image data after receiving x-ray emission from the x-ray source. A detector transport apparatus is actuable in accordance with the first signal to translate the digital radiography detector from at least a first detector position to a second detector position for generating image data at each detector position. A processor in communication with the digital radiography detector obtains the image data of the subject that is generated from the detector. |
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claims | 1. A data processing device, comprising;a gantry including a movable data acquisition device;a plurality of emitter antennas arranged on the movable data acquisition device and adapted to emit electromagnetic radiation including data acquired by the movable data acquisition device;a plurality of receiver antennas arranged on the gantry, each of the plurality of receiver antennas adapted to receive the electromagnetic radiation emitted by each of the plurality of emitter antennas; anda data processing unit coupled to the plurality of receiver antennas and adapted to extract the data acquired by the movable data acquisition device from the electromagnetic radiation received by the plurality of receiver antennas. 2. The data processing device of claim 1,wherein the movable data acquisition device is a rotatable data acquisition device. 3. The data processing device of claim 1,wherein the number of emitter antennas is larger, equal or smaller than the number of receiver antennas. 4. The data processing device of claim 1,wherein the plurality of emitter antennas, the plurality of receiver antennas and the data processing unit are designed as a multiple-input multiple-output communication system. 5. The data processing device of claim 1,wherein the plurality of receiver antennas are adapted for a wireless two-way communication with the plurality of emitter antennas. 6. The data processing device of claim 1,wherein at least a part of the plurality of receiver antennas are arranged spatially fixed to a stationary portion of the gantry and apart from the movable data acquisition device. 7. The data processing device of claim 1,wherein at least a part of the plurality of receiver antennas are arranged movably on the movable data acquisition device. 8. The data processing device of claim 7,wherein the data processing unit is coupled to the plurality of receiver antennas by means of a wired communication channel. 9. The data processing device of claim 1,wherein the data processing unit is arranged spatially fixed and apart from the movable data acquisition device. 10. The data processing device of claim 1,the gantry comprising a shielding element for shielding the plurality of emitter antennas and the plurality of receiver antennas with respect to the data processing unit and/or an environment. 11. The data processing device of claim 1,wherein the plurality of receiver antennas are adapted to emit electromagnetic radiation including feedback information concerning the communication between the plurality of receiver antennas and the plurality of emitter antennas;wherein the plurality of emitter antennas are adapted to receive the electromagnetic radiation emitted by the plurality of receiver antennas. 12. The data processing device of claim 1,wherein a communication between the plurality of receiver antennas and the plurality of emitter antennas is bidirectional. 13. The data processing device of claim 1,wherein a communication between the plurality of receiver antennas the plurality of emitter antennas includes estimating properties of a transmission medium. 14. The data processing device of claim 1,wherein a communication between the plurality of receiver antennas and the plurality of emitter antennas is based on a periodic channel. 15. The data processing device of claim 1,wherein a communication between the plurality of receiver antennas and the plurality of emitter antennas includes at least one of the group consisting of estimating a channel matrix, estimating a channel distortion, and estimating a channel equalization. 16. The data processing device of claim 1,configured as one of the group consisting of a baggage inspection apparatus, a medical application apparatus, a material testing apparatus and a material science analysis apparatus. 17. A tomography apparatus for examination of an object of interest, the tomography apparatus comprising:an electromagnetic radiation source adapted to emit a beam to the object of interest;a detecting device adapted to detect the beam having passed the object of interest; anda data processing device including:a gantry including a movable data acquisition device,a plurality of emitter antennas arranged on the movable data acquisition device and adapted to emit electromagnetic radiation including data acquired by the movable data acquisition device,a plurality of receiver antennas arranged on the gantry, each of the plurality of receiver antennas adapted to receive the electromagnetic radiation emitted by each of the plurality of emitter antennas, anda data processing unit coupled to the plurality of receiver antennas and adapted to determine structural information concerning the object of interest based on an analysis of detecting signals received from the detecting device which is coupled to the plurality of emitter antennas of the data processing device. 18. The tomography apparatus of claim 17,adapted as one of the group consisting of a computed tomography apparatus, a positron emission tomography apparatus, and a single photon emission computed tomography apparatus. 19. The tomography apparatus of claim 17,wherein the gantry comprises a rotatable portion on which the electromagnetic radiation source, the detecting device and the plurality of emitter antennas are mounted. 20. A method of processing data, the method comprisingemitting, by means of a plurality of emitter antennas arranged on a movable data acquisition device of a gantry, electromagnetic radiation including data acquired by the movable data acquisition device;receiving, by means of a plurality of receiver antennas arranged on the gantry, the electromagnetic radiation emitted by each of the plurality of emitter antennas;extracting, by means of a data processing unit coupled to the plurality of receiver antennas, the data acquired by the movable data acquisition device from the electromagnetic radiation received by the plurality of receiver antennas. 21. A computer-readable medium, in which a computer program of processing data is stored which, when being executed by a processor, is adapted to control or carry out:emitting, by means of a plurality of emitter antennas arranged on a movable data acquisition device of a gantry, electromagnetic radiation including data acquired by the movable data acquisition device;receiving, by means of a plurality of receiver antennas arranged on the gantry, the electromagnetic radiation emitted by each of the plurality of emitter antennas;extracting, by means of a data processing unit coupled to the plurality of receiver antennas, the data acquired by the movable data acquisition device from the electromagnetic radiation received by the plurality of receiver antennas. 22. A program element of processing data, which program element, when being executed by a processor, is adapted to control or carry outemitting, by means of a plurality of emitter antennas arranged on a movable data acquisition device of a gantry, electromagnetic radiation including data acquired by the movable data acquisition device;receiving, by means of a plurality of receiver antennas arranged on the gantry, the electromagnetic radiation emitted by each of the plurality of emitter antennas;extracting, by means of a data processing unit coupled to the plurality of receiver antennas, the data acquired by the movable data acquisition device from the electromagnetic radiation received by the plurality of receiver antennas. 23. A data processing device, comprising:a gantry;plurality of emitter antennas arranged on the gantry and adapted to emit electromagnetic radiation in which data are encoded;a plurality of receiver antennas each adapted to receive the electromagnetic radiation emitted by each of the plurality of emitter antennas, wherein the plurality of receiver antennas are arranged on a movable portion of the gantry;a data processing unit coupled to the plurality of receiver antennas and adapted to extract the data from the electromagnetic radiation received by the plurality of receiver antennas. |
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052681289 | abstract | Particulate material is treated by first washing the contaminated material with a contaminant mobilizing solution comprising a leaching agent, a surfactant or a mixture thereof. Large particles, typically greater than 5 mm are mechanically separated, washed with water and returned to the site as recovered soil. Fines, along with contaminants dissolved or dispersed in the contaminant mobilizing solution are separated from intermediate sized particles by a countercurrent flow of the contaminant mobilizing solution, preferably in a mineral jig. The intermediate sized particles are then abraded in an attrition scrubber to dislodge attached mineral slimes or fines. These additional fines are separated from the intermediate sized particles with a countercurrent flow of wash water in a second mineral jig. The preferred oxidizing agent is chlorine, and hydrogen is the preferred reducing agent. |
description | The present invention relates to a control rod for a nuclear boiling water reactor wherein the reactor comprises a plurality of fuel assemblies, elongated channel boxes forming outer casings of each fuel assembly and enclosing fuel rods, and a plurality of control rods which are configured to be insertable in respective control rod positions between the fuel assemblies. Fuel assemblies and control rods are positioned in the core of the nuclear boiling water reactor, BWR. The channel boxes of the fuel assemblies in the BWR usually consist of a corrosion resistant material with a low neutron absorption capacity, such as a zirconium based alloy. The environment in the core of a BWR is demanding for the components positioned therein. The environment is for example highly oxidative. One of the consequences of this demanding environment inside the core of a BWR is that the channel box of the fuel assemblies may be distorted. The channel box may for example bulge or bow. Channel box bow is due to elongation of one channel box side relative the opposite channel box side. Channel box bow is known to arise for different reasons, e.g. initial manufacturing, residual stress relaxation under irradiation, differential irradiation growth and hydration as a consequence of shadow corrosion. The problem of shadow corrosion on components comprising zirconium based alloys in the core of a BWR has been known for a long time. Shadow corrosion is a local corrosion enhancement and can appear on a zirconium based alloy component when the component is in close contact with another metal. Referring to the above, shadow corrosion on the outer side of a channel box can occur when a control rod blade is inserted next to the channel box, i.e. when the channel box consisting of a zirconium based alloy is in close contact with a control rod blade usually having an outer surface of stainless steel. The degree of shadow corrosion depends on the distance between the two components. Direct contact gives most shadow corrosion and at a distance of about 5 mm it effectively disappears. Most of the effect has disappeared already at a distance of 2 mm. Shadow corrosion early in the life of a fuel assembly, i.e. shadow corrosion on the fuel assembly due to an inserted control rod next to the fuel assembly during the first several months of operation, is generally believed to drive the problem of enhanced channel bow of the channel boxes in a BWR. The shadow corrosion can result in increased absorbed hydrogen-induced growth of the outer side of the channel box being closest to the control rod. The increased absorbed hydrogen-induced growth can lead to bowing of the channel box towards the control rod late in the life of the fuel assembly. The bow of the channel box towards the control rod may for example slow down or stop the control rod insertion in an emergency situation, or may lead to channel box-control rod interference, which may for example cause the fuel assemblies to lift due to friction when the control rods are inserted into the core. The control rods are tested periodically with regard to proper functioning and interference. In case some interference is noted or suspected the testing intervals are shortened. With severe interference the control rod is parked in the safe position, i.e. fully inserted and disarmed. In either case the intended operation is disturbed as an ultimate result of the shadow corrosion. Studies have shown that shadow corrosion strongly depends on the distance between the zirconium based alloy component and the component comprising another metal. The occurrence of shadow corrosion is therefore most significant in the case of a large control rod blade and a small distance between the control rod blade and the channel box. The European patent application EP 0 986 069 describes a device and method for preventing shadow corrosion. The device comprises means arranged to electrically insulate at least a part of a second component from a first component in order to avoid shadow-corrosion on the first component. An object of the present invention is to mitigate shadow corrosion on the channel box of a fuel assembly of a nuclear BWR, thereby reducing the risk of shadow corrosion enhanced channel box bow. Another object of the present invention is to provide a control rod with which the shadow corrosion on the channel box is mitigated. At least one of these objects are achieved with a control rod according to the appended claims. Further advantages of the invention are achieved with the features of the dependent claims. According to a first aspect of the invention a control rod for a nuclear boiling water reactor is provided. The reactor comprises a plurality of fuel assemblies, elongated channel boxes forming outer casings of each fuel assembly and enclosing fuel rods, and a plurality of control rods which are configured to be insertable in respective control rod positions between the fuel assemblies. The control rod has a longitudinal centre axis and control rod blades, each control rod blade having a first and a second side and being substantially parallel to the longitudinal centre axis. Each control rod blade comprises an absorber material which extends along the longitudinal centre axis from a first absorber end to a second absorber end, the distance between the first absorber end and the second absorber end defining an active length of the control rod blade. The control rod is characterized in that the control rod blades are provided with distance means on the first and second sides of the control rod blades, the distance means extending a distance of at least a third of the active length of the control rod blade. The distance means are configured to ensure a minimum distance between the control rod blade and the channel box, and to enable the control rod to easily slide along the channel box. The minimum distance ensured by the distance means mitigates the shadow corrosion on the channel box. As the distance means protrude from the control rod blades the distance means may be close to the channel boxes. Thus, the shadow corrosion may be enhanced in the areas of the channel box being closest to the distance means. These areas are only a small portion of the channel box. It is, however, still favourable to minimize the extension of the distance means. Thus, it is an optimisation problem to design the extension of the distance means. A large extension of the distance means along the longitudinal centre axis ensures a minimum distance between the control rod and the channel box. At the same time it is desirable to minimize the extension of the distance means in order to minimize the area of the distance means being close to the channel box and thereby minimizing the shadow corrosion. The distance means may protrude 0.5-2 mm, preferably 0.5-1.5 mm, or more preferably 0.8-1.2 mm, from the sides of the control rod blades. The protrusion of the distance means may be optimized so that a minimum distance is ensured simultaneously with the shadow corrosion. The distance means is distributed on the sides of the control rod blades. Such a distribution of the distance means may be accomplished in many different ways. One way of accomplishing a distributed distance means is to provide the distance means as one or more protrusions. Advantaguously, the protrusions may have a smooth shape, seen in the direction of the centre axis, the smooth shape facilitating the sliding of the control rod along the channel box. In such a way the risk of the control rods getting stuck in the reactor is minized. According to a further embodiment of the invention, at least one protrusion is provided on the upper portion of the control rod blade and at least one protrusion is provided on the lower portion of the control rod blade. This ensures a minimum distance between the control rod blade and the channel box in the lower portion as well as the upper portion. According to a further embodiment of the invention, each protrusion comprises an elongated ridge. Preferably, 1-3 ridges may be provided on each side of the control rod blade. With such ridges the control rod may be arranged to be easily inserted in the reactor as each one of the ridges only have one edge that may get stuck on protruding objects in the reactor. To further minimize the risk of the ridges getting stuck during insertion of the control rod in the reactor the ridges may be provided along lines extending in parallel to the longitudinal centre axis and distributed along the width of the first and second sides, radially out from the longitudinal centre axis. Shadow corrosion is strongest during high power operation of the reactor. This corresponds to the situation in which only a small portion of the control rods are inserted into the reactor. Thus, it is most important to have the distance means on the part of control rod blades that are first introduced into the reactor. The control rod may be arranged to be inserted in the control rod position with the first absorber end first and the ridges may extend from the first absorber end on the first and the second sides of the control rod blades. This ensures a minimum distance between the control rod blades and the channel box when the reactor is operated at a high power. The ridges may extend along the full active length on the first and the second sides of the control rod blades in order to ensure a minimum distance between the control rod blade and the channel box over the entire active length of the control rod blade. Alternatively, the ridges or one or more of the ridges may extend along a part of the full active length. As an alternative to ridges each protrusion may comprise a spot-like protrusion. With spot-like protrusions the area of the distance means being close to the channel box may be minimized. The control rod may comprise four control rod blades, the control rod blades being arranged to form a control rod having a cruciform cross section. This is the standard form of a control rod in a BWR reactor. The distance means may be a homogeneous part of the control rod blades. This may be achieved for example by press forming of the control rod blade. Alternatively the distance means may be welded onto the control rod blades. The control rod may be made mainly of a first material which is different from the absorber material. This is the normal way of designing a control rod as most suitable absorber materials have mechanical properties making them unsuitable for the load bearing parts of the control rod. The design of the control rod according to the invention thereby prevents the control rod blades from coming too close to the channel box of the fuel assembly. Moreover, the design of the protrusions ensures a smooth insertion of the control rod between the fuel assemblies, preventing the control rod from being damaged. In the following description of preferred embodiments of the invention similar features in different figures will be denoted by the same reference numeral. An embodiment of a nuclear boiling water reactor, BWR, which may comprise the control rod according to the invention will first be described with reference to FIG. 1. FIG. 1 shows part of a nuclear plant. The nuclear plant comprises a reactor 1. The reactor 1 comprises a core 2 having a plurality of fuel assemblies 3. Each fuel assembly 3 includes a plurality of elongated fuel rods 7 enclosed in a channel box 6, see FIG. 3. Each fuel rod 7 comprises nuclear fuel 7a and a cladding 7b enclosing the nuclear fuel 7a. The fuel rods 7 are held in predetermined positions relative one another with the help of a number of spacer grids, not shown. The nominal gap between adjacent fuel assemblies 3 in the core is 12 to 20 mm. The reactor 1 further comprises a plurality of control rods 4 schematically indicated in FIG. 1. Each control rod 4 has a longitudinal centre axis z, see FIG. 3. The control rods 4 are located between the fuel assemblies 3 and are connected to drive members 5. Each control rod 4 has four control rod blades 4a, see FIGS. 2 and 3, disposed in a cruciform arrangement. The drive members 5 are able to move the control rods 4 up and down in a vertical direction x into and out from a respective position between the fuel assemblies 3. The nominal thickness of the control rod blade 4a is 7-8 mm. An absorber material 9 is arranged in each one of the control rod blades 4a. The absorber material 9 extends from a first absorber end 11 to a second absorber end 12, wherein the first absorber end 11 is closest to the top end of the control rod 4 and the first absorber end is thus inserted first into the control rod positions between the fuel assemblies 3. The distance between the first absorber end 11 and the second absorber end 12 defines an active length L of the control rod blade. Preferably, the extension of the absorber material in the different control rod blades 4a correspond to each other so that the first absorber end 11 and the second absorber end 12 are common to all control rod blades 4a on a control rod 4. The absorber material may be chosen from the group comprising boron carbide and alloys with hafnium as a main alloying material. These materials all have favourable neutron absorbing properties. It is to be noted that the invention is applicable also to so called grey control rods, i.e. control rods without or with a reduced amount of absorber material. The control rod blades 4a are provided with distance means 13 which in the shown embodiment has the form of protrusions. The protrusions protrude 0.5-2 mm, preferably 0.5-1.5 mm or more preferably 0.8-1.2 mm from the sides of each control rod blade 4a. In the embodiment shown in FIGS. 2 and 3, the protrusions are formed as ridges 14 that are evenly distributed radially. All ridges 14 extend essentially parallel to the longitudinal centre axis from the first absorber end 11 along half the active length of the control rod blades 4a. Only the upper half of the control rod is shown in FIG. 2. It is of course possible to let the ridges 14 extend along the entire active length of the control rod blades 4a. The ridges 14 have a smooth shape to facilitate the sliding of the control rod along the box. There might be protrusions on the channel box on which the ridges 14 on the control rods may get stuck in case they are not smoothly shaped. The ridges 14 may be an integral part of the control rods 4, in which case they may be press formed into the control rods 4. FIG. 4 shows a part of a control rod 4 according to an alternative embodiment of the present invention. In the embodiment shown in FIG. 4 the protrusions are in the form of spot-like protrusions 15, and ridges 14 with a relatively short longitudinal extension. The spot-like protrusions 15 may for instance have the form of smooth cones. Also the spot-like protrusions 15 may be an integral part of the control rods 4, in which case they may be press formed into the control rods 4. The present invention is not limited to the shown embodiments but can be varied and modified within the scope of the following claims. |
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claims | 1. A wavelength-dispersive X-ray spectrometer of a straight moving ray-collection type, the spectrometer being fitted to an X-ray microarea-analyzer such as an electron probe microanalyzer, said X-ray spectrometer comprising:a curved analyzing crystal having a crystalline lattice plane curved to have a curvature equal to the diameter of a Rowland circle in a direction of angular dispersion; andlimitation means mounted integrally with the curved analyzing crystal, the limitation means acting to limit at least one of an incident region of a surface of the analyzing crystal on which incident X-rays impinge and an exit region of the surface of the analyzing crystal from which diffracted X-rays exit toward an X-ray detector such that only X-rays diffracted by an effective diffractive region of the surface of the analyzing crystal are detected by the X-ray detector in response to variation of the effective diffractive region of the surface of the analyzing crystal contributing to actual diffraction when a spectral analysis position of the X-ray spectrometer varies. 2. A wavelength-dispersive X-ray spectrometer of a straight moving ray-collection type as set forth in claim 1, wherein said limitation means is made of an X-ray blocking plate made to upstand from a position on the surface of the analyzing crystal at an end in the direction of angular dispersion of the crystal toward inside of the Rowland circle, and wherein said X-ray blocking plate blocks parts of at least one of incident X-rays going from a point X-ray source toward the analyzing crystal and X-rays diffracted by the analyzing crystal toward the X-ray detector. 3. A wavelength-dispersive X-ray spectrometer of a straight moving ray-collection type as set forth in claim 1 or 2, wherein said limitation means is made of an X-ray blocking plate made to upstand from an end of a crystal support member supporting the analyzing crystal in the direction of angular dispersion of the crystal toward the center of the Rowland circle in the X-ray spectrometer or toward the center of curvature of the curved analyzing crystal, and wherein said X-ray blocking plate blocks parts of at least one of incident X-rays going from a point X-ray source toward the analyzing crystal and X-rays diffracted by the analyzing crystal toward the X-ray detector. 4. A wavelength-dispersive X-ray spectrometer of a straight moving ray-collection type as set forth in claim 1 or 2, wherein said limitation means is made of an X-ray blocking plate made to upstand perpendicularly to a plane defined by the Rowland circle in the X-ray spectrometer and parallel to a straight line which passes through the center of the analyzing crystal and through the center of the Rowland circle from an end of a crystal support member supporting the analyzing crystal in the direction of angular dispersion of the crystal, and wherein said X-ray blocking plate blocks parts of at least one of incident X-rays going from a point X-ray source toward the analyzing crystal and X-rays diffracted by the analyzing crystal toward the X-ray detector. 5. A wavelength-dispersive X-ray spectrometer of a straight moving ray-collection type as set forth in claim 1 or 2, wherein(A) said limitation means has an X-ray blocking plate disposed at an end of a crystal support member for the curved analyzing crystal in the direction of angular dispersion of the analyzing crystal,(B) a part of said X-ray blocking plate covers an appropriate, substantially rectangular region of an end portion of the surface of the analyzing crystal,(C) a front-end portion of the X-ray blocking plate is made to upstand toward the center of the Rowland circle in the X-ray spectrometer or toward the center of curvature of the curved analyzing crystal, and(D) the upstanding portion of the X-ray blocking plate blocks parts of at least one of incident X-rays going from a point X-ray source toward the analyzing crystal and X-rays diffracted by the analyzing crystal toward the X-ray detector. 6. A wavelength-dispersive X-ray spectrometer of a straight moving ray-collection type as set forth in claim 1 or 2, wherein(A) said limitation means has an X-ray blocking plate disposed at an end of a crystal support member for the curved analyzing crystal in the direction of angular dispersion of the analyzing crystal,(B) a part of said X-ray blocking plate covers an appropriate, substantially rectangular region of an end portion of the surface of the analyzing crystal,(C) a front-end portion of the X-ray blocking plate is made to upstand perpendicularly to a plane defined by the Rowland circle in the X-ray spectrometer and parallel to a straight line passing through the center of the analyzing crystal and through the center of the Rowland circle, and(D) the upstanding portion of the X-ray blocking plate blocks parts of at least one of incident X-rays going from a point X-ray source toward the analyzing crystal and X-rays diffracted by the analyzing crystal toward the X-ray detector. 7. A wavelength-dispersive X-ray spectrometer of a straight moving ray-collection type as set forth in claim 1 or 2, wherein said curved analyzing crystal is a Johann crystal having a spherically-curved concave surface having a curvature equal to the diameter of the Rowland circle in the direction of angular dispersion on the lattice plane of the analyzing crystal and in a direction perpendicular to the direction of angular dispersion, and wherein the portion of the X-ray blocking plate which upstands from an end of a crystal support member supporting the spherically-curved Johann crystal in the direction of angular dispersion of the crystal toward the inside of the Rowland circle is substantially rectangular in shape. 8. A wavelength-dispersive X-ray spectrometer of a straight moving ray-collection type as set forth in claim 1 or 2, wherein the height of said X-ray blocking plate upstanding from an end of a crystal support member supporting the analyzing crystal toward the inside of the Rowland circle such that a region contributing to diffraction is set based on data indicating incident angle error of X-rays incident on the surface of the analyzing crystal. 9. A wavelength-dispersive-X-ray spectrometer of a straight moving ray-collection type as set forth in claim 1 or 2, wherein said curved analyzing crystal is an analyzing element made of a layered synthetic microstructure having a lattice spacing of less than 2 nm, and wherein said limitation means is mounted integrally with the analyzing element of the layered synthetic microstructure. |
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054250641 | claims | 1. A nuclear reactor comprising: a reactor vessel having included therein a core which includes fuel bundles containing fissionable material and control rods and which is immersed in water, a water inlet and a steam outlet and a measuring device disposed in the vicinity of said core for measuring the water flow in said core, which core produces in its vicinity a radiation field acting as a radiation source for a measuring device, wherein said measuring device includes: a turbine assembly having a rotor which is rotatably mounted in a housing and which contains at least one radiation field-modulating material for modulating the radiation field produced by the core, at least one radiation detector for converting the radiation field modulation into a detector signal modulation, and a converter coupled to said radiation detector for converting said detector signal into a value corresponding to the water flow. at least one radiation detector for converting the radiation field modulation into a detector signal modulation, and a converter coupled to said radiation detector for converting said detector signal into a value corresponding to the water flow. placing a turbine assembly which includes a rotor, which rotor includes a radiation field-modulating material which can cause spatial variations in a radiation flux in the coolant path such that the flow of said coolant sets said rotor of said turbine device into rotation; detecting with a detector temporal variations in said radiation field which correspond to spatial variations in said radiation field; determining an oscillation frequency from variations in the output of said detector; and calculating the liquid flow rate of said coolant from said oscillation frequency. 2. The nuclear reactor as claimed in claim 1, wherein said radiation field-modulating material is a neutron-absorbing material. 3. The nuclear reactor as claimed in claim 2, wherein said neutron-absorbing material includes 113-Cd and 176-180-Hf. 4. The nuclear reactor as claimed in claim 1, wherein said radiation field-modulating material is a material emitting gamma radiation through neutron activation. 5. The nuclear reactor as claimed in claim 4, wherein said gamma radiation-emitting material comprises 45-Sc, 59-Co and 110-Ag. 6. The nuclear reactor as claimed in claim 1, wherein said rotor contains a neutron-absorbing material and a material emitting gamma radiation through neutron activation. 7. The nuclear reactor as claimed in claim 6, wherein both radiation field-modulating materials form a composite material. 8. The nuclear reactor as claimed in claim 7, wherein said composite material comprises 113-Cadmium/Indium/110-Silver. 9. The nuclear reactor as claimed in claim 1, wherein said radiation field-modulating material is included in a rotor blade. 10. The nuclear reactor as claimed in claim 1, wherein said housing of said turbine assembly connects into a water inlet of said core. 11. The nuclear reactor as claimed in claim 1, wherein guide plates are received in said housing of said turbine assembly. 12. The nuclear reactor as claimed in claim 1, wherein said radiation detector is received in a detector tube extending axially along said core. 13. A measuring device for measuring in a nuclear reactor the water flow in a core, which core produces in its vicinity a radiation field acting as a radiation source for a measuring device, comprising a turbine assembly having a rotor which is rotatably mounted in a housing and which contains at least one radiation field-modulating material for modulating the radiation field .produced by the core, 14. A method for measuring a liquid flow in a nuclear reactor comprising a heat-producing core and a coolant flowing along a coolant path through the core, which method comprises the steps of: 15. The method as claimed in claim 14, wherein said radiation field-modulating material absorbs neutrons so that said radiation field-modulating material modulates the neutron field. 16. The method as claimed in claim 14, wherein said radiation field-modulating material is activated by neutrons so that said radiation field-modulating material emits prompt or delayed gamma rays only after placing of turbine assembly. |
056235291 | abstract | In an X-ray lithographic system comprising a plurality of X-ray exposure apparatus which use an SOR radiation source apparatus as a common illumination light source, an exposure apparatus usable for duplicating an X-ray mask is connected to at least one beam line. The beam line is longer than the other wafer exposure beam lines so that the divergence angle is small (i.e., the resolving power for exposure transfer is higher). Thus, the X-ray mask can precisely be manufactured at low cost. |
summary | ||
claims | 1. A multilayer optic device having an input face and an output face, the optic device comprising:a high-index material layer having a first real refractive index 1−δ1 and a first absorption coefficient β1;a low-index material layer having a second real refractive index 1−δ2 and a second absorption coefficient β2; anda grading zone disposed between the high-index material layer and the low-index material layer, the grading zone comprising a grading layer having a third real refractive index 1−δ3 and a third absorption coefficient β3, such that 1−δ1>1−δ3>1−δ2,wherein at least a portion of one or more of the high-index material layer, the grading zone and the low-index layer comprise one or more corrugations along a first direction, and wherein at least a portion of one or more of the high-index material layer, the grading zone and the low-index layer comprises one or more non-corrugation portions along a first direction. 2. The multilayer optic device of claim 1, wherein at least one corrugation is curved along a second direction. 3. The multilayer optic device of claim 2, wherein the second direction is about perpendicular to the first direction. 4. The multilayer optic device of claim 2, wherein the second direction is a fan beam direction. 5. The multilayer optic device of claim 2, wherein the corrugations are curved outwardly in the second direction with respect to a central axis of the optic device. 6. The multilayer optic device of claim 1, wherein the high-index layer, the grading zone and the low-index layer are conformally disposed. 7. The multilayer optic device of claim 1, comprising two or more multilayer sections, each multilayer section having a high-index material layer, a low-index material layer and a grading zone disposed between the high-index layer and the low-index layer, wherein the high-index layers of the multilayer sections are disposed adjacent each other to form a core. 8. The multilayer optic device of claim 1, comprising two or more multilayer sections, wherein the two multilayer sections have a common high-index layer. 9. The multilayer optic device of claim 1, wherein at least a portion of the high-index material layer comprises a core, and wherein the grading zone is disposed on at least a portion of the high-index material layer, and the low-index material layer is disposed on at least a portion of the grading zone. 10. The multilayer optic device of claim 9, comprising a first grading zone conformally disposed on a first surface of the core, and a second grading zone conformally disposed on a second surface of the core. 11. The multilayer optic device of claim 1, wherein at least one of the layers of the multilayer optic device comprises continuous corrugations in the first direction. 12. The multilayer optic device of claim 1, wherein the one or more corrugations comprise a relatively higher refractive index material compared to a material of the non-corrugation portion. 13. The multilayer optic device of claim 1, wherein the one or more corrugations comprise corrugation comprises a relatively lower refractive index material compared to a material of the non-corrugation portion. 14. The multilayer optic device of claim 1, further comprising a filtering region disposed on the input face of the optic device, wherein the filtering region is configured to filter out X ray beams of determined energy levels from X-ray beams incident on the multilayer optic device. 15. The multilayer optic device of claim 14, wherein the filtering region comprises a pointed conical structure. 16. The multilayer optic device of claim 14, wherein an exterior surface of the filtering region comprises a multilayer stack. 17. The multilayer optic device of claim 16, wherein the multilayer stack comprises alternating layers of high- and low-refractive index materials. 18. The multilayer optic device of claim 16, wherein the multilayer stack comprises a low-index layer, a high-index layer and a grading zone disposed between the high- and low-index layers. 19. A method of making an optic device, comprising:providing a first multilayer section having a high-index layer, a grading zone and a low-index layer conformally disposed, wherein the high-index layer, the grading zone and the low-index layer comprise one or more corrugations along a first axis, and wherein at least one corrugation is curved along a second axis;providing a second multilayer section having a high-index layer, a grading zone and a low-index layer conformally disposed, wherein the high-index layer, the grading zone and the low-index layer comprise one or more corrugations along the first axis, and wherein at least one corrugation is curved along the second axis; anddisposing the first multilayer section on the second multilayer section such that the high-index layer of the first multilayer section and the high-index layer of the second multilayer section are disposed adjacent each other. 20. The method of claim 19, wherein providing the first and second multilayer sections, comprises:providing a first template and a second template, wherein the first and second templates comprise one or more grooves corresponding to the corrugations in the first and second multilayer sections; andconformally depositing one or more multilayers of the first and second multilayer sections on the first and second templates, respectively. 21. The method of claim 20, comprising:providing a high-index material layer having a first surface and a second surface, wherein the first and second surfaces comprise the one or more grooves corresponding to the corrugations in the multilayers of the first and second multilayer sections;conformally depositing grading zones on the first and second surfaces of the high-index material layer; andconformally depositing low-index material layers on corresponding grading zones. 22. A method of making an optic device, comprising:providing a multilayer optic device comprising multilayers, comprising:a high-index material layer having a first real refractive index 1−δ1 and a first absorption coefficient β1;a low-index material layer having a second real refractive index 1−δ2 and a second absorption coefficient β2;a grading zone disposed between the high-index material layer and the low-index material layer, the grading zone comprising a grading layer having a third real refractive index 1−δ3 and a third absorption coefficient β3, such that 1−δ1>1−δ3>1−δ2, wherein at least a portion of one or more of the high-index material layer, the grading zone and the low-index layer comprises one or more corrugations along a first direction;removing at least a portion of the low-index layer, and at least a portion of the grading zone along a plane to form grooves in the multilayers of the multilayer optic device; anddepositing a material in the grooves to form the corrugations. 23. The method of claim 22, further comprising removing at least a portion of opposite surfaces of the high-index material layer. |
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summary | ||
047770090 | description | DESCRIPTION OF THE PREFERRED EMBODIMENT FIG. 1 is a schematic illustration of a typical recirculating steam generator installation in a NSSS while FIG. 2 schematically shows a single steam generator 10. Feedwater is, under certain operating conditions, delivered to the generator via the downcomer nozzle 12 and flows into the downcomer 14 where it mixes with the recirculating saturated liquid. A downcomer valve 15 is located upstream of nozzle 12. The combined flows move down the downcomer and enter the tube bundle region 16 at the bottom of the generator. As the fluid rises through the tube bundle 18, it absorbs heat from the primary loop, exiting the bundle region as a two phase fluid. It then flows upward through the riser region 20 to the separators 22. The separators remove the saturated liquid from the steam, returning the liquid to the downcomer 14 and allowing the steam to rise to the dryers 24, before exiting the steam exits the steam generator 10 and enters the main steam line 26. As is well known in the field of nuclear power, the nuclear reactor (not shown) and the associated piping to and from the steam generator are usually referred to as the primary system, and the reactor vessel and associated piping contain the primary coolant volume. The hot leg of the reactor contains water which has been heated by the reactor and which enters the steam generator through inlet nozzle 30. The steam generator output nozzle 32 returns water from the steam generator through the cold leg piping to the reactor vessel. The recirculation process which includes the steam generator 10 is sustained by the imbalance in the hydraulic heads of fluid between the downcomer 14 and the tube bundle 18 and riser region 20. During high power operation, the difference in these driving heads is significant and leads to relatively stable operation. As the power is dropped, however, the amount of boiling is reduced in the tube bundle 18, causing a reduction in the quality, i.e., the amount of steam in the mixture flowing through the tube bundle, of the fluid and thus an increase in its density. This in turn reduces the amount of driving head which, in turn, reduces the amount of recirculation. As this occurs, the generator 10 approaches a manometer type condition, where the hydraulic head of the downcomer 14 and the hydraulic head of the riser 20 and tube bundle 18 approach each other. Under these conditions, the downcomer water level becomes very difficult to control. The normal water level in the steam generator is indicated at 36. The instrumentation for water level measurement is conventional and is indicated in FIG. 1 at 40. Steam flow is measured at, for example, 42. Feedwater is also and primarily supplied to steam generator 10 through a feedwater nozzle 44 which is connected to an economizer 50. An economizer valve 52 is located upstream of nozzle 44. A principal objective of the steam generator water level control system is to prevent the water level from rising too high, and causing a high level plant trip, or dropping too low, causing a low level plant trip. The water level is controlled primarily by regulating the feedwater flow through nozzles 12 and 44 through modulation of the settings of valves 15 and 52, and by varying the speed of pumps 54. With reference to FIG. 3, a feedwater flow and steam generator water level control is generally illustrated. A set point 54, representing the desired water level, is a first input to a comparator 56. The difference between the set point and the actual steam generator water level, as provided by level instrumentation 40, is outputted from comparator 56 and fed to a controller 60. Controller 60 generates an electrical control signal which, after suitable processing, to convert the electrical signal into a fluidic signal is delivered to one or more valve actuators 64. Each actuator operates a valve or the like, having characteristics as represented at 66. System operation is also influenced by flow relationships, as represented at 68, and by the feedtrain characteristics, as represented at 70. Both the individual flow relationship 68, as influenced by the valve characteristics, and the feed train characteristics 70 determine the steam generator performance represented at 72. Steam generator performance is also influenced by the operating parameters and design of the primary system 74. The changes in the steam generator performance 72 are measured by the level instrumentation 40 and the signal is returned to the comparator 56 to complete the closed control loop. Turning now to FIG. 4, the improved extended range automatic feedwater control system of the present invention will be described in detail. In the following description, it should be appreciated that the invention comprises apparatus and a method for accomplishing novel control techniques. The particular components necessary to implement these techniques are known in the general art of control engineering, but the particular variables used in the control algorithms, and the manner in these variables are employed are novel. Accordingly, where only method steps are described, the associated hardware is either conventionally found in a nuclear power plant, or the selection thereof is self-evident from the function it is to perform. Three parameter steam generator control is known in the art. The present invention uses the known three parameter controller as the base upon which the present invention is constructed. A comparator 76 receives a steam flow signal WS from the instrument 42 that measures steam flow and a feedwater flow signal. The feedwater flow W.sub.fw /signal is provided by a flow sensor 77 (FIG. 1) that measures the feedwater flow. The difference between steam and feedwater flow is a mass flow difference signal which is dynamically compensated in a filtered derivative network 82 and the resulting flow error signal is delivered as a bias signal to a comparator 86. Network 82 has a steady state gain equal to zero so that, when the flow difference is not changing, the compensated output signal will be zero and thus will not contribute to the flow demand signal. A signal commensurate with the measured water level L is delivered to an adaptive lead/lag network 90, and the dynamically compensated level signal from network 90 is delivered as the second input to comparator 86. The network 90 gives higher amplification to rapidly changing inputs than to slowly changing inputs and has a steady state gain of unity. The output of comparator 86 is a water level signal biased by the input to output mass flow difference. This biased level signal is delivered as the first input to another comparator 96 where it is compared with a water level set point (LSP). The result of the comparison performed in comparator 96 is a compensated level error signal. This signal has previously been employed as the primary control signal to maintain steam generator water level. While the prior control scheme works reasonably well for relatively high power operation, i.e., above 20% of rated power, the control becomes unreliable as the power level of the NSSS decreases. In accordance with one aspect of the invention, the lead/lag circuit 90 varies the water level signal L as a function of reactor power level to compensate for dynamic characteristics of the steam generator. The level error signal from comparator 96, i.e., the sum of the compensated mass flow difference and level error signals, is passed through a proportional integral controller 102, where the gain and reset time constant are also adjusted as a function of power to further compensate for the characteristics of the steam generator. The output of controller 102 is a flow demand signal and is the sum of the level error signal plus the integral of the level error signal. Circuit 102 has steady state characteristic such that when the summed error input signal is zero, the controller output signal is constant. A non-zero summed error input signal is integrated causing the output signal to move toward its maximum (for a positive summed error signal) or minimum (for a negative summed error signal) valve. Thus, any level error is forced in steady-state to equal zero. A signal proportional to nuclear reactor power is filtered in circuit 106. The thus processed power signal is delivered to function generators 110 and 112, where the lag coefficient T4 and the ratio of the lead and lag coefficients T3/T4 is determined as a function of power. The thus determined ratio and lag coefficient are inputted to circuit 90. The filtered power signal is also delivered to function generators 114 and 116 which in turn generate the coefficients T8 and K1 for delivery to the proportional-integral controller 102, the gain control signal K being filtered in circuit 117 before inputting to controller 102. The power adjusted set point for cirucit 90 and controller 102 make it possible to control steam generator level over the full power range by accounting for the non-linear characteristics of the steam generator. The feedflow demand signal from controller 102 is the main signal for establishing the set points for the feedwater pump speed, downcomer valve position, and economizer valve position. The steam generator water level is thus controlled, as a function of the measured level and reactor power level. The control scheme is further refined by utilizing a logic scheme that depends on where the reactor power lies relative to three power operating regimes, i.e., a low power regime, typically covering the range of about 0-15% of the reactor power, an intermediate power range, preferably in the range of about 15-50% power, and a high power range, preferably in the range of about 50-100% power. The power range determines the combination of feedwater pump speed, downcomer valve position, and economizer valve position utilized to control water level. The reactor power signal from circuit 106 is also delivered to switchover control circuits 126 and 128. In the manner to be described below, the state of switchover control circuits 126 and 128 in part determines the analog input signals to the pump speed and valve position function generators 130, 132, and 134 and also exercises on-off control over the economizer valve position control signal outputted from function generator 134. It is to be noted that at high power levels the steam flow rate is a reliable measure of power. The steam flow rate signal W.sub.S, in addition to delivery to comparator 76, is filtered in circuit 170 and applied to a switch control circuit 172. Control circuit 172 provides a digital control signal for a switching circuit 174 when the system power exceeds a predetermined level. In one reduction to practice, this control level, on rising power, was 55% of rated power. The circuit 172 simulates a hysteresis effect such that the switching control signal is discontinued, on decreasing power, when the power falls below 50% of rated. The presence of the control signal at the output of circuit 172 causes application of a steady state demand signal from downcomer bias signal generator 176 to the input of downcomer valve position demand function generator 132 via switching circuit 178. From the discussion above, and as will be described in greater detail below, under high power level conditions the downcomer valve will be open but will not be modulated. Also, the economizer valve will be open and modulated. The feedwater pump(s) speed will similarly be modulated. In the intermediate power range, the downcomer valve will be closed and the economizer valve position and pump speed will be modulated. In the low power range, the economizer valve will be closed, the pump speed will be constant and the downcomer valve will be modulated. Closing of the economizer valve during low power operation is achieved by exercising control over a switching circuit 180 to select a "zero" level economizer valve position demand signal when either of the binary inputs to an OR gate 182 are positive. It may thus be seen that the present invention is effectively a single element system at low power levels and a three-element system at high power levels. It should be noted that the steady state steam generator liquid inventory is greater when feeding the downcomer than when feeding the economizer. For this reason, the switchover control circuit 128 is programmed to provide a hysteresis effect so that there is a delay in the valve action to compensate for the two different inventories. The output of circuit 128 is a digital command signal. Below a predetermined power level, fifteen (15) percent in one reduction to practice, the steam flow and feed flow signals are unreliable. For this reason, the maximum demand signal provided at the output of controller 102 is limited in circuit 140 when the power level is below the predetermined level. This limiting action is obtained by employing the digital control signal from switchover control circuit 126, which functions as a level detector to control a switching circuit 142. Thus, when the NSSS is operating at less than 15% rated power in the example being discussed, the output of controller 102 will be limited. When the NSSS power is above 15% of rated power, the limiting circuit 140 is bypassed. Similarly, the output of level detector 126 is used to control a switching circuit 144 so that the steam flow/feed flow bias signal is removed from comparator 86 in the low power operating regime. Another feature of the invention is the automatic control of refilling the steam generator. After a reactor trip, a tripped override signal will appear, after a delay, at a first input of AND gate 152 and a first input of a trip set/reset circuit 190. If the water level is below a threshold level as established by a level detector 156, the level detector output being delivered as the second input to gate 152, refill of the steam generator is accomplished by a control scheme based on the difference between the average temperature T.sub.AVG of the primary loop and a constant commensurate with the average temperature under no load conditions T.sub.AVGNL. The difference in these temperature signals is generated by a comparator 160 and processed in a proportional/integral controller circuit 162. The output of controller 162 will be a flow demand signal. If this demand signal is between upper and lower limits, as set in a limiting circuit 168, it will be passed to switch 166. The state of switch 166 is controlled by the output of circuit 190 to pass only the limited output of controller 162 until circuit 190 is reset by the change of state of one of the inputs to AND gate 152. The input to gate 152 provided by level detector 156 will, of course, change when the sensed water level exceeds the threshold level. Thus, if the reactor has been tripped and the water level is below normal, the primary loop average water temperature controls the refill rate with the water being supplied via the downcomer valve. Once the water level reaches normal, the control is returned to the automated system described above. The output of set/reset circuit 190 is also applied as an input to OR gates 182 and 192. Thus, until circuit 190 is reset after having been set in response to a reactor trip, gate 182 will apply a signal to switching circuit 180 which will close the economizer valve. Similarly, gate 192 will apply a signal to switching circuit 178 which will cause the output of controller 162 to be applied to function generator 132. At power levels in excess of that predetermined as being in the low range, the output of switchover control circuit 128 will operate a switching circuit 194 to cause closing of the downcomer valve until the bias signal from source 176 is applied in the manner described above. The output of circuit 128, after inversion and delay, is applied to OR gate 182 and closes the economizer valve at low power. The delay in opening and closing the economizer valve provides compensation for mass changes. A high level (HLO) will result in a zero flow demand signal and both the economizer and downcomer valve will be closed. The pump(s) will, however, continue to operate. The disclosed control provides for operator intervention, M/A, at various points in the system whereby full or partial manual control may be exercised. It can be appreciated that, in order to implement the control system described above, the dynamic characteristics of the steam generator and other components have to be considered in order to provide appropriate compensation in the various circuits described above. This information can be obtained by analytical studies or field testing, where the steam generator is subjected to primary side perturbations, including steps, ramp and sinusoidal perturbations. The perturbation techniques are intended to characterize the non-linear behavior of the steam generators. The steps and sinusoids provide an indication of increasing responsiveness of the steam generators at low power and also provide an indication of the delays in the system. FIG. 5 demonstrates some of the steam generator characteristics which adversely affect the controlability of the steam generator water level at low power. The initial influx of feedwater tends to cause a small level rise due to the mass increase in the downcomer. As fluid is accelerated to the tube bundle, because of the increase in the downcomer head, the amount of fluid entering the tube bundle is increased. The net effect is that the steam generator reacts opposite to intuition, at least initially, and has an inherent delay which provides a negative phase shift in an uncompensated control scheme. The present invention takes this phenomena into account. A slow sinusoidal perturbation of feedwater flow rate results in a large phase lag in the steam generator water level response. Thus, as shown in FIG. 6, a relatively slow sinusoidal flowrate perturbation at low powers is manifested by a change in water level delayed by a phase lag approaching 180 degrees, indicating a need for compensation in the controller design. This phonomona is also represented by FIG. 8. FIG. 7 shows the rapid increase in the delay time (time for level to recover back to the initial level) as feedwater temperature is reduced. Since most pressurized water reactors feed relatively cold water into the steam generator at low powers (before there is adequate steam to bring the main turbine on line) the delay time is large and can lead to an instability if the control system is not designed to compensate for the delay. FIG. 8 shows the process gain increasing at low power. For an optimum control system, variable gain is required to maintain the controller performance without affecting stability. Particularly, gain should be varied to compensate for the steam generator response characteristics, including the response characteristics of components such as valves which affect the feedwater flow in the system. Other considerations which should be taken into account include the downcomer level versus water level relationship and the state of the fluid where the feedwater is being injected into the steam generator. Of course, the variation of steam generator response as a function of power level is of major importance and the function generators which provide control signals for power level compensation purposes may need to be adjusted during life of the NSSS. FIG. 9 graphically shows actual startup of a typical NSSS where the thermal power of the system is increased from hot standby to 12%. Initially, steam generator level is maintained manually using an auxiliary feedwater system. At approximately 2% power, feedwater control is manually transferred to the feedwater control system of the present invention. Throughout this period, the steam generator level oscillates as the operator attempts to maintain it. After the system is placed in automatic, at about 3% power, the steam generator level stabilizes at its setpoint. Furthermore, it continues to operate in this stable manner, as power is increased and disturbances are imposed on it due to the placing steam reheater in service, driving control rods and changing blowdown rates. FIG. 10 demonstrates how maintaining the feedwater system under the automatic control of the present invention can help to avoid a steam generator level trip. In this case, the operator took manual control of the system and closed the downcomer valve on the steam generator without simultaneously opening the economizer valve. As noted above, the downcomer valve is used during low power operation and the economizer valve is used for operation between 15 and 50% power. Since both valves were closed the steam generator water level dropped rapidly. The transient was mitigated by opening both control valves and placing the system in automatic. In this mode the system was again able to maintain steam generator level. FIG. 11 demonstrates how the present invention is able to restore steam generator water level following a large perturbation. During the transient, feedwater flow is transferred from the economizer valve to the downcomer valve causing the steam generator operating characteristic to change. This causes an initial swell in steam generator level due to the non-equilibrium condition. The feedwater control system limits the overshoot in level and then successfully restores it to its normal setpoint during this low power operation. While a preferred embodiment has been shown and described, various modifications and substitutions may be made thereto without departing from the spirit and scope of the invention. Accordingly, it is to be understood that the present invention has been described by way of illustration and not limitation. |
051065720 | description | DESCRIPTION OF PREFERRED EMBODIMENT FIG. 1 shows schematically the vessel 1 of a pressurized water nuclear reactor with including above the core (not shown) a closing plate, called core upper plate 2, placed horizontally. Vessel 1 extends above plate 2 via an enclosure 3 terminating in a flange 4 on which a cover 5 is placed and latched. Nozzles 6 of large diameter are provided in the side wall of enclosure 3 for the outflow of the core cooling water. In enclosure 3, a guide assembly 7 is mounted above the upper core plate, each guide being provided at its lower portion with a circular guide flange 8 adapted for bearing on plate 2 while forming with it a narrow transverse clearance 9. Guides 7 extend with their vertical axes 10 perpendicular to plate 2 and are provided for supporting clusters of absorbant rods (not shown in detail), the extent of whose insertion inside the reactor core through plate 2 via an operating member 11 permits the nuclear reaction to be controlled. In FIG. 1, reference 12 denotes schematically the circulation direction of the cooling water which, flowing from bottom to top, flows through the upper plate 2 prior to being discharged outside the guides 7 and then from enclosure 9 through nozzles 6. In a conventional construction of the above type, the accurate positioning of the flanges 8 of the guides 7 for the absorbing clusters with respect to the upper core plate 2 is, whereby each guide has to notably positioned perfectly with respect to a corresponding opening of plate 2 for the passage of the absorbing rods. Simultaneously, the transverse dimension of clearance 9 between plate and flange must be rigorously determined, irrespective of the vibrations created by the flow of cooling water and by the hydrodynamic stresses exerted on the connection means which are used. In particular, the assembly formed by the core plate 2 and the flanges 8 which are above it must be such that possible excessive forces can be absorbed by subsequent return of the two parts to their optimal and predetermined relative positions, and this without prejudice to the mechanical structures used. To this effect and according to the invention, each flange 8 of any guide 7 is positioned with respect to the core plate 2 by means of two diametrically opposite massive spindles 13 and 14 (FIG. 2), these spindles being implanted in the thickness of the flange while protruding outside said flange in the direction of plate 2 and extending inside housings 15 and 16 provided in register in plate 2 with a convenient mounting clearance 17. In a particular embodiment of the invention, spindles 13 and 14 emerge from the lower face of each flange 8 along a height which is sufficient for to permit precise positioning of guide 7 connected to this flange with respect to core plate 2, of the order of about 0.5 mm. These spindles 13 and 14 extend into their housings 15 and 16 with a diametrical clearance 17 which in practice is about 0.3 to 0.4 mm. The two positioning spindles 13 and 14 are associated with an assembly of four self-locking shoes 18, placed as shown in FIG. 2. These shoes are so designed that, as will be seen hereafter, they can exert on flange 8 suitably distributed forces, while allowing in combination with the centering obtained by spindles 13 and 14 the indispensable precise positioning and the permanent alignment of the vertical axes of the passages reserved for the absorbing clusters. Preferably, the four self-locking shoes 18 are distributed by pairs, two on either side of the diametrical plane connecting spindles 13 and 14, the shoes of the same pair exerting on flange 8 forces which are respectively in directions 19a and 19b, converging at a point 19c situated on the transverse axis perpendicular to the direction of the plane connecting spindles 13 and 14. The conjunction of the forces of the four shoes 18 and of spindles 13 and 14 allows them to resist displacement of the cluster guide with respect to upper core plate 2 without permanent and fluctuating dragging effects, thereby ensuring the desired centering of axis 10 of guide 7 connected to flange 8, in alignment with that of the passage (not shown) formed in plate 2 underneath flange 8, through which the cluster of absorbing rods will be more or less permanently inserted in the core underneath the plate, for control of the nuclear reactivity. FIGS. 3 and 4 show a particular embodiment of the self-locking shoe 18 and of the control mechanism associated therewith. This assembly includes a block or shoe 22 as such, mounted with a clearance in a transverse bore 21 formed in flange 8, said shoe 22 having a plane lower face 23 in contact with the upper surface 24 of the core plate 2. Shoe 22 protrudes slightly underneath flange 8 over a height corresponding to the clearance 9 provided between plate 2 and flange 8. At its end which is opposite face 23, shoe 22 has a slanting face 25 for cooperating with a push-piece 26, also mounted with a clearance in bore 21. Shoe 22 and push-piece 26 are connected to one another through a non-rigid connection, permitting their relative axial displacement while also permitting the shoe to be suspended underneath the push-piece, rendering it captive. To this effect, the push-piece is extended downwardly by a flat central rib 27, engaging between the two parallel sides 28 and 29 of a clevis (FIG. 4) formed in the corresponding upper portion of the shoe, the connection between rib 22 and sides 28 and 29 of the clevis being provided by a transverse peg 30, extending through a cylindrical hole 31 of larger diameter formed in rib 27. Push-piece 26 is extended upwardly, particularly toward the upper portion of transverse bore 21, by a narrower stem 32, extending over a distance outside this bore, the stem 32 including two successive axial portions 33 and 34 of different diameters, portion 33 having a diameter larger than that of portion 34 so as to form a shoulder 35 at the junction of the two portions. At its upper end opposite shoulder 35, stem 32 is formed with a slit 36 allowing the relative orientation of the stem with respect to the vertical to be indentified and, the direction of the force exerted by push-piece 26 on stem 22 by the cooperation of the respective surfaces of the two elements mutually bearing on one another to be varied. To this effect and in order to allow an orientation of the resultant force created on shoe 22, push-piece 26 is formed at its lower end with a convex face 37, substantially as a portion of cylinder, and adapted to come to bear against the slanting surface 25 of the shoe, the application of these surfaces on one another producing a resultant of the forces created, particularly for the four associated shoes of the same spindle, according to directions 19a and 19b of FIG. 2. Portion 33 of stem 32 extends with a clearance through an axial passage 38, formed in an end-piece 39 the lower end 40 of which engages over a short distance inside bore 21 of flange 8, the end-piece being then secured against movement with respect to the flange, e.g., by a weld 41. End-piece 39 is extended upwardly by a sleeve 42, defining a chamber 43 in which is also mounted, with a clearance, a spring cartridge 44, including a lower support washer 45 through which extends the smalles diameter portion 34 of stem 32 and which bears at the base of this lower portion on shoulder 35. A pack of conical washers 46 is disposed about portion 34, these washers being thus contained between the lower ring 45 and a complementary washer 47. A variable force can be exerted on the spring cartridge 44, and due to the abutment of washer 45 on shoulder 35, on stem 32 and thense on push-piece 36. To this effect, sleeve 42 is formed with an interior screw thread 48, cooperating with a screw thread 49 having the same pitch 49 formed on the outer surface of a calibration bushing 50. The latter is formed with an axial bore 51 through which extends with a clearance the portion 34 of stem 32, and which includes at its upper end a head nut 52 allowing control of the rotation of the bushing in the sleeve screw thread, by exerting on spring cartridge 44 the required force, of the order of 1000N per shoe. The structure of the spring cartridge may be varied, for example by using, as shown schematically in FIG. 4, double washers 46a or other means allowing an elastic transmission of the efforts of the bushing toward the push-piece while ensuring, in the case of a force in the opposite direction, the absorption of the latter by simple compression of the washer pack between washer 47 which remains immobile and washer 45 which is subjected to a telescoping movement within sleeve 42 and which follows the corresponding limited displacement of stem 32. Shoes 22 carry an appropriate surface coating, which is preferably different for the plane lower face 23 and for the slanting upper face 25 of these shoes. Notably, the surface coating of face 23 is chosen in such manner that it imparts to the surface 24 of upper plate 2 a coefficient of friction which is the highest possible. On the contrary, the coating of convex face 37 of face 25, as well as that of push-piece 26, is chosen with a coefficient which is as small as possible in order to facilitate sliding motion of the respective faces on one another. Preferably also, the outer surfaces of shoe 22 and of push-piece 26, as well as the inner surface of bore 21, carry a surface coating similar to that of faces 25 and 37 for facilitating in the same manner their relative sliding motion when the device is used. The operation of the centering and fixation device according to the invention is the following: Flange 8 with its two fixed spindles 13 and 14 (FIG. 2) is positioned on the upper plate 2 with a clearance in their positioning which corresponds to the clearance 17 of these spindles in their respective housings 15 and 16. After mouthing of the spindles and self-locking shoes on the flange of the cluster guides, the spring cartridges are calibrated, and the cluster guides are then mounted on the support structure above the upper core plate. After mounting of the two fixed spindles 13 and 14 on flange 8, one disposes on end-piece 39 the push-piece 32, the washer 45, the spring cartridge 44, the washer 47, the calibration sleeve 50 (without tightening it), the shoe 22 on push-piece 32, and its axis 30. The four end-pieces thus fitted out are positioned on flange 8 and attached, e.g., by soldering. Underneath flange 8 is fixed a wedge of a thickness corresponding to the predetermined space 9, then the tightening is carried out with the assistance of a calibration sleeve 50 of spring cartridge 44 to a predetermined value, taking into account the relaxation of the service elements, the calibration tolerances, the variations of clearance 9, so that the force exterted by the push-piece on the shoes always remains greater than to a calculated value, i.e., about 600N. The tightening of the calibration sleeve is carried out by one of the conventional methods for obtaining a preload: either a torque or an angular stroke after being in contact, or a force cell placed underneath the shoe, or a combination of these methods. The shoes 23 and the push-pieces 32 are oriented with the assistance of slit 36 according to directions 19a and 19b, and then the rotation of calibration nut 32 is braked. The setting plate is then removed, to release the preload of the spring cartridges. The cluster guides are then mounted on the support structure; This positions flanges 8 above the core plate 2. About 20 mm prior to reaching clearance 9, the fixed spindles 13, 14 engage into the respective housing 15 and 16 with a small clearance 17, thereby ensuring a correct orientation of the cluster guide on the upper core plate 2. About 5 mm prior to reaching clearance 9, face 23 of the shoes comes in contact with the core plate 2, the shoes push back push-pieces 32 upwardly, and the latter compress the spring cartridges to the preestablished mounting value. The four shoes 18 are controlled so that the resultant of the forces exerted on these shoes taken two by two, as is also shown in FIG. 2, converges at points symmetrical to one another in a median diametrical plane, perpendicular to that which joins spindles 13 and 14. To this effect, shoes 22, the lower plane faces 23 of which are in contact with surface 24 of plate 2, are subjected by the calibration bushings 50 to an appropriate force in the direction of the plate, stems 32 being previously oriented by their slit 36 in the desired direction. The screwing of bushings 50 is pursued until, due to the lateral displacement of shoes 22 created by the cooperation of the convex face 37 of push-piece 26 with the slanting face 25 of the corresponding shoe, the initially defined diametrical clearance between the shoe and its bore is adjusted, the resilient washers 46 of cartridge 44 being compressed within the corresponding limit. In this position, flange 8 is perfectly positioned with respect to the upper core plate 2, while leaving between these members only the clearance 9 which is necessary for the differential expansion of flange 8 and of the structures connected thereto. The calibration of spring cartridge 44 takes into account the relaxation of the elements in service and the calibration tolerances, so that the force exerted by the push-piece on the self-locking shoes remains substantially uniform, even when clearance 9 varies during the thermal cycles resulting from the operation of the reactor and of the differential expansions resulting therefrom. This axial force, multiplied by the coefficient of friction between the shoes 22 and the upper core plate 2, thus creates a lateral component opposing the transverse displacement of flang 8 under the effect of the hydrodynamic stresses of which these members are the seat, particularly due to the flow of cooling water through them. The respective positions of the shoe and of the push-piece change from those of FIG. 6 to those of FIG. 5, due to the inclination of the bearing faces 25 and 37, the effect of which is to cancel the clearance of shoe 22 and of its housing within flange 8. The invention is not limited to the embodiment more especially described and shown hereabove. In particular and as is shown in FIG. 7, the mounting in two portions of shoe 22 and of push-piece 26 connected via their transverse peg 30 can be replaced by another embodiment shown schematically in FIG. 7, where the lower end of portion 33' of the stem again includes a push-piece 26' acting on the shoe 22' bearing on surface 24 of plate 2, but in which the shoe is made captive due to a retracted bearing surface 53 formed in the bottom of the flange and against which bears, notably in the flange transportation position, a collar having the same profile 54 as said shoe. |
abstract | An apparatus that acts as a shield for radiopharmaceuticals and protects individuals from radioactivity includes a first body with a first hollow core, a second body with a second hollow core and a third body with a third hollow core. The first hollow core, second hollow core and third hollow core collectively house an insert. The insert houses a hypodermic syringe. A first connection means releasably communicates the first body with the second body. A second connection means releasably communicates the first body with the third body. The second body comprises a piston actuator. The piston actuator can be operated to depress the piston of the hypodermic syringe while the first body is in communication with the second body, and while the third body is removed. |
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claims | 1. An apparatus for a scanning microscope, in particular a scanning force microscope, comprising a measurement probe which defines a near field, and having a scanning unit which allows the measurement probe to move relative to a sample in all three spatial directions, in conjunction with a mass spectrometer with an ionization unit, an extraction unit and an analysis unit, wherein the measurement probe has a hollow tip so that the near field of the measurement probe can be used by the ionization unit in such a way that ions are formed only in the near field of the measurement probe, and the shape of the measurement probe allows an essentially axially symmetrical field distribution of the extraction unit with respect to the axis of the analysis unit. 2. The apparatus as claimed in claim 1, wherein the measurement probe is a cantilever. 3. The apparatus as claimed in claim 1, wherein the sample can be moved in all three spatial directions by means of the scanning unit. 4. The apparatus as claimed in claim 1, wherein the ionization unit has a laser, light beams which are indicated by the laser are focused off-axis and are then deflected by means of a mirror in an axial direction, with the mirror having an axial hole which allows the ions to pass through to the analysis unit. 5. The apparatus as claimed in claim 1, wherein the ionization unit has a laser, and light beams which are indicated by the laser are deflected by means of a mirror in an axial direction and are then focused by means of a focusing device, with the mirror and the focusing device each having an axial hole which allows the ions to pass through to the analysis unit. 6. The apparatus as claimed in claim 1, wherein the ionization unit has a laser, and light beams which are indicated by the laser are passed to the measurement probe and cause ionization in the near field of the measurement probe by means of field amplification. 7. A method for high-resolution examination of a measurement sample using a combined scanning probe microscope, in particular a scanning force microscope, wherein the scanning probe microscope is first of all used to record an image of the measurement sample, in particular of the topography of the measurement sample, and wherein a mass spectrometer is then used for destructive, chemical characterization of at least subareas of sections of the measurement sample which are covered by the image,wherein the information from scanning probe microscopy and from mass spectrometry can be compared with high lateral resolution. 8. The method as claimed in claim 7, wherein the selected areas are chosen successively such that the entire area imaged by the scanning probe microscope is analyzed, thus additionally resulting in a chemical image of the sample. 9. The method as claimed in claim 7, wherein further ablation of the measurement sample leads to high-resolution depth information. 10. The method as claimed in claim 7, wherein the distance between two points for ionization can be chosen by analysis of the area ablated by an ionization process, such that this leads to uniform ablation of the measurement sample. 11. An apparatus for a scanning microscope, comprising a measurement probe which defines a near field and a scanning unit which allows the measurement probe to move relative to a sample in all three spatial directions in conjunction with a mass spectrometer with an ionization unit, an extraction unit and an analysis unit, wherein the measurement probe has a hollow tip so that the near field of the measurement probe can be used by the ionization unit such that ions are formed only in the near field of the measurement probe, and a shape of the measurement probe allows an essentially axially symmetrical field distribution of the extraction unit with respect to an axis of the analysis unit. 12. The apparatus as claimed in claim 11, wherein the measurement probe is a cantilever. 13. The apparatus as claimed in claim 11, wherein the sample can be moved in all three spatial directions by means of the scanning unit. 14. The apparatus as claimed in claim 11, wherein the ionization unit has a laser, and light beams from the laser are focused off-axis and are then deflected by means of a mirror in an axial direction, the mirror having an axial hole which allows the ions to pass through to the analysis unit. 15. The apparatus as claimed in claim 11, wherein the ionization unit has a laser, and light beams from the laser are deflected by means of a mirror in an axial direction and are then focused by means of a focusing device, the mirror and the focusing device each having an axial hole which allows the ions to pass through to the analysis unit. 16. The apparatus as claimed in claim 11, wherein the ionization unit has a laser, and light beams from the laser are passed to the measurement probe and cause ionization in the near field of the measurement probe by means of field amplification. 17. A method for high-resolution examination of a measurement sample using a combined scanning probe microscope, comprising the steps of:using the scanning probe microscope to record an image of the measurement sample; andusing a mass spectrometer for destructive, chemical characterization of at least subareas of sections of the measurement sample which are covered by the image; andfurther comprising a step of comparing information from the scanning probe microscope and from the mass spectrometer with high lateral resolution. 18. The method as claimed in claim 17, further comprising a step of selecting areas successively such that an entire area imaged by the scanning probe microscope is analyzed, thus additionally resulting in a chemical image of the measurement sample. 19. The method as claimed in claim 17, further comprising a step of ablating the measurement sample, leading to high-resolution depth information. 20. The method as claimed in claim 19, further comprising a step of choosing a distance between two points for ionization by analysis of an area following said step of ablation, resulting in uniform ablation of the measurement sample. |
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046541837 | description | DESCRIPTION OF THE PREFERRED EMBODIMENTS At the outset, it should be understood that the novel process of the invention involves the principle that H.sup.- ions can be produced with spin polarized protons. Utilizing selective laser neutralization of only one proton spin state of such a multi-ampere H.sup.- ion beam, while the beam is subjected to a magnetic field, results in an H.sup.- ion beam and an atomic hydrogen beam that has polarized protons. Those two beam states are then easily separated from one another by bending the H.sup.- ions with a curved magnetic field, as will be more full described below. Such an essentially one-step process has the obvious advantage that multi-ampere H.sup.- ion beams can be produced having either pulse or steady state. Furthermore, commercially available lasers can be used in the process to achieve essentially 100 percent efficiency in the neutralization of the H.sup.- beam. It should be understood that negative deuterium, D.sup.-, ions can be produced with spin polarized protons, using the principles of the invention, as they are disclosed herein with particular reference to H.sup.- ions. Accordingly, for fusion application requiring D.sup.- ions, the present invention is of particular value. FIGS. 1 and 2 illustrate alternative arrangements of various commercially available components that are used in practicing the preferred process of the invention. Considering first the arrangement shown in FIG. 1, it will be seen that a suitable negative ion source 1 is arranged to produce a multi-ampere beam of H.sup.- ions through an adjacent trimmer and focusing means 2. The ion beam is then introduced into a magnetic field, which is arranged in conventional manner to define a suitable magnetic field line 3, that is shown schematically by the curved arrows in FIG. 1. A commercial available laser 4, which will be more fully described below, is positioned to direct a beam of laser light 4A through an apperture 5A in convex mirror 5 and against the convex mirror 6. The beam of laser light 4A is oriented co-linear with the vertical section of the magnetic field line 3 to define a reflecting cavity, as shown in FIG. 1. As will be more fully discussed below, the magnetic field line 3 is bent in the area 3A, so that the magnetic field is effective to bend H.sup.- ions in an intense ion beam away from the H.sup.o atoms that are shown by the depicited arrow in FIG. 1 as being directed in a beam that maintains its alignment with the vertical portion of the magnetic field line 3. A suitable conventional particle accelerator 7 is positioned to receive the H.sup.- beam and to focus and accelerate it for any desired high energy research experimentation purpose. The arrangement of components used in practicing the process of the invention shown in FIG. 2 is quite similar to that illustrated in FIG. 1, thus the same call-out numbers will be used in FIG. 2, for designating essentially identical component parts. There is shown in FIG. 2 a suitably grounded negative ion source 1 that directs a multi-ampere beam 1A of H.sup.- ions through a focusing and trimming device 2 into a magnetic field line designated by the arrowed line 3. In this arrangement of the process of the invention, a laser 4 is positioned to direct a beam of laser light 4A in a direction that causes it to intersect the beam of each H.sup.- ions at an angle transverse thereto. Thus, the convex mirrors 5 and 6 are positioned to define a cavity that has its longitudinal axis essentially perpendicular to the vertical section of the magnetic field line 3. Otherwise, the process implemented with the component parts shown in FIG. 2 is arranged to operate in essentially the same manner as that shown in FIG. 1. Accordingly, H.sup.- ions are bent by the curved section 3A of the magnetic field line 3, thereby to direct the spin polarized H.sup.- ions in an intense beam toward the accelerator 7, as explained above with reference to FIG. 1. From the foregoing description of the process of the invention it should be apparent that a variety of suitable conventional components can be used for the negative ion source 1, the focusing and trimming means 2, the means for defining the magnetic field line 3, the laser 4 and associated laser cavity mirrors 5 and 6, as well as for the accelerator 7, as discussed above. However, particular examples of the types of lasers that are most useful in practicing the invention will be set forth more fully below. The magnetic field line 3 used in either arrangement shown in FIGS. 1 and 2, in practicing the preferred process of the invention, is formed as a uniform solenoidal magnetic field by use of any well known conventional array of suitable permanent or electro-magnets. When H.sup.- ions are provided in the multi-ampere beam of H.sup.- ions 1A that enters the magnetic field line 3 and is subjected to the uniform solenoidal magnetic field therein, one of the electrons in each H.sup.- ion will have its spin aligned in the direction of the magnetic field, while the other electron in each H.sup.- ion will have its spin aligned in the opposite direction. The proton spin in each of the H.sup.- ions will also be aligned either with, or opposite to, the magnetic field. For magnetic fields below 10.sup.8 Gauss, the H.sup.- ions have only one bound state, accordingly, there will be two populations of H.sup.- ions between which the sole difference is the orientation of their respective proton spins. The principle applied in practicing the method of the invention is to selectively neutralize only one population, or given spin polarized group, of H.sup.- ions, thereby forming a population of H.sup.- ions each of which has its proton spin down, and a second group or population of H.sup.o atoms having proton spin up. After such selective neutralization of the two groups of ions is achieved, the negative ions in one group are readily separated from the neutral ions in the other group by passing the polarized ion beam through the curvature of an appropriately arranged magnetic field, such as through the bent portion 3A of the magnetic field line 3 shown in FIGS. 1 and 2. The desired selective neutralization of ions works when there are proton spin dependent states that are preferentially formed, i.e., selective neutralization of H.sup.- ions will be the result of selective formation of H.sup.o atoms. Various possibilities could be explored in perfecting alternative arrangements of the method of the invention, for example, formation of H.sup.o atoms in the ground state, and formation of excited hydrogen atoms, in the n=2 level. A major advantage of operating near the threshold, i.e. with H.sup.o atoms in the ground state, is the more ready commercial availability of suitable lasers; however, for such atoms the photodetachment cross section is low. On the other hand, while the cross section for the formation of excited hydrogen atoms is 11 orders of magnitude higher, presently available commercial lasers that operate at that wavelength of 1135 .ANG. have short pulses. To further explain the preferred embodiment of the inventive process disclosed herein, the description of the process will now consider photodetachment of H.sup.- ions which result in ground state H.sup.o atoms, because more data exists for that case, including experimental clarification of it. As pointed out above, it will be seen that D.sup.- ions can also be as readily formed by practicing the invention as taught herein. There exists in the prior art extensive data about photodetachment of H.sup.- ions, but the effects of an external magnetic field on the photodetachment cross section has not been incorporated in that pre-existing data, as was done with respect to the photodetachment of S.sup.- in the presence of a magnetic field, per the explanation in the above referenced paper by Blumberg et al. Selective neutralization for achieving nuclear polarized beams has not been discussed anywhere, so far as the present inventor is aware. The main difference realized between achieving photodetachment in the presence of a magnetic field, versus achieving such photodetachment in the absence of a magnetic field, is the resultant quantization of the energy of the free electrons formed in a magnetic field into the well known Landau levels. When photodetachment is effected in a magnetic field, the electron dipole magnetic moment .mu. and the external magnetic field B combined to produce an energy shift from each Landau level of .mu..B. Accordingly, in each such shifted level, there can be either an electron spin up from one Landau level, or an electron spin down from a higher Landau level. Because the lowest energy level in the ground state hydrogen atom H.sup.o is produced with electron spin down and proton spin up, the lowest energy state of the detached H.sup.- ion is an H.sup.o ion with electron spin down, proton spin up, and a free electron spin up. Referring to FIG. 3(a,b), it will be seen that there is shown combined energy levels of a stripped H.sup.- ion in a magnetic field of a few hundred Gauss. These illustrated combined energy levels show the added energy levels of the free electron and the H.sup.o atom in the ground state. The depicted Landau levels refer to the energy levels of a free electron in a magnetic field, with the levels being quantized perpendicular to the field. As the term "degenerate levels" is used in FIGS. 3a and 3b, it means that the depicted levels are either truly degenerate, or are very closely spaced. As shown by the equations in FIGS. 3a and 3b, energy levels of a stripped electron and an H.sup.o atom is the total energy of a free electron, which is shown to equal the sum of the Landau levels, the Zeeman energy and the energy of the H.sup.o atom. FIG. 3a shows those energy levels for magnetic fields below the critical field, whereas those energy levels are shown in FIG. 3b for magnetic fields above the critical field. As shown by FIGS. 3a and 3b and the foregoing discussion, there exists the possibility of selectively detaching a spin up electron from H.sup.- ions with a proton spin up. In terms of frequency units, the energy difference between hydrogen atoms with proton spin down and those with proton spin up (with the electron spin down), is over 400 MHz, for magnetic fields of a few hundred Gauss up to fields of about 1000 Gauss. Thus, it can be seen that from a qualitative standpoint a resolution of about 100 MHz is sufficient to achieve such selective detachment. Presently available commercial lasers have much better resolutions than 100 MHz. Because the beam of laser light 4A that is used in the process of the invention, can be made arbitrarily narrow, the effective Doppler broadening of the laser beam, as seen by H.sup.- ions due to their thermal spread, becomes a limiting factor. In the absence of any acceleration, the Doppler width of a beam from a conventional H.sup.- ion source is much larger than any hyperfine separation or splitting attainable, e.g. at one electron volt the Doppler width is about 8.5.times.10.sup.9 Hz. However, if the beam is accelerated a phenomenom known as kinematic compression occurs. In other words, there is a reduction in Doppler width due to Doppler shift, which can be easily compensated for by adjusting the laser frequency. The factor by which Doppler width is reduced, R=1/2(kT/eU).sup.1/2 where T is the beam temperature and U is the accelerating potential. Accordingly, the reduced Doppler width becomes .DELTA.V=.DELTA.V (O) R. In the case of a typical currently available H.sup.- ion source, such as that available from the accelerator now in operation at Brookhaven National Laboratory, Upton, N.Y., which produces H.sup.- ions having a thermal spread of about 4 eV, extracted at 20 kV, .DELTA.V approximates 120 MHz. Such Doppler broadening is already acceptable for use in the process of the present invention, and it can readily be further reduced if desired. In practicing the preferred arrangement of the process of the invention, a Penning source whose thermal spread is about 1 eV is used. Thus, the selective neutralization effected in the process can be done by making the H.sup.- ion beam 1A from the negative ion source 1 shown in FIGS. 1 and 2 have energy of about 1 KeV. To select a suitable laser, for the laser 4 used in practicing the preferred arrangement of the process of the invention, as shown in FIGS. 1 and 2 it should be understood that the photon flux i.e., P (photon/cm.sup.2 /second) can be estimated from the equation: EQU p.perspectiveto..phi.Pt (1) Where, .phi. is the photodetachment cross-section and t is the intersection time. At one KeV for an intersection region of about 4 meters, t approximates 10.sup.-5 second. At the threshold, the photodetachment cross section when extrapolated from experimental and theoretical results is seen to be only 10.sup.-24 cm.sup.2. Solution of equation (1) yields P approximately 10.sup.29 photons/cm.sup.2 /sec. Assuming a beam cross section A approximating 0.1 cm.sup.2, the required Power for 0.75 eV photons is: EQU Power=APhv=10.sup.11 watts (2) In the foreseeable future there seems little hope of achieving such a Power level. The purpose of showing this calculation is to indicate that the primary problem in this case of a single photon absorption stems from low cross section near threshold while it peaks at about 1.5 eV photons. By contrast, the double photon absorbtion cross section, which requires the use of a 3.2 micron laser, peaks at threshold. Since the H.sup.- ion has a very high degree of dynamic polarizability, the cross section for double photon absorption must be high at threshold. An alternative approach is to use 10.93 eV photons (i.e. use a 1135 .ANG. laser), whose photo detachment products are a free electron and an H.sup.o atom excited in the n=2 level. At this photon energy, the photodetachment cross section has a very sharp resonance whose magnitude is 1.4.times.10.sup.-15 cm.sup.2. Using this value of cross section in equation (1), above, and a hv of 10.93 eV in equation (2) the Power needed becomes: EQU Power=12.5 watts (3) Lyman Alpha lasers are available having Power levels of 100 watts in pulses of 10's of nanoseconds. Utilization of such a laser for the laser 4 shown in FIG. 2 is the preferred arrangement for practicing the process of the invention. In alternative arrangements of the process of the invention, a commercial 32000 .ANG. laser can be used for the laser 4 shown in FIGS. 1 and 2, in a double photon absorption case having a large cross section, i.e. where the H.sup.o atom is left in the n=1 level. In a further alternative process, a 1135 .ANG. laser could be used for the laser 4 shown in FIG. 2, with the resulting H.sup.o atom in the n=2 level. With presently available commerical lasers, the latter approach would limit use of the process to short pulses in the 10's of nanoseconds. If free electron lasers become available in the future, the limits of the process would be greatly expanded. The operation of the preferred arrangements of the process of the invention, as illustrated for example in FIGS. 1 and 2, will now be briefly summarized to better explain the steps of the process. To achieve selective neutralization of H.sup.- ions in a magnetic field, thereby to produce an intense negative hydrogen ion beam with spin polarized protons, it will be understood from the foregoing description of the invention that a first step of the process, as it is illustrated in FIG. 4, is to provide a suitable multi-ampere beam 1A of H.sup.- ions from the negative ion source 1, which may comprise any suitable source, such as one of the available H.sup.- ion beam lines now in operation at Brookhaven National Laboratory. In the next step of the process a beam of laser light 4A is provided from a suitable laser 4. As explained above, such a suitable laser 4 is a 1135 .ANG. laser whose photodetachment products are a free electron and an H.sup.o atom excited in the n=2 level, in the most preferred arrangement of the process of the invention. In alternative arangements of the process, the beam of laser light 4A is produced by a 32,000 .ANG. laser, or in still other alternatives of the process, the laser 4 that is used is in the range of 1135 .ANG. to 32,000 .ANG.. Furthermore, if the laser light beam 4A is polarized other advantageous selection rules apply; thus, for 1135 .ANG. and 16,000 .ANG. polarized laser light the process works well without requiring a finely tuned laser. In the next step of the preferred process, a uniform solenoid magnetic field, as indicated by the magnetic field line 3 in FIGS. 1 and 2, is provided around a portion of the beam of H.sup.- ions 1A, to effectively spin polarize the H.sup.- ions in that beam and thereby to produce a first group of ions having their proton spin aligned with the magnetic field designated by magnetic field line 3, and to produce a second group of ions with their proton spin opposite to the magnetic field. Then, the beam of laser light 4A is directed through the spin polarized H.sup.- ions, either co-axially therewith as shown in FIG. 1, or transverse thereto as shown in FIG. 2, in order to selectively neutralize a majority of the ions in one of the above mentioned groups of ions, without neutralizing the ions in the other group. Finally, one of the groups of ions is separated from the other group of ions and then directed in an intense H.sup.- ion beam toward a predetermined objective, such as the accelerator 7 shown in FIGS. 1 and 2. Of course, the ion beam may be further directed by the accelerator 7 to a desired end use. In the most preferred process of the invention, as mentioned above, the arrangement of components shown in FIG. 2 is utilized, and the magnetic field designated by the magnetic field line 3 is below 10.sup.8 Gauss. Most preferably, that magnetic field is at least 200 Gauss and is sufficiently strong to result in Zeeman hyperfine splitting of the H.sup.- ion beam into two energy states that are solely dependent on the polarization of the respective nuclei of the H.sup.- ion beam. Also, in the most preferred arrangement of the process of the invention, the selected negative ion source 1 is operable to produce an intense beam of H.sup.- ions 1A that is a multi-ampere beam. In such an arrangement of the process of the invention, the first group of H.sup.- ions is preferably neutralized, while the second group of H.sup.- ions is separated from that first group of ions and formed into an intense beam of H.sup.- ions that is then directed into the accelerator 7. In that operation of the invention, the second group of ions is separated from the first group of ions by curving the longitudinal axis of the magnetic field line 3, as shown by the bend 3A therein, in order to bend the intense beam of H.sup.- ions in a path that diverts the second group of ions away from the neutralized ions, which are shown by the symbol H.sup.o and the associated arrow in FIG. 2, so that the H.sup.- ions, as shown by the symbols adjacent the accelerator 7 in FIG. 2, are directed into that accelerator. From the foregoing description it will be recognized that further alternatives and modifications of the invention may be practiced without departing from its true scope. Accordingly, it is my intention to include all such alternatives and modifications within the limits and spirit of the following claims. |
summary | ||
046577254 | claims | 1. A nuclear reactor having a core and water coolant flowing therethrough, said core consisting of prismatic assemblies of identical cross-section arranged vertically and side by side and each comprising a cluster of vertical fuel rods spaced to permit circulation of the water coolant, said fuel assemblies comprising: a plurality of first fuel assemblies whose fuel rods contain initially contain enriched uranium only, certain of the fuel rods being substituted which guide tubes in the cluster and said fuel rods being sufficiently spaced for the neutron energy spectrum within said furst fuel assemblies to be thermal at least when there is water within said guide tubes, and a plurality of second fuel assemblies in which all fuel rods initially contain recovered plutonium only and are so spaced that the neutron energy spectrum within said second fuel assemblies is higher than thermal, said first and second fuel assemblies being distributed throughout the reactor core. 2. A nuclear reactor as claimed in claim 1, wherein said plurality of first fuel assemblies includes fuel assemblies whose guide tubes are arranged for receiving spectral shift control rods only and fuel assemblies whose guide tubes include guide tubes arranged to receive reactor reactivity control rods and guide tubes arranged to receive spectral shift control rods. 3. A nuclear reactor as claimed in claim 1, wherein said fuel assemblies are of hexagonal cross-section and wherein each of said second fuel assemblies is surrounded by first fuel assemblies, in the central part of said core. 4. A nuclear reactor as claimed in claim 1, wherein the distance between adjacent fuel rods in said second fuel assemblies is substantially equal to one third of the distance between the fuel rods in the first fuel assemblies. 5. A nuclear reactor as claimed in claim 1, wherein said first fuel assemblies and second fuel assemblies are distributed evenly throughout the core. |
claims | 1. A collimator for use in single photon emission computed tomography (SPECT), which collimator comprises:a first layer comprising at least three spaced apart elongated slats forming a first array extending in a first direction; anda second layer comprising at least three spaced apart elongated slats forming a second array extending in a second direction orthogonal to said first direction,said first array having a width extending across said second direction, said second array having a width extending across said first direction, wherein each elongated slat of each array has a length extending across the entire width of the other array,each of said slats constructed of a radiation attenuation material. 2. The collimator of claim 1, wherein the space between said slats is fixed and non-variable. 3. The collimator of claim 2, wherein the space between said slats is fixed by foam. 4. The collimator of claim 2, wherein the space between said slats is fixed by guide plates having grooves into which ends of said slats are positioned. 5. The collimator of claim 2, wherein the space between said slats is fixed by grooves in the top of said first layer and grooves in the bottom of said second layer. 6. The collimator of claim 2, wherein each of said slats in a layer are tilted at an angle greater than zero and all of said slats in a layer are tilted in the same direction. 7. The collimator of claim 1, wherein the space between said slats is variable. 8. The collimator of claim 7, wherein the space between said slats at one end of said slats is less than the space between said slats at the other end of said slats. 9. The collimator of claim 8, wherein the space between the slats is varied by application of a force to both sides of the layer at one end of said slats. 10. The collimator of claim 8, wherein each of said slats in a layer are tilted at an angle greater than zero and all of said slats in a layer are tilted in the same direction. 11. The collimator of claim 7, wherein the space between said slats is varied through use of springs. 12. The collimator of claim 7, wherein the space between said slats is varied through use of plastic having air bubbles. 13. The collimator of claim 7, wherein the space between said slats is varied through use of magnetic force. 14. The collimator of claim 1, wherein each of said slats in a layer are tilted at an angle greater than zero and all of said slats in a layer are tilted in the same direction. 15. A nuclear imaging acquisition system for use in single photon emission computed tomography (SPECT), which system comprises:a collimator comprising a first layer comprising at least three spaced apart elongated slats forming a first array extending in a first direction and a second layer comprising at least three spaced apart elongated slats forming a second array extending in a second direction orthogonal to said first direction,said first array having a width extending across said second direction, said second array having a width extending across said first direction, wherein each elongated slat of each array has a length extending across the entire width of the other array, each of said slats constructed of a radiation attenuation material; anda detector having a side which detects radiation emanating from an object after passing through said collimator. 16. The nuclear imaging acquisition system of claim 15, wherein the space between said slats is fixed and non-variable. 17. The nuclear imaging acquisition system of claim 16, wherein the space between said slats is fixed by foam. 18. The nuclear imaging acquisition system of claim 16, wherein the space between said slats is fixed by guide plates having grooves into which ends of said slats are positioned. 19. The nuclear imaging acquisition system of claim 16, wherein the space between said slats is fixed by grooves in the top of said first layer and grooves in the bottom of said second layer. 20. The nuclear imaging acquisition system of claim 16, wherein each of said slats in a layer are tilted at an angle greater than zero and all of said slats in a layer are tilted in the same direction. 21. The nuclear imaging acquisition system of claim 15, wherein the space between said slats is variable. 22. The nuclear imaging acquisition system of claim 21, wherein the space between said slats at one end of said slates is less than the space between said slats at the other end of said slats. 23. The collimator of claim 22, wherein the space between the slats is varied by application of a force to both sides of the layer at one end of said slats. 24. The nuclear imaging acquisition system of claim 21, wherein the space between said slats is varied through use of springs. 25. The nuclear imaging acquisition system of claim 21, wherein the space between said slats is varied through use of plastic having air bubbles. 26. The nuclear imaging acquisition system of claim 21, wherein the space between said slats is varied through use of magnetic force. 27. The nuclear imaging acquisition system of claim 21, wherein each of said slats in a layer are tilted at an angle greater than zero and all of said slats in a layer are tilted in the same direction. 28. The nuclear imaging acquisition system of claim 15, wherein each of said slats in a layer are tilted at an angle greater than zero and all of said slats in a layer are tilted in the same direction. |
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description | This application claims the benefit of priority from U.S. Provisional Patent Application Ser. Nos. 61/136,134, filed Aug. 14, 2008, and 61/193,510, filed Dec. 4, 2008, the contents of both of which are incorporated herein by reference in their entireties. The present invention relates to radiation sources and to methods of generating radiation. A lithographic apparatus is a machine that applies a desired pattern onto a substrate, usually onto a target portion of the substrate. A lithographic apparatus can be used, for example, in the manufacture of integrated circuits (ICs). In that instance, a patterning device, which is alternatively referred to as a mask or a reticle, may be used to generate a circuit pattern to be formed on an individual layer of the IC. This pattern can be transferred onto a target portion (e.g. comprising part of, one, or several dies) on a substrate (e.g. a silicon wafer). Transfer of the pattern is typically via imaging onto a layer of radiation-sensitive material (resist) provided on the substrate. In general, a single substrate will contain a network of adjacent target portions that are successively patterned. Lithography is widely recognized as one of the key steps in the manufacture of ICs and other devices and/or structures. However, as the dimensions of features made using lithography become smaller, lithography is becoming a more critical factor for enabling miniature IC or other devices and/or structures to be manufactured. A theoretical estimate of the limits of pattern printing can be given by the Rayleigh criterion for resolution as shown in equation (1): CD = k 1 * λ NA PS ( 1 ) where λ is the wavelength of the radiation used, NAPS is the numerical aperture of the projection system used to print the pattern, k1 is a process dependent adjustment factor, also called the Rayleigh constant, and CD is the feature size (or critical dimension) of the printed feature. It follows from equation (1) that reduction of the minimum printable size of features can be obtained in three ways: by shortening the exposure wavelength λ, by increasing the numerical aperture NAPS or by decreasing the value of k1. In order to shorten the exposure wavelength and, thus, reduce the minimum printable size, it has been proposed to use an extreme ultraviolet (EUV) radiation source. EUV radiation sources are configured to output a radiation wavelength of about 13 nm. Thus, EUV radiation sources may constitute a significant step toward achieving small features printing. Such radiation is termed extreme ultraviolet or soft x-ray, and possible sources include, for example, laser-produced plasma sources, discharge plasma sources, or synchrotron radiation from electron storage rings. Radiation may be produced using a plasma. The plasma may be created, for example, by directing one or more laser beam pulses at a fuel, such as droplets of a suitable material (e.g. tin), or a stream of a suitable gas or vapour, such as Xe gas or Li vapor. The resulting plasma emits radiation, e.g., EUV radiation, which is collected using a collector such as a mirrored collector, that receives the radiation and focuses the radiation into a beam. The radiation emitting plasma and the collector may together be considered to comprise a radiation source. A radiation source which generates plasma by directing a laser onto fuel droplets, as described above, is often referred to as a laser produced plasma (LPP) source. It is desirable to provide, for example, a method and/or apparatus that increases the time taken for a droplet of fuel to pass through a laser beam pulse. Such an increase in time results in an increased time during which a plasma, and therefore radiation, may be generated. Alternatively or additionally, it is desirable to provide, for example, a method and/or apparatus that allows the control of the angle between the direction of movement of a stream of droplets of fuel and a laser beam. According to a first aspect of the invention, there is provided a radiation source configured to generate radiation, the radiation source comprising: a fuel droplet generator constructed and arranged to generate a stream of droplets of fuel that are directed to a plasma generation site; a laser constructed and arranged to generate a laser beam that is directed to the plasma generation site, an angle between the direction of movement of the stream of droplets and the direction of the laser beam being less than about 90°, and a collector constructed and arranged to collect radiation generated by a plasma formed at the plasma formation site when the beam of radiation and a droplet collide, the collector being configured to reflect the radiation substantially along an optical axis of the radiation source, wherein the laser beam is directed to the plasma generation site through an aperture provided in the collector. The radiation source may further comprise a controller for controlling the direction of the laser beam and/or the direction of movement of the stream of droplets to control the angle between the laser beam and the stream of droplets. The controller may be configured to control the position or orientation of the fuel droplet generator and/or the laser to control the angle between the direction of the laser beam and the direction of movement of the stream of droplets. The stream of droplets may be directed such that the stream of droplets has a component of movement along the optical axis of the radiation source which is pointing away from the collector. The laser beam and the stream of droplets may be directed such that the laser beam and the stream of droplets have a component of movement along the optical axis of the radiation source which is pointing away from the collector. The radiation source may further comprise a catchment arrangement for catching droplets that are not directed toward the plasma generation site. The laser beam may be directed along the optical axis of the radiation source. The stream of droplets may be directed in a direction that is not perpendicular to the optical axis of the radiation source. The laser beam and the stream of droplets may be directed to the plasma generation site through an aperture provided in the collector. The collector may be a normal incidence collector. The radiation source may be configured to generate EUV radiation. According to a second aspect of the invention, there is provided a method of generating radiation, comprising: directing a stream of droplets of fuel to a plasma generation site; directing a laser beam to the plasma generation site, an angle between the direction of movement of the stream of droplets and the direction of the laser beam being less than about 90°, and using a collector to collect radiation generated by a plasma formed at the plasma formation site when the beam of radiation and a droplet collide, and reflecting the radiation substantially along an optical axis of the radiation source, wherein one of the laser beam and the stream of droplets are directed to the plasma generation site through an aperture provided in the collector. The second aspect of the invention may include, where appropriate, one or more features of the first aspect of the present invention. According to a third aspect of the invention, there is provided a radiation source configured to generate radiation, the radiation source comprising: a fuel droplet generator constructed and arranged to generate a stream of droplets of fuel that are directed to a plasma generation site; a laser constructed and arranged to generate a laser beam that is directed to the plasma generation site, and a controller for controlling the direction of the laser beam and/or the direction of movement of the stream of droplets to control the angle between the laser beam and the stream of droplets. The controller may be configured to control the position or orientation of the fuel droplet generator and/or the laser to control the angle between the direction of the laser beam and the direction of movement of the stream of droplets. The direction of the laser beam and/or the stream of droplets may be controlled to control the time taken for a droplet of the stream of droplets to pass through at least a part of the laser beam. The direction of the laser beam and/or the stream of droplets may be controlled to control a property of radiation generated by the radiation source. An angle between the stream of droplets and the direction of the laser beam may be less than about 90°. The third aspect of the invention may include, where appropriate, one or more features of the first or second aspects of the present invention. According to a fourth aspect of the invention, there is provided a method of generating radiation, the method comprising: directing a stream of droplets of fuel to a plasma generation site; directing a laser beam to the plasma generation site; and controlling the direction of the laser beam and/or the direction of movement of the stream of droplets to control the angle between the laser beam and the stream of droplets. The fourth aspect of the invention may include, where appropriate, one or more features of the first, second or third aspects of the present invention. According to a fifth aspect of the invention, there is provided a fuel droplet-laser beam collision time increasing apparatus, that is also arranged to generate radiation, the apparatus comprising: a fuel droplet generator constructed and arranged to generate a stream of droplets of fuel that are directed to a plasma generation site; and a laser constructed and arranged to generate a laser beam that is directed to the plasma generation site, an angle between the direction of movement of the stream of droplets and the direction of the laser beam being less than about 90°. The fifth aspect of the invention may include, where appropriate, one or more features of the first, second, third or fourth aspects of the present invention. According to a sixth aspect of the invention, there is provided a method of increasing the collision time between a droplet of fuel and at least a part of a laser beam, the collision resulting in the generation of radiation, the method comprising: directing a stream of droplets of fuel to a plasma generation site; directing a laser beam to the plasma generation site, an angle between the stream of droplets and the direction of the laser beam being less than about 90°. The sixth aspect of the invention may include, where appropriate, one or more features of the first, second, third, fourth or fifth aspects of the present invention. According to a seventh aspect of the invention, there is provided a lithographic apparatus comprising: a radiation source according to the first or third aspects of the present invention, or a fuel droplet-laser beam collision time increasing apparatus according to the fifth aspect of the present invention; a support constructed and arranged to support a patterning device, the patterning device being configured to pattern radiation that passes through the intermediate focus; and a projection system constructed and arranged to project the patterned radiation onto a substrate. The seventh aspect of the invention may include, where appropriate, one or more features of the first, second, third, fourth, fifth or sixth aspects of the present invention. According to a eighth aspect of the invention, there is provided a device manufacturing method comprising: generating radiation using the method according to the second, fourth or sixth aspects of the present invention; patterning the radiation with a pattering device; and projecting the patterned radiation onto a substrate. The eighth aspect of the invention may include, where appropriate, one or more features of the first, second, third, fourth, fifth, sixth or seventh aspects of the present invention. FIG. 1 schematically depicts a lithographic apparatus 1 according to one embodiment of the invention. The apparatus comprises an illumination system (illuminator) IL configured to condition a radiation beam B (e.g. EUV radiation); a support structure (e.g. a mask table) MT constructed to support a patterning device (e.g. a mask or a reticle) MA and connected to a first positioner PM configured to accurately position the patterning device; a substrate table (e.g. a wafer table) WT constructed to hold a substrate (e.g. a resist-coated wafer) W and connected to a second positioner PW configured to accurately position the substrate; and a projection system (e.g. a reflective projection lens system) PS configured to project a pattern imparted to the radiation beam B by patterning device MA onto a target portion C (e.g. comprising one or more dies) of the substrate W. The illumination system may include various types of optical components, such as refractive, reflective, magnetic, electromagnetic, electrostatic or other types of optical components, or any combination thereof, for directing, shaping, or controlling radiation. The support structure holds the patterning device in a manner that depends on the orientation of the patterning device, the design of the lithographic apparatus, and other conditions, such as for example whether or not the patterning device is held in a vacuum environment. The support structure can use mechanical, vacuum, electrostatic or other clamping techniques to hold the patterning device. The support structure may be a frame or a table, for example, which may be fixed or movable as required. The support structure may ensure that the patterning device is at a desired position, for example with respect to the projection system. The term “patterning device” should be broadly interpreted as referring to any device that can be used to impart a radiation beam with a pattern in its cross-section such as to create a pattern in a target portion of the substrate. The pattern imparted to the radiation beam may correspond to a particular functional layer in a device being created in the target portion, such as an integrated circuit. The patterning device may be transmissive or reflective. Examples of patterning devices include masks, programmable mirror arrays, and programmable LCD panels. Masks are well known in lithography, and include mask types such as binary, alternating phase-shift, and attenuated phase-shift, as well as various hybrid mask types. An example of a programmable mirror array employs a matrix arrangement of small mirrors, each of which can be individually tilted so as to reflect an incoming radiation beam in different directions. The tilted mirrors impart a pattern in a radiation beam which is reflected by the mirror matrix. The term “projection system” may encompass any type of projection system, including refractive, reflective, catadioptric, magnetic, electromagnetic and electrostatic optical systems, or any combination thereof, as appropriate for the exposure radiation being used, or for other factors such as the use of an immersion liquid or the use of a vacuum. It may be desired to use a vacuum for EUV or electron beam radiation since other gases may absorb too much radiation or electrons. A vacuum environment may therefore be provided to the whole beam path with the aid of a vacuum wall and vacuum pumps. As here depicted, the apparatus is of a reflective type (e.g. employing a reflective mask). Alternatively, the apparatus may be of a transmissive type (e.g. employing a transmissive mask). The lithographic apparatus may be of a type having two (dual stage) or more substrate tables (and/or two or more mask tables). In such “multiple stage” machines the additional tables may be used in parallel, or preparatory steps may be carried out on one or more tables while one or more other tables are being used for exposure. Referring to FIG. 1, the illuminator IL receives a radiation beam from a radiation source SO. The source and the lithographic apparatus may be separate entities, for example when the source is an excimer laser. In such cases, the source is not considered to form part of the lithographic apparatus and the radiation beam is passed from the source SO to the illuminator IL with the aid of a beam delivery system comprising, for example, suitable directing mirrors and/or a beam expander. In other cases the source may be an integral part of the lithographic apparatus, for example when the source is a mercury lamp. The source SO and the illuminator IL, together with the beam delivery system if required, may be referred to as a radiation system. The illuminator IL may comprise an adjuster for adjusting the angular intensity distribution of the radiation beam. Generally, at least the outer and/or inner radial extent (commonly referred to as σ-outer and σ-inner, respectively) of the intensity distribution in a pupil plane of the illuminator can be adjusted. In addition, the illuminator IL may comprise various other components, such as an integrator and a condenser. The illuminator may be used to condition the radiation beam, to have a desired uniformity and intensity distribution in its cross-section. The radiation beam B is incident on the patterning device (e.g., mask) MA, which is held on the support structure (e.g., mask table) MT, and is patterned by the patterning device. After being reflected from the patterning device (e.g. mask) MA, the radiation beam B passes through the projection system PS, which focuses the beam onto a target portion C of the substrate W. With the aid of the second positioner PW and position sensor IF2 (e.g. an interferometric device, linear encoder or capacitive sensor), the substrate table WT can be moved accurately, e.g. so as to position different target portions C in the path of the radiation beam B. Similarly, the first positioner PM and another position sensor IF1 can be used to accurately position the patterning device (e.g. mask) MA with respect to the path of the radiation beam B. Patterning device (e.g. mask) MA and substrate W may be aligned using mask alignment marks M1, M2 and substrate alignment marks P1, P2. The depicted apparatus could be used in at least one of the following modes: 1. In step mode, the support structure (e.g. mask table) MT and the substrate table WT are kept essentially stationary, while an entire pattern imparted to the radiation beam is projected onto a target portion C at one time (i.e. a single static exposure). The substrate table WT is then shifted in the X and/or Y direction so that a different target portion C can be exposed. 2. In scan mode, the support structure (e.g. mask table) MT and the substrate table WT are scanned synchronously while a pattern imparted to the radiation beam is projected onto a target portion C (i.e. a single dynamic exposure). The velocity and direction of the substrate table WT relative to the support structure (e.g. mask table) MT may be determined by the (de-)magnification and image reversal characteristics of the projection system PS. 3. In another mode, the support structure (e.g. mask table) MT is kept essentially stationary holding a programmable patterning device, and the substrate table WT is moved or scanned while a pattern imparted to the radiation beam is projected onto a target portion C. In this mode, generally a pulsed radiation source is employed and the programmable patterning device is updated as required after each movement of the substrate table WT or in between successive radiation pulses during a scan. This mode of operation can be readily applied to maskless lithography that utilizes a programmable patterning device, such as a programmable mirror array of a type as referred to above. Combinations and/or variations on the above described modes of use or entirely different modes of use may also be employed. FIG. 2 shows the apparatus of FIG. 1 in more detail, including the source SO, the illuminator IL, and the projection system PS. The source SO generates EUV radiation from a plasma 2. The plasma 2 is created by directing a laser beam 4 onto droplets of a suitable material such as Sn or Gd which are generated by a droplet generator 6. The laser beam 5 causes the droplets to be vaporized, thereby generating the plasma 2. Radiation 7 emitted by the plasma 2 is collected by a collector 8, and is directed onto a grating spectral filter 10. The radiation 7 then passes from the grating spectral filter 10 to an intermediate focus 12. The intermediate focus 12 acts as a virtual source point at an aperture in the source SO. From source SO, the radiation 7 forms a beam of radiation 14 that is reflected in the illuminator IL via first and second normal incidence reflectors 16, 18 onto a patterning device (e.g. a mask) positioned on support structure MT. A patterned beam 20 is formed which is imaged in the projection system PS via first and second reflective elements 22, 24 onto a substrate (not shown) held on a substrate table WT. More elements than shown may generally be present in the illuminator IL and projection system PS. Greater or fewer elements than shown in the FIG. 2 may generally be present in the illuminator IL and projection system PS. For example, there may be greater or fewer mirrors present than those shown in FIG. 2. FIG. 3 shows an embodiment of a radiation source SO, in cross-section, including a normal incidence collector 8. The collector 8 has an elliptical configuration, having two natural ellipse focus points F1, F2. Particularly, the normal incidence collector includes a collector having a single radiation collecting surface 30 having the geometry of the section of an ellipsoid. The ellipsoid radiation collecting surface section 30 extends along a virtual ellipsoid (part of which is depicted by as dotted line E in the drawing). As will be appreciated by the skilled person, because the collector mirror 8 is ellipsoidal (i.e., including a reflective radiation collecting surface 30 that extends along an ellipsoid), the collector mirror 8 focuses radiation from one focal point F1 into another focal point F2. The focal points are located on the long axis of the ellipsoid at a distance f=(a2−b2)1/2 from the center of the ellipse, where 2a and 2b are the lengths of the major and minor axes, respectively. The long axis of the ellipsoid defines an optical axis OA for the radiation source SO. If the embodiment shown in FIG. 1 includes a laser produced plasma (LPP) radiation source SO, the collector 8 may be a single ellipsoidal mirror as shown in FIG. 3, where a plasma formation site 32 is positioned in one focal point (F1) and an intermediate focus IF is established at the other focal point (F2) of the mirror. Radiation 7 is shown as emanating from the plasma generated at the plasma formation site 32, located in the first focal point (F1), towards the reflective radiation collecting surface 30 and being reflected by that surface towards the second focus point F2. For example, according to an embodiment, a mentioned intermediate focus IF may be located between the collector and an illuminator (see FIG. 1 or FIG. 2) of the lithographic apparatus, or be located in the illuminator, if desired. In the present embodiment, the source SO is an LPP source, that is associated with a laser source configured to generate a laser beam of coherent light, having a predetermined wavelength. A beam of radiation (e.g., light) from the laser is focused onto a fuel (the fuel for example being supplied by a fuel supplier such as a fuel or droplet generator, and for example including fuel droplets) to generate a plasma and radiation therefrom, in a laser produced plasma process. The resulting radiation may be EUV radiation, in this embodiment. In a non-limiting embodiment, the laser source may be a so-called CO2 laser, and the fuel may be tin (Sn), or a different type of fuel, as will be appreciated by the skilled person. The plasma formed at the plasma formation site 32 may be configured to emit diverging radiation, and the collector 8 may be arranged to reflect that diverging radiation to provide a converging radiation beam that converges towards the intermediate focus and along an optical axis of the source (as shown in FIG. 3). In an embodiment, the plasma is created by a pulsed high energy CO2 laser source which shoots (i.e. is arranged to direct) a radiation beam onto a fuel target in the form of a droplet. FIG. 4 shows that droplets 34 may be formed by a (for example, tin) droplet generator 6. In an embodiment, a contamination barrier (not shown), which may also be called a debris mitigation system and may be in the form of a foil trap, may be present between the plasma formation site 32 and the collector 8. In the embodiment illustrated in FIG. 4, a droplet of fuel 34 may move out of focus of the laser beam 4 for faster droplet speeds and long laser pulses, such that not all available laser pulse energy can be used to convert the fuel droplet 34 into radiation. The droplet 34 and the laser beam pulse 4 are desirably synchronized in time so that the efficiency of the radiation source may be improved. It is desirable to have the position of the plasma very stable over time for optical reasons. For example, the positioning of the reflective surface of the collector 8, as well as other optical elements that are optically downstream of the source, may be more optimally positioned if the position of the plasma is stable over time, which may improve the efficiency of the radiation source and the lithographic apparatus. Advanced feedback loops may be desired, both for time synchronization and for position stability of the plasma (these feedback loops are desirably coupled). Even with control systems in place, light loss may occur for faster droplets and/or for longer laser pulses. FIG. 5 shows a laser beam pulse 4 and a fuel droplet 34. The droplet 34 is shown as moving toward the laser beam pulse 4 in a direction that is perpendicular to the direction of the laser beam pulse 4. The time that a plasma from the droplet 34 can be created is given by: t plasma = l droplet - laser v droplet if t laser _ pulse > t plasma ( 2 ) Where tplasma is the time that the plasma can last, ldroplet-laser is the path length of the droplet 34 through the laser beam pulse 4 (of sufficient intensity), vdroplet is the speed of the droplet 34 and tlaser—pulse is the duration of the laser beam pulse 4. FIG. 6 shows that the laser beam pulse 4 can be approximated as a cylinder with a diameter dlaser and an effective length in the direction of the laser beam pulse of llaser. FIG. 6 also shows the droplet 34 moving toward the laser beam pulse 4 in a direction that is not perpendicular to the direction of the laser beam pulse 4. This may be advantageous, as discussed below. FIG. 7 again shows the droplet 34 moving toward the laser beam pulse 4 in a direction 36 that is not perpendicular to the direction of the laser beam pulse 4. Specifically, the Figure shows the droplet 34 moving toward the laser beam pulse 4 at an angle α with respect to the direction of the laser beam pulse 4. The path length ldroplet-laser of the droplet 34 through the laser beam pulse 4 is then approximately: l droplet - laser = d laser sin α , as long as l laser > d laser tan α ( 3 ) Provided that the duration of the laser beam pulse 4 is long enough, better use can be made of laser beam pulse so as to increase the plasma lifetime and to increase the optical output of the radiation source. For example, when α is greater than or less than about 90°, it can be seen that the path length of the droplet 34 through the laser beam pulse 4 is greater in comparison with the situation shown in and described with reference to FIG. 5, where the droplet 34 is shown as moving toward the laser beam pulse 4 in a direction that is perpendicular to the direction of the laser beam pulse 4. It will be understood that the path length of a droplet passing through the laser beam pulse may also be described as a collision length or distance, since the laser beam is colliding with the droplet over this path length. Similarly, it will be understood that the time taken for a droplet to pass through the laser beam pulse may also be described as a collision time, since the laser beam is colliding with the droplet over this time. An embodiment of the present invention may therefore be described as a droplet-laser beam collision time increasing apparatus, and/or as a method of increasing the collision time. In an embodiment, the angle α is less than about 85°. In an embodiment, the angle α is less than about 80°. In an embodiment, the angle α is less than about 75°. In an embodiment, the angle α is less than about 70°. In an embodiment, the angle α is less than about 65°. In an embodiment, the angle α is less than about 60°. In an embodiment, the angle α is less than about 55°. In an embodiment, the angle α is less than about 50°. In an embodiment, the angle α is less than about 45°. In an embodiment, the angle α is less than about 40°. In an embodiment, the angle α is less than about 35°. In an embodiment, the angle α is less than about 30°. In an embodiment, the angle α is less than about 25°. In an embodiment, the angle α is less than about 20°. In an embodiment, the angle α is less than about 15°. In an embodiment, the angle α is less than 10°. In an embodiment, the angle α is less than about 5°. In accordance with an embodiment of the present invention, because the time taken for a droplet of the stream of droplets to pass through at least a part of the laser beam pulse is longer (i.e. the path length of the droplet through the laser beam pulse is longer), the total optical output of the radiation source may be higher, provided that the laser beam pulse lasts long enough. For example, the EUV output may be increased by a factor of or proportional to 1/(sin α). It will therefore be appreciated that the direction of the laser beam and/or the stream of droplets maybe controlled to control a property of radiation generated by the radiation source (e.g. the duration or intensity of the radiation). Another potential advantage of embodiments of the present invention is that the synchronization between the timing of the laser beam pulse and the generation of droplets may become less demanding, since the droplet is in the laser beam pulse for a longer period of time. As a result of embodiments of the present invention, the EUV emitting plasma volume (i.e., plasma formation site) may become larger. As long as the larger volume has no negative impact on the optical properties of a so-called optical column of the lithographic apparatus, this may relax the position stability specification, since the shadows in the far field may become less sharp. Less sharp shadows mean that small plasma position variations may lead to less intensity gradients on a far field mirror when the plasma position moves. Any shift in the position of the plasma position may become relatively smaller, since the plasma itself is larger. The laser (and thus the direction of the laser beam pulses) may be aligned with the optical axis of the source such that the beam of radiation provided by the laser is substantially in line with the optical axis of the source. However, the laser (and thus the direction of the laser beam pulses) does not need to be aligned with the optical axis of the source such that the beam of radiation provided by the laser is substantially in line with the optical axis of the source. The laser may direct laser beam pulses at any appropriate angle, and/or the droplet generator may direct droplets at any appropriate angle. FIGS. 8a and 8b show two such examples. FIG. 8a shows that the droplets 34 and the laser beam pulses 4 may be directed through an aperture 38 provided in the collector 8. This arrangement may, for example, save space in and around the collector. In other embodiments, only one of the laser beam pulses and the droplets may be directed through such an aperture. FIGS. 8a and 8b also show that, for example, a controller CO may be provided. The controller CO may be in communication with the droplet generator 6 and/or the laser that provides the laser beam pulses 4 to control the angles at which the droplets and laser beam pulse are directed. Alternatively, the controller CO may be in communication with arrangement which control the position or orientation of the droplet generator 6 and/or the laser that provides the laser beam pulses 4 to control the angles at which the droplets and laser beam pulse are directed. The controller CO may be used to control the angle α referred to above, in order to control the time taken for a droplet to pass through the laser beam pulse. The controller CO may thus be used to control a to control, for example, the amount of radiation that is generated by the radiation source. The controller may be a part of the radiation source. The controller may be a part of the lithographic apparatus, and be in communication with the droplet generator and/or laser. The controller may obtain or be provided with data at least indicative of a direction in which the droplets and/or laser beam is/are to be directed. The controller may, for example, be a computer or the like. The droplets may be directed in a direction that is perpendicular to the optical axis of the radiation source. In another embodiment, the droplets may be directed in a direction that is not perpendicular to the optical axis of the radiation source. The laser beam pulse and/or the droplet may have a component of movement along the optical axis of the radiation source which is pointing away from the collector. This may reduce the amount of contamination that is incident on the collector during or after the formation of the plasma. This may alternatively and/or additionally increase the time during which the droplet may pass through the laser beam pulse. In an embodiment, the laser may be a CO2 laser. Other lasers may be used, as is known in the art. The laser may generate a continuous laser beam, and/or one or more laser beam pulses that make up the laser beam. In order to ensure that the droplets are directed in a certain direction, the droplets may not be dropped. Instead, the droplets may be fired, projected or the like. The firing or projection may be such that the direction of movement of the droplets does not significantly change from the time of firing/projection until the time at which the laser beam pulse is incident on the droplet. Such constant and consistent direction may not always be possible, for example during the start-up time and shut-down time of the droplet generator. During these times, droplets may fall short of the plasma formation site. A catchment arrangement may thus be provided for catching droplets that are not directed at the desired angle (i.e. toward the plasma generation site) and which might otherwise contaminate one or more parts of the lithographic apparatus or radiation source. The catchment arrangement may be a receptacle or a gutter or the like. Although specific reference may be made in this text to the use of lithographic apparatus in the manufacture of ICs, it should be understood that the lithographic apparatus described herein may have other applications, such as the manufacture of integrated optical systems, guidance and detection patterns for magnetic domain memories, flat-panel displays, liquid-crystal displays (LCDs), thin-film magnetic heads, etc. Although specific reference may have been made above to the use of embodiments of the invention in the context of optical lithography, it will be appreciated that the invention may be used in other applications, for example imprint lithography, and where the context allows, is not limited to optical lithography. The terms “radiation” and “beam” used herein encompass all types of electromagnetic radiation, including ultraviolet (UV) radiation (e.g. having a wavelength of or about 365, 355, 248, 193, 157 or 126 nm) and extreme ultra-violet (EUV) radiation (e.g. having a wavelength in the range of 5-20 nm), as well as particle beams, such as ion beams or electron beams. While specific embodiments of the invention have been described above, it will be appreciated that the invention may be practiced otherwise than as described. For example, the invention may take the form of a computer program containing one or more sequences of machine-readable instructions describing a method as disclosed above, or a data storage medium (e.g. semiconductor memory, magnetic or optical disk) having such a computer program stored therein. The descriptions above are intended to be illustrative, not limiting. Thus, it will be apparent to one skilled in the art that modifications may be made to the invention as described without departing from the scope of the claims set out below. |
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055241310 | summary | FIELD OF THE INVENTION AND RELATED ART The present invention relates to an alignment apparatus for aligning a mask having a semiconductor chip pattern and a semiconductor wafer in a predetermined positional relationship, more particularly to a semiconductor chip manufacturing SOR (synchrotron orbital radiation) X-ray exposure apparatus wherein after the mask and the semiconductor wafer are aligned, a resist on the semiconductor wafer is exposed to the orbital radiation rays (SOR X-rays) in the form of a pattern corresponding to the pattern of the semiconductor chip on the mask. Lithography using X-rays for semiconductor chip manufacturing has been noted as a fine lithographic technique to realize high density semiconductor chips and was proposed by Spears and Smith in "Electron Lett. Vol. 8, No. 4: P.102, 1972" in 1972. Since, however, there has not been a small size and high power X-ray source, it has been difficult to install a semiconductor chip manufacturing X-ray exposure apparatus using X-ray lithography in the semiconductor manufacturing plant. In addition, it has not met the needs for mass-production, because of low throughput and the like. Recently, however, a very small size SOR ring has been developed, capable of providing high power X-rays with the use of a normal conductor or super conductor magnet. Therefore, the X-ray source is no longer a major problem. Referring first to FIG. 2, in the X-ray exposure apparatus, pattern exposure is effected in a so-called proximity exposure method. The mask 2 comprises a substrate 201 having a thickness of several microns with high X-ray transmissibity, an absorber material 202 having a high X-ray absorptivity formed into a pattern to be printed on the wafer and a supporting frame 203 for supporting the substrate 201. The mask 2 is opposed to the wafer 3, which is the workpiece, with a predetermined gap (approximately several tens of microns) with precision and stability. The X-rays 1 are applied to the semiconductor wafer 3 through the mask 2, so that the resist 5 applied on the wafer 3 is exposed to the pattern determined by the absorber material 202, by which the pattern is transferred onto the wafer 3. As for the pattern transfer process, there have been proposed a scanning exposure system shown in FIG. 3A, scanning mirror exposure system shown in FIG. 3B and a whole surface exposure system shown in FIG. 3C. The scanning exposure system and the whole surface exposure system have been proposed in "J. Vac. Sci. Technol. B1 (4) 1984, p.1271" and in "IBM Research Report RC 8220, 1980", respectively Referring to FIG. 3A, the scanning exposure system is such that after the mask 2 and the wafer 3 are aligned in a predetermined positional relation, the mask 2 and the wafer 3 are moved as a unit in a direction indicated by an arrow relative to the X-rays 1 in the form of a sheet beam produced by the SOR source 4 to transfer the pattern of the mask 2 onto a predetermined area of the wafer 3. In the scanning mirror exposure system shown in FIG. 3B, after the mask 2 and the wafer 3 are aligned in a predetermined relation, a mirror 301 disposed between the SOR source 4 and the mask 2 is swung in the direction indicated by an arrow to scan the exposure area (the entire area of the mask pattern to be transferred) with the X-rays 1 from the SOR source 4, thus transferring the pattern. In the whole surface exposure system shown in FIG. 3C, a mirror 302 having a convex reflecting surface is disposed between the SOR source 4 and the mask 2 to diverge the X-rays 1 from the SOR source 4. The diverged X-rays 1 are projected simultaneously onto the entire exposure area. In this manner, the mask pattern is transferred onto the wafer 2 after alignment therebetween. E. S. Piller proposes in "JT Applied Physics, Vol. 47, No. 12, p. 5450" that the mask 2 and the wafer 3 are disposed within a predetermined closed ambience in an X-ray exposure apparatus, and then the pattern exposure is carried out. It is also proposed therein that the mask 2 and the wafer 3 are disposed in a He gas ambience from the standpoint of temperature. Furthermore, common inventors have proposed in Japanese Laid-Open Patent Application Publication 178625/1985 that in an X-ray exposure apparatus using an X-ray tube, a state apparatus, mask and wafer transportation apparatus, an alignment apparatus or the like are disposed in a plurality of closed ambiences. Referring to FIGS. 4A and 4B, the apparatus disclosed in the above-mentioned Japanese Laid-Open Patent Application will be described. The wafer is contained in a cassette, which is contained in turn in a wafer loading cassette accommodating chamber 402. The exposed wafers are contained in a cassette in a wafer unloading cassette accommodating chamber 403. The mask having a pattern to be transferred is contained in a cassette, which is in turn contained in a mask cassette accommodating chamber 404. The mask and the wafer are aligned with each other by an electron beam device 411 shown in FIG. 4B in the main chamber 401, and thereafter they are shifted as a unit into an illumination chamber 405, where the pattern exposure is effected with the use of the radiation source, that is, the X-ray rube 410. Between the chambers, shut-off valves are disposed between respective chambers to maintain the ambiences thereof independently from each other, although FIG. 4B shows only the shut off valve 409 between the illumination chamber 405 and the main chamber 401. An SOR X-ray exposure apparatus is proposed in "Proceeding of SPIE, Vol 448, 1983, p 104", for example. FIG. 5 shows this apparatus schematically. This apparatus includes a so-called vertical stage which is movable in the vertical direction. The pattern exposure to the X-rays 1 is performed with the mask 2 and the wafer 3 being supported on the stage. SUMMARY OF THE INVENTION Taking as an example, a dynamic random access memory, which will hereinafter be called "DRAM" is recently a standard of the density of the semiconductor chip or device, 64 MB DRAM requires the line width of 0.3-0.4 micron, and more than 100 MB DRAM requires 0.25 microns of line width, and therefore, an exposure apparatus capable of pattern transfer of this dimension. However, no exposure apparatus capable of such a fine pattern exposure as not more than 0.25 micron has been made practical. In order to accomplish such an exposure apparatus, it is necessary to establish (1) ambience control for stably maintaining the mask and the wafer, (2) removal of contamination such as dust, (3) X-ray exposure for uniformly exposing the pattern transfer area, (4) an alignment with precision on the order of 1/100 micron and the resolution on the order of 1/1000 micron and others. Accordingly, it is a principal object of the present invention to provide an exposure apparatus which executes key processes in the semiconductor device manufacturing, capable of manufacturing semiconductor devices of 64 MB or higher density. It is another object of the present invention to provide an SOR X-ray exposure apparatus which uses X-rays from the SOR source and which can transfer a fine pattern on the mask onto a resist on a semiconductor wafer. It is a further object of the present invention to satisfy the requirements of the above-described (1)-(4) to accomplish a practical X-ray exposure apparatus. In order to realize manufacturing of the semiconductor devices having a pattern including lines of approximately 0.25 microns width, very high precision is required for the exposure apparatus. Table 1 shows various factors of the registration accuracy which is the major item of accuracy required for the semiconductor exposure apparatus and allocations thereof. TABLE 1 ______________________________________ (unit: micron) Items Reg. Precision ______________________________________ Mask manufacturing error 0.025 Wafer processing error 0.025 Stage accuracy 0.025 Alignment accuracy 0.025 Optical system 0.025 Other errors in exp. apparatus 0.025 Registration accuracy 0.06 ______________________________________ The "other errors in the exposure apparatus" in the above Table contains deformation or strains of the mask and the wafer due to heat. As will be understood from Table 1, the deformation of the mask and the wafer permitted in the exposure apparatus, the deformation permitted during the exposure, are 0.01 micron at maximum. Referring to FIGS. 6A and 6B, aspects of a typical SOR source 4 will be described. In FIG. 6A, the SOR source 4 is schematically shown as an orbit of electrons. The X-rays 1 are produced by brmsstrahlung of electrons accelerated to an extent of a relativistic velocity. The X-rays thus produced have a power profile which is Gaussian, in the vertical direction V with a divergence angle of several mRad. at maximum and with substantial uniformity in the horizontal direction H over a length open to the orbit of the electrons. The profile is generally trapezoidal. This profile is the one detected by an X-ray detector 1551 shown in FIG. 15 for example, at an exposure position where the mask and the wafer are present during exposure. A wavelength used in the SOR X-ray lithography according to the present invention is within a range of 5-15 angstroms. The electron energy accelerated to the relativistic velocity is approximately 0.5-1.0 GeV. FIG. 6B shows spectral strength characteristics for the respective wavelengths at various points on a path of the exposure beam, more particularly, a position immediately before the mirror, the position immediately after the mirror, a surface of the Be window, a mask surface and a resist surface in the order from the SOR source 4, when the X-ray exposure apparatus is constituted by disposing the mirror for reflecting the X-rays and the Be window between the SOR source 4 and the mask. It is understood from this Figure, 5-15 angstrom wavelength range is preferable in SOR X-ray lithography. From the same, it is understood that the SOR has continuous spectral characteristics. FIG. 7A shows a mask using inorganic material for the substrate 201, and FIG. 7B shows the same using organic film. In FIG. 7A, the substrate 201 joined to the frame 203 is constituted by a silicon wafer, and a silicon nitride, for example, is laminated in thin film in a pattern area 701 in which an absorber material pattern is to be formed. Or, the silicon wafer is laminated by etching. In this case, the thickness of the substrate in the pattern area 701 is approximately 2 microns. In the example of FIG. 7B, an organic film having a thickness of 20-10 microns is bonded to the frame 203. An example of the material of the organic film, there is polyimide. Referring back to FIG. 2 showing the proximity exposure system, the distance between the mask and the wafer during the exposure, which will hereinafter be called "proximity gap", is approximately 10-50 microns. Table 2 shows temperature rise of the silicon nitride mask of FIG. 7A and a polyimide mask of FIG. 7B when the ambience in the space between the mask and the wafer is vacuum, air or He. TABLE 2 ______________________________________ (Unit: degree) Ambient Medium Silicon Nitride Mask Polyimide Mask ______________________________________ Vacuum .sctn.60 .sctn.60 Air 1.17 1.18 He 0.343 0.351 ______________________________________ The data of the above Table is on the basis of the X-ray power of 120 mW/cm on the mask surface and on the thermal emissivity of 0.5 in the vacuum. In view of the facts that the sensitivity of the resist material at present is approximately several tens mJoule--one hundred mJoule/cm.sup.2 and that the transmissivity of the mask substrate to the X-rays is approximately 50%, the above data are appropriate. As will be apparent from the Table, in order to prevent thermal strain, it is preferable that the space between the mask and the wafer is filled with He gas. Therefore, in the present invention, the alignment between the mask and the wafer and subsequent pattern exposure are performed while the mask and the wafer are within a He ambience chamber. By containing the mask and the wafer in the chamber, the problems of contamination, such as dust, can be solved. Further, in the apparatus of the present invention, the temperature, pressure and purity of the gas functioning as a thermal conduction medium in the chamber are controlled with high precision so as to stabilize the X-ray transmissivity in this ambience, as in a vacuum. Referring to FIGS. 8A, 8B, 9A and 9B, the differences between the whole surface exposure system and the scanning exposure system (scanning mirror exposure system) will be described. FIG. 8A illustrates the whole surface exposure system wherein the X-rays 1 are applied simultaneously to the entirety of the pattern area 701. FIG. 8B illustrates strains of the mask resulting from this system. FIG. 9A illustrates the scanning exposure system wherein the X-rays 1 in the form of a sheet beam scan the pattern area 701 sequentially. FIG. 9B shows the mask strains or deformations resulting from this system. The X-rays produced from the SOR have a very small divergence in the vertical direction, and therefore, the half peak width of the X-ray power profile is only approximately 10 mm even if the exposure position is away from the SOR emitting point by a distance of 10 m. Therefore, in order to reduce the exposure period of the scanning exposure system (the same in the scanning mirror exposure system) to such an extent as being comparable to that of the whole surface exposure system, the strength of the X-ray applied to the mask should be several times that of the whole surface exposure system. This increases the mask strain. Table 3 shows the temperature rise and the mask deformation in the whole surface exposure system as compared with those in the scanning mirror exposure system. TABLE 3 ______________________________________ (Unit: degree, micron) Silicon Nitride Mask Polyimide Mask Exposure Temp. Max. Temp. Max. System Rise Strain Rise Strain ______________________________________ Scanning mirror 0.5 Hz 1.34 0.016 1.40 0.094 8 Hz 1.22 0.015 1.22 0.082 1000 Hz 0.343 0.0084 0.351 0.046 Whole 0.343 0.0084 0.351 0.046 Surface Exp. ______________________________________ The data of this Table are based on the material of the wafer chuck 1807 (FIG. 18) being alumina (Al) having a thickness of 0.5 mm under the condition that the back surface temperature is constant. With respect to the scanning mirror exposure system, the mirror swing is taken as a parameter. As will be understood from the above Table, the mask strain can be made under the tolerable level (0.01 micron) only by the whole surface exposure system and the scanning mirror exposure system, the latter being possible only when the mirror is swung at a high frequency. Also, it is understood that a usual polyimide film is not usable as a mask substrate. However, in consideration of the situation wherein the mirror 301 (FIG. 3B) is located in a high vacuum ambience of approximately 10.sup.-9 Torr., it is very difficult to realize that the mirror is swung at a high frequency over several tens of Hertz. Further, the difficulty is more significant in the scanning exposure system wherein the mask and the wafer are moved together. Therefore, the present invention employs the whole surface exposure system. In the exposure apparatus, the accuracy of the pattern line width is required to be 5-10% of the minimum width of the line to be transferred. In the exposure apparatus capable of exposing 0.25 micron of the minimum line width, the pattern line width accuracy is 0.012 micron. This means that the uniformity of the X-ray illuminance over the entire exposure area (pattern area 701) or the uniformity of the amount of the exposure over the entire exposure area is required to be approximately +2.5%. On the other hand, the strength profile of the SOR X-rays, as described hereinbefore, is in the form of the Gaussian function in the vertical plane, and therefore, it is not easy to profile the uniformity of approximately .+-.2.5% in the amount of the exposure over the entire exposure area in the whole surface exposure system. If the illuminance is made uniform over the entire exposure area, it cannot be avoided to use only the central portion of the Gaussian distribution, with the result that the efficiency of the X-ray energy is low. The present invention, adopts the exposure system shown in FIGS. 10A and 10B so as to solve those problems. In this system, the X-rays from the emitting point 1001 in the SOR source are incident on the X-ray mirror 1002 at a low glancing angle, and the X-rays diverged by the mirror 1002 are applied on the mask. The mirror 1002 is designed such that the minimum illuminance of the X-ray strength profile in the exposure area is maximum under the set conditions of the exposure apparatus. FIG. 11 shows the X-ray strength profile in the exposure area in this example. As will be understood, the illuminance is significantly different at the central area and the marginal areas of the exposure area. In this system, the illumination distribution can include .+-.10% difference in the exposure area. However, in the exposure system, the non-uniformity of the illuminance is corrected by the shutter mechanism 1003 during the exposure, so that the uniform exposure is effected over the entire exposure area. The shutter mechanism will be briefly described in conjunction with FIG. 10B. A steel belt 1009 is trained between a driving drum 1009 and an idler drum 1001 and is provided with a rectangular aperture 1012 having a leading edge 1004 and a trailing edge 1005. The Y axis is perpendicular to the optical axis of the SOR X-ray and is substantially vertical. A t-axis represents time. A curve 1006 indicates movement of the leading edge 1004, and a curve 1007 indicates the movement of the trailing edge 1005. The shutter mechanism drives the driving drum 1009 so that the exposure period .DELTA.T(y) at each point on the Y axis is different corresponding to the illuminance profile shown in FIG. 11, by which the amount of exposure (=exposure period.times.illuminance) is uniform over the entirety of the exposure area. In the exposure apparatus (FIG. 5) disclosed in the above-mentioned "Proceeding of SPIE, Vol. 448, 1983, p 104" the path of the X-ray from SOR source is once stopped by a Be window, and thereafter, the wafer is exposed to the X-ray through the wafer in the air. FIG. 12 shows the dependency of the thermal conductivity to pressure for air and He. It will be understood from this Figure that air has a lower thermal conductivity than He under the same pressure. Therefore, in consideration of the strain of the mask described with respect to Tables 2 and 3, it is difficult to achieve the object of the present invention by exposure in air. In consideration of this, the present invention adopts exposure in the closed He ambience for which the temperature, pressure and purity are controlled with high precision. In addition, in order to meet the SOR X-ray source, the conveying mechanism and the exposure stage are of vertical type, and the mask and the wafer are conveyed within a completely closed ambience. By this, in the SOR X-ray exposure apparatus, a high throughput and reduction of influence by particles (dust) and contamination are achieved. Furthermore, in the present invention, a wafer stage is controlled in 6 axes (X, Y, Z, .THETA. (=.omega.z, .omega.x, .omega.y)), and the mask stage is controlled in the .THETA. axis only, so as to achieve the high accuracy of the mask stage, in view of (1) that the optical axis of the X-ray from the SOR source hardly changes, (2) that the accuracy of the mask is sufficient as described hereinbefore and (3) that in the SOR X-ray exposure, the stability in the relation between the optical axis of the X-ray from the SOR source and the mask is most important. In the present invention, the mask stage is rotatable about the .THETA. direction in order to align the orientation of the mask with the movement direction in X and Y axes of the wafer which is step-and-repeat-exposed. In order to accomplish this mask alignment with great precision, a reference mark is provided on the wafer stage, and the mask alignment is effected using the reference mark. These and other objects, features and advantages of the present invention will become more apparent upon a consideration of the following description of the preferred embodiments of the present invention taken in conjunction with the accompanying drawings. |
summary |
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